RESEARCH REACTOR UTILIZATION

IAEA-147

RESEARCH REACTOR UTILIZATION

PROCEEDINGS OF A STUDY GROUP MEETING ON RESEARCH REACTOR UTILIZATION

SPONSORED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY AND HELD IN BANDUNG, INDONESIA.

FROM 2 TO 6 AUGUST 1971

A TECHNICAL REPORT PUBLISHED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1972

T h e IA E A does not m aintain stocks o f reports in th is series. How ever, m icrofiche copies o f these reports can be obta ined from

IN IS M icro fich e C learinghouse In ternational A tom ic Energy Agency Kärntner R ing 11 P .O . Box 590 A - 1011 V ienna, A u stria

on prepaym ent o f US $ 0 .6 5 or aga in st one IA E A m icrofiche serv ice coupon.

PLEASE BE AWARE THAT ALL OF THE MISSING PAGES IN THIS DOCUMENT

WERE ORIGINALLY BLANK

This Study Group Meeting on research reactor utilization was the

first such meeting held "by the IAEA in Indonesia. The meeting was convened

in Bandung from 2 - 6 August 1971* A total of 36 scientists and engineers

from 10 countries, including three lecturers from Australia, Prance and the

U.S.A. participated.

Of special interest is the fact that among the subjects discussed

at this meeting the topic of engineering studies was included in an Agency

study group meeting on research reactor utilization for the first time.

These proceedings constitute an informal record of the meeting and

of the main points "brought out during the discussions. Pinal preparation

of these proceedings has "been done by the Agency’s staff. While the papers

have not "been submitted to the authors for a final check, in the interest

of expediting early publication, the papers have "been reviewed carefully and

edited accordingly.

It is expected that this record of the meeting will provide useful

information regarding possibilities and new approaches in carrying out

engineering and physics studies in research reactors typical of the smaller

research centre.

The Agency wishes to express its appreciation to the Government of

Indonesia for hosting the meetings to Prof.G.A. Siwabessy, Mr. B. Sudarsono,

Mr. S. Soepadi and their collaborators from the Indonesian Atomic Energy

Commission for their many efforts in the organizing of the meeting; to the

session chairmen Mr. B. Sudarsono, Dr. R. Ramaflna and Dr. L. Ibe; and,

finally, to all the Study Group participants for helping to ensure a productive

and useful meeting.

Foreword.

Status Reports:

Research and. Development in the Indian Nuclear Energy Power Programme, R. Ramanna, India

Status Report on the Current Activities and Future Plans of the Office of Atomic Energy for Peace, Bangkok, 1971» R* Pumlek, Thailand

A Report on the Status and. Functions of the TSING-HUA Open Pool Reactor. Chen-Hwa Cheng and. Chio-Min Yang, Republic of China

Status Report of the Bandung Reactor Centre.S. Soepadi, I. Subki, and. A. J. Surjadi,Indonesia

Utilization of Pakistan’s Research Reactor (PARR). N. M. Butt, Pakistan

Status Report on PRR-1. L.D. Ibe, Philippines

Research Reactor Utilization in Engineering:

Status Report of the Engineering Programmes and Proposed Use of the Korean TRIGA Research Reactors in Support of Power Reactor Fuel Development.B. Whie Lee, Korea

Certain Engineering Problems in the Thai Research Reactor. R. Pumlek, Thailand

The Boron-Stainless Steel Shim Safety Rodsand their Worths in TRR-I Core as Compared to theB^C-Filled Rods. S. P. Kasemsanta, Thailand

Engineering et Utilisation des Reacteurs de Recherche au Centre d fEtudes Nucléaires de Grenoble, (original version). F. Merchie, France

Engineering and Use of Research Reactors at the Grenoble Nuclear Research Centre (English transla­tion). F. Merchie, Prance

Utilization of a Research Reactor as Preparation for the Introduction of Nuclear Power. A.C. Wood, Australia

Engineering Programmes Involving Research Reactors. L. J. Koch, USA

Research Reactor Utilization Engineering Work in Support of a Nuclear Power Programme.S. K. Mehta and. S. R. Sastry, India

Power Upgrading of the TRIGA Mark II Reactor from 25O kW to 1000 kW. S. Soepadi, I. Subki and K. Linggoatmadjo, Indonesia.

Operating Experience with PRR-1. L. D. Ibe, Philippines

Six Years Operating Experience with the TRIGA Mark II Reactor at the Bandung Reactor Centre.I. Subki and K. Linggoatmadjo, Indonesia

Vietnam TRIGA Mark II Reactor Maintenance:Troubles with the Rotary Specimen Rack Assembly. Ton-That-Con and Ngo-Dinh-Long, South Vietnam

Use of Reactor Neutron Beams for Research:

Current Studies Utilizing the Neutron Crystal Spectrometer. S. Chatraphom and T. Nimwanadon, Thailand

Status Report on Experiments Utilizing Reactor Neutron Beams at AERI. H. J. Kim, Korea

Plan for the Construction of Slow Neutron Spectrometers in the AERI (Design of an Inverted Filter Spectrometer). H. J. Kim, H. K. Kim and B. G. Yoon, Korea

Neutron Beam Experiments at Trombay.B. A. Dasannacharya, India

Neutron Crystal Spectrometers at the Bandung Reactor Centre. K. Linggoatmadjo and Z. Amilius, Indonesia

Design and Possible Utilization of a Neutron Guide Tube Bi-Filter on a Beam Hole Experiment at the 1 MW TRIGA Mark II Reactor. S. Jatiman, A. J. Surjadi and S. Supadi, Indonesia

Neutron Spectrometry Research at PARC. M.G. Natera and Q. 0. Navarro, Philippines

A Report on the Beam Ports Utilization of the TSING HUA Open Pool Research Reactor. Chen-Hwa Cheng and Chio-Min Yang, Republic of China

Conclusions and Recommendations

249

259

271

281

289

295

301

323

331

3 53

363

373

377

List of Participants 383

Research and Development in the IndianUnclear Energy Power Programme

by R. Ramanna

Bhabha Atomic Research Centre

Trombay, Bombay-85

ABSTRACT

The status of the entire Indian nuclear research and develop­

ment programme ie outlined. The present status and the 10 year

profile of the nuclear power programme is described. Existing

and planned facilities, research and development activities and

manpower requirements are discussed. Research reactors at

Trorabay, nuclear power reactors in India; and the characteristics

of a variable energy cyclotron (under construction), a pulsed

fast reactor (under design) and a fast breeder test reactor are

listed.

In describing the status of reactor utilization in India,

I would like to outline the status of our entire nuclear power

programme, so that the work on the utilization of research reac­

tors is seen in the proper context of the entire programme of

nuclear power we have set ourselves for the next 10 years. A

more detailed status situation concerning the utilization of

research reactors in engineering and use of neutron beams in

physics will be given by my colleagues Mr. S. K. Mehta and Dr.

B. A. Dasannacharya.

I will first describe very briefly the present status of

our programme and'the 10 year profile which describes our nuclear power programme up to the year 198O. I will then indicate how

the Bhabha Atomic Research Centre has played a vital role in its

development and point out the R & D effort required for the ful­

fillment of the 10 year profile.

As you may know from previous reports, the Bhabha Atomic

Research Centre is the national centre for all matters concerning

research and development of nuclear energy. Another centre near

Madras called the Reactor Research Centre is now being set up

and this centre will be mainly devoted to research and develop­

ment in fast reactors in all its aspects.

The Bhabha Atomic Research Centre is now more than 15 years

old and has strong groups in the basic sciences such as Physics,

Biology and Chemistry, but its immediate value to the reactor

programme is the fact that it has divisions working in problems

of Fuel Development^Reprocessing, Heavy Water production, Health

Physics, Waste Treatment and Technical Physics (High Vacuum

Technology, Development of special materials, etc.)- I do not

make any mention of our production activities like Isotopes,

Electronics, etc., as they are not directly connected with the

reactor research and development programme as such.

Existing Facilities and the 10 Year Profiles

At the moment there are three research reactors operating

at Trombay. Besides these, a fast reactor facility is in an

advanced stage of construction and an isotope producing reactor

is under design. The design details and the utilization of these

five reactors are summarized in Appendix I. Some details of the

six nuclear power reactors, two in operation, the other four

under construction are given in Appendix II. Other special

radiation facilities being mainly created for university research

are given in Appendix III. These include a Variable Energy

Cyclotron (60 MeV protons) and a Pulsed Fast Reactor (both

entirely built in India). The first project of our fast reactor

programme will be a Fast Breeder Test Reactor, work on which has

already started at Kalpakkam, near Madras. Some details of this

reactor are given in Appendix IV.

It is seen that by 1975 India will have about 1200 Mwe of

nuclear power out of the country's total of about 23,000 Mwe.

While the Tarapur station was a turnkey contract awarded to

General Electric, USA, and the Indian participation in its construc­

tion was mainly in the form of sub-contracts, it has been estimated

that nearly 40% of the components made for the Rajasthan station

will be of Indian origin and the Madras station will have about

80$ of its components of indigenous origin. A programme of this

magnitude calls for a special effort of R & D both from our

research institutions like Bhabha Atomic Research Centre and

Indian industries.

Since the plan essentially envisages the setting up of

natural U - DgO reactors, much of the research and development

revolves around, fuel development using zirconium alloy cladding,

the economic production of D2®’ coolant technology, the develop­

ment of reactor control systems, water chemistry, waste manage­

ment and safety. The recovery and utilization of Pu produced in

the fuel calls for another programme of research and development,

especially if it is related to the development of fuel for fast

breeder type reactors.

The ten year programme of nuclear power is summarized in

Appendix V and the cost estimates are given in Appendix VI.

It is seen from Appendix V, that a self sufficient nuclear

power programme with a target of 2J00 Mwe involves the construc­

tion of several additional facilities such as opening of new

mines, heavy water plants, a nuclear fuel complex, etc.

Research and Development

I now mention some research and development problems with

reference to fuel development, reprocessing and waste treatment.

Other engineering problems are referred to in the paper by

Mr. Mehta.

Research and development in reactor fuels covers a wide

range of activities since it involves fundamental studies on the

materials of interest to the entire nuclear energy programme.

The studies carried out at Bhabha Atomic Research Centre include

the beneficiation of low grade uranium ores, development of

ceramic nuclear fuels, production technology of reactor grade

zirconium and zircaloys, development of suitable zirconium—base

alloys with desired mechanical properties and studies in corro­

sion behaviour of alloys of interest. An important part of the

work on fuel development is the fabrication of plutonium oxide

fuel elements for the fast reactor facility, pulsed fast reactor

and the fast reactor programme in general and the fabrication of

Plutonium - Aluminium alloy fuel elements for Swimming pool reactors.

From the experience gained from the beneficiation of ores

and extraction of materials like zirconium, vanadium, niobium,

etc., a plant has been set up at Hyderabad for the large scale

production of zirconium. Fundamental studies on the behaviour

of metals and their alloys have been carried out to evaluate

their suitability in reactor environment. The development of

zircaloy and. the new zirconium based alloys is an important part

of the programme.

The reprocessing programme commenced, with the setting up

of the demonstration reprocessing plant at Trombay for recovery

of plutonium from the irradiated fuel from CIRUS a 40 MW heavy

water moderated research reactor. This plant has been in

operation since 1965» The setting up of this plant was pre­

ceded by a very limited amount of studies on the partitioning

of uranium and plutonium carried out on micrograms scale. The

operation of the plant, however, indicated the need for develop­

ment in process and engineering aspects pertaining to reprocessing.

Thus the development work in reprocessing followed the setting

up of a large scale facility as contrasted with the usual pattern

of preceding it. With the setting up of the power stations at

Tarapur, Ranapratapsagar and Kalpakkam the necessity for inten­

sive development work in reprocessing of thermal reactor fuels

was felt. Apart from collection of equilibrium data and stage-

wise separation information pertaining to the co-extraction of

uranium and plutonium from fuel solutions obtained from boiling

water reactor fuel and candu type fuel, the work so far carried

out and in progress includes investigations on the preparation

and use of uranous nitrate for the partition of plutonium from

uranium, the use of TLA for extraction of the plutonium, the

recovery of neptunium from irradiated fuel solutions and the

separation of uranium-233 from irradiated thorium and thorium

oxide.

The future programme in research and development in the

field of reprocessing will be mainly oriented towards reprocessing

of fast reactor fuel and the long term programme in the reprocessing

of fast reactor fuels include the studies on non-aqueous methods

like salt metal reduction transfer, fluoride volatility and

electrolytic reduction.

Since we are interested in the Molten Salt Breeder Reactor

concept, wherein reprocessing of fuel on line is a basic design

feature, studies on these aspects particularly in view of the

fact that our Molten Salt Breeder Reactor fuel cycle has to be

started with Pu, is in progress.

Por the purpose of investigations in the advanced, areas

of reprocessing especially pertaining to fast breeder fuel

reprocessing a development laboratory is being constructed at

the Reactor Research Centre.

Research and Development work in the field of radioactive

waste management was initiated even in the early stages of the

Centre's development. Initial studies were in the field of

ion exchange and chemical treatment for decontamination of low

active effluents. Several Indian clay minerals were studied

for their use in separation of cesium from the active effluent

streams by ion exchange - these studies included a complete

analysis of their physical and chemical properties, their cation

exchange capacities, and their mineralogical characteristics.

Removal of strontium and other hazardous isotopes from the

active streams was studied through laboratory investigations and

chemical precipitation techniques including pilot plant models

under various process conditions. As a result of these studies

a 50*000 gallons/day plant has been set up, using a combined chemical précipitâtion-cum-ion exchange process. Research and

Development activities were also initiated during this period

in the fields of gas cleaning, solid waste management and decon­

tamination of materials.

Subsequent phases of research have been directed towards

insolubilization of intermediate and high active wastes in

stable, solid media for ultimate disposal. A method has been

developed to incorporate intermediate level liquid wastes in

bituminous or high density polyethylene matrices, and on the

basis of laboratory studies, that included product evaluation

for leaching and radiation stability, a plant is proposed to be

put up at Tarapur to handle active effluents from the Power

Reactor Fuel Reprocessing Plant. Towards solidification of very

highly radioactive effluents, various glass compositions have

been developed for a wide range of waste compositions expected

at the power reactor sites and their properties under possible

conditions of operation and storage have been studied with

respect to environmental surveillance aspects. Research is

also in progress on recovery of fission products like cesium

and. strontium from the high active waste streams for possible

use as heat and power sources.

Manpower

The manpower growth of the Bhabha Atomic Research Centre

organization is shown in figure 1. It is seen that it generally

follows an S curve as is expected of such organizations leading

to a growth rate given by

§ ■=* " C o - " )

Nearly 50% of the graduate staff of the organization has been

provided by the Bhabha Atomic Research Centre Training School

started in 1957» In the case of power projects or production

plants the growth rate follows the shape given in figure 2. If

one knows the staff requirements at the initiation of the project,

its peak period of construction and its final maintenance staff

level, it is possible to estimate the manpower required in various

disciplines for such projects. A computer analysis making use

of available data on the start time of projects, transfers,

promotions and resignations, etc., to obtain the size of our

manpower requirements in the next few years is given in figure 3*We believe our universities, the new institutes of technology,

and our training school can provide the necessary trained personnel.

Appendix I RESEARCH R EACTORS AT TROMBAY

NAME OF REACTOR

D A T E OF CRITICALITY

T Y P E(F U E L ,MOD. & COOLANT)

POWER( t h e r m a l )

PEAK CORE FLUX

FACILITIESAVAILABLE

U S E S

A P S A R A AUGUST 1 9 5 6

E N R IC H E D U (¿0 7 . ) , L IG H T W A T E R M O DERATED iC O O L E D S W IM M IN G POOL T Y P E

1 MW ~10,3n/cm2/

sec.

7 NEUTRON BEAMS FAST NEUTRON IRR -ADiATiON FACIUTY ( lO ^ fn /c m ^ /s e c ) , THERMAL COLUMN, S H IE L D IN G CORNER

IS O T O P E PR O D U C T IO N SOLID STATES FISSION PH YSICS R E S E A R C H W IT H NEUTRON BEAMS

Cl R U S JULY 1 9 6 0 NAT.(J-D20 MODERATED

H20 c o o l e d40MW 6x10t3n/cm2/

sec

7 LOOP POSITIONS IN CORE25 NEUTRON SEAMS THERMAL COLUMN R A P ID IRRADIATION FACILITY

ISOTOPE PRODUCTION S O L ID STATE &NUCLEAR PH YSICS RESEARCH W IT H N EUTRO N BEAM S. A LSO N U C LEA R CHEMISTRY Z ENGINEERING LO O P E X P E R IM E N T S

Z E R L IN A JA N . 1 9 6 1 N A T. U -0 20 MODERATED,

(N O C O O LA N T )

ZERO POWER (¿100 watts)

Í108 n/cm2/ sec

U S E D FOR STUDYING N A T. U - D 2 0 L ATTiCELS

O F IN T E R E S T TO T H E IN D IA N A T O M IC

ENERGY P R O G R A M

FA S T CRITICAL

f a c i l i t y

OCT. 1.971(e x p e c t e d )

P u -0 2 .FU£tLEp-

ZE R O E N ER G Y FA S T R E A C T O R ( 3 L ITR E COR E )

ZERO POWER (^ 1 0 watts)

Sl09fn/cm2/sec

ZERO ENERGY M O C K UP OF

P R O P O S E D P U L S E D F A S T REACTOR

IS O TO P EPRODUCTIONr e a c t o r

i r r a d i a t i o n

•j w e i i o t . , -

L A T E 19 74 (S C H E D U L E D )

NAT.U-D20 MODERATED

H20 o r D20C O O L A N T

100 MW —B E IN G S E T UP TO A U G M E N T G R O W IN G R A D IO IS O T O P E P R O D U C T IO N AND IRRADIATION PROGRAM

Appendix I I N U C L E A R POWER R E A C T O R S IN INDIA

NAME LOCATION D A T E O F CRITICALl TY

POWER( E L E C T R I C )

T Y P EF U E L MODERATOR C O O L A N T

TAPP I

TAPP n

TARAPUR (NEAR BOMBAY MAHARASTRA)

APRIL ,1969

AUGUST,1969

200 MW(e)

200 MWIe)

ENRICHED

U 0 2

it

BOILING LIGHT WATER

11 1 «

RAPP I

RAPP II

RANAPRATAPSAGA (NEAR KOTA R AJASTHAN)

R

LATE 1971

LATE 19 7 3

200 MW (e)

200 MW (e)

Nat. U O 2

11

d 2o

s *

PRESSURISEDd 2o

J*

MAPP I

M APPII

K A L P A K K A M (n e a r MADRAS

TAMIL NADU)LATE 1974

LATE 1975

235 MWie)

235 MW( e)

Nat. U02

Ü

0 20

JJ

PRESSURISEDD2O

J4

Appendix III-A

VARIABLE ENERGY CYCLOTRON (UNDER CONSTRUCTION)

Location

Proton Energy

Beam Current w "

Magnet Pole Diameter

Magnet Pole Gap

Total Weight of Main Magnet

Spiral Sectors (Pole tips)

Magnetic Field (azimuthally varying)

RF System Frequency Range .

Energy gain per turn

Oscillator Power

Ion source filament current

Expected Date of Commissioning

Total Project Cost :

Uses :

Calcutta

6 - 6 0 MeV

Internal s 1 ma External : 100/u A

224 cm (88")

min = 19 cm, max = 30 cm ✓— > 27O tonnes

3 nos. 55° max. angle each

I7.I KG (average)

5.5 MHz I6.5 MHz 140 KeV (max value)

400 KW (max)

5OO A (max)

1973

^ R s . 6 Crores

NUclear Reactions Radiation Damage Studies Proton Rich Isotope Production Radiation Biology

Appendix III-B

PULSED FAST REACTOR (UNDER DESIGN)

Location

Type

Fuel

Mode of Pulsing

Coolant

Average Power

Peak Power

Pulse Width

Repetition Rate

Yield per Burst

Expected Date of Criticality

Cost Estimate

Reactor Research Centre, Kalpakkam, Tamil Nadu

Repetatively Pulsed Fast Reactor

Plutonium Oxide

Reflector rotation

Forced air

30 KW

14 MW

*■* 50 sec

50 pulses/sec 13■— '2 x 10

1974

Rs. 2 Crores

neutrons

FAST BREEDER TEST REACTOR

Location

Type

Fuel & Enrichment

CoolantThermal Power Electrical Output Volume of Core Critical Mass

Blanket (radial and axial) Reflector (radial) Breeding Ratio

Reactor Vessel Control Rods

Expected date of criticality Estimated Cost

: Reactor Research Centre, Kalpakkam,Tamil Nadu

: Liquid Metal Cooled Fast Breeder Reactor (LMFBR)

: UO2 -PUO2 (weight percent of PuO£ is 30%) u235^u233 enrichment in uranium 5. 6% Pu239+Pu241 enrichment in plutonium /*-'74% Liquid sodium 42. 5 MW(t)

13 MW(e)55 litres (62 fissile sub assemblies)

40 Kg of fissile plutonium and 6 Kg of fissile uranium isotopesTh02 (414 sub assemblies)Ni (138 sub assemblies)

internalexternal

= .0 3 = .47

total = . 50 (breeds mainly U233)

236 cm dia x650 cm ht, (stainless steel)6 nos, made of enriched boron carbide

1975R5. 35 crores

The reactor is designed to serve more as a materials and engineering test reactor and for obtaining experience with sodium coolant rather than for breeding more fuel. However a small quantity of U^33 ^11 be produced as shown by the breeding ratio figure above.

LEGENDPREPARATION

CONSTRUCTION

COMMISSIONING

- OPERATION

COST ESTIMATES OF THE ATOMIC ENERGY PROGRAMME

Funds Requiredo. no. i»m -

1970-80 1970-76 1976-80

(Figures in Rs. erares)1. 2700 MWe

a) 1000 MWe constructed or under construction 130.00 101.00 29.00b) 1700 MWe new

3 x 235 MWe 230.00 44.00 186.002 x 500 MWe 275.00 5.00 270.00

2. Design of 500 MWe advanced thermal reactors .. 5.00* 5.00 —

3. Fast Breeder Test Reactor 8 Reactor Research Centre

a) Fast Breeder Test Reactorb) Sodium Coolant Technology > 50.00* 29.00 21.00c) Thorium Bred U 233 fuel Jd) Reprocessing .. .. .. .. 5 .00 3.0 0 2.0 0

Heavy Water 400 T/year including 167 under construction and 233 additional .. 95.00 75.00 20.00

6. 500 MWo Fast Breeder Reactor .. . . .. 125.00 — 125.00

6. Development of gas centrifuge technology and special materials (carbon filament) . . . . 110.00* 10.00 100.00

7. Development of Narwapahar Uranium Mines .. 18.00 4 .00 14.00

8. Nuclear Fuel Complex .. . . . . .. 13.00 13.00 —

9. Fuel Reprocessing Plants for Plutonium .. .. 23.00 9 .0 0 14.00

10. Bhabha Atomie Research Centre .. .. .. 165.00 65.00 100.00

11. Isotope Applications .. .. . . .. 6.00* 2 .00 4 .0 0

Total .. .. 1250 .DO 365.00 885.00

* Ad hoc estimates.

1.2.

3.

4.

Anticipated Revenue from industrial Projects in a Full Year

Rs. Crores126.00

20.00

Sale of power

Heavy Water

Fuel Production

Plutonium

Total

20.004 .0 0

170.00

STAF

F ST

RENG

TH4500 r*

t4 0 0 0

3500

3000

2500

2000

1500

1000

500

-TECHNICAL•SCIENTIFIC•ADMINISTRATIVE■MAINTENANCE & AUXILIARY

y

y

t

yy

1959 '60 '61 '62 '63 *64 *65

YEAR*66 '67 '68 '69 '70 '71 '72

Figure 1

NUMBER OF SCIENTIFIC OFFICIERS frOR ALL THE DISCIPLINES ÀT THE END OF EACH YEAR FOR POWER PROJECT MAPP-I1 (2 35 Mffl

cn

YEAR

Figure 2

NUM

BER

OF OF

FICE

RS

( » SC

í )

NUMBER OF OFFICERS AT THE END OF EACH YEAR FOR THE PROJECTS (RAPP I* II,MAPP M I, HWPK 100, HWPB 70, NFC, RRC )

DISCIPLINE WISE

BOO

YEAR

Figure 3

STATUS REPORT ON THE CURRENT ACTIVITIES AND FUTURE PLANS OF THE OFFICE OF ATOMIC ^¡NERGY FOR PEACE, BANGKOK, 1971

R» PumlekOffice of the Atomic Energy for Peace

Bangkok, Thailand

a b s t r a c t

This report provides information concerning the current scientific

activities and future plans utilizing the Thai Research Reactor (TRR-l).

Presently radioisotope production seems to be the main utilization of the reactor.

I. Reactor Operation

The Tahi Research Reactor is routinely operated 8 hours per day, 5

days per week. However, continuous (24 hrs) runs for 3 or 4 consecutive days

per week had been performed when requested by the users. Our long-term planning

is being contemplated with continuous operation.

As the requirements for sample irradiations are increasing both in

number and in the level of activity involved, it is felt that the installation

of more irradiation facilities with good accessibility are needed in the

reactor core.

II. The Research Reactor as a graining Facility

Thailand is planning to acquire a nuclear power plant in the near

future. Realizing that the technical personnel in the field of nuclear

technology would be inadequate, the Office of Atomic Energy for Peace (OAEP)

together with the Electricity Generating Authority of Thailand (EGAT) and

Chulalongkorn University set up jointly a basic nuclear training programme

to produce personnel with the preliminary technical training for the super­

vision and operation of the nuclear power plant. A one year nuclear training

course began in 1970 and the training programme included main subjects such as nuclear physics, reactor theory, reactor instrumentation and control, nuclear

electronics and instrumentation, reactor materials, thermal aspects of nuclear

power plants, health physics and reactor shielding. Experiments on radiation

measurement and reactor physics are also included in the course. The experi­

ments on radiation measurement are designed to familiarize the trainees

with radiation counting techniques using various kinds of radiation detectors

and nuclear electronic equipment. The reactor physics laboratories are aimed

at providing students with background on reactor physics measurements.

17

Several topics on reactor laboratories such as neutron flux measurements,

void coefficient measurement, and control rod calibration have been included.

III. Chemi str.y

The activities of the Chemistry Division in utilizing the TRR-I

for neutron activation analysis (NAA) may be divided into two general areas:

research and service. Chemists at the OAEP have been working in closed

cooperation with those in other Government laboratories to meet requirements

in their scientific investigations.

The current work on NAA is briefly stated below:

Rice-Soi1-Plant Study

Na. Al, Mn, Cu, and Zn are determined from agricultural samples.

It is expected that Ga, As, Co and Mo will be included in the analysis of

future routine samples.

Inorganic Contents in Human Stones and other Biological Samples

Ca, Mg, Na and P were determined in the human stones in the milligram

per gram range. The amount of Hg and Se were also found in the human stones

in the microgram range. The analysis of phosphate in urine of calculi-suspected

patients is being continued.

Archeological Samples

We had been requested to perform NAA on small samples of archeological

value, particularly pieces of 0.2 to 0.5 g from Buddha images of different

periods.

Human Hair

Occasionally, the Chemistry Division was called upon to perform hair

matching for the Police Department. This is being systematically studied as

it requires more time to gain experience.

FUTURE PROGRAMME

The Chemistry Division plans to concentrate and strengthen work on the

following areas:

1. Producing more qualified scientists and technicians to work on NAA.

2. Implementation of work in the field of forensic activation analysis.

3. Application of nuclear reactions based on secondarily produced

particles and using the delayed neutrons counting technique for mineralogy and

geochemistry.

4. Development of fast radiochemical separation.

5« Cooperation with other laboratories on projects of analyticalquality control services.

IV. Radioisotopes Production

The utilization of the TRR for radioisotope production began in I962

when the reactor first went critical. Presently radioisotope production seems

to be the main utilization of the reactor at OAEP and more than half of the

amount of the radioisotopes needed in the country are produced locally.

The radioisotopes routinely produced include Br-82 in an aqueous NH^Br

sterilized solution, Au-198 in form of gold grains. 1-131 in the form of Nal in

a dilute NaOH or ^ 2820^ solution is the most requested radioisotope and is

delivered in gelatine capsuls. P-32 in the form of Na^PO^ in an isotonic

phosphate buffer solution. K-42 in the form of KCL in an isotonic solution and

Na-24 in an isotonic solution of NaCL are also common radioisotopes requested

from other laboratories in the country. In 1970, the total capacity of radio­

isotope production was 21.313 curies. The amount of radioisotopes produced is

limited by the number of irradiation facilities and the operating time of the

reactor. By mid 1972, a new hot cell will be installed for the sole production

of 1-131 in order to meet the increasing demand of this particular isotope in

the country.

V. Future Plans

Reactor Instrumentation. Since semiconductor components are gaining

ground in electronic circuit designs, the OAEP plans to transistorize the

reactor control system in the near future. A programmable control system

utilizing digital equipments is also of interest.

Industrial Application. Attempts have also been made to utilize the

reactor in the study of neutron radiography to complement the activity of the

X and Y ray radiographic service which is already available at the OAEP.

Neutron radiography will be helpful for the quality control of dry cell

production in the country.

Counting Techniques. The OAEP is planning to set up a computer-based

pulse height analyzer using solid state detectors. A fast pneumatic irradiation

facility will be installed thus providing an analytical tool with higher

sensitivity and rapidity in neutron activation analysis.

A REPORT ON THE STATUS AND FUNCTIONS OP THE TSING-HUA OPEN-POOL REACTOR

BY CHEN-HWA CHENG and CHIO-MIN YANG

NATIONAL TSING HUA UNIVERSITY, HSINCHU, TAIWAN, REPUBLIC OP CHINA

ABSTRACT

This is a report on the various functions and applications of

the Tsing Hua Open-pool Reactor which has "been in use in Taiwan since

I96I. The report is divided into three sections discussing the three

main functions of the reactor:

(1) education and training purposes,

(2) research and development,

(3) practical applications, especially radioisotope

production and irradiation services.

Details are given under each of these headings and a list of

thesis work "based on use of the reactor is given in an appendix.

At the National Tsing Hua University in Taiwan, for the past ten

years we have been utilizing the nuclear reactor bothfor educational and

practical purposes. The Tsing Hua Open-pool Reactor (THOR) reached its

initial criticality in April 1961. Its first core used 20$ enriched uranium,

and the present core enrichment is 90$» Since its initial operation THOR

has been used for training and education, research and development, radio­

isotope production and irradiation services. A brief account of these three

functions is given below:

(i) Education and Training:

The Institute of Nuclear Science at. the National Tsing Hua University

was established in 1956* It began by offering M.S. courses in Nuclear Physics,

Nuclear Chemistry and Nuclear Engineering. THOR has been used extensively

after its completion for their laboratory work and research for theses in

connection with these courses. As the faculty grew larger, the burden of teaching

courses was transfered to newly established branches of the Tsing Hua University

and the Institute was able to concentrate on more special programs. The new

branches of the university were the Institute of Physics, the Institute of

Chemistry and the Institute of Nuclear Engineering. They were established

in 1966, 1968 and 1970 respectively. In addition, the undergraduate Department

of Nuclear Engineering was established in 1964» Reactor Laboratory is a

required course both for senior students majoring in nuclear engineering and

for graduate students who did not specialize in nuclear engineering as

undergraduates. As prerequisites for this course in Reactor Laboratory,

students must pass courses in principles of Nuclear Engineering and Reactor

Physics. The course itself consists of the following;

1) Reactor operation and control,

2) Reactor Engineering and Technology,

3) Reactor Physics,

4) Neutron Moderation and Diffusion, and

5) Neutron Physics.

There are twenty experiments included within these categories,

however only twelve of them are offered to all students. The rest are

only carried out as special projects (for details see Appendix A).

The Institute of Nuclear Science continued to offer training courses

of special kinds to other branches of the university responsible for the formal

academic program. These were in fields not covered by the academic courses

of the university and were designed to meet various practical needs. To give

you an idea about what sort of courses there were I have listed some of them.

1.) Health Physics Training Course, 2.) Radio-isotope Basic

Technique Course, 3») Radiation Instrument Maintenance Training Course,

4«) College Students Summer Seminar on Atomic Energy and 5«) Nuclear Power

Technology Training Course. All these training courses need a reactor for

the experiments. The Nuclear Power Technology Training course is especially

worth mentioning. Its contents includes Elementary Reactor Physics, Energy

Removal in Reactor, Reactor Kinetics, Reactor Control and Instrumentation,

Reactor fuel, Reactor Safety and Reactor Laboratory besides Nuclear science

fundamentals, applied mathematics and electronics, (for details see appendix B).

The duration of the course is six months, however, each session varies according

to its special requirement.

Number of classes Number of Studentsor Sessions______ or Trainees

Nuclear Engineering Undergraduate Students

Nuclear Engineering Graduate Students

Radiation Instrument Basic Technique Training Course

Radiation Instrument Maintenance Training Course

Nuclear Power Technology­Training Course

College Students Summer Seminar on Atomic Energy

Health Physics Training Course

II. Research and Development

Besides using the reactor in the ordinary course work for the

training of our students, we have also applied it to experiments and programs

aimed at research and development. Prom the beginning we had two aims in mind

for the reactor. The first was the general upgrading of our science education.

This included the use of the reactor for the training of students in the ways

I have just discussed and also included use for research and development which

will be the next topic.

Following are some of the activities which have been or are carried

out at THOR:

1.) Reactor Operation and Control16

a. Feasibility study using the N power meter as a power

control channel.

b. A quick and precise technique to determine control rod

worth.

c. Reactor loading effect.

d. Transfer-function study.

e. Optimal control for a nuclear reactor in a distributed

parameter model.

4

11

152

174

14

3

5

5

2

144

67

139

260

58

2.) Fuel Management

a. Neutron flux measurement and fuel burn up calculation.

b. Spent fuel b u m —up study.

(i) by measuring the fission neutrons in the thermal column

(ii) by measuring delayed neutrons

(iii) by measuring the change of reactivity

(iv) by mass spectrometer

_ c. Spent fuel handling technique and transportation.

3.) In Pile Dosimetry

a. Neutron energy spectrum.study using

(i) Activation detectors/ \ 6( n ) Li semi-conductor

(iii) He^ spectrometer

(iv) Fission track detectors

b. Neutron Temperature Measurement Using

(i) Activation detectors .

(ii) Danger coefficient method

(.iii) Mass spectrometer

(iv) Fission track detectors

4.) Reactor Physics

a. Noise analysis

b. Rossi-Alpha Method

. 5») Neutron Moderation, Diffusion and Absorption.

To measure the age, diffusion coefficient, diffusion length

and absorption cross section on some special materials such

as deuterium compound.

6.) Reactor Engineering and Technology

a. Heat Transfer Study

b. Neutron Irradiation Facilities Study

(i) Under water irradiation container design

(ii) Cooling system design for irradiation tube.

M.S. theses work carried out by graduate students since 1961 are

listed in appendix C.

III. Radio-isotope Production and Irradiation Services:

Ever since its establishment one of the main functions of the

Institute of Nuclear Science has been to act as a national laboratory in

the field of nuclear science. In this capacity it has applied THOR to many

practical uses mainly in the field of radioisotope production and irradiation

Services.

Since 1962, one year after the completion of the reactor, the

institute has been engaged in the production and supply of short— }.ived

radio-isotopes to domestic users. At present the following nuclei are

regularly produced and supplied locally: F-l8, Na-24, Mg-28, P-I32, S-35, K-42,

CA-45, Cr-51, Fe-59, Cu-64, Zn-65, AS-76, Br-82, Rb-86, Tc-99, Mo-99, 1-131,

RISA, ROSE BENGAL, Hg-197, Au-198 (colloid), and Hg-203.

Neutron Irradiation is also an important service rendered by THOR.

The Institute had always been busy answering requests from all over the

country to utilize the THOR for irradiation purposes. They include seed

irradiation for mutation purposes, rice boar and fruit fly for eradication

study, solid state physics-material damage by irradiation. Neutron physics -

capture gamma study, Hot atom chemistry - applied to isotope production,

EÖgineering study - engine wear study, electrical wire characteristic change

stude, In-pile dosimetry study.

Activation analysis is also a very important application of THOR.

At present, the following tasks are being carried out through the use of THOR:

1.) Ancient bromee analysis - in cooperation with the National

Palace Museum,

2.) Tunna fish mercury content analysis - in cooperation with

the Food Processing Research Institute,

Mercury content in rice to determine its origin - in cooperation

with .the Agriculture College, National Taiwan University,

Analyzing Cu, Zn, Cr, As contents in human tissue - in

cooperation with Naval Medical Research Unit No. 2

5 .) Surface water and air pollution study - in cooperation with

the Taiwan Institute of Environmental Sanitation

6.) Fissile mineral determination by using the delayed neutron

countering method.

Prom the work described above it is quite clear that experiments

utilizing the reactor are indispensible in our nuclear program. Because of

the multi-function nature of THOR, its operation schedule is quite busy.

The regular scheme is as follows:

Monday - student reactor laboratory,

Tuesday through Wednesday - continuous full power operation for

radioisotope production, irradiation services and

experiments requiring high power level.

Thursday and Friday - Research and theses work, the operation

procedure is determined by experimenter’s requirements.

Saturday - Preventive maintenance.

Sunday - Cooling period for low power reactor laboratory .

Four weeks are allocated for overhaul annually, usually two weeks

each time during both summer and winter vacations. This operation schedule

proved to be satisfactory. However, the present overloaded situation could

be improved if some of the students reactor laboratory and training courses

laboratory were shifted to use a sub-critical assembly. Reactor power

upgrading is also under study. The main idea is to increase the available

flux: and machine time for research work to cope with future needs.

APPENDIX A

REACTOR LABORATORY

A. Reactor operation and control

1.#* Approach to critical experiment of THOR

2. Determination of the exact critical mass

3.#* Calibration of the regulating blade worth

4. * Negative reactivity measurement by the blade drop method

5.#* The Measurement of in-core neutron flux by the induced activity

method and power level calibration

6. Measurement of the transfer function of THOR by means of a pile

oscillator.

B. Reactor engineering and technology

7. * Measurement of the reactor importance function

8. * Measurement of the absorption cross section by the danger

coefficient method.

C. Reactor physics

9. Uranium-235 delayed neutron parameters

10. * Past fission factor.

D. Neutron moderation and diffusion

11. Measurement of neutron and gamma attenuations in water

12.# «Measurement of thermal neutron diffusion length in water

13.#* Age of Pu-Be neutron source

14« Measurement of the neutron temperature

15. * Removal cross section and fast neutron shielding.

E. Neutron physics

16. * Total neutron cross section by the transmission method

17.#* Resonance absorption integral

18. Threshold detectors for fast neutrons

19» Neutron time-of-flight spectrometry

20. Neutron diffraction.

APPENDIX B

NUCLEAR POWER TECHNOLOGY TRAINING COURSE

1. Introductory atomic and nuclear physics ( 30 hrs

2. Introduction of radiation with matter ( 20 hrs

3. Reactor chemistry ( 30 hrs

4. Physics of nuclear detectors ( 15 hrs

5. Basic nuclear electronics ( 20 hrs

6. Introductory reactor physics ( 40 hrs

7. Thermal aspects of reactors ( 30 hrs8. Reactor heat transfer ( 20 hrs

9. Reactor tonetics ( 40 hrs

10. Reactor control and control instrumentation ( 30 hrs

11. Radiation protection and monitoring ( 30 hrs

12. Reactor shielding ( 20 hrs

13. Reactor safety ( 30 hrs

# 6 experiments for nuclear power technology training course.

* 12 experiments offered to all students.

14. Applied mathematics ( 60 hrs. )

15. Digital computer ( 30 hrs. )

16. English (100 hrs. )

17. Electronics laboratory *

18. Radiation measurement laboratory **

19» Chemistry laboratory

20. Health physics laboratory

21. Reactor laboratory

22. Reactor operation ( l8 hrs. )

Electronics laboratory -*

(1) Oscilloscope operation

(2) Transistor and diode characteristics

(3) Pulse amplifiers

(4) Power supplies

(5) Regulators

(6) Pulse shaping

(7) Rate Meters

(8) Scaler

(9) Discriminators

(10) Coincidence and Anticoincidence circuits.

RADIATION MEASUREMENT LABORATORY **

(1) Characteristics of a G-M counter.

(2) Gas-flow proportional counter.

(3) Beta range determination.

(4) Neutron detection by a long counter.

(5) Gamma-ray spectrum and calibration of a multichannel analyzer.

(6) Single channel analyzer gamma-ray spectrometer.

(7) Absorption of gamma-rays.

(8) Characteristics of scintillation counters.

(9) Semiconductor detectors.

(10) +Neutron flux mapping by foil and wire activation.

(11) +Absolute counting by 2 IT proportional, G-M, and scintillation counters.(12) Positron anihilation by 2 scintillation detectors.

(13) +Beta-thickness gauge. •

(14) Delayed neutron measurement.

(15) Time to pulse-heigh converter measurement.

Those with + take two afternoons, the rest take one afternoon for each experiment.

HEALTH PHYSICS EXPERIMENTS ***

(1) Environment monitoring for a reactor facility.

(2) Radiation survey and detection for a reactor facility.

(3) Calibration and application of survey meters.

(4) Reading of Tsing Hua film badge.

(5) Urinalysis.

(6) Gamma-ray attenuation experiment.

(7 ) Fast neutron dose evaluation.

The number of students for each experiment is limited to six.

APPENDIX C

M. S. THESES BY GRADUATE STUDENTS

1961 Class:

1. Determination of Dysprosium and Samarlium in Bare Earth Minerals by Activation Analysis

2. Calorimetric Dosimetry of Gamma Radiation by Thermistors

3. A Study of thermal Aspect of Tsing Hua Nuclear Reactor

4. Measurements of Neutron Spectra by Threshold Detectors

5. Angular Distribution of Fission Fragments from U-238 with Neutrons of Moderate Energy

6. Ellipsoidal Reactor Analysis

1962 Class:

7. Measurements of Neutron Spectra of Tsing Hua 1 MW Open» pool Type Research Reactor

8. Investigation of Neutron Inelastic Scattering

9. Szilard-Chalmers Reactions on Copper Compounds

Yu-Wen Yu Pei-Hsin Yu

Hsing-Chi Yu

Chang-Ping Wang

Chi-Chung Wang Sheh-Chun Chou

Shau-Jin Chang Yuan-Li Wang

Tien-Chen Liu Chung-Ching Liu

Ching Lu-Shiu Hua-Ching Tong

Ko Ton

Yin Moon-Lung Chov-Kin-Lian

10. Absolute Determination of Neutron Source Strength and the Measurement of the Space Distribution of Theraal-Neutron Flux

11. A Study of Tsing Hua Reactor Operation Characteristics

12. A Study of fuel Element Geometry

13» Study of Natural Convection betveen Parallel Plates for Steady Uniform Wall Heat Flux

14. Measurement of Transfer Function of the Tsing Hua Reactor

1963 Class*

15» Experimental Studies of Energy Responses of a Boron- compound Neutron Scintillator

16. Measuring of the Effect of Void on the Thermal Diffusion Length

17. Calculation of the Themal Flux and Importance Function of the Tsing Hua Beactor

18. Determination of the Tsing Hua Beactor Neutron Temperature

19. Investigation of Neutron Inelastic Scattering

1964 Classt

20. Neutron Inelastic Scattering up to 10 Mev by the Ellipsoidal Rotator

21. Fast Neutron Spectrometer

22. The Measurement of the Energy Spectrum of Neutron Beam from the Reactor Beam Port

23» The measurement of Neutron Temperature in Tsing Hua Reactor Core mrd the Study of Its Deviation due to Neutron Leakage

24, A Study of Resonance Escape Probability in Various System

25, A General Investigation of Nuclear Properties of Fuel- moderator Mixture

26, A Study of Natural Convection in Thin Channel with known Heat Flux Input

1965 Class:

27, Themal Neutron Absorption Cross Sections by Modified Two Group Danger Coefficient Method

28, Determination of Tairig Hua Nuclear Reactor Transfer Function and Transient Analysis by Analog Computer

Weng Pao-Shan

Lu Yang-Shen

Chen-Hu-HsiULiu-Yang-Kan

Hsu Kuan-Ling Chen-Ka-Wei

Yang Chio-Min

Che-Wen Mao

Hsing-Shou Cheng

Chian-Yeh Ho

Su-Tien Hsu

Ma Ta-Tao

Ghhi-Chong Wu

Hsin-YÜ Wang

Chin-Kuei Wen

Jium-Kuen Koo

Chei-Chung Ho

Kueng Yeh

Nai-Chen Ho

Chi-Kang Cheng

Yeh-Chin Ko

29« Study of Theimal Neutron Energy Spectrum in the Reactor Core

30. Thé Analysis of Gamma Bays Spectra

31* Flux Monitoring Fuel Element

1966 Classt

31. Gamma Ray Penetration Through & Backscattering from Concrete Slabs

33« Convective Heat Transfer in Parallel Plate Channel with Sinusoidal Heat Flux Distribution

34. Shielding Design for a 300 Mw Thermal Pressurized Water Reactor

35» A Study of y -Ray Dosimetry by Polarographie Method

1967 Class:

36. The Effects of Fast Neutron Irradiation on Transistors

37» Studies of Reactor Transients Using an Electronic Simulator

38. Measurement of Themal Neutron Spectrum by Using a Slow Neutron Chopper

39* The Effects of Gamma Radiation on Transistors

1968 Class:

40. Chemical Behavior of Iron-59 Recoil Atoms

41. Chemical Behavior of Chromium -51 Recoil Atoms

1969 Class:

42. Feasibility Study of Fuel Burn-up Measurement in the THOR Thermal Column

43# Optimal Control for a Nuclear Reactor in Distributed Parameter Model

1970 Class:

44. Effects of Irradiation Damage by Ganaaa-rays in Silicon Surface-barrier Detector

45. The Measurements of THOR Fast Neutron Spectrum at E-l Beamport by Li® Semi-conductor Spectrometer

46. The Feasibility Study of Fast Neutron Conversion

Kuo-Hung Chang

Sy-Ming Shy

Tsu-Chung Wu

K. C. Wu

Shing-Tai Chen

Cheng-Min Tseng

Sung-Tsuen Liu

Ynng-Chau Yen

Lung-Rui Huang

Wei-Hsiang Teng

J. B. Shao

Yih-Hsiung Chen

Ting Gann

Hwei-Yen Yang

Baw-Lin Liu

Shin-Shyong Wu

Chen-Sbyong Yeh

Si-Jzei Yang

47. Beactor Noise Analysis

48. A Study of the Intermediate Neutron Energy Spectrum

49. Measurements of Thermal and Fast Neutron Fluxes in Beactor with Nuclear Track Detector

50. Determination of Fuel Bttm-up by Using Flux Variation Method

51. A Statistical Feature of Nuclear Beactor Transfer Function

52. Fuel Bura-up Measurement by Perturbation Theory

Tzun-Ben Chang

Ming-Huei Lee

Taer-Fu .Huang

Hsien-Mo Lee

Kuo-Ting Tao

Ta-Ming Lai

I97I Class:

53* Neutron Activation Analysis on Some Chinese Antique Pieces

Ha i-Yung Huang

54. The Effects of Gamma-rays on the Function Field Effect Jen-Shien Chung Transistors

55. Bas si- &L Experiment

56* Investigation of Thermal Neutron Energy Spectrum of THOB

Yang-Ho Sun

Hung-Jen Yang

STATUS REPORT OP THE BANDUNG REACTOR CENTRE

by

S. Soepadi, I. Subki, A. J. Surjadi Bandung Reactor Centre

Bandung-Indonesia

Abstract

The Bandung Reactor Centre was commissioned in the beginning of 1965 and it took alraœst two years since then to be actively engaged in its present activity.

The delay in the initiation of the programme is due mainly to the infamiliarity with the potentiality of atomic energy, the lack of competent engineers and scientists, and the limited funds available.

The initial activities were emphasized on training, isotope production and research on related problems. These efforts were then followed by adaptation to research and to a smaller extent research in physics and chemistry.

The current activities cover: continuous operation of thereactor of three days a week to support isotope production and beam-port experiments; health physics activities comprising radiation protection in hot laboratories, waste management, fall­out measurements, intercalibration of film badges with the IAEA and the Bhabha Atomic Research Centre/lndia; production of radio­isotopes for nuclear medicine and nuclear hydrology activities; besides, a hot cell facility is under construction to handle high activity gamma sources; routine neutron activation analysis of mineral ores and deposits; and neutron spectrometry.

Some of the above mentioned research activities have been executed under cooperation with various departments and also with the IAEA.

In the last two years the interest of the Sovernment has been increasing, this is reflected in the increase of the annual budget and the cooperations with governmental bodies. With this favourable condition long-term research projects can then be undertaken.

Introduction

The Bandung Reactor Centre has been commissioned in the be­

ginning of 1965 and it took almost two years since then to be

actively engaged in its present activity.

All those developing countries facing multifacetted problems

and activities in the field of science and technology will un­

doubtedly encounter the same epxerience as we did during the two

years before we started with our activities. The many problems

which arose compelled us to determine a definite choice of activi­

ties, bearing in mind the existing needs and the available faci­

lities.

At the early start of our programme it was felt necessary to

start activities in different fields. This was mainly due to the

infamiliarity with the potential uses of atomic energy. It re­

quires constant efforts and patience to convince the responsible

people about the potentiality of nuclear energy and at the same

time to adopt the basic modern nuclear techniques.

Apart from the facts mentioned and the limited funds available,

the incorporation of competent engineers and scientists to the pro­

gramme also requires a tireless and continuous effort.

Gradually the situation became more favourable and it was

realised that nuclear energy was of great significance for the overall

development of the country. The government took steps to raise the

annual budget which entailed the upgrading programme of our reactor,

and the inflow of trained scientists and engineers from abroad is

making the utilization of nuclear energy in Indonesia rather promising.

After three successive years during which all our activities were

directed toward the practical application of atomic energy, and the

training of scientists and engineers was being accomplished, time has

finally come to start with fundamental research and undertake long-term

projects.

As the first reactor was being installed in this country the main

objectives have been to explore existing problems for initial activities

with emphasis on training, research and isotope production, and studies

on related problems. ,

Training geared for this purpose has produced qualified engineers

and technicians through long term university education or short

courses for professionals wishing to adopt nuclear techniques to their

respective disciplines.

Bearing in mind the general climate that prevails and the need that

has been felt, all efforts are directed to adaptation research, and in

a smaller proportion research in physics and chemistry has also been

carried on, and simultaneously familiarization with nuclear instruments.

Anyhow, these phases have enabled the development of researchers and

made it possible to carry on research at a higher level in various fields.

The isotope production with its initially limited programme is

expanding rapidly, this is attributable to the adaptation research

mentioned earlier; the quality and quantity of isotope needed for the

different studies depend heavily on the capabilities of our upgraded

TRIGA Mark-II reactor being the only reactor existing in this country.

Current Activities

The reactor has been scheduled to run on a weekly programmes

Monday through Wednesday continuous operation, Thursday and Friday

8 hours operation and Saturday is left for maintenance work.

Other activities are described elsewhere.

The Health Physics Division is routinely engaged in basic radiation

protection work, e.g. radiation protection in hot laboratories, radio­

active waste management, calibration of Health Physics instruments,

radioactive fall-out measurements and it also maintains a film badge

service for the users exposed to radiation outside the centre.

Among other scientific activities which have already been carried

on in the past we mention the followings

1 . film badge intercalibration services wiîbïi the IAEA and the

B.A.R.C./lndiaj

2. intercomparisonal glass dosimetry with the IAEA.

Above all the production of radioisotopes for domestic use in

the field of medicine, hydrology, and research has been emphasized.

A hot cell equipped with accessories, capable of handling up to 100 Ci

equivalent Co-60 is under construction, this laboratory enables us to

produce high activity isotopes. Routinely 30 kinds of isotopes and

their labelled compounds have been produced with a net activity of

about 145 curies last year, most of which were used for experiments in

Hydrology and Medicine.

It was hard to accomplish a rapid increase in isotope utilization

without a close cooperation with the organizations or bodies concerned

with their use.

An agreement has been reached with the Ministry of Public Works

and the Ministry of Health to establish a close collaboration on

radio-isotope applications in the fields of Hydrology and Nuclear

Medicine.

In the last three years 10 research projects on discharge measure-

mentff, dam seepage investigation, permeability mapping and sediment

gauging, have been successfully conducted.

A research contract on isotopic methods in tropical soil hydrology

has been granted by the IAEA and work on that subject is being carried out.

Last year on the premises of the centre, while waiting for the

completion of the building in the general hospital, a clinical ward was

officially inaugurated, including a variety of activities such as

uptake studies, liver, brain and renal scanning studies, therapy of

cases of hyperthyroidism and carcinoma. A number of 649 patients

received treatment during that period and various isotopes with a total

activity of 3*7 Ci were used.

The reactor has also been routinely used for irradiation purposes,

e. g. neutron activation analysis; among the targets that have been

irradiated are mineral ores ar deposits. The availability of a Ge-Li

detector encouraged us to start our programme of instrumental neutron

activation analysis.

At the present time a research contract on "Geochemical and Geo-

botanical Prospecting for Gold and Copper by Neutron Activation Analysis"

has been granted by the IAEA. While another research contract on

"Some Aspects of the Utilization of Tritium" has just been completed.

Neutron spectrometry has been initiated in this country since the

participation of our staff members in the I.P.A* projects. The results

of this project are reported elsewhere.

Apart from this, neutron radiography and in-pile dosimetry are

being carried out.

The programme on radiobiology has emphasized mainly agricultural

and allied branches of science using the Standard Triga Irradiation

Facility (STIF). Mutation breeding studies on rice grain and soy beans

are being carried out, and at the same time the possibility of achieving

pest control is being studied.

A variety of subjects in the nuclear sciences have been proposed to

the various faculties of the Bandung Institute of Technology making

available the facilities of the Centre to the students from the

Institute to carry on research leading to study reports or theses.

Some members of the Centre are given the opportunity to teach at

the Institute, and annually 35 students on the average complete their

study reports and/or theses.

Programme and Resources

The Bandung Reactor Centre is staffed with 38 scientists, 32 technicians,

and 61 administrative officers, twice the number at the time of

commissioning.

Though a sound scientific tradition does not prevail yet, in

general it can be said that recruitment of additional personnel pro­

ceeds smoothly without difficulty; concurrently these people may either

choose a far better paid job and yet they remain and contribute with

their full dedication to the centre. Most of the staff members have

* I.P.A. India-Philippines-IAEA regional research cooperation agreement

received their professional training abroad, however a more specialized

training is needed. Young graduates for specialized training are

generally supplied by the universities.

The basic instrumentation which is available at the centre was

supplied by a U.S. Government grant. More sophisticated instruments

designed to be used for advanced studies were required during the last

few years.

Since the nuclear centre is a governmental establishment, the

sources of income are from the government funds and are therefore

dependent on national events and development; and this may some time

hamper further planning for development.

The government has taken more interest and provided better support

for research and development, including atomic energy matters during the

last years.

In 1969, the first neutron diffractometer set up was ready for use

and the first experiment on elastic scattering was carried out.

In the light of the upgrading programme, studies on inelastic

scattering are made feasible using a specially predesigned Beryllium

Detector System, which the IAEA decided to support through a project of

regional cooperation.

The engineering aspects that have been covered to date are:

- neutron radiography, a vertical beamport has been designed and its

construction is underway;

- design of the secondary cooling tower for 1000 kW operation;

- programming for reactor code applications and for research reactor

control ;

- reactor chemistry, accentuated on coolant chemistry and failed

fuel element detection.

A further engineering programme should be selected appropriately,

the selection of which depends upon joint efforts with other national

authorities involved with the nuclear power programme. But the

general interest will cover the following topics: reactor chemistry,

materials study, instrumentation development and thermal hydraulics.

The basic requirements for advancement and development of the

centre are well established.

The results of the studies on the use of isotopes and radiation

were promising, this is reflected in the confidence and the increase

of the annual budget and the contracts with other governmental bodies.

However,a large increase in the annual budget should not be

expected within the next few years. It can be expected that the govern­

ment will finance quick yielding research studies while the financing

of long term research projects should be secured from other sources.

Utilization of Pakistan’s Research Reactor (PARR)

by S. M. Butt

(Neutron Diffraction Group)

Pakistan Institute of Nuclear Science and Technology

Nilore, Rawalpindi, Pakistan

ABSTRACT

The research programme under execution at PARR (Pakistan

Research Reactor) at the Pakistan Institute of Nuclear Science

and Technology, in Nilore, Rawalpindi is described.

The utilization of the 5 MW Swimming Pool Research Reactor

in the field of Solid State Physics, Nuclear Physics, Radio­

isotope production and activation analysis is discussed.

Some recent results of the various research projects

currently under investigation are reported.

Further research work envisaged is briefly mentioned.

Introduction

The Pakistan Atomic Energy Commission started its programme

with a lot of vigour and enthusiasm after Dr. I. N. Usmani, the

present chairman, took over this organization. The object was

the peaceful uses of atomic energy.

Several Atomic Energy Centres were planned and the Pakistan

Institute of Nuclear Science and Technology (PINSTECH), was

started at Nilore, about 15 miles outside Rawalpindi/Islamabad,

the capital.

The country’s first reactor, the 5 MW Swimming Pool research

reactor was planned at this Institute. The reactor became

critical in December 1965 and the full power of 5 MW was

attained in June 1966.

The maximum thermal neutron flux at the core at full powern S g Ô 2

is about 3 x 10 n/cm sec. and more than 10 n/cm sec at the

wall of the reactor on a typical radial beam tube.

Reactor Facilities and Utilization (Fig. l)

Beam Tubes (Fig. 2)

There are six horizontal radial beam tubes with one through

tube passing tangentially at the reactor-core and ending on the

opposite faces of the reactor shield.

Three of the radial tubes are 6" in diameter and the other

three are 8" in diameter.

Vertical Tube

This is a 2.4 inch diameter aluminium tube filled with water

and extends from the reactor grid-plate to the reactor bridge.

The tube can be inserted in any hole of the grid-plate so that

any desired flux could irradiate the sample placed in this tube.

The tube is used for irradiation of small samples with a1 2

flux of the order of 10 n/cm sec.

Thermal Column

A graphite thermal column of 4'x4' cross section and 5'

depth is provided. The thermal column is closed by a M g

concrete door of 5'x5' cross section and 5' deep having four

wheels which move on parallel rails. The door has four holes

of 6" diameter each, which are normally closed with removable

concrete plugs.

The thermal column is used where thermal neutrons are

required for some irradiations. It offers a high cadmium ratio^

about 5OO at the graphite face.8 / 2

The thermal neutron flux at the graphite face is 10 n/cm-

sec. By removing some graphite a thermal flux of as much as10 2

10 n/cm sec can be obtained. It is intended to set up a

single axis/double axis crystal neutron spectrometer at the

thermal column, mainly for the measurement of total neutron

cross sections of materials.

Pneumatic Rabbit System

This system has two stations, one near the Hot Cell and the

other in the Chemistry and Isotope Laboratory at the ground floor

of the rector-hall. The system consists of a net-work of 2”

diameter aluminium tubes connecting the stations to the places13 13 2

near the core, where fluxes of about 1 x 10 and 6.5 xlO n/cm

sec can be obtained.

This facility is used for making radioisotopes and also to

study some short lived isotopes. A polythine tube known as

"Rabbit" of 2" diameter with a sliding contact on the inner of

aluminium tubes, carries the material of which the radioisotope

is to be made and is transmitted to the core with pneumatic

control and moves with a speed of 30 to 40 feet/sec. A maximum

weight of the rabbit can be 16 ounces including the material.

The system is used for making radioisotopes and for delayed

neutron fission experiments.

Hot Cell

It is a 9 by 6 feet heavily shielded room provided with a

facility of slave manipulators. The inside of the room is

visible through a lead-glass. The room is connected through

the transfer-port to the reactor pool so that big irradiations

can be handled. Large amounts of materials can be irradiated

in the reactor-pool and then transported to the Hot-Cell through

the mechanism of transfer-port of 2’x2’ cross section, the 3

feet thick heavy density wall of the hot cell permits safe

handling of about 1000 curies of radioactive samples.

Utilization for Research and Training

The research programme around the reactor at the beam ports

is mainly being carried in Physics. However, in addition some

activation-analysis of some materials after irradiation in the

reactor is being carried out by the Nuclear Chemistry Division.

We shall now discuss the Physics programme in some detail.

Physics

The physics research programme consists mainly of two branches,

namely, the Solid State Physics and Nuclear Physics. There are

four main research groups in existence at present which are

engaged in these fields. The Nuclear Physics groups engaged in

Fission Physics and Neutron Capture Gamma Ray Spectroscopy, started

their programme in 1966 soon after the reactor became critical.

The Solid State Physics groups engaged in neutron diffraction

and scattering from solids and liquids and the radiation damage

studies started somewhat later.

Nuclear Physics

Fission Studies (G.D. Alam, M. A. Shaukat,T. A. Khan, M. Zafarullah Khan)

The group has the following programme of research: '

1. Studies of tertiary fission.

2. Study of X-rays from the fission fragments.

3. Study o f Y - rays from the fission fragments.

The group has recently done an experiment on ’’fission

fragment energy-correlation measurements for thermal and reson­

ance energy neutron induced fission of Pu”. Experimental results

of the double energy measurements using solid-state detectors

are obtained for thermal and resonance energy neutron induced

fission of Pu. A monoenergetic neutron beam of 0.297 e.v was

obtained through reflection of the incident neutron beam from

the (002) planes of the Zinc single crystal. Mass and energy

distributions have been obtained containing 1.6 x 10 events4

for thermal fission and 4 x 10 events for resonance fission.

Preliminary results indicate increase in the symmetric

yield for resonance fission compared to thermal fission.

Introduction

Low energy neutron induced fission cross-sections show

pronounced resonances in the electron volt region. These resonances

may correspond to different "transition states of the compound

nucleus. According to the theoretical ideas of Wheeler based

on the collective model, the compound nucleus undergoing fission

is relatively ’’cold" due to large deformations involved, conse­

quently few well defined rotational and vibrational quantum

states are available for the fission process. On this basis low

energy neutron induced fission could occur mainly through a

well defined quantum state. It is, therefore, of interest to

investigate fission induced by monoenergetic neutrons of resonance

energies and to study the variations of the mass yields from

level to level.

Experimental Procedure and Data Analysis

A schematic diagram of the experimental set-up and elec­

tronics is shown in figure 3- A monoenergetic neutron beam was obtained through the Bragg reflection of the 1” diameter collimated neutron beam from tne (002)—plane of the Zn single

crystal mounted on a simple single axis spectrometer system. The

diffracted neutron beam was ftirther collimated with a lMxl" Soller

collimator. The overall energy resolution was sacrificed to

obtain a higher flux from the resolved neutron beam.

239The Pu target and two heavy ion surface barrier detec­

tors (D^ jD^) facing the target were placed in a small aluminium

chamber having thin front and back aluminium windows to avoid

excessive scattering of the neutrons. The O .297 eV resonance

was identified by varying the Bragg angle of the crystal and

measuring the fission rate as well as monitoring the neutron

flux with a small BF3 detector placed immediately behind the

chamber. The fission rate normalized to the neutron flux is

shown in figure 4* The maximum of the normalized count rate

falls at a neutron energy of 0.297 eV corresponding to the peak

in the fission cross-section resonance. The Bragg angle was

adjusted corresponding to the maximum of the fission rate.2 -JQ 2

The Pu target was 70/^gm/cm deposited on 3 inch thick

nickel foil and had the following isotopic composition:

239Pu, 99.10$; 24°Pu, 0.888$; 241Pu, 0.014$.

Details of the electronic system (figure 3 ) used in this

experiment, are self explanatory. The data was stored in the

4096 channel multi-parameter analyzer in a 64 x 64 mode. Close

watch was kept on the gain of the system. Stability of the

gain of the system was checked every few hours, by taking puiser

measurements and adjusting the gain of the amplifier and Zero

level of the analyzer for small gain shifts. The data punched

on paper tape after the end of each run was finally processed

on an IBM 360/40 computer. In order to eliminate grid fluctua­

tions in the transformed data the H (x^,x^) events of the corre­

lated pulse heights in a given position (x^,X£) of the array were

treated as N independent events and were processed separately,

by adding random numbers distributed between -0.5 and +0.5 to

each pulse height. The pulse heights were converted to mass9

and energy by the mass dependent calibration and mass and momentum

conservation.

Result and Discussion

The interim results are summarized below:

The fragment mass distirubtion and average total kinetic

energy E . as a function of the provisional mass, for resonance

as well as thermal neutron induced fission are shown in figure 5*5

The thermal and the resonance runs respectively contain 1.7 x 104

and 4 x 10 events. A lisir of average total kinetic energies,

light and heavy fragment masses and the distribution widths is

given in table I. For reference, also are included the corres-239 Í

ponding values for thermal fission of Pu, obtained previously.

Within the resolution, the average total kinetic energy, light and

heavy fragment masses are in agreement. However, due to poorer

resolution of 64x64 channels, the mass distribution width is

higher and the peak to valley ratio is lower in the present

experiment.

The mass distributions for the two cases are plotted on a

log scale in figure 6: to show the variations in symmetry.

These results indicate an increase of symmetric neutron induced

fission compared to the thermal fission. The relative variations

are given by

R - <p/vLs / <p/¥Wwhere (P/v) res and (p/v ) Th are respectively the peak-to-valley

ratios for resonance and thermal fission yields. These values

are,

<p/v>ReS. " 45 i 10<P/V W . ' 97 ± 10

The yield at symmetry is approximately twice compared to the

thermal run. Similarly, the results show deviations in the

average total kinetic energy as a function of mass. The (Ej^p g

is effectively larger than in the symmetric parts of the

mass distribution. Although the measurements clearly indicate

the dependence of fission yield on the state of the compound

nucleus at the saddle point, these results are in disagreement

with the previous radiochemical measurements of Regier and his

co-uorkers.

Neutron Capture Gamma Spectroscopy (A.M. Khan, Irahad Mohammad,J. A. Mirza, Anwarul Islan,C. A. Majid).

The research programme of the group consists of:

Nuclear structure studies involving the accurate

determination of energies and intensities of neutron

capture gamma-rays, measurement of angular correlation

and polarization correlation of cascading (n, f) radia­

tion and the measurement of life-times of the low lying

nuclear levels.

The equipment at present available with the group consists

of a 30 cc Ge (Li) detector, two 3"x3" Nal(Tl) detectors, ORTEC

modular electronics and a 4096 channel analyser. The germanium

detector and the associated electronics give a resolution of 3«7

kev at I .32 Mev. A 6”x6" split Nal(Tl), annular detector, now on

order, will be used together with the germanium detector as a

pair spectrometer.

The experiment is being set up on â 6" diameter tangential

tube. The neutron beam brought out through a service of lead7 2

and paraffin wax collimators has a flux of 1.5x10 n/cm . sec, at

the target position at full reactor power. The cadmium ratio

is 10 and the gamma ray dose is 0.5 r/h.

Preliminary measurements made with a 2 cc Ge (Li) detector

on an Fe-target are shown in figure 7- Modifications are being

made in the system to incorporate the 30 cc Ge (Li) detector

and the Nal annulus when it is received.

Solid State Physics

Neutron Diffraction and Scattering Studies (N.M. Butt, Q.H. Khan,M.M. Beg, Javed Aslam, Miss Attika Rabbani,A.A.Z. Ahmed, M. Afzal).

The work is being pursued by the Neutron Diffraction Group

of this institute.

These studies were proposed by this group in early 19^7

with the following research programme:

1. Study of lattice dynamics by the method of inelastic

scatteringcf slow neutrons from solids and liquids.

2. Crystal structure determination by the method of

Heutron Diffraction.

3. Measurement of total neutron cross sections of

materials using monoenergetic neutrons in the energy range

5 x 10 3 e.v. to 1 e.v.

4» The provision of X-ray diffraction facilities which

are necessary for a solid state physics laboratory.

5- The provision of single crystal growing facilities

To implement this programme it was decided to have the

following experimental facilities:

1. Triple-axis Spectrometer.

This instrument has been purchased from Poland under

the Pakistan—Poland barter agreement. Its main features are

given below:

Main Features of the Triple-axis Spectrometer

The mechanical parts consist of the first-axis where a

single-crystal monocromator is placed (figure 8). The crystal

is surrounded by a very heavy monochromatic shield of about 3

feet radial thickness. The crystal rotation table and the mono­

chromator arm are coupled through gears in the ratio 1:2. The

table has an angular range of rotation of 360° while the mono­

chromator arm (or shield) can rotate over a range of -15° to +90°.

The accuracy of the angle setting is 1* and the backlash in gear

coupling is about 2*.

The second-axis system, the sample table and the sample

arm are placed on the monochromator arm.

The sample table has no gear-coupling with the sample arm

and both can be rotated independently. The angular ranges of

the two are 360° and 90° respectively with an accuracy of angle setting of 1'.

The third-axis, analyser table and the analyser arm are

placed on the sample arm and are coupled with a gear ratio ofo

1:2. The angular range of the table is 360 while that of the

analyser arm is 160°.

The angles of all the sixes can be set automatically by a

programmed paper tape.

The spectrometer operate§ in the automatic mode through

a logic system of the control electronics.

The spectrometer can be used for three different fields in

diffraction studies namely study of lattice dynamics in the

triple-axis mode; structure determination in the double-axis

mode and for determination of total neutron cross sections in

the single-axis mode.

Experiments on the Triple-axis Spectrometer

Before starting the installation of the spectrometer an

in-pile collimator (figure 9) was installed in the Beam tube

No. 3» The collimator provides a beam cross section of 2"x2”.

The collimator is provided with a water-shutter (a stainless

steel rectangular tube of 2"x2” cross section and 4 feet length which can be filled or emptied by water with remote system), a

cooling jacket for the Bi-filter (which can be cooled by circu­

lating liquid nitorgen from outside) and a mechanical-shutter

with an eccentric 2l,x2" hole- The mechanical-shutter can be

operated with remote control from the chain-pulley system.

When the water-shutter and the mechanical shutters are

closed, the neutron and gamma radiation background in the

working area of the spectrometer is quite below the safety

limits. After the installation of the spectrometer in May,

1971 the following three experiments have been performed:

1. Diffraction pattern of Mn Pe2 0^

The spectrometer was used in the double-axis mode. A

Zn (OOOl) monochromator of 0.8"x2.5”x7” was used.

The monochromator arm was set for neutrons of 1.22A0

wave-length. The Mh Fe2 O4 powder was filled in an empty

aluminium cylinder of 1 cm diameter and 4 cm high and was

mounted on the sample table. It was adjusted to be completely

covered by the incident neutron beam.

The diffraction pattern obtained is given in figure 10.

The idea of this diffraction pattern was to compare

the positions and the intensities of the peaks already obtained

on a similar spectrometer in Poland. The results agree well

to substantiate a good calibration of our spectrometer.

2. Phonon dispersion-relation for Al(lll)

A single crystal A1 sphere (2" diameter) was used at

the sample table. Neutrons of wave-length 1.22A° (or an energy

of 0.055 e.v) idffracted by a Zn(000l) monochrometer were used

for the inelastic scattering from the sample. The phonon peaks

were obtained along the symmetry direction Al(lll) of the sample

for q-values of 0.06A° \ 0.08a° 0.12A*"* \ and 0.18a ° \

figure 11 (a,b). The phonon dispersion curve thus obtained is

given in figure 12.

This experiment was also done to caliberate the spectro­

meter in the triple-axis mode. The dispersion curve obtained

was in agreement with the one already obtained with a similar

spectrometer installed at the nuclear centre of Swierk, Poland.

Mosaic spread and reflectivity of crystals

For any single crystal grown in a laboratory, the mosaic

spread is not in an accurate control of the experimenter. Only

by experience one may grow crystals in the neighbourhood of the

stipulated mosaic spread.

Therefore it is necessary to know these parameters of the

crystal by experimentation after these have been grown. The

present experiments on the measurement of the mosaic spread and

the reflectivity of Zn, Cu, Al, Pb, and Ge single crystals will

be carried out.

Al(lll) Single Crystal

An Al single crystal plate of dimensions 6"x3"xl" with

(ill) planes parallel to the flat surface was placed at the

sample table.

The incident neutron "beam falls at an angle 0 = 12.34°

on the crystal and is diffracted at an angle 2© in which

direction the detector is placed. The crystal was adjusted

to get maximum reflection. By keeping the detector fixed at

20, the crystal was rotated over its reflection range.

Results

Figure 13(a) gives the geometrical divergence of the inci­

dent beam due to the collimators.

Figure 13(b) gives the diffraction peak of the Al(lll)

reflection obtained for neutrons of wave-length =1A° (energy

O.O82 e.v). Two overlapping peaks are observed which indicate

that this piece of the crystal is not one single crystal but

is divided into two parts each being a good single crystal.

Further investigations to verify this explanation are in progress.

The incident divergence of the neutron beam is 13* while

the full widths at half height of the observed diffraction peaks

are about 18*. A mosaic spread of 5’ can be attributed to the

crystal. However, these are very preliminary measurements and

further detailed investigations of the mosaic spread and the

reflectivity of this crystal are still in progress.

Investigations on neutron shielding materials

(a) Borated Paraffin:

Although borated paraffin is widely used as a neutron shield

the data is not easily available for a systematic study of the

effect of varying percentage of Boron in paraffin. A systematic

study on these lines was made.

A good collimation set up was arranged at the thermal

column as shown in figure 14* Blocks of 1” thickness and 4"x4"

in cross section and varying in Boric acid concentration from

10$ to 80$ by weight, were made.The attenuation curves thus obtained are given in figure 15»

(b) Shielding properties of indigenous wood

Wood contains a considerable amount of dydrocarbons and

is expected to be a good neutron shield. Also it is convenient

to make various sizes and shapes of wood for shielding purposes

in experiments around the reactor beamports. The results are

shown in figure 16. It is concluded that though borated

paraffin is the best neutron shield among these* nevertheless,

"gurgan"(one kind of wood) is also a good shield. Another kind,

"chir", which is rather cheaper than gurgan and is easily avail­

able in the country is rather a more suitable material if wood

shielding blocks are to be used in the experiments.

Radiation Damage Studies (S. Mansoor Ali, K.A. Shoaib, S.U.Cheema, P. H. Hashmi*)

This group has been working for the last three years. It

has undertaken to study the effect of fast neutrons on the

electrical properties of the low mobility semiconductors with

special emphasis on Zr0o and Th 0_. A fluence of the order of19

10 nvt will be needed so as to have the density of the irradia­

tion induced defects comparable to the density of a free charge

carrier. The following programme will be pursued:

(1) Electrical resistivity as a function of temperature.

(2) Linear heating before irradiation quenching and cold

work.

(3) Linear heating after introducing defects.

(4) Isochronal heating measurements.

(5) Isothermal heating measurements.

Reactor Utilization other than Physics

Radioisotope production (Matiur Rehman, M.Y. Mirza,M. Bashiruzzaman, H. M. Karim)

The programme of isotope production is expanding. A

separate plant for processing bulk quantities of 1-131, S-35, and

P-32, as large as 10 Ci/run is being installed.

At present the isotopes are being supplied mainly to medical

centres, industries and training centres in Pakistan. During

1969-TO, the following supplies were made:

Isotope

Ir-192

Na-24

K -42

Co-58Br-82

Cr-51

Au-198 ) In-ll6m) 1-128 )

Supplied to:

Spancers (Pak) Ltd.

SEATO Cholera Lab, Dacca

SEATO Cholera Lab, Dacca

Atomic Energy Centre, Lahore

Atomic Energy Centre, Lahore

Atomic Energy Medical Centre, Karachi

Reactor School, Pinstech

Strength

2 Ci

60 mCi

35 roCi

2 mCi

20 mCi

2 mCi

Severallow activitysources

In addition to meeting the demands of local users in

Pakistan, this division is planning the export of isotopes on

competitive international rates. After the full scale working

of the above mentioned plants, the division hopes to be able to

export the surplus material.

Further, the research and development programme for the

production and processing of other radioisotopes like Au-198,

Tc-99m, Co-58 and Cr-51 is also being pursued.

Activation Analysis

The reactor is also used by the Nuclear Chemistry Division

is determining the composition of materials by the neutron

activation analysis. The samples are irradiated in the

neutron flux near the core of the reactor through the rabbit

control system. From the spectrum of radiation emitted, one

can infer about the composition of the material.

Reactor School

The school looks after the training requirements of the

operators as well as the engineers for the power reactor programme

of the PAEC in Pakistan. Since its start about 4 years ago,

the Reactor School has organized one year’s course for the

reactor operators which were mostly menât for the 137 MW KAJJUPP

reactor (Karachi Nuclear Power Project) and the 200 MW reactor

for Roopur, East Pakistan. The former reactor has just 'become

critical while the latter is under the active programme of the

PAEC.

For the last three years, the reactor school has concen­

trated on the training of nuclear engineers with a one year

extensive programme consisting of lecture courses in physics,

health physics, engineering and nuclear engineering. Also

several experiments in physics and nuclear engineering are

conducted in a laboratory. In addition to several nuclear physics

and neutron physics experiments, the following experiments on

the 5 MW research reactor are conducted:

(1) Start up experiment on the Pakistan Research Reactor

(PARR) at PINSTECH.

(2) Control rod calibration of the PARR.

(3) Reactivity effect of the thermal column and beam tube

flooding.

(4) To obtain the diffraction pattern of reactor neutrons

and use the monoenergetic neutrons for the determina­

tion of the total neutron cross section of Cobalt and

Indium, etc.

The reactor school trainees who are basically either physics

graduates or engineering graduates, get a MSc degree (Nuclear

Technology) from the University of Islamabad. In the 1970-71

batch of trainees there are more than 30 persons. The University

of Islamabad is responsible for conducting the examination and

for the award of degrees to the successful candidates.

The reactor school also has the provision of imparting

training to the candidates from other countries who have

expressed interest in it.

In addition to the five permanent members of the Reactor

School staff, an IAEA expert and scientists and engineers from

several other disciplines like physics, chemistry, health physics,

electronics, nuclear engineering, etc., give specialized courses

to the trainees.

Concluding Remarks

The utilization of the research reactor at Pinstech is

being made for various disciplines. The activities are further

being expanded by setting more experiments at the remaining

beam tubes and using the irradiation facilities in the Reactor

School.

Recently an arrangement is being made with the University

of Islamabad where the students of the Universities can use the

reactor facility for studies leading to the degree of M.Phil

and Ph.D. of this University.

Acknowledgement

The author is grateful to Mr. Ahmed Ali, Technical Officer

and other technical members of the Neutron Diffraction Group and

to the Workshop Staff for help in technical matters.

Useful discussions with Dr. M. A. Shaukat and Dr. G. D. Alam

are also acknowledged.

TABLE I

Mean Values and Widths of Distributions

..... .... — ................— —" “■ ...—1 — .. T, rn n- , « -

239™.Pu + N Thermal 239 NPu resonance

Quantity This work Ref A*

Ek (MeV) I78.3+2.O I7 7.7+I .8 I76.9+2.O

K (MeV) 12.1 12.2 12.1

“L101.2 100.34 IOO.9

mH 139.6 139.66 I39.2

mL =fflH 6.7 6.01 6.7

*with neutron emission correction

Ref. A.: J. N. Heiler, P. J. Walter and H. W. Schmitt.Phys. Rev. 149, No. 3, 894 (1966).

GAMMA CEIL DOOR

OPEN END•“* ------

BEAM TUBE No.2 8* DIA

BEAM TUBE No.1 8'DIA

THROUGH TUBE 6 'DIA

/Beam Tübe Ho. 3 (6 ’ DIA)

THERMAL COLUMN. DOOR

rh STALLiÜIPe n d JWn

n \\

BEAM TUBE No 6 6TDIA

BEAM TUBE No 5 8'DIA

GRAPHITE THERMAL COLUMN.

BEAM TUBE No U S'DIA

Fig. 1

pARR experimental facilities layout

FIGURE 2. TYPICAL BEAM TUBE SKETCH AT PAKISTA M RESEARCH REACTOR

ta

'

«*■ 1

Éa ..

Ia

1

ACORE

*— -c -

1

C----- (i B

1

1LEAD PLATE-''

>

^-FINISHED POOL WALL LEAD PLATE- ^1

DIMENSIONS ARE IN INCHES

r* g n

BEAM TUBE DIMENSIONS TABLE

BEAM TUBES PLUGS

BE

AM

TU

BE

No

NO

MIN

ALD

IA a b C d e f g ,(app.)

LEAD PLUG A CONCRETE PLUGB CONCRETE PLUG C

DIA LENGTH No OF DIA LENGTH No OF DIA LENGTH No OF

1 8 15.0C 10.00 8.56 13 37.00 63.00 70 U.81 10.43 1 9181 11.81 3 8.31 13.31 22 8 15.00 10.00 8.56 13 37.00 62.00 70 U.81 10.43 1 9.81 11.81 3 8.31 13.31 23 6 H . 75 9.50 6.50 12 37.50 68.50 70 U.50 10.31 1 9.00 12.00 3 6.00 13.50 24 6 H.75 9.50 6.50 12 37.50 62.50 70 U.50 10.31 1 9.00 1100 3 6.00 13.50 25 8 15.00 10.00 8.56 13 37.00 60.50 70 U.81 1(K3 1 9.81 11.81 3 8.31 13.31 26 6 K.75 9.50 6.50 12 37.50 60.75 70 U.50 10.31 1 9.00 12.00 3 6.00 13.50 2

THRUTUBE 6 12.00 10.00 6.00 13 33.50 — 70 U.50 10.31 2 9.00 12.00 6 6.00 13.50 U

FIG .3. BLOCK DIAGRAM OF CIRCUIT AND SCHEME OF EX­PERIMENTAL SET UP

100 200 300 400 500 600NEUTRON ENERGY (MeV) — ►

F1G¿ NORMALISED FISSION RATE AS A FUNCTION ON NEUTRON ENERGY.

90 100 110 120 130 140

FRAGMENT MASS (a mu)

Fig. 5

Mass and Energy distributions for resonance and thermal neutroninduced fission of 2^9pu#

7. YI

ELD 0.1

0.01:

0.001

/r s

\ >

V

•o

o

•o

•O

oY v

oo

oo

o m

• » •o •

o

* oo

239 Pu RESONANCE ENERGY NEUTRON INDUCED FISSION YIELD o

Pu THERMAL ENERGY NEUTRON INDUCED •» FISSION YIELD

o 239

oo

ooo

60 70 80 90 100 110 120 130FRAGMENT MASS (am u) -

U0 150 160 170 180

Pig. 6

Provisional mass distribution for resonance and thermal neutron induced fission of 239Pu, plotted on

a Log/Linear scale.

COUN

TS/C

HAN

NEL

PE

R H

OUR

CHANNEL NUMBER

Fig. 7

REACTOR CORE MONOCHROMATOR (IST-AXIS)

Fig. 8

Layout of the Triple-axis spectrometer.

FIG.9 SHIELDING EXPERIMENTAL SET UP AT THERMAL COLUMN.

180

COUN

TS/1

000

sec

ENERGY TRANSFER 10” rad/sec (w ) -------------- >

FIG.11 (a) PHONONS IN AL(111) DIRECTION,OF SINGLE CRYSTAL SPHERE

PHONON FREQUENCY 10 rad/sec (w)

Pig. 11(b)

Phonons in A1 (ill) direction of single crystal sphere

10 ra

d/se

c

FIG. 12 PHONON DISPERSION CURVE FOR AL(111)

COUN

TS

x 107

2

sec

ANGULAR POSITION OF THE DETECTOR (1 DIV= 2.16)FIG. 13 (a)GEOMETRICAL RESOLUTION OF THE INCIDENT

NEUTRON BEAM

COUN

TS/2

0 se

c

5660 5680 5700 5720 (1 DIV=2.16'IANGULAR POSITION CRYSTAL

FIG. 13(b) REFLECTION PEAK O F A LΠD FO R NEUTRON WAVELENGTH = 1 A#

Fig. 14

Experimental set up to study the neutron shielding properties of indigenous wood .

CON

CRET

E

CURVE 1 PURE WAX CURVE 2 10 7. BORIC ACID CURVE 3 20 BORIC ACID CURVE U 30 7. BORIC ACID CURVE 5 40 7. BORIC ACID CURVE 6 50 7* BORATED WAX

12 15 18

Thermal neutron atenuation of pure wax and horated wax

STATUS REPORT ON PRR-l *

byLibrado D. Ibe

Acting Commissioner Philippine Atomic Energy Commission

A BSTRACT

The PRR-1 is the principal facility used in the furtherance of atomic energy activities in the Philippines. It is utilized for isotope production* sample irradiations, and conduct of experi­ments in the nuclear sciences and engineering and training of per­sonnel. Researches aimed at increasing the utility of PRR*-l and insuring its safety are currently undertaken.

Introduction

The Philippine Research Reactor PRR-1, an open pool type

facility, is the first nuclear reactor in the country. It is located

within the campus of the state-owned University of the Philippines

and is operated and maintained by the Philippine Atomic Energy

Commission. The reactor became critical for the first time on

26 August 1963.

Reactor Utilization

Since the attainment of initial criticality, the PRR-1 has been

the main facility used in the furtherance of atomic energy activities

in the country, particularly in the production of radioisotopes; trace

element determination in different samples; investigations and

* To be read in the IAEA Study Group Meeting on ResearchReactor Utilization in Bandung, Indonesia from 2 to 6 August, 1971

experiments in nuclear sciences and engineering; and the train­

ing in nuclear techniques of scientific and technical personnel of

the Commission as well as other agencies, including university

faculty and students.

The PRR-1 has been operated at different power levels up

to one megawatt for various durations in order to accommodate

requests for sample irradiations and physics experiments. As

of June 1971, the PRR-1 had generated a total of about 1, 200

megawatt-hours of thermal energy and completed more than 4,200

sample irradiations. Isotope production accounted for the majority

of irradiation requests, followed closely by those submitted by the

neutron activation analysis group. The experimental facilities

often used for this purpose include the beam ports, two pneumatic

tubes, in-core radiation baskets and two vertical 2-inch dry

pipes. The dry pipes were added only about two years ago in

order to meet the increasing irradiation requests.

Two of the six beam ports are permanently tied up with

the two neutron crystal spectrometers used for physics experi­

ments. These spectrometers were utilized in the India-Philippines-

Agency (IPA) project, where a number of physicists from Taiwan,

Thailand, Korea, Indonesia and the Philippines received training

in neutron spectrometry. This project was terminated in late 1969.

Another occasional use of the reactor is in the conduct of

experiments or demonstrations given in connection with training

courses for scientific and technical personnel of the various institu­

tions, including faculty and student members from local universi­

ties and colleges. A special training program in reactor engineer­

ing was conducted recently for the engineers of the Manila Electric

Company (MERALCO). The MERALCO, the largest private electric

utility in the country, is seriously considering putting up a nuclear

power plant in the very near future.

Fuel Management

Up to the early part of 1971, the PRR-1 was still using the

original 20% enriched fuel elements loaded ih 1963» The long

period of utilization of these fuel elements could be attributed

to the fuel management procedure adopted by the reactor opera­

tions personnel. For maximum utilization of the fuel, the

reactor operating schedules were programmed to permit simul­

taneous servicing of the irradiation requests submitted by the

research and service units, as well as the isotope production

group. Ten to twelve hours operation at a rated power of one

megawatt were normally scheduled for four days a week with one

day set aside for 100 KW operation. This schedule also afforded

the use of the reactor for fatst neutron irradiation of seeds

with the aid of the IAEA-supplied Standard Neutron Irradiation

Facility(SNIF). Occasionally, full power operation lasting as

much as 40 hours continuously as requested by the reactor

users is also performed.

In 1968, the twenty 93°f< enriched fuel elements fabricated

in the United States arrived. These were intended for partial

replacement of the spent original fuel elements. As mentioned

above, it was only early this year when partial reloading of

the PRR-1 core was donet In the reloading schedule, the out-in

method was adopted. Ten of the 20f enriched elements in the

central section of the core were retired and some of the peripheral

elements brought in. The ten new 93% enriched fuel elements

added were initially mounted on the outer sections of the core.

With this new core configuration and fuel composition, it became

necessary to perform another series of flux and reactivity measure­

ments as well as recalibration of the nuclear instruments.

In the planned second phase of care reloading another set

of ten original fuel elements will be replaced and the remaining

ten new fuel elements brought in. No time schedule has yet been

set for this phase.

Current Researches

Aside from operating the reactor, the PRR-1 operating

personnel are currently engaged in a number of research activi­

ties all of which are aimed at insuring the safety and increasing

the usefullness of the facility. Some of the projects in progress

include:

1. Reactor stability experiment;

2. Improvement of the pool dewatering system; and

3f Transistorizing of the reactor instrumentation system.

In the reactor stability experiment, the objectives are to

(1) measure the fission neutron lifetime in the core, (2) deter­

mine the effect of temperature on the neutron lifetime, and

(3) determine the reactor transfer function. A reactor oscillator

will be used to introduce the sinusoidally varying excess re­

activity in the core.

The slow draining of the pool water in the low-power

section has presented problems on the sealing of the bulkhead

gate« Studies on the rapid dewatering or filling up of the pool

with the use of a centrifugal pump and some means of establish­

ing adequate sealing pressure between the gate and pool divider

wall are being undertaken.

On the instrumentation system, our experience has been

that the reactor iostTttzne&tis not completely transistorized

could not offer prolonged trouble-free services due to heavy

current drain and er pensive heat generated in the vacuum tubes.

In contrast, the fully-transistorized logic and trip actuator

amplifiers of the scram circuit have been giving smooth and

satisfactory performance up to the present» Thus the gradual

conversion of the reactor instrumentation from the vacuum tube

to a solid-state system was decided. The initial phase of the

project involves the transistorizing of the two start-up channels

and the log N and period amplifiers. The conversion of the high

voltage power supplies and gamma radiation remote area monitors

will follow.

In addition to the above, improvement of existing experi­

mental facilities to increase their usefullness is in progress.

This includes the modification of some of the beam ports to ac­

commodate experimental apparatus and the introduction of flexible

handling mechanisms in the vertical dry pipes to permit simul­

taneous and uniform irradiation of several small samples.

Conclusion

Since its initial operation, the PRR-1 has played a major

role in the development and promotion of the atomic energy

program in the Philippines. Besides its scientific and technical

uses, it has helped to generate active interest and awareness in

science from the public so that the Center attends to an average

of 7,000 visitors yearly. These visitors, mainly students from

the high schools and colleges all over the country, are also

briefed on the research, development and training activities of

the PAEC, The PRR-1 also serves as a ready device for the

training of a pool of technical personnel with reactor expertise

and experience, so essential in a country that is already seriously

considering the establishment of a nuclear power plant.

#

Status Report of the Engineering Programs and Proposed

Use of the Korean TRIGA Research Reactors in support of

Power Reactor Fuel Development

Byoung Whie Lee Atomic Energy Research Institute

Office of Atomic Energy Seoul, Korea

-ABSTRACT

The current status of the engineering programs on the use of Korean

TRIGA research reactors is summarized.

The major effort of the research and development for Nuclear Power

in Korea is the power reactor fuel development. In this connection, the

method for the activated sintering of UO2 pellets by TiOg addition was

developed. In order to test any adverse effect on radiation stability and

behaviour of UO^ pellets due to TÍO2 addition, the TRIGA Mark III King

Furnace is proposed for this application.

The paper describes the method for activated sintering and discusses

the prospect of the in-core irradiation using the TRIGA Mark III reactor,

particularly on the effect of TÍO2 addition to UOg.

Introduction rThe Korean Nuclear effort was initiated approximately twelve years

ago when the Atomic Energy Research Institute was first activated under the

Office of Atomic Energy with the purpose of peaceful uses of Atomic Energy.

The first research reactor, TRIGA Mark II, went critical on March,

1962. The power level of the TRIGA Mark II reactor was subsequently upgraded

from 100 Kw to 250 Kwdue to the increase in demand for radioisotope production

and the need for higher neutron flux.

The upgrading of power level was achieved by the following means:

1.) The insertion of six additional fuel elements,

2.) The relocation of control rods,

3.) The capacity increase of the cooling tower, heat exchanger and pumping system,

4») The recalibration of indicators and control instrumentation.

The demand for radioisotopes increased steadily, through the

years. The total number of nuclides produced in the past amounts to

thirtyfour. The production of radioisotopes in this year amounts to

about 18 Ou. A noteworthy trend in radioisotope consumption is the

gradual shift of demand from medical to industrial use. As yet,

approximately SOfo of the total production is used for medical purposes.

However, when the TRIGA Mark III reactor will go critical- in spring 1972,

most radioisotopes for industrial application such as Ir for radiography

will be produced locally. Then, the consumption trend of radioisotopes

would rapidly shift to industrial use. Not only the radiographical

application of radioisotopes, but the tracer and gauging applications

of radioisotopes are expected to increase since the petrochemical, cement,

fertilizer, integrated steel work, related metal working^automotive and

and electronic industries are growing rapidly.

Being a country with short natural energy resources, Korea has

to rely for its major energy supply on imported primary fuel. 'The dependency

on imported fuel would "become more pronounced in the future unless other

domestic energy resources are found. The main source of domestic energy

supply is the anthracite coal deposit for which the economically feasible

maximum supply is only limited to twentyfour million tons annually. The

prospect of the fuel supply and demand in Korea is tabulated in Table I.

Table I.Prospect of the Energy Supply-and Deaand (Unit: 1000 tons of coal eqruiv. )

Year i m . 1980 1985 1990 1995 2000

Total demand 65,200 100,900 146,800 215,800 302,700 424,500

Domestic supply 27,190 28,250 28,050 28,200 28,150 28,500

Coal 21,700 24,000 24,000 24,000 24,000 24,000

Hydropower 1,430 2,250 3,050 3,700 4,150 4,500

Fire wood 4,460 2,000 1,000 500 - -

Fuel to be imported 38,010 72,650 118,750 187,600 274,550 396,000

Fraction ofimported energyto total demand^) 58.3 72 81 87 91 93

Because of the forest preservation effort, the domestic supply

coming from fire wood would decrease rapidly and become nil in 1995.

In view of the large amount of fuel to lie imported, nuclear power

has a definite advantage over the import of oil in the following aspects.

1. Electricity generated from nuclear power is cheaper than

that from conventional imported fuel oil fired in thermal

power plants.

2. Ease of transportation and storage.

3. Less air pollution and public hazards.

Due to the aforementioned advantages of nuclear power over the thermal

power from imported fuel oil, the future increase in electricity demand would

be likely to be met by nuclear power as much as possible. The electricity

demand forecast and the prospect of installed capacity of nuclear power are

tabulated in Table II.

Table II

Year 1975 lgeo 1985 *990 1995 2000

Max.demand (MW) 4,185 7,330 11,590 17,670 25,960 37,270

Installedcapacity (W) 4,604 8,063 12,749 19*437 29,556 40,997

Required new installation

(MW) 2,868 3,459 4,725 6,998 9,384 13,436

New nuclearpower (MW) 600 700 2,600 4,500 8,000 12,800

New conventional power (MW) 2,268 1,759 2,125 2,498 1,354

Total installed capacity of nuclear power(MW) 600 2,300 4,900 9,400 17,400 30,200

Fraction of mxcl. power to total demand (<f0) 13.0 28.5 .38,5 48.2 61 74

As shown in Table II, nuclear power will play à major role in

generating electric power as a base load plant. This role would become more

important in the future. In the year 2000, 74 per cent of the total electric

power demand is estimated to be met by nuclear power plants.

Foreseen nuclear energy as a potentially economically attractive

electric power source the Office of Atomic Energy initiated in 1962 a

Nuclear Power Program. An economic and technical feasibility study was

carried out. IAEA experts on site selection subsequently studied the

prospective sites for the first nuclear power plant. With due processes and

preliminary studies, it was decided to build the first nuclear power plant

at KO-RÏ with a 600 Mwe Westinghouse Pressurized Water type reactor in

January, 1969» The construction of this plant is presently underway. By

the end of 1975» the generated power from the first nuclear reactor will be tied into the distributiorj^etwork.

Current Status of Engineering Programs

The engineering programs in research reactor utilization at the

Korean Atomic Energy Research Institute are divided into two major groups:

Industrial applications and production of radioisotopes, and development

of reactor materials in support of nuclear power programs. The current .

research projects relevant to the aforementioned are listed in appendix I.

Through our past experiences with TRIGA Mark II reactor operations

and the rapid growth in radioisotope demand, the feasibility of the power

level upgrading of the TRIGA Mark III reactor from its original 2 MW to 5 MW

thermal was studied prior to its actual construction in order to provide for

the minor design modification of the coolant system. With proper experiments,

the thermal and hydraulic design parameters were obtained by assuming

appropriate nuclear parameters.

The results are tabulated in Table III.

Heat transfer and Hydraulic Parameters (Predicted) ~ 5 MW(th) TRIGA

Number of fuel elements 120

Diameter 1.47 in.

Length (heated) 5O.O in.

Flow area. 0.695 ft2Wetted perimeter 38.62 ft

Hydraulic diameter 0.0601 ft

Heat transfer surface 58.O ft2

Inlet coolant temperature 90 °F

Exit coolant temperature (avg.) 128 8f

Coolant mass flow approximately 450,000 lb/hr

Avg. flow velocity 3.O ft/sec.

Avg. heat flux 307,000 Btu/hr-ft

Por Optimum operation of the TRIGA Mark II reactor at the power

level of 5 Mw thermal, the required estimated coolant velocity is 3.0 ft/sec.

with a maximum coolant temperature of 128° F. Therefore, 5 Mw thermal

operation of the TRIGA Mark III reactor is not possible with the coolant

natural convection, thus forced convection is required, and in order to

achieve this, three 8 inch down-flow pipes were installed at the "bottom

of the reactor tank.

Fuel Development in support of Nuclear Power

Since nuclear power is expected to play a major role in generating

electricity as a "base load plant in the future, the current engineering

program should be formulated in such a way to support the nuclear power program.

In formulating such a program, the socioeconomic as well as industrial capa­

bility of the country has to be well taken into consideration.

Korean metal producing and metal working industries are now in a stage

of development. Therefore, it would be some time before Korea will be capable

of economically manufacturing the major parts of a nuclear power plant. However,

the fabrication of fuel is estimated to be technically feasible based on the

present state of the art and know-how in Korea. Moreover, the fabrication cost

of fuel is an appreciably large portion of the nuclear power generating cost.

As a result of this philosophy, the fuel fabrication program was started

in 1968 at laboratory scale with assistance from Argonne national Laboratory.

In due course, the method for the activated sintering of U02 pellets by Ti02

addition tías developed. In future, irradiation experiments have to be made

in order to test any adverse effect on radiation stability of U02 pellets due

to Ti02 addition. For this purpose, the use of the TRIGA Mark III King Furnace

is proposed.

Method of Activated Sintering of U02 by Ti02 Addition*

1.) Preparation of U09 powder for cold compaction; As received the U0?

powder is reduced in hydrogen atmosphere at 850°C for 24 hours to

attain a stoichiometric compound usually of U02 0.05 w/o Ti02

powder is mixed to the UOg and 1 w/o polyvinyl alcohol is added as a

binder. Then, the U02 powder is granulated to approximately 20 mesh

size by mechanical mixing. 0.2 w/o of zinc stearate is added and the

UOp is then cold pressed to form a green compact with the density of

6.5 to 7 g/cm .

2.) Presintering of green compact;

The green compact is presintered in a hydrogen atmoshphere furnace

at 850°C for 24 hours in order to drive off the volatile materials

such as binder and lubricant. The density of presintered pellet

decreases about 2'fo from green density.

3.) Pellet sintering;

The presintered pellet is sintered in a hydrogen atmosphere furnace

at 1430°C for 8 hours to form a pellet with the density of 10.1 to

1 0 .3 g /c n A

Comparison of the results:

1.) Sintering temperature suppression due to TÍO2 addition ;

By the addition of 0.05 w/o TÍO2J "the sintering temperature of

UO2 pellets could be lowered as much as 170°C in obtaining the high

density pellets.

2.) Effect of TÍO2 addition on mechanical properties;

There seems to be no apparent adverse effect on compressive stress

of UO2 pellets due to TÍO2 addition.

Proposed use of the TRIGA Mark III King Furnace to test any adverse

effects on radiation stability of UO2 pellets due to TiOg addition :

1.) Description of the TRIGA King Furnace;

The in-core furnace is shown schematically in Figure 1 indicating

the significant parts and components.The heater element (shown in Figure 2),

is an 8-in. long graphite tube, 1-in. 0D x 3/4 in. ID heated by a saturable

core transformer. Surrounding the graphite heater element are two 10-mil

thick molybdenum radiation shields. Outside the shield is the aluminum

containment vessel which is in contect with the reactor pool water. The

dimension and configuration of the outer aluminum containment are the same

as those of the TRIGA fuel elements facilitating the location of the furnace

in a fuel element position. The center of the heater is on the horizontal

center line of the core.

The outer containment tube extends upward from the core to

approximately 5 ft above the water level of the reactor. Inside the

outer containment tube is a concentric aluminum access tube which connects

at the bottom with the graphite heater element and ends at the top with a

flange which provides a viewing window for temperature measurements. The

axial centerline of the graphites heater element coincides with the axial

center line in the TRIGA fuel elements. Samples to be irradiated are

placed inside a graphite container whose inside dimensions are l/2 in. in

diameter and up to 3 in. long. The container is lowered into the furnace

with a special tool. The furnace is constructed so that the lower section

can be disassembled to permit replacement of the graphite heater element

or molybdenum thermal shield. Figures 3,4, 5» 6 show the assembly of

the principal components of the lower heating section. Note in particular

how the molybdenum shield, graphite heater element and lower section of the

furnace itself can be disassembled.

The upper section of the furnace is equipped with an apparatus to

permit temperature measurements. Figure 7 is a close-up view of an apparatus

used in steady-state irradiation with an optical pyrometer to measure the

temperature. An alternative top section is available (not shown) which has

been designed specifically for use with a photomultiplier for temperature

measurements when the sample is exposed to pulsed irradiation. The output

from the photomultiplier tube can be displayed on an oscilloscope which can

then be photographed to provide a permanent record of temperature versus time.

With a dual trace scope, a concurrent trace of the reactor pulse can

also be obtained.

The design of the furnace is arranged so that a sample can be purged

with an inert gas such as helium or run at a static gas pressure during

irradiation. Notice the gauge in Figure 7» Directly behind the gauge,which

records both inches of mercury vacuum and gas pressure, is the vacuum port

and helium fill line. In operation, the furnace is first evacuated and then

purged with helium (also see Figure 1).

2.) Proposed test program using the TRIGA King Furnace;

Since the King Furnace is capable of heating a sample as high as

2000°C, the irradiation of U02 pellets is proposed to be performed at

temperatures from abient to 1500°C at an appropriate constant power level

during one hour to yield 10'*' fissions. The data on fission gas release

as a function of temperature would be obtained and the effect of TiC^

addition on fission gas release could be studied. Through metallographies!

examination and the measurement of irradiated UOg pellets, the effect of

TÍO2 additions on the dimensional stability would be studied.

Summary

The current engineering program at KAERI in utilization of TRIGA

reactors can be summarized as the radioisotope production and the development

of fuel in support of the Korean nuclear power program. If the results of

irradiation experiments on activated sintered fuel are favorable, the

developed fuel would be applicable for nuclear power plants and the application

of the developed method for commercial fuel production would contribute to the

decrease in nuclear fuel costs.

References

1. G.T. Schnürer, A.T. McMain, and P.U. Fischer: IAEA Proceedings N0.I3O

on Engineering Programs in Research Reactors, 231, IAEA, Vienna, 1971

2. E.E. Anderson, S. Langer, N.L. Baldwin, and F.E. Vanslager: •

Nuclear Technology, Vol.11, 259-265, 1971»

3. Hj. Matzke: Journal of Nuclear Materials, Vol.20, 328-331, 1966.

4. S. Naymark and C.N. Spalarist Proceedings of Third International

Conference on the peaceful uses of Atomic energy, Vol.11, 425-435»

United Nations, New York, 1965.

List of Research Projects in Support of Engineering Programsat the A.E.R.I.

I. Research and Development in Applications of Radio-Isotopes

1. Preparation of fiber "board,

2. Industrial radiation source production and development,

3. Radiation grafting on textiles and fibres,60

4. Management of tihe Co gamma irradiation facility,

5. Syntheses of radio chemicals,

6. Radiochemical studies on the Szilard-Chalmers process,

7. Construction and applications of a RI excited X-ray source,

8. A study on the syntheses or tritium luminous compounds,

9. Development of a fire alarm applying radio-isotopes,

10. Development of a radioisotopic power generator.

II* Research and Development of Reactor Materials

1. Measurement of neutron total cross sections of reactor materials,

2. Badiative capture cross section measurements of reactor materials,

3. Past critical assembly criticality calculations,

4. Calculation and measurement of fast neutron energy spectrum and flux,

5. Measurements of reactor parameters by means of a pulsed neutron generator,

6. Heat transfer studies in fast breeder reactors,

7« System design of FCAfs control and instrumentation system,

8. Kinetic studies of fast critical assemblies,

9. Research on fast breeder reactor instrumentation system,

10. Past neutron detector development for FCA,

11. Fuel fabrication for fast breeder reactors,

12. Preparation of nuclear fuels,

13. Fuel cycle analysis for fast critical assemblies,

14. Studies on reactor shielding using domestic minerals (for neutronshielding materials),

15. Nuclear magnetic resonance in UO2 + x

16. Study of microscopic dynamics in reactor materials by neutron scatteringj

17. Structure analysis of reactor materials for FCA’s by neutron scattering,

18. Study of the gamma ray energy level of ferrous oxide by the

Mbssbauer effect,

19. TL effect of natural calcium fluoride,

20. Study to correlate the colour centre and thermoluminescence in LiF,

OVERALL WEfGHT

23 PT.

TËMPÊRATURÊ MONITORfNG MIRROR

CLAMP

UPPERFLANGE

HEATERCOMPRESSION

SPRINGS

POfÈMftÂL ELECTRODE

mSÜLÀTOrt

HÊLÏUM PuftGÈ ÔÜTLÊT

iNTtBNÂλELECTftOÖf

V ë SSÊL

GÀS ÊRÊSSUftE SENSOR

' î » ,

ï . ITIUGA ÏCiftg itittiâce sehetssatite

Figure 2. Furnace section of TRIGA King furnace

Figure 3. TRIGA King furnace--furnace section disassembled

Figure 4. TRIGA King furnace-^furnace section partially assembled

Figure 6. TRIGA King furnace--furnace section assembled

CERTAIN ENGINEERING PROBLEMS IN THE THAI RESEARCH REACTOR

"by 'Ratana Pumlek

Chief, Reactor Operation Division,Office of Atomic Energy for Peace .

Bangkok, Thailand'

ABSTRACT

Temporary means were performed to put the reactor back into operation

when the critical components were defective. The real problem was spare parts

shortage because the original supplier was no longer in the reactor business.

A brief description on the replacement of the Curtiss-Wright designed shim-

safety rods and reactor magnets with solid boron stainless steel rods and

RMG type magnets respectively is given.

The Tahi Research Reactor (TRR-l) is an open pool type facility designed

by the Curtiss-Wright Corporation, USA for operation at a maximum power of 1 MW,

it was made critical on 27 October 1962. Prom the first criticality up till now

two significant problems had occurred.

(1) The swelling of B^ C - filled shim-safety rods which resulted in

the jamming of the rods in the guide track of the corresponding fuel elements.

A brief description of the occurence and subsequent corrective action is

described in a separate report presented at this meeting.

(2) Failures of underwater reactor magnets. Two types of control rod

are used in the TRR-l, namely the shim-safety rods and the regulating rod.

Three shim-safety rods are magnetically coupled to three rod drive systems.

Upon power failure or on receiving a scram signal, the exciting current of

the coupling magnets will be cut off and the rods will fall freely into the

core. The regulating rod is bolted directly to the rod drive assembly instead

of being connected through a magnetic coupling.

Curtiss-Wright Reactor Magnets: Three reactor magnets and a spare

were originally supplied by CW for the TRR-l. The magnet has a spherical

magnet face with a 26 inches radius of curvature through which the magnetic

coupling force is applied to a convex armature with a spherical radius of

26 inches, which is in turn bolted to the top end of a shim-safety rod;

the detailed magnet data is shown in Table I. Magnet currents are supplied

by the magnet amplifier incorporated in the safety amplifier (Honeywell type1908-A8).

On October 30, 1967» "tbe safety rod magnet in the rod drive system

No . 3 was found defective and was replaced by the spare magnet.

On November 5 » 1967» "the safety rod magnet in the rod drive system

No.2 was found during functional testing to have one -side of its coil grounded

to its casing. Since no more spare magnets were available, the safety

amplifier had to be modified to have a floated ground i.e. all its internal

ground wires were connected together but disconnected from its chassis and

from the system ground. The modification successfully served as a temporary

measure to make the defective magnet work^ as proved by subsequent functional

tests of the rod drive system.

The magnet in the SR-drive No.2 failed completely on November 11, 1967

thus putting the reactor out of operation. The magnet problems had been noticed

for some time before the failure took place. Since the ffiagnet of CW original

design was no longer available, we had been forced to study some techniques to

repair our defective magnets at the OAEP. In the meantime, it was necessary

to devise temporary means to put the reactor back into operation as soon as

possible. Direct coupling of the SR-2 to the drive system was considered,that is,

to replace the magnet with an aluminum alloy rod. The particular SR then

could not be dropped when a reactor scram was called for and thus reactor

scramming would be achieved by the only two remaining SRs. The assessment

of the risk involved in such operation was as follows. '

Prom previous rod calibration experiments, it had. been established

that in any core configuration ever employed in the TRR-1, the core excess

reactivity was always less than the shutdown worth of each safety rod.

Particularly with the configuration currently in use at that time, the core

excess reactivity was 2.84$ and the rod worth of each SR was experimentally

found to be 2*SQffo. Prom this reasoning and from actual operating experience,

it was clear that one of the two remaining SR which could be scrammed had

enough negative reactivity to shutdown the reactor. Direct coupling was

therefore adopted as the short-term means of solving the immediate problem.

The first attempt to repair the defective magnet was done by

rewinding the magnet coil with the best locally available #36 SWG enamelled

wire and potting the coil with 3M insulation resin Uo.4« The whole magnet

was again assembled and sealed with the same potting resin. After testing,

it was found that its electrical property was unchanged. This first repaired

magnet last only 59 hours in operation. The major cause of failure was

found to be a leak in the coupling seal. The second attempt was done in

•the same manner as described but the coupling seal had been modified. It

was found that the repaired magnet lasted about 340-700 hours or about 2-5 months

of operation.

The above method was a short term solution for the magnet failure

problem. A long term solution was established by seeking the suppliers which

could furnish the magnet that could completely replace the CW magnet.

RMG Type Reactor Magnets: The proposal to replace the CW magnets were

summarized in two categories. The first one was a dry type magnet (out of

water magnet) which would require a newly designed rod drive mechanism. In

view of limited finance available and other practical considerations, we had

to abandon this attempt. The second was to replace the CW magnet requiring a

high voltage/low current source by a new type of under-water magnet which would

require a low voltage/high current source, i.e. a transistorized magnet

power supply.

The RMG type magnet (the type of magnets as used in the ASTRA-reactor)

designed by the Oesterreichische Studiengesellschaft für Atomenergie Ges.m.b.H.

Reaktorzentrum Seibersdorf, Austria, was found to meet our requirement. With

a minor modification, five RMG type magnets, four magnet seats and one magnet

power supply were fabricated in Austria and they were tested and installed

in the TRR-1 in June 1970.

The RMG-type magnet coils were wound on stainless steel coild formers,

wrapped in glass ’’silk" and embedded in a relatively high radiation resistant

epoxy resin (Aradite P with hardener 972). The leads to the coil were made

of BOSTRAD 19» a wire completely insulated by an inorganic material. The

windings were made of Polyimid-insulated copper-wire. The magnetic material

used was stainless steel type ARW 4006 (Schoeller Bleckmann). The holding

force at the nominal current (250 mA) was close to 20 kg. To get a reliable

indication of the safety rod release, a sealed reed switch was positioned on

the outside of the extension rod. The switch was actuated by a permanent

oxide magnet, which was held in the " on M position by the armature of the

safety rod by means of an extension wire. The detailed magnet data is shown

in Table I .

In order to supply a much higher current for the three HMG-type magnets the

magnet current power supply section of the original Honeywell safety amplifier

was replaced by a new magnet current supply of transistorized design with a

built-in pulse generator (O'- 100 m.sec.) for testing magnet release times. This

new magnet current supply was also provided by the Oesterreichische Studien­

gesellschaft fUr Atomenergie and is known as the Magnet Current Supply type

Mg Str. vs. To retain the original safety features, the a.c. power supply

for the Mg Str Vs magnet current supply was connected via the same series

of slow scram contacts. In the fast scram mode, the supply would be scrammed

by any of the three fast scram signals from two safety channels and the

Log-N period amplifier.

Conclusion The RMG type magnet is an "open" type magnet, thus

avoiding the complicated sealing techniques. The coil could be easily

repaired and replaced.

The EMG type magnet in the SR-drive No.l failed on May 3» 1971 and

the spare magnet was installed. Upon investigation it was found that the

magnet failure was due to the break down of the lead wire insulation by

corrosion but the coil itself was still in good condition. The defective

magnet had been continuously in use for over 10 months.

Table I Shim-gfefety Rod Magnet Data

CW magnet RMG magnet

Coil former Lenin bakeliteCoil data Triple-layer enameled Copper #33

Stainless steel Polyimid-insulated copper

Number of turra 10,000 Max.resistance: 85O ohms Max. current 50 mA potted in Aradite CN 502

wire (0.36 mm dia). No.of turns: 1,000Max.resistance 75 ohmsMax, current 300 mA potted in Aradite P withhardener 972

Yoke Armco magnetic ingot iron Stainless steel type ARW 4OO6 (Schoeller Bleckmann)

Magnet seat Armco magnetic ingot iron with a nickel Stainless steel typeplated curvature surface (26" radius) ARW 4006, self-aligning

Can Stainless steel 304 None

THE BORON-STAINLESS STEEL SHIM-SAFETY RODS AND

THEIR WORTHS IN "TRR-1 " CORE AS COMPARED TO THE B.C-FILLED4

RODS

Sobhak P. Kasemsantá Office of Atomic Energy for Peace

Bangkok, Thailand

Abstract

Swelling of the B^C-filled shim safety rods originally

supplied by Curtiss Wright for the TRR-1 reactor is discussed.

The approach taken to replace these safety rods by rods of a

superior design and the experiments to determine the rod worths

of the new shim-rods are described.

I . Deformations of the B^C-Filled Rods

**The shim-safety rods originally supplied by CW for TRR-1

were of laminated construction, consisting of a stainless steel outer shell and a cadmium inner shell fitted with appropriate endpieces; the cavity inside the cadmium inner shell was filled with boron carbide powder.

Swelling of B.C-filled shim-safety rods which led eventually to jamming of the rod in the guide track of corresponding fuel element were experienced in the operation of some swimming pool reactors (see; for example, Refs. (l-2)). For TRR-1, the first experience of this nature was encountered with SR-3, after 32 months of operation, around the end of July 1965; details of the incident and temporary measures taken to solve the problem were reported in the RSC Meeting on 5 August 1965 (3). Again, around mid-September of 1965, SR-1 began to show similar symptoms; it was taken out of the system, compressed to reduce swelling and replaced in the system (as reported in Ref. (4) )•

* Presently with the IAEA, Division of Research and Laboratories

** Curtiss Wright Corp., USA

It was realized that a proper solution to the problems of swelling of the B-C-filled rods would be to replace them with shim-safety rods of better design. Prom literature and direct consultation with the staff of the Ford Nuclear Reactor who experienced similar problems in I960, a decision was made to replace the original rods with ones fabricated from solid boron- stainless steel. Discussions with t#e Diamond Power Specialty Corporation of Lancaster, Ohio', supplier of replacement rods for Ff¥TR, started in September 1965, resulting in the order of three boron-stainless steel shim-safety rods from the company for $ 2,145 (c.i.f. Bangkok, shipment by air).. The three safety rods were received at TRR in August 1966.

I I . Physical Characteristics of the Boron-SS Rods

Tÿpe 304 stainless steel containing 1.5$ of natural boron

Solid blade with cross section of a flattened cylinder (semi-elliptical);4 grooves milled on each of the flattened sides starting approximately 4 inches from top of blade and running clear to th« bottom to reduce weight.

Shown in comparison with those of the original B^C-filled rod in Table 1.

No calorizing in order to maintain smooth surface and close dimensional tolerances.

Details of construction are such that the original piston can be fitted to top of blade without any modification.

No. 70119 - 1827 (Diamond Power Specialty Corp).

Table 1 - Comparison of Dimensions and Weight

Description Boron-SS Rod B.C-filled Rod 4

(l) Width across Flats 0.875/0*865 in. 0.8?5(+ 0.010) in.

(2) begoss Semi-circular 2 .260/2.250 in. 2 .250(+ 0.010) in.

(3) Length of Blade (approximate)

30 inches total 30.56 inches incl. top & bottom plugs.(25 inches poison length)

(4) Weight(approximate)

10 lbs. 8.5 lbs,calculated incl. top & bottom plugs.

Blade Material:

Blade Construction:

Dimensions and weight:

Finish :

Other Features:

Reference Drawing :

Ill. Comparison of Rod Forths

In December 1966, excessive friction between the SR-3 and its guide track within the corresponding fuel element was again ob­served. Moreover, the SR-1 could only be moved freely up to about 23 inches a.bove its insert limit*; beyond that it inadvertently dropped with a normal magnet current of 40-45 mA again, showing a sign of deformation. The absormal behavior of the SR-1 did not interfere with normal operation; however, since the SR-1 and the SR-3 are diagonally opposite and they are normally calibrated against each other when a rod calibration experiment is performed, it was decided to perform a s e ^ e s of experiments befere the de­formation in either rod became/serious to carry out such experiments.

Consequently, the reactor was shut-down from 22-26 December to allow sufficient decay of the shorter half-life poisons.On December 27, the SR-3 was calibrated against the SR-1, employing the standard in­hour method. On December 28-29, the original SR— 3 was removed from the system and the original B^C-filled blade was replaced by one of the new Boron-SS blade. After reassembly,the new SR-3 was tested for free movement in the guide track of the same fuel element and the whole magnet drive system was also run through standard functional test procedures. After making certain that the new SR-3 with a Boron-SS blade worked properly,the new SR-3 was calibrated against the SR-1 on 30 December,employing the same set of parameters in the experiments.

The results of the two experiments are plotted on the same co-ordinatés in term of the integral rod worth of the SR-3 as shown in Fig. 1.

With the original blade the reactor went critical when the SR-3 was at 12.75 inches; the critical position of the SR-3 shifted to 11.45 inches with the new blade while other control rods were at the same positions (practically at withdrawal limits). The shifting of the critical position of the SR-3 clearly indicates that the Boron-SS blade has less rod worth than the original B.C-filled blade. The core configuration used in the experiments was the configuration No. 5-Gl(a). The remaining core excess reactivity determined from both experiments were 1.81$ and 1.82$ respectively; this agreement of results is a proof of the consistency of the experiments and clearly indicates that the shorter half-life poisons had decayed sufficiently and did not interfere with the measurements.

Comparative worths of the new and original blade can be estimated from Fig. 1, taking only the range from 12.75 to 24 inches.In this range :

The integral worth of B.C-filled rod is 1.81$The integral worth of Boron-SS rod is approximately

(1.82-0.3)$ or 1.52$

Therefore, the loss in the control rod worth when changing from the B C filled to the Boron-SS rod is, in percent of original worth, approximately equal to (0.3/l»8l) x 100 or 16.57$ of original worth.

* Normal travel of control and safety rods in TRR-l is 24 inches.

Since the loss is calculated in percent of original worth, the figure should remain practically unchanged when the core configuration is changed, and also when three safety rods are coupled in gang operation.

CW states that the total worth of the three safety rods in gang operation in a typical water-reflected core is approximately 7.2$ £ k/k, taking into account the shadow effect. This quoted value is in good agreement with our previous experiments, within the experimental limits.

Therefore, taking the quoted value as a basis, it can be estimated that for the new Boron-SS rods, the total worth of the three rods in gang operation in a typical water-reflected core should be approximately 6$ 4 k/k.

It should be noted that the total rod worth is a parameter which depends on fuel loading, core configuration and rod arrange­ment. The estimated value of 6$ A k/k, therefore, could be used in planning a new loading but it must be checked by actual measurement after the core is built.

IV. Conclusions

Owing to the tendency of the B^C-filled shim-safety rods to swell and jam in the guide tracks of the corresponding fuel elements, they are to be replaced with safety rods of better design. Three Boron-SS safety rods are now available for such replacement in the TRR-1.

To obtain the comparative woTths of the two types of rods, the TRR-1 is now (February 1967) operating with one Boron-SS and two B^C-filled safety rods.

Actual experiments reveal that in the TRR-1 the worth of the Boron-SS rod is approximately 83.43$ of that of the B^C-filied rod.

It is known from a private communication that in the FNR, the worth of their new Boron-SS rods is approximately 80$ of that of their original B^C-filled rods. Considering that the FNR original rods were slightly different in design from the TRR's B^C-filled rods, the results with the FNR and the TRR seem to be in good agreement.

Based on information from CW and our previous ex­periments, the following can be expected from the new Boron-SS rods:

(1) Individual rod worth : 2.66% A k/k in a typical waterreflected core

3.16$ A k/k in a typical graphite- reflected core

(2) Worth of three rods in gang : 6$ ^Ak/k in a typical water-reflected core

1 Ricker, C.W. & Dunbar, W.R., "FNR Shim-Safety Rod Deformations”, Nuclear Science and Engineering, Vol. 9, 1961, 410-411.

2 Morris, P.A. "Reactor Safety & Construction Practice”, Reactor Safety & Hazards Evaluation Techniques, Vol.l, 203-204;Proc. Sym., Vienna, 14-18 May, 62

3 Kasemsanta, S.P., ’’Partial Jamming of a Shim-Safety Rod in Control Element Guide Tube", Report to RSC Meeting, Thai AEC,2/25O8 , 5 August 1965» (in Thai)

4 Kasemsanta, S.P., B.I. Report No. 2*, "Back-up Informationin Support of the SS Material Balance Report", 31 March 1966, page 3-5*

*) B.I. Reports are submitted regularly to the Agency at 6-months interval under the Safeguard Agreement.)

ROD

WORTH

(•/•

Ak/fc)

Fig.1. INTEGRA! WORTH O f SR-3 (PARTIAL)

fiBBHfift. M . ™EN GINEERING ET UTI LI S A T I O N DES REACTEURS DE RECHERCHE

A U CENTRE D*ETUDES NUCLEAIRES DE GRENOBLE

par

P. MERCHIE

Chef de l a Section d * Exploitation des Réacteurs

Centre d*Etudes Nucléa ir es - GRENOBLE

RESUM E

L a presque totalité des programmes français

d ' i r ra di at io n est réalisée dans des réacteurs d u type piscine*

Après avoir rappelé les caractéristiques

avantageuses des réacteurs pisoine dans le développement des

techniques nucléaires, nous présentons les caractéristiques

générales des réacteurs SILOE, MELUSINE, SILOETTE du Centre

d 1Etudes Nucléaires de GRENOBLE,

N o u s indiquons les améliorations successives qui

ont été apportées à ces réacteurs et notamment les di fférentes

augmentations de puissance que nàus avons effectuées*

Enfin, nous développons les possibilités

expérimentales offertes, ainsi que les pr in ci p a u x types de

dispositifs d ’irradiation qui ont été réalisés et leur

u t i l i s a t i o n s u cours de ces dernières années.

P o u r terminer, nou s évoquons l a collabo ra ti on que

n o u s avons aveo d'autres pays dans los domaines de

1 ' e ng inaering ot de l ’utilis a t i o n des réacteurs de reaheroho.

Plusieurs dizaines do réacteurs do recherche du

typo pisci ne fonctionnant actuellement dans lo mondo entier.

Conçus à 1 1origino sans objeotif précis, mais so situant

dans dos perspectives essentiellement universitaires et do

r ec herche fondamontalo f l'expérience a montré depuis quo cos

réacteurs convenaient également très b i o n à l a recherche

appliquée ou technologique. C'est ainsi qu'en líLÚJtfCE, l a

p r e s q u e totalité des programmes d ’irradiation est réalisée dan©

dos réacteurs du typ© piscine,

„ T R IT ON et HELUSINE, construits e n 1958 p o u r uno puissance

do 1 MW, fonctionnant actuellement respectivement à 6 et

8 MW.

- SILOS, mise e n service en 1963 à 15 M W et modifiée on 1967

p o u r fo nctionner à 30 MW, représente une amélioration

importante dos pilos piscines dos années 50*»60 ot sort do

r é a d t e u r d* essais do matériaux*

•* P o u r OSIRIS (1 9 6 5 ), le changement do sens do oiroulation do

l ' e a u do refroidissement du c o e u r et certains changements

do s tr ucture d o l a pisc in e ont permis d*atteindre une

p u i s sa nc e do 70 MW,

Cos r é a c t e u r s , auxquels il convient d*ajouter

PEGASE, spécialisé dans les irradiations do grosses boucles,

con stituent u n ensemble complet pour les oseáis de matériaux

sous ra yonnement ot sont l a source d'une expérience très

largo dans le domaine do l'emploi des piles de recherche«

Il ressort de cetto expérience que, p o u r los pays dont

les progra mme s nucléaires sont en cours de définition, ou

d o n t los program mes déjà établis entrent dans leu r phase d©

réalisation* l a p oss e s s i o n d * u n réacteur piscine présente

b e a uc ou p d*intérêt et constitue u n solide point d*ancrage

potar lo développement des techniques nucléaires.

Après avoir b rièvement développé los oaraotérietiquos

générales et originales de ces réaoteurs telles q u J elles

apparaissent à l ‘usage, nous décrirons plus pa rt iculièrement

l eô réacteurs de GRENOBLE ; nous présenterons ensuite

1 *u ti li s a t i o n qui en est faite e n recherohe fondamentale ot

on rocb.croh.0 appliquée et technologique et p o u r lesquelles

des dispositifs d ’irradiation hautement epéçialisés ont été

réalisés*

Nous évoquerons également la collaboration qui existe

depuis de nombreuses années avec plusieurs paye é t r a n g e r s ,

not am me nt dos pa ys dont les programmes nucléaires s ‘élaborent

ot se développent et ceci dans les domaines de 1 * exploitation,

d u développement ot do l ’utilisat io n des réadteurs»

No us présenterons les ligaas générales do notre

collaboration, ainsi que les raisons q u i nous ont conduit

à réaliser u n projet de réacteur, spécialement destiné a ux

pays en v oi e de développement, a f i n d'y promou vo ir l a technologie

nucléaire,

2. P ROPRIETES flENlïïRAJÆg Tfl8S REACTEURS PISCINE

Le SHCoès remarquable dos réaoteurs de recherche

du type pisc ine s'explique p a r 1 ’ensemble des qualités qu'ils

présentent ot qui se sont révélées ou confirmées à l'usage f

— simplicité et sûreté de fonctionnement

— souplesse et polyvalence dans l e u r u t ili sa ti on (recherohe

fo nd amentale appliquée et technologique)

♦X h aut es performances depuis quelques années

— faible p r i x de co ns truction et d *exploitation et p a r

conséquent faible p r i x de r e v i e n t des expériences ot dos

irradiat ions

2.1 » Simplicité öt sûreté

Comparés aux réactours de rocherche à e au lotir do

ou d u type tank, co sont en effet des ré act ours plus simples,

dans lesquels o n ne r encontre pas les installations

particulières, généralement sources de difficulté© que sont

p a r exemple t

-» le cha r gement—déchargement des éléments oombuetibles et

des dispositifs expérimentaux

— l a commande dos barres do contrôle

*• le s installations propres à l*eau lourde | circuits de

p u r i f i c a t i o n et couverture gazeuse au n i v e a u supérieur do l a

cuve, étanchéité totale dos circuits et risques triti um

p e n da nt les opérations d' entretien et de démontage», etc..

— l a pressiorisation des circuits do refroidissement, les

oircuits auxiliaires tels quo refroidissement à l ’arrô-fc,

r ef ro idissement des protection© ou de l a piscine.

L a grande disponibilité et l a bonne sûreté de

foncti onn em en t dos ré ac teurs piscine résultent p o u r vine

bonne par t do l e u r simplicité do r é al is at ion à laquelle

s ’ajoutent des facteurs intrinsèques de sûreté b i o n connus»

Les nombreuses études de développement et do sûreté

effectuées e n M I A N CE, à GRENOBLE (neutronique, thermique,

mécanique, hydraulique) et à CÀDARACHE avec le r é a c t e u r CABRI

(excursions de puissance, accidents de réfrigération, oto..)

ont débouché s u r u n o connaissance approfondie de ce type

do réacteur, oe qui perme t de los u t i l i s e r aveo le m a x i m u m/

d o re ndement ot de s û r e t é •

Cooi s * illustre p a r les améliorations et les

augmenta ti on s de p ui ssance successives réalisées sur les

différents réacteurs français*

2 .2 . Souplesse d 1emploi ot n o l w a l o n o o dans leu r

ut il is a t i o n

L a grande s o u p l e s s e de leurs structures le ur permet

de s ’adapter facilement et rapidement aux besoins variés des

expérimentateurs* L a modifi ca ti on des structures du c oeur ou

112

1 *a djonction d *équipements supplémentaires (cuve à eau lourde,

source froide, oanaux,, e t c ,. ) ne présen te nt pas do difficultés

particulières oar le ur bonne accessibilité perme t des démontages

ot dos interventions faciles ot rapides, môme sous plusieurs

mètres d ’eau»

P a r ailleurs, le grand volume immédiatement

disponi bl e autour d u fcoour permet aux dispositifs d'irradiation

ot a u x expérionoes particulières de cohabiter en très grand

nomb re ot facilite beaucoup l e u r manutention*

Ii*absence de cuve ou de ta nk autour du coeur y

ajoute los p r in cip au x avantages suivants s

-• possibilité de réal is er toute configuration d u coeur qui

convient le m i e u x aux besoins des expérimentateurs

- v isibilité totale et accessibilité remarquable d u coeur,

verticalement ou latéralement, ce qui explique l a rapidité

des interventions sur le coeur (chargement des éléments

combustibles d u coeur e n quelques heures p a r exemple)

** acoès direct des canaux tout contre le coeur et grand

nombre d 1 emplacements périphériques à flux élevés sans

at ténuation n i dégr ad at io n des flux p a r les parois d ’u n

oaisson séparant le c oeur de s o n réflecteur

~ simplicité et facilité de chargement et do déchargement des

dispositifs oaepé riment aux dont certaine sont notanmont

effectuée ¡réacteur en fonctionnement* Sans p én aliser les

irradiations on cours p a r arrôt d u réacteur, cette d e m i è è e

possibilité autorise le retrait ou l a mise on p lace dos

radioéléments^ de certains dispositifs en pajme ou dont

l ’irradia ti on commence ou se termine au cours d ’u n cycle

do fonctionnement, dos boucles froides souvent utilisées

p o u r les irradiations de recherche fondamentale de durée

relativement courte, des dispositifs dont o n veut suivre

pas à pas l ’évolution p a r neutrographie (exemple : examens

suooossifs on cours d ’irradiation de la formation du t r o u

cen tr al dans les crayons combustibles UOg ~ PuOg de l a

filière r a p i d e )*

Cet-fce remarquable souplesse d'emploi explique

l' utilisation polyvalent g des réacteurs pisc in e t

— p r o d u c t i o n de radioéléments

— recheroho fondamentale avec les canaux, boucles froides,

cuve à eau lourde, tubes p n e u m a t i q u e s , etc.*

— recheroho appliquée ou technologique avec des dispositifs

expérimentaux spécialement développés à cet usage (fours et

boucles do différents types décrits ci-aprèB)*

2*3* Ha utes •performances

Los performances atteintes vont également dans le

sens de la polyvalence. Construits à l'origine pour des

puissances relativement faibles, leur ut il i s a t i o n était

principalement réservée à l a recherche f o n d a m e n t a l e • P a r

suite de l'amélioration continue de leurs performances, les

réaotours piscine ont réalisé une remarquable pe rcée ve rs les

flux élevés, ce qui explique l e u r succès e n recherche appliquée

et technologique. Ces réacteurs s'adaptent donc parfaitement

aux programmes évolutifs et peuvent fr anchir p a r étapes des pu is »

sanees de pl us en plus élevées, de quelques M W à quelques

dizaines de M W p o u r répondre à l ’évolution des b e so in s dos

u t i l i s a t e u r s •

P a r exemple, les f l u x disponibles dans SILOE à 30 MW

sont les suivants t

14 i 2— flu x thermique I 4,7 * 10 n / c m »sec

A A Q .

» f l u x rapide f (B > 1 MeV) : 2,3*10 n / o m .soc. (flux rapide

directomont utili sab le à l ‘intérieur d ' u n dispositif

d 1ir radiation placé dans le coeur).

2.4* F aible coftt do construction et d 'exploitation

Enfin, du fait de le ur simplicité ot de l eur grande

disponibilité, ces réacteurs n e sont pas coûteux » "une étude

effectuée en 1969 avec ac tualisation dos p r i x au 1er janvier

1 969 a montré que l es p r i x de construction des réacteurs piscine

étaient les p l u s b a s et que leurs frais d ’exploitation étaient

les plus faibles*

L g d expérimentât ours d ‘ tua réacteur piscine

réaliseront donc uno appréciable économie e ur le coût des

irradiations auxquelles il convient d ’ajouter los économies

s ur les dispositifs expérimentaux eux-mômes qui, à performances

égales, sont généralement plus simples ot donc moins chers

que dans tout autre ré ac teu r à ta nk ou à oau lourde»

3* LES REACTEURS D E GRENOBLE

L o CEN-G- exploite de ux réaoteurs do recherohe et

d ’ossai do matériaux, MEL US INE et SILOE, ot u n réacteur de

b as se puissance, SILOETTE.

Ces 3 réacteurs groupés géographiquement, constituent

u n ensemble h o m o gè ne s u r le p l a n dos performances et des

poss ib il ité s expérimentales offortes a ux u t i l i s a t e u r s • Ces

derniers peuvent aussi bé né fi ci er s ’ils lo désirent, d ’u n

ensemble do eorvicos spécialisés (fourniture do dispositifs

expérimentaux, dosimétrio, neutrographie, calculateur, etc.«),

mis e n p lace dans lo but do l o u r fournir los meilleures

conditions pou r r é a l i s e r louas irradiations «

3*1« SILQB - Caractéristiques générales (fig. f ot 2)

Co r é a c te ur de 30 MW se différonoie dos réaotcura

p i s c i n e construits avant l u i p a r u n certa in nomb ro do

dispositions originales qui ont ensuite servi de modèles

aux modernisations et transformations apportées a u x réaoteurs

p l u s anciens situés soit e n Prance, soit à l ’étranger (bloc

coeur, o rg an isation d u circuit primaire* omplacroment de l a

cellule chaude, équipement dos zones expérimentales,

éléments combustibles, b a r res d e contrôle, etc..).

Ce réacteur, initialement p r é v u p o u r une p ui ssance

do 10 M W a tout do suite fonotionné à 15 MW» Après 4 ans de

fonctionnement, il a subi u n cortain nombre d ’améliorations

qui. ont p e rm is d ’augmenter s a p ui ssance à 30 Mtf o n 1967. E n

1 9 7 1 , après d ’autres améliorations sur le circuit primaire

et soar les éléments combustibles, sa p ui ssance s e r a p or té e

à 35 MW.

Lo coeur, do géométrie rectangulaire avoc plusieurs

crénoaux, ost formé d ’éléments combustibles do différents

typos fabriqués p a r l a Compagnie d 1Etudes ot do R é a l i sations

de Combustibles Atomiques (CERCA) avec le concours d u C.E,A.

Ces àlénonts sont tous à plaquas planes»

a) éléments st andard comportant 2 3 plaquos chargées avoc do

l ’ur a n i u m enrichi à 93 $

b) éléments de contrôle à 1 7 plaques dons lesquels se déplacent

les barres do contrôle du type "fourchetteM

o) éléments spéciaux d 'i rradiation dont les plaquos

combustibles entourent 1 ou 2 emplacements d *irradiation

riches en flux rapides ot dans lesquels viennent se p la ce r

des dispositifs e xpérimentaux (fig* 3)»

Le coeur ost plaoé sur u n tabouret p ar l'intermédiaire

d ’une grille qui comporte 100 positions dont u n e quarantaine

environ est occupée p a r les éléments combustibles ot les

éléments réflecteurs en b e r y l li um qui sont disposés sur une

face d u coeur. Les emplacements qui restent peuvent ôtre

occupés p a r des dispositifs expérimentaux* L 'év ol u t i o n do ces

dispositifs a été t elle q u' il est maintenant courant do p la co r

4 dispositifs dans u n môm e emplacement d ’irradiation, co qui

multiplie d ’autant les positions d ' i rr adi at io n et notamment

les pos it ion s à flux élevés*

L'étude noutronique, h ydraulique et thermique du

coeur ost p ar ticulièrement poussée a f i n do réduiro les portos

de charge du coeur et d ’extraire le m a xi mu m do puissance dos

éléments p o u r obtenir des fl ux élevés (P ^ ~ 270 k W / l -P ' m o y e n '

- 125 w / o n

Le refroidissement d u coour ost assuré p a r l a

c irc ul at io n do l 'e au do la piscine entre les plaquos dos3

éléments oombu stibies à u n débit de 2200 m /h environ. A l a

sortie d u cooUr, cette © a u traverse dos bacs do dé sactivation

et se refroidit ensuite dans dos échangeurs classiques avant

do re to ur ner dans l a p i sc ine dont l a température ost stable

au tour do 3 0 °C*

C g débit p ri ma i r e as s vir o en môme temps <ye lo

re froidissement du coeur celui do t o u s les dispositifs

expérimentaux placés dans ou autour du coeur»

Los fonctionnements à des puissances passant do

15 M W à 30 et 35 M W sont possibles car n oue avons t

•- réduit lo p ic do puissance dans lo coeur

— optimisé les éléments combustibles

•• augmenté lo débit prima ir e do 1500 m^/h, à 15 M W jusqu'à*7

2200 m / h p o u r 35 MW. L a limite os* dans ce cas imposée

p a r 1 ’augmentation de l a porte do charge du coeur et los

ris qu oo de oavitation dos pompes primaires« L*accroissornent

à 3 5 M W de l a capacité d'échange thermique du circuit

pr imaire s'est effectué en rajoutant 3 échangeurs de

chaleur i&ûtt&dqttQS tvux 3 échange rus de chaleur existante à 15 MW*

*• augmenté le débit secondaire do 1200 m ^ / h à 1 5 0 0 m ^ / h

(l*eau froide d u oircuit secondaire est extraite de l a

nappe phréatique à une température no dépassant jamais 13°C ;

elle est ensuite rojotée dans l a rivière qui passe à

p r o x i m i t é )•

L a ré du ction d u p ic de puissance est obtenue on

réduisant la largeur des canaux dans lesquels se déplacent

los barro s do contrôle. U n n o u v e l élément do contrôle et u n

n o u v e a u type de b arre de contrôle, appelé "barre fourchotte"

ont été développés. L lan cienne ba rr e do contrôle d'épaisseur

21 m m est remplacée p a r deu x plaques absorbantes d' épaisseur

3 m m fsituées latéralement dans le n o u v e l élément combustible.

L a distanoo entre les plaques combustibles situées de part

et d * autre du canal de barre est ainsi notablement réduite

avec comme conséquence direoto l a ré du ct io n des pics de

p u i s sa nc e qui se situont toujours sur cos plaques combustiblesy

lorsque l a barre de contrôle est retirée» Les plaques

absorbantes sont o n A g- In — Cd avec u n rovôtemont do N i ck el

électrolytiquo ou b i e n on H a f n i u m qui n la p as b e s o i n d'ôtro

prot ég é contre l a corrosion (fig* 4)«

Les barres "fourchettes“ offrent p a r al lie vira une

me illeure efficaoité que les barres centrales« Ainsi, en

maintena nt le môme nombre de barres de contrôle, il a été

p o s s i b l e d ’augmenter l a charge en TJ5 des éléments combustibles

et du ooeuTt

i

L 'o pt im is at ion de l'élément combustible a oonsisté

dans l a diminution de l a largeur du canal de refroidissement

qui est passé de 2 , 9 m m à 2 , 1 m m ; ainsi, avec l a môme

gé om ét ri e extérieure, le n ouvel élément standard oomporte

23 plaques a u l i e u de 18 et le nouv el élément de contrôle en

comporte 1 7 au l i e u de 1 2 . L ’augmentation du nombre do plaques

et 1 1 au gm entation simultanée de l a oharge en U 5 des plaques

(qui est passée de 1 0 , 9 gr à 1 2 , 2 1 gr ou 1 4 , 7 2 gr suivant les

éléments ou l a p o s iti on dos plaques dans les é l é m e n t ^

signifient que l a oharge t o t a l e en U 5 de l'élément est fortement

augmentée (280 gr — 3 3 0 gr ou 3 4 0 gr suivant lo type d'élément

standard contre 196 gr dans IL'élémeirfc à, 18 plaquos)*

Il e s t ainsi possible d 'o bt en ir des b u m —up très

élevés, de l'ordre do 50 $, ce qui représente dos économies

importantes dans le coût d'exploitation du réacteur»

Les nombreuses études thermiques cffeotuées sur

des boucles ho rs -pile à GRENOBLE ou aveo le ré ac teu r CABRI à

CADARACHE, ont pe rmis do b i e n connaître les phénomènes

da ng e r e u x que sont 1 'ébullition looale ot le renve rs em en t de

débit conduisant au b u m - o u t . Coci nous a permis de mieux

déf in ir les marges de sécurité e n marche normale et d 1 écrire

des codes et des programmes do calcul très préois qui dorment

en f o n ct io n dos différents paramètres d ' u n réacteur les flux

calorifiquos d' ébu llition looale et de renversement de débit*

Enfin, il faut mentionner que le problème de

l'activ it é en surface do la piscine a été résolu dès l a

cons tr uc tio n en maintenant a u sommet do l a piscine une couche

d 1oau ohaudo d'épai ss eu r comprise entre 1 et 2 mètres et dont

l a t em pérature est supérieure de quelques degrés à l a température

d u reste de l a piscine. Cotte couche d ' e a u ohaudo arrôte les

los mouvomontß ascendants d'eau ohaudeactive on provenance

du voisinage du coeur et réduit ainsi l ’aotivité ambiante

à l a surface de la piscine d'u n f a o t e u r 20, Getto méthode a

été également appliquée p a r l a suite auz autres réacteurs

p i s o i n e d u C.E.A* (MELUSINE - TRITON - O S I RI S).

3*2. Possibji 1 -îté« oypérimn ^ a.jir»f> et services spécialisés

offerts aux, utilisateurs de SILOE

SILOE a été conçue pour les expérimentateurs en

tenant compte de 1 *importante expérience acquise avec MELUSINE

dont le chargement on expériences était à saturation avant l a

cons tr uc tio n de S ILOE en 1961.

Les surfaces offertes aux expérimentateurs sont

importantes ot réparties sur 4 niveaux. L'équipement de ces

zones expérimentales a été poussée a u m a x im um p o u r faciliter

1 1 installation des expériences : ali mentation élootrique à

pa r t i r d u r éso au secour u et du résea u à sûreté totale,

li aisons de séourité avec le t a b l e a u de contrôle d u r é a o t o u r

et aveo le calculateur, al im entation en fluides divers,

évacuation d*effluents de toute nature, etc,«

A 30 MW, les flux offerts a u x expérimentateurs sont

les suivants t

a) flux^rapidea^ (E > ^ 1 M o V 2

- 6 emplacements dans les éléments d 'irradiation d u coeur

avec des flux utiles dans les dispositifs d ’irradiation14 / 2

compris entre 1 , 8 ot 2 ,3 * 1 0 n / o m soc

« parmi les emplacements autour d u coeur, une quinzaine ont13 1 4 / 2

des flux dans l ' e a u compris entre 2 ,5 . 1 0 et 1 ,5 * 1 0 n / o m sec*

b) fltux thermiques

Parmi tous les emplacements utilisables, u n o

vi ng ta in e ont des fl ux compris entre 1 , 5 et 4 , 7 * 1 0 ^ n/cm^ sec.

o) Bdhauffement gamma

Les emplacements d u coour o u immédiatornent au tour

d u coour ont des éohauffementa compris entre 2 et 12 Vf/g

(dans le graphite)*

D o u x carnuz sont installés p oux la re cherche

fonda me nta le j situés derrière l a face d u ooeur réfléohie

p a r le béryllium, ils libèrent les trois autres faces p o u r les

dispositifs d'irradia ti on verticaux» Les flux thermiques

disponibles à l a sortie dos collimateurs (S = 70 X 30 mm) sont 8 2

de 3,5 à 4.10 n / o m sec. Les canaux sont actuellement

u t i lis és p o u r l a D i f fr ao ti on Neutroniq.ue (2 goniomètres p ar

canal). L'un des goniomètres est adapté à l ’étude des

Btruotures cristallines de eubstanoos organiques à p a rt ir do

monocristaux, les trois autres sont destinés à l'étude de

substances magnétiques, soit sur monocristaux, soit soir

poudres. Dans ce dernier cas, le trav ail est largement

facilité p a r 1 ' emploi do m u l t i d é t e c t e u r s • L e h a l l et les

bâ ti me nt s extérieurs permettent de travailler on temps do vol

si nécessaire,

3 ,2,2 , Tubes d ’irradia tion

U n tube p ne umatique à deu x v oies d 'irradiation, dont

l a distance d u coeur est réglable, est relié à doux laboratoires

situés hors do l ’enceinte du réacteur. Ce tube est utilisé pou r

les beso in s do l ’analyse p a r activation, de la Chimie

N u c l é a i r e , e t c •,

3*2,3» Boucl es f r oides à azote liquide

Ces boucles peuv en t occuper des positions fixes

sur une f a c e du coeur et permettent des irradiations dans des13 1 4 / 2

f l u x thermiques compris entre 1 , 1 0 et 2 ,3 » 1 0 n / o m .sec et12 1 3 / 2

des fl ux rapides compris entre 1 , 1 0 et 2 ,8 , 1 0 n / o m .soc,

C o b b o u c le s s o n t m is e s e n p l a c e o u r e t i r é e s d u c o e u r , r é a c t e u r

on f o n c t i o n n e m e n t . Le d é f o u m e m o n t des échantillons se fait de

f a ç o n se mi -automatique et sans réchauffement, afin de maintenir

et de p o u vo ir étudier les défauts créés p a r les neutrons

r a p i d e s ,

Do n ombreuses études de Physique du Solide sont

effectuées e ux dos m ét au x purs, s emi— cond uc te urs , eto. •

(traînage magnétique, frottement interne, eto). E n reoherohe

technologique, oes b o u d e s permettent l'étude des propriétés

des ma tériaux irradiés et utilisés à basse température

(source froide du réacteur à Ha ut —Plux, supra-conducteurs)*

3»2*4* Cellule ohaudo

Sa conception originale, directement en por te -à -

fau x et ouverte s u r l a piscine, l a rond extrêmement facile

d ’utilisation, L'opération, généralement diffioile, qui

consiste à transférer les objets actifs d'un r éacteur vers la

c e llu le chaude peut se faire directement sans rupture do

protection, sans l'intermédiaire d ' u n sas, eto.«, oe qui

appo rt e u n gain de temps appréciable avec une meilleure

sécurité.

S o n équipement interne est adapté aux différentes

opérations qui doivont s* y dérouler t travaux sur les

conteneurs de radio-éléments, interventions, réparations ot

démontages de dispositifs d'irradiation, démantèlement des

dispo si tif s usagés, défournemont ot enfournement spécialisés

(ex. échantillons de matériaux do structure ou do combustible

irradiés dans du Na£), examens métrologiques, eto.,

3.2.5. H outr o g raphie

U n appareil do nsutrographio immergé offre â.

l'expérimentateur l a possibilité d' un examen d u type r a di o-

graphique de s o n disposit if expérimental à tout moment a u cours

do son irradiation, quelle qu'en soit l'activité, D 'uno qualité

identique à celle do l a radiographie X o u de la gammagraphie

olassiquo (impuissantes dans ce cas étant donnée l'aotivité y

des échantillons), ce contrôle v i s u e l permet d'une part do

suivre l'évolution d'uno expérience en oourô d'irradiation

(fissures, gonflements, déformations des échantillons, état

dos pièces mécaniques d u dispositif, eto) et d'autre part de

p r é p a r e r les examens physiques à effectuer on laboratoire

chaud. L a d ime ns io n de la photographie est de 3 0 4 0 cm et

p e rm et de couvrir en une ou do ux images l a totalité do l a aono

irradiée» L a dtirée d ' u n examen est d 'environ 30 minutes (Pig.5-6) •

121

U n calculateur spécialement consacré aus: expériences

assure notamment lo pilotage et la surveillance des dispositifs

ot soulage l ’expérimentateur d'un grand nombre de préoccupations

ot de travaux longs ot fastidieux.

E n effet, ce calculateur travaille on temps ré el

ot peut centraliser environ 500 mesures en provenance dos

dispositifs, mesures qu'il surveille et régule p ar a ction directe

s u r les baies de contrôle et do commande associées à chaque

dispositif. A chaque instant, l'expérimentateur peut aussi

interroger le calculateur sur 1 1 état de son <H spositif ot du

ré ac te ur au m o y e n d 'interrogateurs-répondeurs ot, pa r

ailleurs, le calculateur lui fournit systématiquement les

relevés horaires do chacune de ses mesures ; il peut aussi

tracer 3.os courbes d ' évo lu ti on des paramètres mesurés (fig. 7)*

Enfin, on temps partagé, le calculateur effectue

le dépouillement do certaines mesures, p a r exemple* les

m e sur es de dosimétrie associées aux expériences*

Le calculateur est très p e u connecté au réacteur

l u i - m ô m e . S on rôle est alors, par exemple, de f o u rn ir à chaque

e xpérimentateur les informations générales qui complètent ses

informations p articulières : information générale telle que

l a p uis sance du r é a c te ur , directement calculée à p a rt ir des

signaux délivrés p a r les capteurs d u réacteur.

Depuis quelques mois, nous développons dos baies de

co ntrôle numérique, p o u r l a survoillance et l a régulat io n

des dispositifs expérimentaux«.

3»3« Conclusions

Depuis s a mise e n service, Siloé a rendu d ’importants

services si l' on o n juge p a r le nombro important do dispositifs

irradiés. Il a subi dos transformations qui ont sensiblement

amélioré ses performances et il est possible que co réacteur

qui n * a pas encore atteint ses limites supérieures reçoive

d'autres améliorations dans u n proc he avenir»

122

4. LES REACTEURS D E GRENOBLE t MELUSINE

4.1• Caractéristiques

E n 1958, MELUSINE était 1 g premier r éacteur du type

pisci na à ooour ouvert construit on PRANCE.

Depuis cott g dato, los flux disponibles ont été

progressivement augmentés en faisant p a ss er l a puissance à

4 M W (déo. 1 9 6 5 ) et do 4 à 8 M W on sept. 1970.

De nombreuses modifications ont été apportées au

réa ot ou r p o u r fa ci liter s on exploitation du point de vue de

l a sûreté et de l a régularité de fonctionnement et pour

au gm enter ses performances et sos possibilités expérimentales t t a bl ea u do contrôle - circuit do refroidissement - bloc coour —

cuvolage d 'étanchéi t é - accès expérimentaux - h a l l étanche.

L a piscine oomporte trois compartiments t

— lo compartiment coeur, équipé de 3 canaux ho rizontaux

ra d ia ux et d*u n canal tangentiol à doux accès

*- lo compartins©nt m é d ian qui sert do dégagement ot qui pout

Ôtro utilisé p o u r des irradiations sous flux gamma en

utilisant los éléments combustibles irradiés déchargés du

coour (exemple t étude du comportement o n dynamique do

roulements et graisse)

- lo compartiment arrière pou r stockage, essais hors flux ot

d é f o u m o m o n t s dos dispositifs d'irradiation*

Les conditions d'irradiation demandées p a r les

expérimentateurs ont beaucoup évolué depuis 1 ' entrée on

service do MELUSINE. Ceci a conduit à adapter plusieurs fois

lo coour aux besoins ot aux moyens d'irradiation.

Lo coour est actuellement composé d 'uno trentaine

d'éléments combustibles comportant 23 plaques chargéos d'U

enrichi à 93 Uno face du coour est réfléchie p a r du

b é r y l l i u m derrière le quel se trouvent les nea dos canaux.

A I'or&giJXG, l e 0&ÔV& éttë&t .¿&mpoïxâ3X -à ¡utiJpOïit mobile, -ce qui o n diminuait considérablement 1 ’ acc e ss ib ili té *

Depuis 1965, loa é lé me n t s cot»bu£ t ib ie s posés .sur'-iun

t a b o u r e t p a r 1 1 intormédiaire d ’une j£i?l¡y& à. 110 positions

(77 "X 81 îom).« ¿De-débit priaeià^ô d e ï?@froîdies ein ont est d o

5^0 nP/h. {écoulement .descendant).

- -lie coeur, s n f or me gré&êrale e.arrée, compoarte - -

quelques c r é n e a u x on périphérie» A u centre d u coeur, des

é l é men ts spéciaux d ’irradiation, permettent d ’obtenir des"13 2flux rapides relativement élovée ( 0 7*5*10 21/cm #a)

(fig. 3)« ■ -

4»2. ^m eliorations successives ot augmentations do puissance

On tr ouvera ci—dessous 1* h i s torique des principales

modifications ot améliorations apportées a u réac teu r MELTTSIUE.

lïous dirons ensuite quelques mot s s u r l a récente

au gm entation do pu is sa nce de 4 M W à 8 M W (essais effectués

jusqu'à 10,5 MW).

1959 - D é m arrage à 1 M W

1960 — A ug me n t a t i o n à 1,4 M W

1961 — A u g m e n t a t i o n à 2 M W - Au gm en t a t i o n .du débit e t mise

en place d ’u n 2ème échangera?

1 9 6 5 - A u g m e n t a t i o n à 4 M W ~ Mise e n place do vola nts d ’inertie

sur les pompes primaires. A u g m e n t a t i o n du débit

secondaire. I nst allation d ’une couche chaude.

M o di fi ca ti on des structures de l a piscine, d u coeur,

du tableau do contrôle.

1968 - Montage d ’une nouvelle tuyauterie sortie coeur. Mis e

on place dos barres " f o u r o h o t t e s * (fie. 4)

19 7 0 -* Mod if ic at ion complète du -circuit de ré frigération

1971 - Marche n o rm ale à 8 MW,

Coi:-to dernière augmentai;ion de puissance a été

envisagée dans lo contexte suivant t

enveloppo budgétaire : 300.000 F

«*• r éalisation des travaux pendant l a période annuelle d 1 arrôt

du réa ct eu r (août 1970)

*• modifications mineures des dispositifs expérimentaux pou r

les a d a pte r au doublement de puissance*

4.2,1« Etudes__et réali sa tio n de ^ a u g m e n t a t i o n d©

p u i s san oo à 8 M WM « t m » m * m * mm mm m» mm mm.mm mm mm

Les objectifs fixés plus haut ont été atteints

essentiellement p a r la refonte complète de l a partie hors

p i l e du ciresuit de r éfrigération afin d 1 obtenir u n débit

globa l dans lo coeur de 560 m ^ / h et vine capacité de transfert

de 1 * échangeur de chaleur de 8 MW. Cependant l ’étude d u

coour a dû ôtre reprise.

4.2.1.1. Coeug

L a configuration d u coeur a subi de légères

modifications p o u r obtenir le flux max im um cfens les dispositifs

exp ér iment aus:.

1) Les doux boîtes d'irradiation dans le coeur sont remplacées

p a r do ux éléments combustibles avec cavité double

d ' á -rrra d ia -fc io n , d ’ où. u n g a i n d o l ' o r d r e do 2 0 s u r l e s

f l u x r a p i d e s .

2) Les éléments oombustiblos neufs sont placés on p é ri ph éri e

ot n o n plus a u contre du coeur. Le gain sur le flux

thermique autour d u coeur est de 6 % en moyenne. Lo gain

sur le flux rapide autour d u coeur est de 20 % en moyenne*

E n contre-partie le chargement d u ooeur avec

élémontB neufs on périphérie nécessite, à réactivité constante»

vin apport supplémentaire d* U 235 de 10

Lee p ri nc i p a u x résultats des calculs thormiquos

d u coeur sont indiqués dans le tableau suivant p o u r le débit

pri ma ir e n o r m a l * 5 6 0 m /h^et po ur le débit do sécurité i

450 m /h, avec u ne p i sci ne à 30°C.

Tempé ra tur e de gaîne au p o i n t le plus chaud p o u r P s= 8 MW,

= 560 m - V h

C a lc ul avec les valeurs nominales

Ca lc ul avec cumul des coefficients d 1incertitude au point ohaud

83°C 109°C

Puissan ce correspondant à 1 1 é bullition au point chaud

<

§«a*es

15,7 Mtf 10 M W

« &S-4-Icy*'

14,3 M W 8, 75 M W

P u i s s a n c e correspondant à l a r e d i s tr ib uti on de déb it dans le canal ohaud

-c" s

§1

CLG f

20,4 M W 11,4 M W

es?-*•

t0.O

16,3 M W

I

9,3 M W

4.2.1.2. Ech an ge r^

D e u x options furent envisagées t ajou te r u n

tr oi sième éc hangeur ou installer u n échangeur uniq ue do 8 MW.

Cette dernière solution avec u n échangeur à plaques a été

r e t e n u e p o u r les raisons suivantes t

- fa cilité d' im plantation dans u n local exigu. Ce type

d 1é changeur ost très remarquable p a r sa compacité (volume

de l a partie active î 1 m )

- gai n financier important p a r rapport à la première s olution

p a r suite do l a simplification oxtrômo doe tuyauteries

— faible duréo do montage

Cos échangeurs à plaques sont intéressants aussi

à d ’autres points de vue t

- possibilité d ’augmenter l a capacité d'échange en rajoutant

des plaques sur le môme bâti

- coefficient d'échange important, d ’où faible consommation

d' ea u indus tri olie (pour une surface d'échange et une

différence do température primaire-secondairo données)

— faible encrassement grâce aux turbulences provoquées p a r

le dessiii des plaques

— exam en et nettoyage aisés des plaques côtés p rimaire et

secondaire t les plaques sont serrées dans u n b â t i p a r s i x

tiges filetées*

le problème d'étanchéité, lié aux joints entre

p l a q u e s i paraît parfaitement résolu. L a mise en communication

du primaire ot du secondaire, p a r rupture d'un joint, est

rendue impossible p a r la présence de 2 joints entre les

circuits avec mise à l ’air libre de l'espace entre ces

2 joint«*

4 . 2 , 1 , 3 , Pomps s ^ e £ ¿ i r c u i t p r i m a i r e

Les groupes moto-pompes utilisés sont les azxoiene

groupes Siloé 15 MW* Caractéristiques nominales d ’une pompe :

560 m /h— 19 m de CE. Moment d'inertie du volant t 20 m kg.

D e u x pompes sont installées t une en servia©, l ’autre on

secours.

Toutes les conduites d ’origine (fi 150) extérieures

a u bloc piscine ont été remplacées p a r des conduites do ¡6 250

et 0 3 0 0 p o u r réduire les pertes do charge ot ainsi r é a lis er

lo débit ôésiré tout en évitant l a mise en dépres si on du bac

do désactivation.

U n n o u v e a u diffuseur de re tour pisoine a é t 6

installé p o u r obtenir u n e faible vitesse de sortio do l ’eau

(25 om/s) ot évite r ainsi de p e r t ur be r l a oouoho chaude»

L ’utilisa ti on d'une seule pompe entraîne, en oas do

r u pt ur e de l ’aooouploment entre pompe ot volant d ’inertie,

l ’annul a t i o n brutale d u débit, donc le passage on convection

naturelle, quelques secondes après l a chute dos barres

p r o v oq ué e p a r les sécurités “manque débit". L a pu is sance

résiduelle, b i e n q u ’inférieure à 1 M i ost encore rela ti ve me nt

importante. C'est p o u r retarder d'une minute le ronvorsomont

d u débit dans le coeur qu'un bao t a mp on do 1 m a été placé

à l a sortio du bao do désactivation. Ce bac est à une altitude

comprise entre H (hauteur du p l a n d ’eau de l a piscine) oto

H o «A H (AH t pertes de charge entre le coeur et le b ac d©

d é s a o t i v a t i o n ) , L e b ac ta mpon est donc vide en marche normale

et se remplit à l ’arrôt du primaire, jouant ainsi le rôle do

nv o l a n t d 1e a u”.

4,2,1,4. R ésultat!»

Les essais ot les mesures thermiques ont été

effectués dans les conditions nominales do fonctionnement du

r é a c t e u r on déplaçant notamment dans le ooeur trois éléments

combustibles équipés c h a cun do p lu sieurs thermocouples qui

d o nn en t les températures de gaine des p l a qu es combustibles

(un élément de chaque type t élément standard, do contrôle

et d'irradiation),

XI on ost ainsi à chaque augmentation do puissance,

que ce soit à M EL US I N E ou à SILOE,

Les mesures ont confirmé l a po si t i o n dos points

chauds ot la validité des calculs et elles ont montré que la

m a r g e p a r rapport à 1* ébullition locale (128°C) était

importante.

Des ossais o n surpuissanoe (10,5 MVi) ont permis de

s ’assurer que les conditions de sécurité étaient satisfaites.

L ’augmentation de puissance de MELUSINE, son

adaptati on à 1 ' environnement expérimental, ses possibilités

de transformations n o n encore épuisées, illustrent les

princi pal es qualités des réacteurs de recherche d u type

pis ci ne î simplicité, souplesse et polyvalence.

4*3* Possibilités expérimentales de MELUSINS

Melusine est u n réac teu r dont l a po lyvalence est

b i e n marquée. Ses caractéristiques de flux, ses équipements

tels que canaux, cuve à eau lourde, tubes pneumatiques,

boucl es froides, sa souplesse do fonctionnement, lo rendait

très attrayant pour l a recherche fondamentale et p o u r la

recherche appliquée. Depuis s a construction, d'importants

aménagements ont été réalisés pour lo rendre aussi complet

que p ossible p a r rapport aux besoins des utilisateurs*

Parmi le a emplacements utilisables simultanément

a uto ur d u coeur, les créneaux fournissent à l 'i ntérieur de

mandrins d 'aluminium représentant les expériences des flux13 2

thermiques atteignant 6 à 7*10 n / c m ,s et des flux rapides

entre 2 et 4 , 3 * 1 0 ^ n / c m 2 s (E > 1 MeV).

E n p remière rangée autour d u coeur, le flux14 / 2

thermique varie entre 3 et 1.10 n / o m , s, tandis que le

flux gamma donne des échauffements compris éntre 0,6 et

1,6 w /g dans le g r a p h i t e •

Des écrans en plomb de 2 cm d'épaisseur ont été

insérés entre le c oeur et oertains dispositifs expérimentaux

p o u r réduire 1 1échauffement gamma d ' u n facteur 2 environ»

4*3*1• Canaux

Les canaux radiaux comportent depuis 1965 u n e

"chaussette1* amovible* Nous donnons ci-après leur utili sat io n

pr in ci p a l e on 1970, uti li s a t i o n qui va ri e évidemment avec

les programmes de r e c h e r c h e en oours.

-* Canal ra dial n° 1 i faisceau sorti équipable d'un, dispositif

de "ähauffage” des neutrons pour le diffractomètre à

neutrons polarisés (fig. 8).

« Canal r ad ia l n° 3 î neutrographie industrielle à énergies

à neutrons variableô (êpithermiques, thermiques et froids)

réservée à l'examen n o n destructif d'objets n o n iri*adiég

pou r lesquels l a radiographie et l a gammagraphie sont

inadaptées (fig* 9). le faisceau de neutrons froids, un ique

en so n genre, ouvre de nouvelles perspectives dans l ’examen

des aciers et la recherche des matériaux sous forme de traces.

— Canal r a dia l n° 2 : faisceau sorti potar l ’expérimentation

d'une méthode visant à améliorer la résolution dos mesures

p a r temps de vol.

- Canal teoigenticl s u n diffuseur (béryllium) placé dans le

canal au n i v e a u du coeur permet d'augmenter l'intensité du

flux do neutrons thermiques, les flux do neutrons rapides

et gamma restant relativement faibles»

L a sortie T2 â u canal tangentiel est équipée d'un

collimateur très efficace* Le fai sc ea u est utilisé pour deux

études préliminaires concernant le r é a ct eu r à haut flux franco*»

allemand (mise au point de mono chromât eurs ot étude de l'effet

M O S S B A U E R ) .

L a soirfci© sert à l'étude dos produits de fission

à vie courte au m o y e n d ' u n fouir associé à u n spootromètre de

masse.

4.3»2* Cuve à e au lourde

U n e cuve à eau lourde est plaoée à 8 cm d ’u n e face

d u coeur« Lo flux dans l a cuve est riohe en neutrons

thermiques et p résente u n b o n rapport ^CG ^ ^ P P 03^

variant de 86 à 250 suivant les emplacements d ’irradiation)«

U n tube pneumatique et une boucle à n é o n liquide

sont associés à cette cuve. U n faisceau, caractérisé p a r des rapporte

<p_j_ /<p et élevés pe ut Ôtre sorti verticalement de la

cuve«

4 « 3 »3 » T ubos p n e u m a tiques

Srois -tubes pneumatiques sont roliés au Laboratoire

do Moyenne Activité (LMA) o ù sont implantés plusieurs

laboratoires d u CEN— G *

~ tube à aocès direct dans l a ouve à eau lourde, utilisé d'une

fa ço n très intensive p o u r les analyses p a r activat io n p a r

neut ro ns thermiques

« tube à accès direct contre le coeur dans l ’eau légère,

utilisé do façon classique

Oes doux tubes sont pourvus d ' u n système pneumatique de

ro t a t i o n de óa r t o u ches afi n d ’homogénéiser les doses reçues

dans tout le volume des é c ha nti ll on s•

*• tube de transfert connecté à u n e machine de chargement

automatique on p a rt ie immergée dans le compartiment médian»

Ce tube permet d 1 expédier a u Laboratoire de Moye nn e

Ao ti vi té après u n certain temps de désactivation on p i s o i n e 9 les échantillons irradiés dans dos containers type r a di o­

éléments } il supprime ainsi toutes les contraintes liées

aux transports de ces radioéléments p a r chateaux do plomb»

4.3.4» Boucles — Dispositifs d 'irradiation -»

Neutrographie

— Sans qu'il soit besoin d 'arrôtor 1© réaotour, los bouclas

froides sont mises en plaoe ou retirées do leu r p o s i t i o n

d ' i rr ad ia ti on et les défournements se font de façon

semi-automatique et sans réchauffement des échantillons»

A u nombre do trois, elles sont u til isées p o u r des études

fondamentales en p hysique chimie, eto...<1 A O

. 1 boucle à H é l i u m liquide (5°K - cp_ »= 2*10 n / c m s)

(E > 1 MeV)

• 10 o. 1 b o u d e à n é o n liquide (28°K — = 9,5.10 n / c m s)

^th ~ 6» 1 0 ^ n / c m 2 s)

1 9 O» 1 boucl e à azote liquide (78°K - a 9>5»10 n / o m s)

( 9 ^ = 6 , 1 0 ^ n / c m 2 s)

131

(Rappelons qu'une bouol o à, n é o n liquidé est également en

service dans la cuve à eau lourde).

- P o u r les beso in s de l a Physique d u Solide, dos fours

spéciaux sont en service (T° de 30 à 50Ç°C), équipés de

solénoSdes créant des champs magnétiques de l'ordre de

5000 oersteds.

<*« Enfin, p o u r les irradiations technologiques, les lispositifs

d ’irradiation classiques (fours CHOUCA, HEBE, HP, etc..)

ocoupont les différentes positions autour ou dans lo coeur

(voir description et performances des dispositifs plus l oi n

dans ce rapport)

— Comme à SILOE, u n a ppareil de noutrographio de mSmo

caractéristiques, immergé dans lo compartiment milieu,

est principalement réservé à l ’examen dos dispositifs

d*irradiation,

4 » 3 « 5* Cell u les^chaudes

Deu x cellules chaudes ont été installées p o u r

effectuer certains trava ux (examen, découpage, conditionnement,

réparation) sur les échantillons, les dispositifs d *irradiation

et le combustible. L a plus importante cellule est blindée

p o u r des activités (y de 1 MeV) de 10 kCuries. L a p l u s

p et it e est principalement utilisée p o u r les radioéléments,

q u ’elle reçoit directement à p a r t i r du dessus piscine.

4.4. Conclusion

Mélusine, grâce à l a di sposition p ar ti culière de

ses équipements a ut ou r d u coeur, offre des neutrons de

caractéristiques, spectres et flux, adaptées à une large -,

gamme d ’oxpéri. enees à l a fois de recherche fondamentale et de

recherche appliquée.

Tout en conservant ces qualités, des flu x d e plias

en pl us intenses sont proposés aux utilisateurs p a r

au gm entation successive de l a puissance de fonctionnemont,

comme cel a v i e n t d ’être encore le cas en 1970~*1971»

Ceoi on fait u n réaoteur particulièrement

intéressant et dont les possibilités sont complémentaires do

oelles d© SBiloé, oes deux réaoteurs constituant ainsi toa

ensemble homogène et o o m p l e t .*

5. SILOETTE

Ce rapide tableau des réacteurs do Grenoble serait

inoomplet sans u n mot sur le réacteur Siloette.

N e dépassant pas 100 kW, les servioes rendus ot

offerts p a r oe réa ct eu r maquette sont irremplaçables» Toutes

les études o o n o e m a n t les coeurs de Siloé et d© Hélusino

sont offeotuées dans Siloette, ce qui évi£e leur im mo bilisation

périod iqu e i études des configurations, établissement dos

cartes dô flux, étude de l ’effet des dispositifs d'irradi at ion

(réactivité, dépression de flux, oto..).

Dans le oadre môme du projet de ré acteur à h a u t flux frano o— a l l em an d, Siloette a permis d 1étudier l'intérôt d'une

source ohaude et de différentes sour oes froides p o u r la

modif ic ati on des spectres de neutrons. Les doux canaux Offrent

des possibilités intéressantes : expérienoos sur conduits do

neutrons, études de choppers de différents types, n e u t r o graphies

d * o b jets n o n irradiés, etc., (fig. 10),

C * est également u n excellent outil mis à l a

d isp os it io n des u.-fcxlxsateurs pour los éirudeS d * app aareilZLages

n o u ve au x | appareils de mesure de flux, de radio-protection,

chaînes d ’éleotr o n i q u e , e t c . .

L a d o u b l e v o o a t i o n de Siloette p eut ainsi se

r é s u m e r de l a f a ç o n suivante : support des réactours

pr in o i p a u x d' irradiation et collaboration aveo les ut ilis at eure

intéressés p a r u n réaoteur aux flux relativement p e u intenses

ma is au ré gime de fonctionnement très souple.

6. DISPOSITIFS D »IRRADIATION P O U R L À R E CH ERCHE TECHNOLOGIQUE

Outre les dispositifs simples spécialement adaptée

à l a p r o d u c t i o n do n o m br eu x radioéléments et n ot amment du

133

cobalt, du nitrure de magnésium, ote, les dispositifs

d * i r ra dia ti on proprement dits occupent les 3 faces libres du

coour, L eu r nombr e m o y e n autour d'un coeur est d ’une

quarantaine environ (fig* 1l)fc- û ¿

Avant mise en pile, les dispositifs peuvent être

essayés dans des installations hors piscine ou en piscine

a f i n de vér if ier leurs caractéristiques de fonctionnement

(débit de refroidissement, caractéristiques de chauffage

p a r e x e m p l e ) «

Des postes do travail sont aménagés autour des

p i s c i n e s de façon à permettre le stockage et l'intervention

sur les dispositifs irradiés ou à effectuer les défournements

des porte-échantillons en fin d'irradiation. Ces défournements

se font alors à l'aide de hottes ou de chateaux de plomb

spécialement adaptés pou r le transfert en cellule chaude, pour

le transfert vers les laboratoires chauds sur le site do

GRENOBLE, on M lA NC E ou à l'étr an ge r (fig. 12).

Nous décrivons ci-dessous les dispositifs

pol yv a l e n t s les plus fréquemment utilisés ot qui sont également

u t i l i s é s dans les autres réacteurs français et étrangers.

D'autres dispositifs expérimentaux originaux

seraient à citer, tels que : boucles à liquides organiques,

bo u c l e s à gaz, boucl es à sodium, fours à lame de gaz,

spectromètre do résonance magnétique, fours spécialisés de

Phy si qu e du Solide avec ou sans champ magnétique, boucle à

n à o n et hydrogène liquidos, bouclo à h é l i u m liquide, etc,

ot qui sont légalement utilisés dans ces réacteurs.

6.1. Recherche effectuée avec les fourB CHOUCA

Les fours CHOUCA sont utilisés pour l'irradiation

de ma té ri au x de structure ou d ’échantillons combustibles à dos ,,

températures allant do 150 4 1000°C. L a température ost

r é g ul ée p a r 6 éléments chauffantb indépendants à — 2°C et

1 ’écart relatif maxi mal do température sur la longueur utile

d u f o u r ost de l ’ordre do 1 fo,

L a longueur utile est de 400 m m ot pout atteindre

600 m m si l ‘uniformité de température est moins impérative » Le

diamètre utile est de 2 5 cm» L a p r e ssi on intérieur*© est réglable

de 0 à 8 0 bars* Le contrôle du chauffage est fait p a r 12

thermocouples et 1*implantation de 18 thermocouples est prévue

p o u r les échantillons *

Le tableau oi-dessous récapitule quelques expériences

réalisées aveo ces dispositifs dont 200 exemplaires ont été

fabriqués à ce jour et utilisés dans los réacteurs français

et é t r a n g e r s •

M a t é r i a u irradié But de l ’expérienceNatu re p o r t e ­é chant ilions

A c i e r inox

Cylindre de graphite

Fils d'acier inox

Be, Mg, Zr, Fe, A l

Eprouvettes de résilience

Métrologie avant et après irradiation (effet Wigner)

Eprouvettes de traction

Eprouvettes de traotion

P ort e - éc hantil— Ions simples

Combustibles dispersés

2Graphite C0

Graphite bore

Silice sous CO2

Comportement sous flux rétention des gas do fission

E tude de oorrosion r ad io — lytique

Etude de l a fabrication do l ’héli u m créé

Comportement eous f l u x

Capsules à gaa

A c i e r inox Zr, Cu, Be

A c i e r boré

Carbure d ’uran i u m

P l u t o n i u m m ag nésium

Pastilles d 1U r a ni um

Mesures de résilience résistance à l a traotion conductibilité thermique

Examens métallurgiques après irradiation

Etude de d if fusion du P l u t on iu m ûans le mag né si um

Etude des déformations sous contrainte et sous flux

Capsules N a K

Graphite U r a n i u m

A c i e r

Métrologie sous flux

Effet Wigner-oroissanoe fluage en traction

Dispositifs de p a l p a g o à cavité réson na nte

De prin ci pe analogue a ux fours CHOUCA, ils ont u n

diamètre u tile plus grand (53 m m ou 49 m m suivant qu * ils sont

u t i l i s é s avec ou sans éc ran thermique)* Ile permettent ainsi

1 * irradi ati on de quantités plus importantes de mat ér ia ux et

1 1échauffenent nucléaire permet d'obtenir des températures

élevées, de 350°C à 1400°C. Le chauffage électrique d ’appoint

est obtenu p a r 4 éléments chauffants. L a longueur u til e est de

4 00 mm* L a p r e s si on peut v a ri er de 0 à 60 bars.

Ces fours servent à l ’irradiation de ma tériaux de

structure l graphite à haute t e m p é r a t u r e * gluoine (BeO) à

1300°C, maté riaux réfractaires divers, stéatite à 250 et 450°C,

etc.., et de petits équipements S capteurs de pression, câbles

coaxiaux, relais électriques, microrupteurs, chambres à

fission, etc«*

6.3* Rec her ch e effeotufle avec les fours CYELfiHO

Ces dispositifs sont destinés à l ’irradiation de

combustibles à haute puissance linéaire (1700 W / c m ) • Ils

sont é quipés d ' u n dispositif de mesure de l a p ui ssance nucléaire

d é ga gé e p a r 1 1 échantillon* Les irradiations se font sous des

conditions do puissance, do température et de pr es s i o n

équivalentes à celles des réacteurs de puissance« Le«

échantillons sont placés dans d u N a K ou du so dium afin de

b i e n assu re r les échanges thermiques et la ba rrière thermique

p r i n c i p a l e est constituée p a r une é pa isseur d'inox fritté

et une lame do gaz. L a p r e s s i o n p e u t être choisie entre 1 et

60 bars*

Ces dispositifs permettent de régler la température

do gaine de l ’échantillon combustible p o u r une puissance

dégagée imposée, ou d' évacuer l a puissance maximale possible

p o u r une température de gaine imposée.

Exemplasd' ut ili sa ti on de ces dispositifs dont 80

exemplaires ont été réalisés } v o i r tableau on page suivante.

N o m expér. EchantillonsPuissancelinéaire

Température P ression But expérience

CÏRANO 1 à 7 U 02 gainé ino x ou Zr Cvj/

350 à 500 w/cm

Gaine 6 0 0°C Coeur 2000 °C

60 bMesure d u taux de libéra tio n dos gaz do fission» M esure de l'intégrale do conductibilité thermiqua

ICARE U C gainé inox

I7 OO w/ om Gaine 650°C Co e u r 1200°C

2 0 b

CYPRES et CIRCE

U 02 gainé Zr Cu

4 5O W/cm Gaine 600°C 60 b Mesure da p r e ss io n interne du combustible (gaz de fission) et mé tr o l o g i e sous flux (mesure de v a r i a t i o n de longueur) à l'aide d ' u n c a p t e u r de déplacement à cavité résonnante

N AD IA et V ENCA

Combustibles à évents

50 0 à 600 W/cm

G-aine 300° C à 600°C

0 , 8 b Etude d u déchargement dos gaz de fis si on dejas tua b a i n de sodium

P A C U02 gainé Zr 9 5O W/cm G-aine 330 °C 40 b Compo rt emont sous flux d ' u n combustible avec f u s i o n à 0 0 b u t

POL¥CARPE U 02 gainé ino x

600 W/cm G-aine 650°C 2 b 6 crayons irradiés simultanément étude d u comportement do différents types de combustibles

CÜRSUM U 02 ot U02-Pu02 Gaine inox o u Zr

4 0 0 w/cm à 600 W/c m

G-aine 450- 500°C

2 b Mesure do dé formation directement s u r le combustible» A v e c o u sans contrainte ot mesure conductibilité thermique

Ces fours, chauffés p ar haut© fréquence, permettent

d© réa li se r ot d© r é g u l e r sous flu x nucléaire de -très hautes

températures atteignant 2000°C. U n inducteur alimenté on

courant alternatif (50 à 500 kHz) engendro dans u n blindage

dos courants qui. induisent à leur t o u r dans u n suscopteur qui

pe ut Ôtro 1 1 échantillon lui-môme ou une partie d u p o r to -

échantillons* Lo diamètre des porto-échantillons varie do 16

à 21 m m ; l a puissance H P atteint 200 W p a r centimètre do

h a u t e u r ot l a puissance nucléaire 400 W / c m ; l a r ég ul ati on de4*

température peut 8tre faite à 5°C ou — 0|5°C selon les

spé c i f i c a t i o n s •

Exemples d ’expériences réalisées 1

a) Etudes do d if fusion des gaz de f i s s i o n hors des combustible»,

les gaz de f i s s i o n étant entraînés p a r u n courant d * h él ium et

analysés hors pi le (combustibles U02, combustibles TJC,

combustibles dispersés)*

b) Essais d 1endurance et d© vieillissement acoéléré de s

graphite, combustibles U02 ou TJC, combustibles spéciaux

p o u r conversion t h e r m o- io ni que •

c) Etudo et réa li s a t i o n d'une source de neutrons chauds p ar

chauffage H P d ’u n cylindre de graphite à 2300°E*

6*5* Re ch erche effectuée avec les cavités résonnantes

Le comportement dimensionnel des ma té riaux de

structure ou des combustibles soumis a ux rayonnements résulte

d© l ’action de divers phénomènes tels que t le fluage, la

croissance* le gonflement (des combustibles p a r e x , ), l a

de nsification (silice vitreuse p a r ex.), les cyolages,

1*effet Wigner, etc**

Dans beaucoup de cas, ces mesures sont faites

après irradiation avec tous les désavantages que cela comporte

(interprétation d u phénomène plus délioate, nécessité do

m u l t i p l i o r les essais p o u r obtenir plusieurs p o i n t s )*

L a miso a u point d'un appareillage qui permet de

mes ur er directement sous flu x les variations dimensionnoU.es

dos é c h a n t i l l o n s , élimine les inconvénients cités plus haut.

Les déformations de 1 1 échantillons pendant l'irrad ia ti on provoquent

u n changement do volume d'une cavité résonnante on hyper-

fréquonoe, oavité associée à l'échantillon p a r l'intermédiaire

d ' u n piston. L a mesure consiste alors à mesurar la v a r i a t i o n

de l a fréquence do résonanoc do la oavité, v a r i a ti on qui ost

direotomont liée à l a déformation do l ’échantillon, - Le

prooédé de mesura a les caractéristiques suivantes :

— gamme do mesure * 0 à 1 m m et 0 à 20 m m

— p o u v o i r de ré so lu ti on : < 0,01 % do la gamme do mesure

— p r é c i s i o n t < 0,05 $ de l a gamme do mesure

— sensibilité j < 1 mi cron

6,6, R oc horche effectuée avoo 1 ^ f< p our études

do fluago

Les cellules de fluago sont utilisées p o u r des

irradiations d 'éprouvettes maintenues en tr ac ti on ou oc mprossion

avec mesure on oontinu des paramètres suivants : température,

oontrainto, flux noutronique, variations dimonsiormelles et

vite ss e do fluage.

Ces cellules comportent u n vér in de traction o u de

co mpression à soufflet métallique déformablo, dos thermocouples,

dos détecteurs de flux noutronique, u n captour do déplacement

à oavité résonnante* L a compensation des déformations

élastiques ot dos effets do d ilatation dus aux températures

est a s s u r é e « L 1h omo gé né is at ion dos températures peu t 8tro

ré alisée p a r remplissage de m étal liquide (NaZ) .

Ces oollulos snt permis jusqu*à présent des études

do fluago sur : aoier inoxydable, graphite, tube en p ression

i n t e r n o .

Le circuit sodium à thermopompe est destiné à

ir radier dos éléments combustibles dans des conditions

theriviiques analogues à celles dos réacteurs rapides do

p u i s g anco*

L ’utilisa ti on d' un t el dispositif en r é a o te ur do

re c h e r c h e doit Ôtre compatible avec les accès exp é mont aux

disponibles ot la sûreté générale des installations (pile

et expérimentations voisines).

De ux éléments sont importants t la circulation du

sodium est a s s u r é e p a r des éléments spéciaux thermopompes

qui présentant une grande sécurité intrinsèque de fonctionnement

car ils ne nécessitent aucune alimentation électrique ou

l i a i s o n hors pile et p a r ailleurs, lo sodium est en tout point

séparé du mili eu extérieur (notamment de 1' eau dans u n r é a c t e u r

type picicino o u tank) p a r u ne double paroi d ’acier ino ydablo.

Caractéristiques^techniques

- puissance totale évacuable

— s e ct io n d ’essai

- température du s o d i u m

— vitesse du s o d ium

7. COI-LABORATIOU AVE C D ’AUTRES PAYS

L'u ti li sa tio n intensive des réacteurs d u C.E.A.,

tant en recherche fondamentale q u ’on recherche appliquée»

illustre b i e n l'intérôt présenté p a r les réaotcurs piscines

po ur lo développement des techniques nucléaires*

Depuis do n ombreuses années, ooci fait l ’objet d'une

co ll ab oration aveo plusieurs pays étrangers dont dos pays

en voie do développement, coopération qui s ’exerce dans

pl us ieurs domaines.

60 k W

0 20 mm, longueur 1400 mm

300 à 700°C

2 à 5 m/s

a) A c o u o i l do stagiaires (ingénieurs, universitaires) soit

dans u n but préois do formation aux problèmes do construction

et d 1 exploitation dos réacteurs et dos dispositifs

expérimentaux, soit dans des perspectives plus largos do

pa rt ic ip ati on directe au travail dos équipos do réacteurs

ou do p ré par at io n de thèses universitaires. C'est ainsi

que oortaines améliorations techniques originales ou

certaines mé thodes de nosuro ou de calcul ont été

développées o u r é a liséos p a r dos collaborateurs étrangers p o u r

lo mei lle ur profit mutuel*

b) Rô le do conseiller tochnique principalement dans l a

construction, l'exploitation et l'amélioration dos

po rf ormancos dos réacteurs pisoino. Disposant d'une large

expérience dans oo domaine, ot notamment dans les réalisations

d' au gmentation de puissance, nous participons aux projets

correspondant do plusieurs pays étrangers qui veulent

adaptor, transformer ou améliorer leur réacteur po ur on

faire 3.‘outil de base de leur programme de développement

des techniques nucléaires.

U n e coopération étroite et régulière est particulièrement

établie avoo plusieurs pays d'Amérique du Sud»

û) U n exemple intéressant de coopération pout Ötro citó I c'est celui du jomolag© entrepris dopuis p lu sieurs années

avec u n contre Su d-Américain po ur réaliser plusieurs

objectifs : améliorer Igb performances d u r éacteur de oe

contre, l'utiliser pour dém ar rer u n programmo do ph ysique

du solide, constituer une é q u i p o mixte do physiciens du

solido travaillant da no une première phase sous l a d i r e c t i o n

eciontifiquo ot avec d u maté ri el dos laboratoires

correspondants do G-ronoblo, ot devant évoluer, clans une

deuxième phase, vers l a fo rmation d'un laboratoire autonome

et préparé à défini«? ses propres programmes ot à mottro

on oouvro sos propres moyens de roohorcho*

d) Enfin, les actions do collaboration quo nous avons menées

n o u s conduisent à précon is er u n type do réa ct eu r spécialement

adapté aux pays en voie de développement ot p o u r l e q u e l nous

avons établi u n p rojet de base.

141

Il s'ag it d ' u n ré acteur p i sci ne do 100 k W a u d é p a r t ,

comportant 6 ca naux et oonçu pou r p o u vo ir fonctionner j u s q u’à

5 M W dans une étape ultérieure. L ’augmentation de puissance

de 100 k W à 5 M W no r e m e t pas o n cause ni le bloo piscine,

n i 1 ' architecture générale des installations ot peut

i n t e r v e n i r facilement et aux moindres frais, uniquement en

remplaçant oertains matériels et en complétant quelques

installations «

Dans une première phase, la puissance de 100 k W

est h abi tuellement considérée comme suffisante p o u r le

la nc ement d ’u n programme nucléaire de formation do spécialistes,

do p r o d u c t i o n de radio-isotopes à vie courte, de recherche do

p h y s i q u e de réacteurs et d 1 applications d i v e r s e s •

Dans u n e deuxième phase, une puissance de 1 à 5 M W

pe rm et l ’extension du programme ini tia l à dos rocherohes et

des cssaiB d ’irradiation s ’intégrant dans u n pr og ramme général

plus avancé ot plus orienté de développement dos techniques

nucléaires* Los performances à 5 M W sont alors voisines de

cellos de MELUSINE et TRITON, ce qui offre dono les mômes

po ss i b i l i t é s d ’utilisation.

L a troisième phase qui consiste à réa li se r u n

programme de re che rc he ot d ’irradiations technologiques très

évolué néoossito la mise o n service de r é a c t e u r s de recherche

et d 'essais à hautes p e r f o r m a n c e s , type SIL0E ou OSIRIS ;

elle no peut guère ôtre envisagée dans des pays no disposant

p a s d ’une infrastructure scientifique et teohnique

suffisamment développée et préparée.

Au tr ement dit, l ’économie des deux p re mières phases

ost difficile à concevoir avant de p a s s e r à l a dernière phase.

Dans le cadre de possibilités financières modestes,

ce p r o je t fournit u n môme outil do trava il po ur los doux

p re mi èr es phases, ce qui permet donc do faire avanoor plus

r api de me nt le programme de développement sans Ôtro retardé

p a r 1 1 a ttribution de crédits plus importants pour le passage

do l a première à l a deuxième phase.

E n of fot, l a conception initial© d u r é a c t e u r ost

tollo quo 1* augmentation do puissance do 100 k W à 5 M W peut

se faire a u x moindres frais tandis que la plu s- v a l u e payée

au départ p o u r b én éf i c i e r ensuito do oot important avantage ost

p o u élevée, contrairement aux a p p a re no o s *

XI v a de soi que le réacteur peut également Ôtr©

oonstruit p o u r une puissanoe initiale comprise entro 0 et 5 MW,

Sa conception, qui résulte do l'expérience acquise

depuis do longues années, en fait u n réacteur simple, oonpaot

et spécialement pensé p o u r constituer l'élément do base a u

début d*un programme nucléaire do recherohe fondamentale ot do

roohoroho appliquée. ,

8 * CONCLUSIONS GENERALES

Le groupement des 3 réacteurs Mélusine, Siloé,

Siloette sur u n môme sito mot à la disposition des

exp éri mont at ours une gamme do possibilités capable do satisfaire

sin on l a totalité, du moins une bonne partió de lours besoins

d ’irradiation»

Cependant, et indépendamment de l a reoherch©

continuelle de l'amélioration des performances des réaoteurs,

il faut aussi p o u vo ir orienter 1 * expérimentateur dans la

me il leure façon d ’utiliser les réacteurs mis à le ur disposition»

Ceci peu t se faire en lo ur proposant u no gamme de services

complémentaires et spécialisé» qui ont tua double b u t *

- les aider dans la ré alisation de dispositifs ot d o mesures

spécifiques aux réacteurs

- les libérer ainsi de préoccupations techniques avec lesquelles

ils no sont pas toujours familiarisés.

C 'ost dans oos perspectives que lo Sorvioo dos

Pilos d u CEN-G- a mis l ' accent sur le développement }

a) d'équipements et do dispositifs d' irradiation adaptes aux

réaoteurs ot à la mageure p artie des besoins usuols dea

expórinentatGU2vs

b) de méthodes de dosimétrie neutrons et y adaptées aux

dispositifs

o) d'examens tels q u e la n e u t r o graphie

d) de la E3urvoillan.ee et de la régulation des oxpérienoes

par oaloulateur

e) do d é f o u m o m e n t s d'échantillons, oto,.«

Et finalement, nous pouvons dire que les nombreux

p r o gr ès réalisés on matière d'irradiations depuis plusieurs

années sont le fruit de oe courant d 1 éohange incessant entre

les Qzsp é riment at ours et les Piles.

AT THE GRENOBLE NUCLEAR RESEARCH CENTRE

PRESENTED AT THE IAEA STUDY GROUP

MEETING ON RESEARCH REACTOR UTILIZATION

BANDUNG, INDONESIA, 2-6 AUGUST 1971

*>y

P. MerchieChief of the Reactor Operation Section Nuclear Research Centre Grenoble, Prance

ABSTRACT

Almost all irradiation programmes in Prance are carried

out in swimmingpool-type reactors.

After-summarizing those characteristics of swimming­

pool reactors which are advantageous in the development of

nuclear techniques, we give the general characteristics of

the SILOE, MELUSINE and SILOETTE reactors at the Grenoble

Nuclear Research Centre.

We mention the successive improvements made in these

reactors and, in particular, in their power.

Lastly, we elaborate on the experimental possibilities

which they offer and on the principal types of irradiation

devices which have been developed over the last few years,

together with the uses to which they have been put.

In conclusion, we describe our collaboration with other

countries in the engineering and use of research reactors.

1. INTRODUCTION

Several dozens of swimming-pool-type research reactors are now

in operation in the world. Although not designed originally for any

specific purpose, "but being suitable essentially for university work

and basic research, these reactors have since been found by experience

to be highly useful also for applied Or technological research. Thus,a great

part <f the irradiation programmes in Prance are carried out in swimming­

pool reactors, for example :

- TRITON and MELUSINE, built in 1958 with a power of 1 MSf,

are now operating at 6 and 8 MSf respectively.

- SILOE, commissioned in 1963 with 15 MW and modified in 196?

to operate at 30 MW, is a great improvement on the swimming­

pool type of the 1950-1960 period and is used for materials

testing.

- la OSIRIS (1965), it has been possible to attain a power

of 70 MW by changing the direction of circulation of the cooling

water for the core and by carrying out some structural changes in

the pool.

These reactors, including also PEGASE, which is specialized in

irradiating large loops, form a complete range of facilities for

materials testing under irradiation and provide a source of wide

experience in the use of research reactors.

The experience gained shows that, in the case of countries

which are in the process of defining their nuclear programmes or

whose programmes, having already been prepared, are entering the

stage of implementation, the possession of a swimming-pool reactor is

of great interest and acts as a sound basis for developing nuclear

techniques.

After dealing briefly with the general and special character­

istics of these reactors, as they appear in practice, we shall give

a more specific description of the Grenoble reactors. We shall then

describe the use of these reactors in basic, applied, and. technological

research, for which purpose highly specialized, irradiation devices

have been designed.

We shall also describe the collaboration existing over a number

of years with several foreign countries, especially those whose nuclear

programmes are being prepared and implemented in regard to the operation,

development and use of reactors.

We shall give the general lines on which this collaboration is

carried out together with the reasons that led us to implement a

reactor project intended specially for promoting nuclear technology

in developing countries.

2. GENERAL PROPERTIES OF SWIMMING-POOL REACTORS

The remarkable success of swimming-pool research reactors is

due to the combination of properties which they possess and which

have been demonstrated or confirmed in practice :

- operational simplicity and safety;

- flexibility and multiplicity of use (basic, applied and

technological research);

- high performance for several years;

- low construction and operating costs and hence low cost of

experiments and irradiation runs.

2.1. Simplicity and Safety

In comparison with heavy-water or tank-type research reactors,

swimming-pool reactors are simpler, having no special facilities

(which are generally sources of trouble) such as those for :

- loading and unloading fuel elements and experimental devices;

- control rod devices;

- in the case of heavy water facilities : purification circuits

and gas blanket at the upper level of the vessel, full leak-

proofing of circuits and tritium hazards during maintenance

and dismantling operations, etc.;

- pressurisation of cooling circuits, auxiliary circuits such as ’

cooling at shut-down, cooling of shielding or of tbs jpooi;

The high degree of availabilityand "the operational safety .of

swiannicg-pool-reactors is due largely to the ^sigiplicity -öf -theit .

design and tb the existénçe of well-^known Intrinsic s a f e t y factors, *•

The numerous development and safety stu&iee-carried Out in

Prance and particularly at Grenoble (neutron, thermal., mechaftical and hydrau­

lic studies ) and at Cadarache with the -C4BRÏ. reactor (power excursions,

cooling accidents, etc.) have yielded extensive information on this

type of reactor, enabling us to use it with .the maximum efficiency .

said safety.

This is reflected in the successive improvements and power

increases effected in the various French realtors.

2.2. Flexibility and multiplicity of use

The high structural flexibility of swimming-pool reactors makes

it possible to adapt them easily and rapidly to the diverse needs of

researchers. Structural modifications in the core or the addition

of supplementary equipment (heavy-water vessel, cold source, channels,

etc.) do not entail particular difficulties, since the good accessib­

ility permits dismantling and easy and rapid intervention even under

several meters of water. i

Furthermore, because of the readily available large volume around

the core, irradiation devices and devices for special experiments can

be operated simultaneously in large numbers and can be handled quite

easily.

The following advantages further accrue from the absence of a

vessel or tank around the core s

- the core configuration most suitable for the needs x>f the

researcher can be arranged;

- full visibility and high accessibility of the -core, vertically

or laterally, which explains the speed with which operation

can be carried out on the core (fuel elements can he loaded

in a few hours, for example);

- direct access to the channels right against the core and a

large number of peripheral high-flux sites without flux

attenuation or thin-down due to vessel walls separating the

core from the reflector;

- simplicity and facility of loading and unloading experimental rigs

and loops,some of such jobs are carried out when the reactor

is in operation. Without upsetting the irradiations under

way by shutting down the reactor, it is thus possible to remove

or introduce radioisotopes, rigs which are out of order

or whose irradiation begins or ends during an operating cycle,

cold loops often used for relatively short irradiations in

basic research, devices whose behaviour has to be studied step

by step by n e u t r o n ^ M ? S y (e.g. successive examination, during

irradiation, of the formation of a central hole in UO^-PuO^

fuel pins of the fast-reactor type, etc. )

This remarkable flexibility explains the multiple use of swimming­

pool reactors :

- Production of isotopes;

- Basic research with channels, cold loops, heavy-water vessel,

irradiation facilities, etc.

- Applied or technological research with experimental devices

developed specially for this purpose (furnaces and loops of

different types described hereafter).

2.3* High performance

The performance achieved likewise promotes multiple use.

These reactors were built originally for relatively low power, and were

intended chiefly for basic research. However, as result of continuous

improvement in their performance, they have achieved a remarkable

break-through in regard to high fluxes, and this accounts for their

success in applied and technological research. These reactors are perfectly

therefore /suited to expanding programmes and can gradually attain in­

creasing high powers, from a few MW to several tens of MS1/, in keeping

with the increasing needs of users.

Por example, the fluxes available in SILOE at 30 MW are :

- Thermal flux : 4*7 x 1 0 ^ n/cm sec;

- Past flux (E ^ 1 MeV) : 2.3 x 1 0 ^ n/cm? sec (the fast

flux can be used directly inside an irradiation rig

placed in the core).

2.4« Low construction and operating costs

Because of their simplicity and high degree of availability,

these reactors are not costly. A study carried out in 1969, with

prices converted to the level of 1 January 1969, showed that the

construction and operating costs of swimming-pool reactors were the

lowest.

Researchers using a swimming-pool reactor will therefore make

an appreciable saving on the cost of irradiation work and also on

the experimental devices which, for the same performance, are generally

simpler, and consequently cheaper, than those in any other tank - or

heavy—water—type reactor.

3. THE GRENOBLE REACTORS

The Grenoble Nuclear Research Centre operates two research and

materials testing reactors, MELUSINE and. SILOE, and a low-power reactor,

SILOETTE.

These three reactors, which are grouped in the same area, con­

stitute a homogeneous complex from the standpoint of performance and

the experimental possibilities offered to users. The latter, if they

so desire, can also benefit from a complex of specialized services

(supply of experimental devices, dosimetry, neutronfgraphy, computer, etc.),

which are made available so as to ensure the best conditions for carrying

out their irradiations.

3*1• SILOE ; General characteristics (Figures 1 and 2)

This 30-MW reactor differs from the earlier swimming-pool reactors

in that it contains a number of novel features, which later acted as

models for modernizing and transforming the older reactors in Prance

and abroad (core block, arrangement of the primary circuit, site of

the hot cell, equipment of experiméntal zones, fuel elements, control

rods e t c .) .

This reactor, which was initially designed for a power of 10 MW,

operated immediately at 15 MW. After four years operation it underwent

various improvements, which enabled its power to be increased to

30 Wf in 1967• in 1971» after further improvements to the primary

circuit arnd-.fuel elements, its power will-be raised t o -35 MW.

;r. -The oôrei having a ïectanguiar geometry with several indentations

is composed of fuel eietoettts öf différent types made by the Compagnie

d*Etudes et de Réalisations de Combustibles Atomiques (CERCA).They all have

plane plates* -

(a) Standard elements containing 23 plates loaded with 93$~enriched

uranium; -

(b) Control elements with 17 plates, in which the-*fork"- type

control rods move ; .

(c) Special irradiation elements, the fuel plates of which surround

one or two irradiation sites rich in fast fluxes and in which

experimental devices are placed (Pig. 3)«

The core is placed on a raised platform over a grid containing

100 positions, about 40 of which are occupied by fuel elements and

beryllium reflector elements located on one face of the core. The

remaining sites can be used for experimental devices. These devices

have been developed in such a manner that it is now usual to place

four devices in one irradiation site, providing a correspondingly

increased number of irradiation positions, especially high-flux

positions.

The neutron, hydraulic and thermal design of the core is

particularly thorough so as to reduce piessure losses from the core

and obtain the maximum power from the elements with a view to achieving

high fluxes (P _ 270/®tf/litre, j » 125 W/cm ). iH6stn / ni8>x

The core is cooled by circulating water from the pool between

the fuel element plates at a rate of approximately 2200 m^/h. On

leaving the core, this water passes through decay tanks and

is then cooled in conventional heat-exchangers before returning to the

pool, which has a stable temperature of about 30°C.

This primary flow ensures cooling of the cote and also of all

the experimental devices placed inside or around i t .

Operations at powers increasing from 15 MW to 30 and 35 MW

are possible because we have î

- reduced the power peak in the core;

- optimized the fuel elements;

- increased the primary flow from 1500 m^/h at 15 MW to

. 2200 m^/h at 35 MW. The limit in this case is imposed

by the increasing pressure loss from the core and by cavitation

hazards in the primary pumps. The increase of the primary

circuit heat exchange capacity to 35 MW is effected by adding

three heat-exchangers identical to the three used for 15 MW;

- increased the secondary flow from 1200 m^/h "to 1500 m^/h

(the cold water from the secondary circuit is drawn from the

ground water source at a temperature never exceeding 13°C;

it is later discharged into the nearby river).

The power peak is reduced by decreasing the width of the channels

in which the control rods move. A new control element and a new type

of control rod* called a "fork" rod, have been developed. The old

control rod 21 mm in thickness is replaced by two absorbent plates

3 mm thick placed laterally in the new fuel element. The distance

between the fuel plates situated on either side of the rod channel is

thus decreased appreciably, the direct result being a reduction of the

power peaks, which are always found on these fuel plates when the control

rod is withdrawn. The absorbent plates are made of Ag-In-Cd with

electrolytic nickel coating or else are fabricated of hafnium, which

needs no protection against corrosion (Figure 4)•

The "fork" rods also offer better efficiency than central rods.

With the same number of control rods it was thus possible to increase

the TJ charge of the fuel elements and of the core.

Optimization of the fuel elements consisted in reducing the

width of the cooling channel from 2.9 mm to 2.1 mm. Thus the new

standard element, with the same external geometry, contains 23 plates

instead of 18, and the new control element 17 plates instead of 12.

The increase in the number of plates and the simultaneous increase

in their U,. weight (from 10.9 g to 12.21 g or 1 4 »72 g, depending on

the elemeits or the position of the plates in the elements) mean that

the total U^ weight of the element is raised considerably (280 g to

330 g or 340 g, depending on the type of standard element, as compared

to 196 g in the 18-plate element).

It is thus possible to obtain very high burn-ups of the order

of 50$ - representing considerable saving in reactor operating cost.

The numerous thermal studies carried out on the out-of-pile

loops at Grenoble or with the CABRI reactor at Cadarache have enabled

us to obtain a better knowledge of t>e hazardous phenomena of local

boiling and reversal of flow (or flow reduction) leading to burn-out. This

has permitted a better definition of the safety margins in normal operation

and enabled us to write very precise codes and calculation programmes

giving the calorific fluxes of local boiling and reversal of flow

as functions of the different reactor parameters.

Lastly, it should be mentioned that the problem of activity at

the pool surface was solved at the construction stage by maintaining

at the top of the pool a hot-water layer 1-2 m thick with a temperature

a few degrees warmer than that of the remainder of the pool. This

layer stops ascending movements of the active hot water coming from

the neighbourhood of the core and thus reduces environmental activity

at the pool surface by a factor of 20. This méthod was applied sub­

sequently to the other swimming-pool reactors operated by the French

Atomic Energy Commission (MELUSINE, TRITON and. OSIRIS).

3.2. Experimental possibilities and specialized services offered

to the users of SILOE

SILOE was designed for use by researchers on the basis of the wide

experience obtained with MELUSINE, whose experimental load had reached the

saturation point before SILOE was built in 1961.

The areas offered to researchers are large and are distributed

over four levels. These experimental zones were equipped as elaborately

as possible in order to facilitate the setting up of experiments -

electric supply from the normal grid with stand-by arrangements and from

the total reliability network, safety connections with the reactor

control panel and. the computer, supply of various fluids, removal of

effluents of all kinds, etc.

At 30 MW, the fluxes available to researchers are :

(a) Fast fluxes (E 1 MeV) :

- six sites in the irradiation elements of the core with useful

fluxes between 1.8 and 2.3 x 10 n/cm. sec in the irradiation

devices.

- of the sites around the core, about 15 have fluxes between

2.5 x 1013 and 1.5 x 1 0 ^ n/cm? sec in water.

(b) Thermal fluxes t

Of the usable sites, about 20 have fluxes between 1.5 and14 / 2

4.7 x 10 n/cm. sec.

(c) Gamma heating :

The sites in the core or immediately round it have heating

between 2 and 12 W/g (in graphite).

3 .2.I. Channels

Two channels are installed for basic research. Situated behind

the core face reflected by beryllium, they make available three other

faces for vertical irradiation devices. The thermal fluxes available8 2

at the collimator exit (S « 70 x 30 mm) are 3.5 - 4 x 10 n/cm. sec.

The channels are now utilized for neutron diffraction (2 goniometers

per channel). One of the goniometers is adapted, to the study of the

crystal structure of organic substances, using single crystals, the

other three being intended for magnetic substances, using single

crystals or powders. In the latter case, the work is made much easier

by the use of multidetectors. The reactor hall and the ancillary

buildings make it possible to use the time-of-flight method,if needed.

3.2.2. Pneumatic irradiation facility

A pneumatic irradiation facility with two irradiation channels

(with adjustable distance from the core) is connected to two laboratories

located outside the reactor enclosure. The pneumatic facility is used

in activation analysis, nuclear chemistry, etc.

154

3.2.3. Liquid nitrogen cold loops

These loops can occupy fixed positions on one face of the core

and permit irradiations in thermal fluxes between 1 x 10 and

2.3 x 1014 n/cm? sec and fast fluxes between 1 x 1 0 ^ and 2.8 x lO^^n/m^si«»

These loops can be introduced into or withdrawn from the core when the

reactor is in operation. Samples are unloaded semi-automatically and

without heating in order to preserve and leave available for study the

defects caused by fast neutrons.

A large number of solid-state studies have been carried out

on pure metals, semiconductors, etc. (magnetic viscosity, internal

friction and so on). In technological research, these loops provide

the means for studying the properties of materials which are irradiated

and used at low temperature (cold source of the high-flux reactor and

supe r-conduc t ors).

3.2.4. Hot cell

Its novel design, directly overhanging and opening on to the

pool, makes it extremely easy to use. The generally difficult operation

of tranferring active objects from a reactor to the hot cell can be

carried out directly without removing the shielding or without the use

of an airlock and so on. This gives an appreciable saving of time

together with greater safety.

Its internal equipment is adapted to the different operations

to be carried out - work on radioisotope containers, tunnels, inter­

ventions, repair and dismantling of irradiation devices, dismantling

of worn-out devices, specialized loading and unloading operations (for

example, samples of structural materials or of fuels irradiated in

NaK), metrological examinations, etc.

3.2.5* Neutron radiographyradio-

An immersed neutron/graphy aparatus offers the researcher the

possibility of carrying out a radiographic-type examination of his

experimental devices at any time during irradiation, regardless of

the activity. This visual check, the quality of which is identical

to that of conventional X- or gamma-radiography (which cannot be

carried out in this case because of the gamma activity of the samples),

makes it possible to follow the course of an experiment during

irradiation (cracks, swelling, deformation of samples, state of the

mechanical parts of the devices, etc.) and to prepare the physical

checks to be carried out in the hot laboratory. The dimension of the

photographs is 20 x 40 cm, and the whole of the irradiated zone can

be covered by one or two pictures. The duration of the examination

is about 30 minutes (Figures 5 and 6).

3*2.6. Computer

A computer assigned specially to the experiments carries out,

in particular, regulation and monitoring of the devices and relieves

the researcher of much trouble and long and tedious work.

The computer operates in real time and can centralize about

500 measurements from the devices. It monitors and replaces these

measurements by directly acting on the monitoring and control pannels

attached to each device. The researcher can interrogate the computer

at any moment on the state of his device and of the reactor through

the question - answer stations. Furthermore, the computer routinely

provides himyliourly extracts from each measurement process. It can also

plot curves showing the development of the parameters measured ( W g . 7).

Lastly, operating in shared time, the computer analyses certain

measurements, for example, dosimetric measurements associated with

the experiments.

It is not normally connected to the reactor itself. When it is

so connected, its role is, for example, to provide each researcher with

general information supplementing his own special data, such as reactor

power calculated directly from the signals obtained by the reactor sensors.

For some months we have been developing digital control pannels

for the monitoring and regulation of experimental devices.

3«3* Conclusions

Since its commissioning, SILOE has been of great service, as

attested by the large number of devices irradiated. It has undergone

changes which have appreciably improved its performance. It has not

yet reached full potential and it is possible that further improvemets

will be effected in the near future.

156

4. THE GRENOBLE REACTORS: MELUSINE

4*1 Characteristics

In 1958» MELUSINE was the first open--core swimming-pool type

reactor in Prance.

Since then, the fluxes available have "been increased progressively,

raising the power to 4 MW (December 1965) an<l from 4 to 8 MW in

September 1970.

A large number of modifications have been made in the reactor

to facilitate its use from the standpoint of operational safety and

reliability and to improve its performance and experimental

facilities. These modifications include the control panel, cooling

circuit, core block, leakproof lining, experimental access ports

and leakproof reactor hall.

The pool consists of three compartments:

- Core compartment, equipped with three radial horizontal channels

and a tangential channel with two access ports;

- Middle compartment, which is used for relief of congestion and

can also be employed for gamma irradiations using the irradiated

fuel elements unloaded from the core (for example, dynamic

study of the behaviour of bearings and grease);

- Rear compartment used for storage, out-of-flux testing, and

unloading of irradiation devices.

The irradiation conditions required by researchers have changed

much since MELUSINE went into operation. This has meant that several

times the core has been adapted to the irradiation requirements and

facilities.

The core now consists of about 30 fuel elements comprising 23

plates loaded with 93Í° enriched uranium. One face of the core is

reflected by beryllium, behind which are the channel ends.

The core was originally suspended from a travelling crane, and

this considerably restricted accessibility. Since 1965» the fuel

elements have been placed on a raised platform over a grille with

110 positions (77 x 8l mm). The primary coolant flow is 560 m^/h

(descending flow).

The core, the general shape of which is square, contains some

indentations at the edge. Special irradiation elements at the centre

of the core make it possible to obtain relatively high fast fluxes

(0r = 7*5 x lO1^ n/cm2 sec) (Fig. 3).

4.2 Successive improvements and power increases

Below we give a chronological description of the principal

modifications and improvements made in the MELUSINE reactor. We shall

then say a few words on the recent power increase from 4 to 8 MW

(tests were carried out up to 10.5 MW).

1959 - Start-up at 1 MW

1960 - Increase to 1 .4 MW

1961 - Increase to 2 MW. Increase of flow rate and installation

of a second heat-exchanger.

1965 - Increase to 4 MW. Installation of flying wheels in

primary pumps. Increase of the secondary flow rate.

Introduction of a hot^î^fir. Structural modifications

to the pool, core and control panel.

1968 - Installation of a new piping system at the core outlet.

Installation of "fork" rods (Fig. 4).

1970 - Complete modification of the cooling circuit.

1971 - Normal operation at 8 MW.

The last power increase was planned in the following context:

- within a budgetary allocation of 300 000 francs, .

- the work had to be carried out during the annual shutdown of

the reactor (August 1970),

- minor modifications in the experimental devices in order to

adapt them to the doubling of power.

4.2.1. Planning and execution of power increase to 8 MW

The objectives set forth above were attained essentially by

completely redesigning the out-of-pile part of the cooling circuit

in order to obtain an overall flow of 5^0 m^/h in the core and a

heat transfer capacity of 8 MW. The core had, however, to undergo

some design modifications.

4*2.1.1. Core

The configuration of the core was slightly modified to yield the

maximum flux in the experimental devices.

(1) The two irradiation boxes in the core were replaced by two fuel

elements with a double irradiation cavity, providing a gain of the

order of 20$ on the fast fluxes.

(2) The new fuel elements were placed at the edge and not at the

centre of* the core. The average gain on the thermal flux around the

core was &fo and that on the fast flux 20$.

On the other hand, the location of new elements at the edge of thep *5 c

core requires, for the same reactivity, a 10$ increase in U.

The principal results of the thermal calculations of the core are

given in the following table for a normal primary flow of 560 m^/h and

a safety flow of 450 m^/h with the pool at 30°C.

Cladding temperature at the hotte

point for P = 8 MW, Qp = 5^0 m^/

c+ Calculation

viith rated

values

Calculation

with

cumulation of

uncertainty

coefficients at

the hot point

83°C 109°C

Power corresponding to "boiling

S'*#sLO,11

15.7 MW 10 MW

at the hot pointeo

n

I4 .3 MW 8.75 MW

Power corresponding to

so

VOLO,Hc#1

20.4 MW II .4 MW

redistribution of flow in ~

the hot channel 6OLPi

II

16.3 MW 9.3 MW

4.2.1.2. Heat-exchangers

Two options were considered - addition of a third exchanger or

installation of a single 8-MW exchanger. The latter solution with

a plate-type heat-exchanger was adopted for the following reasons:

- facility of installation in a small area. This type of exchanger

is highly compact (volume of the active part 1 m^),

- high financial saving in comparison with the first solution

owing to extreme simplification in the piping, .

- short assembly time.

The plate-type heat-exchangers are interesting also from other

standpoints:

- the exchange capacity can be increased by addition of more plates

to the same frame,

- the high exchange coefficient and hence low consumption of

industrial water (for a given exchange surface and a given

primary-secondary temperature difference).

- low sediment formation as a result of the turbulence due to the

plate design.

- easy inspection and cleaning of plates on the primary and

secondary sides; the plates are arranged in the frame in groups

of six threaded rods.

The problem of making the joints between plates leakproof appears

to have been fully solved. Communication between the primary and

secondary through a broken joint is ruled out, because the circuits are

separated by two joints with an air gap “between them.

4.2.1.3. Pumps and the primary circuit

The motor-pump groups ased are those of the old 15-MW SILOE. The

rated characteristics of a pump are 5^0 m^/h with a 19~m water head.2

The moment of the inertial mass is 20 m kg. Two pumps are installed,

one in operation and the other on stand-by.

The original external piping (0 150) in the pool block was fully

replaced by <f> 250 and 0 300 pipes to reduce losses of head and obtain

the desired flow without causing a pressure loss in the decay

tank.

A new diffuser for return to the pool was installed in order to

obtain a low water outflow (25 cm/sec) and thus prevent disturbanceof the hot layer.

The use of a single pump, in the case of breakdown of the coupling

between the pump and the flying wheel leads to sudden stoppage of

flow and hence to transition to natural convection a few seconds after

the fall of the rods following triggering of the "no flow" safety

systems. The residual power, although less than 1 MW, is still

relatively high. In order to delay the reversal of flow in the core

by one minute, a 1-m^ buffer tank was placed at the outlet of the

deactivation tank. This tank is at a heig&fc between Hq (height of

water surface in the pool) and Hq-AH ( A H represents losses of head

between the core and the deactivation tank). The buffer tank therefore

remains empty in normal operation and is filled when the primary stops,

thus playing the role of an "inërtial mass" of water.

4.2.2. Results

The thermal tests and measurements were carried out under the

normal operating conditions of the reactor by moving about in the core

three fuel elements, each equipped with several thermocouples giving

the temperatures of the füel plate cladding (the three elements

comprised one standard, one control and one irradiation type element).

The same operation was performed for each power increase in

MELUSINE and SILOE.

The measurements confirmed the positions of the hot points and

proved the validity of the calculations, demonstrating the high margin

over local boiling (l28°C).

Tests at over-power (l0.5 MW) showed that the safety conditions

were satisfactory.

The power increase in MELUSINE, its adaptability to the experimental

environment and its still-unexploited potential for modification

illustrate the principal qualities of swimming-pool research reactors

- simplicity, flexibility and versatility,,

4*3» Experimental possibilities for MELUSINE

MELUSINE is a reactor with pronounced versatility. Its flux

characteristics, equipment such as channels, heavy-water vessel, rabbits

and cold loops, and its operational flexibility make it very attractive

in basic and applied research. Since its construction, it has been

subjected to considerable modification to make it as complete as possible

in regard to users' needs.

Among the sites around the core which can be used simultaneously,1 2

the indentations supply thermal fluxes of up to 6 - 7 x 10 n/cm .secI 2

and fast fluxes between 2 and 4*3 1 10 n/cm .sec (E >■ 1 MeV) inside

aluminium mandrels representing the experiments.

In the first row around the core, the thermal flux varies between

3 and 1 x 10'*'4 n/cm2 .sec, while the gamma flux produces heating between

0.6 and 1.6 W/g in the graphite.

Lead shields, 2 cm in thickness, have been inserted between the

core and some of the experimental devices in order to reduce gamma

heating approximately by a factor of 2.

4.3*1» Channels

Since 1965 the radial channels have contained a removable "thimble".

Below we describe the main use of these channels in 1970, which

obviously varies with the research programmes in hand:

- Radial channel No. 1: ejected beam, can be equipped with a

neutron "heating" device for the polarized-neutron diffracto­

meter (Fig. 8).radio-

- Radial channel No. 3s industrial neutron/graphy at variable

neutron energies (epithermal, thermal and cold) for non-destructive

testing of unirradiated objects, to which X- and gamma-radiography

cannot be applied (Fig. 9)» The cold-neutron beam, the only of

its kind, opens up new prospects for the testing of steels and

for research on trace materials.

- Radial channel No. 2: ejected beam, for testing a method to

improve the resolution of measurements by time of flight.

- Tangential channel: a (beryllium) diffuser placed in the

channel at the level of the core makes it possible to increase

the intensity of the thermal neutron flux, the fast-neutron and

gamma fluxes remaining relatively low. Outlet T2 of the

tangential channel is equipped with a very efficient collimator.

The beam is used for two preliminary studies relating to the

Franco-German high-flux reactor (development of monochromators

and study of the Mossbauer effect). Outlet T1 is used for

studying short-lived fission products with a furnace connected

to a mass spectrometer.

4.3.2. Heavy-water vessel

A heavy-water vessel is placed at a distance of 8 cm from one face

of the core. The flux in the vessel is rich in thermal neutrons and

shows a good <P V . j . ratio (this ratio varies from 86 to 250,

depending on the irradiation sites).

A pneumali/ancLa a^iqu^d-ne^n^l&op are connected with this vessel. A

beam with high <P ,, / , and 9 <P ratios can be extractedun rasx un y

vertically from the vessel.

4.3.3. Pneumatic irradiation facilities

Three pneumatic irradiation facilities are connected to the Medium

Activity Laboratory area, where several laboratory facilities of the

Grenoble Nuclear Research Centre are located :

- A pneumatic facility with direct access to the heavy-water

vessel and used intensively for thermal neutron activation

analysis ;

- A pneumatic facility with direct access to the core in the

light water and used in the conventional manner.

These two facilities are provided with a pneumatic system for

rotating the cartridges in order to homogenize the doses in the whole

volume of the samples.

- A pneumatic transfer facility connected with an automatic

loading machine partly immersed in the middle compartment.

This facility makes it possible to convey irradiated samples

in standard radioisotope containers to the Medium Activity

Laboratory after a period of cooling in the pool. It thus

removes the limitations imposed on the transport of such

radioisotopes in lead containers.

4.3.4. Loops, irradiation devices and neutron Tsatliography

Without shutting down the reactor* the cold loops can be

introduced or withdrawn from their irradiation position, and unloading

can be carried out semi- automatically without heating the samples.

These loops, which are used for basic studies in physics, chemistry,

etc., are three in number:

- 1 liquid-helium loop (5°K - 7 faQ^ = 2 x lO^2 n/cm2 .secj ’ E > 1 MeV)

* 1 liquid-neon loop (2 8°K - f = 9 .5 x 1 0 12 n/cm2 .sec;

f1^ = 6 x lO"*- n/cm2.sec)

- 1 liquid-nitrogen loop (78°K - ? fag± = 9*5 * 1 0 12 n/cm2 .sec;

f ^ = 6 x lO1^ n/cm2 .sec)

(it may be recalled that a liquid-neon loop is also in operation

in the heavy-water vessel).

- Special furnaces (T° 30-500°C) equipped with solenoids to generate

magnetic fields of the order of 5000 Oe are available for solid-

state physics investigations.

- Lastly, for technological irradiations, the conventional irradiation rigs

ard loops(CHOUCA, HEBE, high-frequency furnaces, etc.) occupy

different positions around or in the core (see the description

and performance of devices presented later in this article.

radi o—- As in SILOE, a neutron/graphy aparatus with the same characteristics

and immersed in the middle compartment, is used mainly for testing

irradiation devices. „

4.3«5» Hot cells

Two hot cells were installed in order to carry out certain

operations (examination, cutting, processing and repair) on samples,

irradiation devices and fuel. The larger cell is shielded for activities

(l MeV gamma) of 10 kCi. The smaller is used mainly for radioisotopes,

which it receives directly from the top of the pool.

4 »4» Conclusion

Because of the special layout of the equipment around the core,

MELUSINE produces neutrons with characteristics, spectra and flux

suitable for a wide range of experiments for both basic and applied

research.

Fluxes of increasing intensity, without detriment to these

qualities, have beèn offered to users by successively increasing the

operating power, and this continued to be done in 1970-71.

As a result, MELUSINE is a particularly interesting reactor

offering facilities complementary to those of SILOE, the two reactors

thus constituting a homogeneous and complete set.

This rapid survey of the Grenoble reactors would be incomplete

without some reference to SILOETTE.

The services rendered by this model reactor of power not exceeding

100 kW are vital. All the core layout studies for SILOE and MELUSINE

are carried out in SILOETTE without the need to shut the former down

periodically - configuration studies, plotting of macroscopic flux

variations, study of the effect of irradiation devices (reactivity,

flux depression, etc.).

Under the Franco-German high flux reactor project SILOETTE provided

the means for studying the importance of a hot source and various cold

sources for modifying neutron spectra. The two channels offer

interesting possibilities - experiments on neutron channels, studies of

choppers of various types, neutron^frapSy of non-irradiated objects,

etc. (Fig.10).

In this reactor users also find an excellent tool for studying new

equipment including flux measurement and radiation protection devices,

electronic chains, etc.

The two-fold purpose of SILOETTE can be summed up as follows:

support of the main irradiation reactors and collaboration with users

interested in a reactor of relatively low flux intensity and very

flexible operating conditions.

6. IRRADIATION BIGS AM) LOOPS FOR TECHNOLOGICAL RESEARCH

Apart from the simple devices specifically designed for producing

various radioisotopes and in particular cobält, magnesium nitride, etc* the

irradiation devices proper occupy the three free faces of the core. The

average number of such devices around the core is about 40 (Fig. 11) ,in S ILOE.

Before being placed in the reactor, the devices can be tested

in facilities in or outside the pool in order to check their operating

characteristics (e.g. cooling rate and heating characteristics).

The work stations are located around the reactor pools in such

a way as to permit storage of and work on irradiated devices or unloading

of sample holders at the end of irradiation. The latter operation is carried

out by means of hood-type unloading devices or lead containers specially

adapted for transfer in the hot cell or for transport to the hot laboratories

at the Grenoble site or elsewhere in Prance or abroad (Fig.12).

Below we describe the multi-purpose devices, which are used

most frequently in French and foreign reactors.

We should also mention other novel experimental devices, such

as organic-liquid loops, gas loops, sodium loops, gas-film furnaces, magnetic

resonance spectrometer, specialized furnaces for solid-state studies with

or without a magnetic field, liquid-neon and-hydrogen loop, liquid-ljelium

loop, etc., which are also used in these reactors.

6.1 Research with the CHOUCA furnaces

The CHOUCA furnaces are used for irradiation of structural materials

or fuel samples at temperatures from 150 to 1000°C. The temperature is regulated

by means of six independent heating elements to within +_ 2°C and the maximum

relative temperature deviation over the effective length of the furnace is

of the order of 1$.

The effective length is 400 mm and can be increased to 600 mm if a

strictly uniform temperature is not required. The effective diameter is 25mm.

The internal pressure can be regulated from 0 to 80 bars. Heating is monitored

by means of 12 thermocouples, and 18 thermocouples are provided for the samples.

The table below summarizes some experiments carried out with these

devices, 200 which have so far been produced and used in French and foreign

reactors.

Irradiated, material Hirpose of experiment Nature of sample holders

Stainless steel Impact strength tests

. Graphite cylinder Metrology before and after irradiation (Wigner effect)

simplesampleholders

Stainless steel wires Tensile strength tests

Be, Mg, Zr, Pe, Al Tensile strength tests

Dispersion fuels Behaviour under flux;

retention of fission gases

Graphite COg Study of radiolytic corrosionGas

Boron-filled graphite Study of helium production capsules

Silica under CO^ ’ Behaviour under flux

Stainless steel, Zr,Cu,Be Measurements of impact strength, tensile

Boron steel Strength, thermal conductivity!

i

Metallurgical examinations after irradiation 1 NaX

Uranium carbide

Plutonium magnesium Diffusion study of plutonium in magnesium

capsules

Uranium pellets Study of deformation under stress and under flux

Graphi t e-urani urn

Steel

Metrology under flux

Wigner effect, growths, creep under tensile stress

Resonance—cavitjcensordevices

These furnaces, which operate on the same principle as CHOUCA

furnaces, have a bigger effective diameter (53 or 49 mm, depending on

whether they are used with or without a heat shield). They thus permit the

irradiation of larger amounts of material, and nuclear heating makes it

possible to obtain high temperatures, from 350° to 1400°C. Additional

electrical heating is provided by 4 heating elements. The effective length

is 400 mm, and the pressure can be varied from 0 to 60 bars.

These furnaces are used for irradiating structural materials

(high-temperature graphite, glucine (BeO) at 1300°C, various refractory

materials, steatite at 250° and 450°C, etc0) and small items of equipment

(pressure sensors , co-axial tables, electrical relays, micro— circuit-

breakers, fission chambers, etc.).

6.3 Research with the CYRANO furnaces

These devices are designed for irradiation of fuels at high linear

power (17OO W/cm). They are equipped with a device for measuring the

nuclear power released by the sample. Irradiation is carried out under

power, temperature and pressure conditions equivalent to those obtaining

in power reactors. The samples are placed in NaK or in sodium in order to

ensure heat exchange, the main heat barrier being constituted by a layer

of sintered stainless steel and a gas film. The pressure can be selected

between 1 and 60 bars.

These devices make it possible to regulate the fuel sample cladding

temperature for a given power release or to extract the maximum possible

power for a given cladding temperature.

Examples showing the use of these devices, of which 80 have been

made, are given in the following table.

Name of experiment Samples Linear Power, W/cm

Temgerature, Pressure,bars Purpose of experiment

CYBAHO 1 to 7 UOg with stainless steel or Zr Cu cladding

35O-5OO Cladding600

Core2000

60 Measurement of the rate of re­lease of fission gases. Measure­ment of the heat-conductivity integral•

ICARE UC with stainless steel cladding

1700 Cladding65O

Core1200

20

CYPRES and CIRCE

UOg with Zr Cu cladding 450 Cladding600

60 Ifeasuremènt of the internal pressure of fuel (fission gases) and metrology under flux (measure­ment of length variation) by means of a resonance-cavity displace­ment sensor.

NADIA and VENCA

Vent-type fuel 500-600 Cladding300-600

0.8 Study of discharge of fission gases in a sodium bath.

P A C UOg with Zr cladding 950 Cladding330

40 Behaviour, under flux, of a fuel with core melting

POLYCARPE UOg with stainless steel cladding

600 Cladding65O

2 6 simultaneously irradiated fuel pencilsj study of the behaviour of different types of fuel.

SIR SUM IJOg and UOg-PuOg with stainless steel or 2r cladding

500-600 Cladding45O-55O

2 Direct measurement of deform­ation of fuel, with or without stress, and measurement of heat conductivity.

6o 4 Besearch with high frequency furnaces

These furnaces with high-frequency heating make it possible to

obtain and regulate, under nuclear flux, very high temperatures of up

to 2000°C. An alternating-current inductor (50-500 kHz) generates, in a

shield, currents which in turn induce currents in a susceptor, which can "be

the sample itself or a part of the sample holder. The latter*s diameter

varies "between 16 and 21 ram; the high-frequency power attains a value

of 200 W per centimetre of height and the nuclear power a value of 400 W/cm;

the temperature can be regulated to within 5°C or +_ 0.5°C, according

to specifications.

Examples of experiments:

(a) study of fission gas diffusion outside fuels, the fission

gases being entrained by a helium stream and analysed outside the reactor

(UOg, UC and dispersion fuels).

(b) Endurance and accelerated-aging tests of graphite, U02

or UC fuels and special fuels for itermoiordc conversion.

(c) Design and production of a hot neutron source by hi^i-

frquency heating of a graphite cylinder at 2300°IC.

6.5 Besearch with resonance cavities

The dimendional behaviour of structural material or fuels

exposed to radiation results from the action of various phenomena such

as creep, growth, swelling (of fuels, for example), compacting (vitreous

silica, for example) cycling, Wigner effect, etc.

In many cases, these measurements are carried out after irradiation

with all the disadvantages that this involves (interpretation of the finer

phenomena and the need to multiply tests in order to obtain a number of

points).

The disadvantages mentioned above can be eliminated by developing

an apparatus which can be used for direct measurement, under flux, of the

dimensional variations of samples. The deformation of a sample during

irradiation causes a change in the volume of a hyper-frequency resonance

cavity, the latter being connected to the sample by means of a piston.

The process consists of measuring the variation of the resonance frequency

of the cavity, and this variation is related directly to the deformation

of the sample. The measurement procedure has the following characteifcistics*

measurement range : 0 to 1 mm and 0 to 20 mm;

resolutions : ¿.0.01# of the measurement range

accuracy : ^ 0 .05$ of the measurement range;

- sensitivity s K * 1 ]im.

6*6. Research with cells for creep studies

The creep cells are used for irradiation of samples kept under

tension or compression with continuous measurement of the following

parameters •: temperature, stress, neutron flux, dimensional variations

and creep rate.

The cells contain a tension or compression jack with deformable

metal bellows, thermocouples, neutron-flux detectors and a resonance-

cavity displacement sensor. Elastic deformations and the effects

of temperature expansion are compensated. The temperatures can be

homogenized by liquid metal filling (NaK).

These cells have so far made it possible to carry out creep

studies on stainless steel, graphite and tubes under internal pressure.

6.7* Sodium loop

This loop is intended for irradiating fuel elements under

heat conditions similar to those of fast power reactors.

The use of such a device in a research reactor should be

subject to compatability with the access parts available for ex­

periments and with the general safety of the facilities (reactor and

experiments) in the vicinity.

Two aspects are important : the circulation of the sodium is

ensured by special heat-pump elements, which provide a high degree of

intrinsic operational safety since they need no electric supply or

connection outside the reactor; in addition, the sodium is separated

at all points from the external medium (especially from the water in

a swimming-pool or tank-type reactor) by a double stainless steel wall.

Technical characteristics :

Total extractable power

Test section

Sodium temperature

Sodium flow-rate

60 kW

p 20 mm, length 1400 mm

300-700°C

2-5 m/sec

7. COLLABORATION WITH OTHER COUNTRIES

The intensive use made of the reactors belonging to the French

Atomic Energy Commission for both basic and applied research,

illustrates the importance of swimming-pool reactors in

the development of nuclear techniques.

For many years past, this development has been the subject of

collaboration with several foreign countries, including developing

countries, in a number of fields :

(a) Training of engineers and university graduates either specifically

in the problems of construction and operation of reactors and ex­

perimental devices or, within a wider framework through direct part­

icipation in the work of reactor teams or through provision of facilities

for preparation of university theses. In this manner, various novel

technical improvements or measurement and calculation methods have been

developed by, or entrusted to, foreign collaborators with the greatest

mutual benefit-

(b) Supply of technical advice mainly in the construction, operation

and improvement of the performance of swimming-pool reactors. With our

wide expérience in this field and particularly in effecting power in­

creases, we participate in the relevant projects of several foreign

countries which wish to adapt, modify or improve their reactors with

a view to securing thereby a basic tool for their programmes on the

development of nuclear techniques. Close and regular co-operation has

been established in particular with several South American countries.

(c) An interesting example of co-operation can be cited here - that

of a joint project in progress for several years past in collaboration

with a South American research centre for the purpose of improving the

performance of the reactor operated by the centre, using it for launch­

ing a solid-state physics programme and forming a mixed team of solid-

state physicists, working in the first phase under the scientific

direction and with the equipment of the corresponding laboratories

at Grenoble. One aim, in the second phase, is to establish an in­

dependent laboratory capable of defining its own programmes and de­

veloping its own research facilities.

(d) Lastly, based on the experience of our collaboration activities, we

are in a position to recommend a type of reactor which is particularly

suitable for developing countries, and for which we have developed a

basic project.

This is a 100-kW swimming-pool reactor with six channels,

designed to operate at a power of up to. 5 MW at Itelater stage. The

power increase from 100 kW does not affect either the pool block or

the general layout of the facilities and can be easily achieved

at minimum cost by merely replacing certain items of equipment and

supplementing a number of others.

In the initial phase of operation, a power of 100 kW is usually

considered to be adequate for launching a nuclear programme involving

training of personnel, production of short-lived radioisotopes, re­

search in reactor physics, and various applications.

In the second phase, a power between 1 and 5 MW enables the

initial programme to be extended to irradiation research and tests

within the framework of a more advanced general programme with emphasis

on the development of nuclear techniques. The performance at 5 MW is

close to those of MELUSINE and TRITON, and the possibilities of use

are therefore the same.

The third phase, which consists in carrying out an extensive

technical research and irradiation programme, requires the commission­

ing of high-performance research and test reactors of the SILOE or

OSIRIS type. This can hardly be considered feasible in countries with­out a sufficiently developed and equipped scientific and technical

infrastructure.

In other words, the economy of the first two phases is difficult

to evaluatebefore proceeding to the last phase.

Within modest financial means, this project envisages one and

the same working tool for the first two phases, thus making it possible

to accelerate the development programme without awaiting the grant of

substantial financial allocations to permit the transition from the first

to the second phase.

The initial design of the reactor is such that the power can be

increased from 100 kW to 5 MW at very low cost, while the extra ex­

penditure incurred at the outset in order to benefit subsequently from

this valuable feature is, contrary to appearances, not high.

The reactor can naturally also be built for an initial power

of between 0 and 5 MW*

Because of its design, which is the result of many years of

experience, this reactor is simple, compact and specifically adapted

for constituting the foundation of a basic and applied nuclear re­

search programme.

8. GENERAL CONCLUSIONS

The grouping of three reactors - MELUSINE, SILOE and SILOETTE -

at the same site makes available to researchers a range of facilities

which can satisfy, if not the whole, at least a good part of their

irradiation requirements.

Independently of continuous research aimed at improving the

performance of reactors, there must, however, also be provision for

guiding researchers in making the best use of the reactors placed at

their disposal. This can be achieved by offering them a range of

additional special services for the double purpose of s

- assisting them in the matter of designing devices and taking

measurements specific to the reactors; and

- relieving them thereby of technical details with which they

are not always familiar.

It is with these objectives in mind that the Reactor Service of

the Grenoble Nuclear Research Centre has stressed on the development

of (a) equipment and irradiation devices to suit the reactors and to

meet most of the usual needs of researchers; (b) neutron and gamma

dosimetric methods adapted to the devices; (c) examination methods such

as neutron radiography; (d) computerized monitoring and regulation of

experiments; (e) unloading*©f samples, etc.

In conclusion, it can be said that the considerable progress

achieved in the sphere of irradiation over the past few years is the

result of this continuous collaboration between researchers and reactor

personnel.

Pig. 1 Vue générale du hall de SILOE General view of SILOE hall

Fig. 2 Tableau de contrôle de SILOE Control panel of SILOE

Elément d'irradiation à deux trous avec un four CHOUCA dans l'un des deux trous

Irradiation element with two holes with a CHOUCA furnace in one hole

Pig. 4 Barre fourchette dans l’élément de contrôle Pork rod in the control element

Pig. 5 Neutrographie d’une rose, illustrant la sensibilitéde cette méthode dans l'examen non destructif des matériaux hydrogénés, un des aspects parmi d'autres de la complémentarité de la neutrographie et de la radiographie.

Neutron radiography of a rose, illustrating the sensitivity of the method in the non-destractive examination of hydro­genated materia, one aspect of the complementary nature of neutron radiography and radiography.

Dispositif d'irradiation à examiner

Collimateur

Vérin de blocage du dispositif

Diaphragme

Fig. 6 Neutrographie d'un dispositif d'irradiation trèsradioactif

Dispositif d ' étanchéitè

( p a r joint d « glaceou élastique )

Vérin de mise en place de la cassette porte- convertisseur

Parois neutrophages

fenêtre1 5 0 x 4 0 0 Mél. 3 0 0 X 4 0 0 Sil .

Cd + In

A P P A R E I L DE N E U T R O G RAP HIE I M M E R G É ( Siloé - Mélusine)

CONTROLE DE DISPOSITIFS RA D IO A C TIFS

Fig» 6 Neutron radiography of a highly radioactive irradiation device

JACKING FOR CLAMPING THE DEVICE

DIAPHRAGM 8-30 mm

Cd*In NEUTRON ABSORPTION WALLS

SCALING DEVICE (WITH FROZEN OR

LASTIC PACKING)

IM M E R S E D N E U TR O N R A D IO G R A P H Y (S IL O E M E L U S I N E )

CHECKING OF RADIOACTIVE DEVICES

'JACK FOR POSITIONING HE CONVERTER HOLDER

CASSETTEWINDOW 150 x 400 Mel 300x400 Sil

A P P A R A T U S

INSTALLATION SILOEPig. 7 Schéma de l'installation du calculateur de surveillance et de

régulation des dispositifs expérimentaux

SHOE FACILITYPig. 7 Sohematio layout of the computer for monitoring and regulating experimental devices

Fig» 8 Diffractomètre à neutrons polarisés sur un canal radial de MELUSINE

Polarized-neutron diffractometer on a radial channel of MELUSINE

Pig. 9

Industrial neutron

radiography faoility

on the

HSLUSIIiE channel

G EN ER A L LAYOUT OF APPARATUS

MOVABLE FILTER BISM UTH SIN G LE CR YSTAL

M O VABLE DIAPHRAGM / 5 a 20 m m

R E A C TO RE N C L O S U R E

• • &K.A ♦ A

! -X &*. • .

* A ’ * * •â .

BORON

FILTRE FOR THERMAL NEUTRON BEAM

INDUSTRIAL NEUTRON RADIOGRAP

INSULATIO N G AP G A S E O S U S N2 O UTLET

&Cd + ln

E X H A U S T FO R (p f j CON TACT B ETW EEN ÍS

m m mBORON'

LEAD

m i mLIQUID N GAP

FILTER FOR COLD NEUTRON BEAM

HY APPARATUS MELUSINE 1969GRENOBLE NUCLEAR RESEARCH CENTRE(FRANCE)

Pig. 9

Installation de

neutrograpMe

industrielle

sur le

canal de

MELUSIHE

I M P L A N T A T I O N G E N E R A L E DE L ’ A P P A R E I L

Diaphragme mobile

Filtre amovi.ble m o n o c r is ta l b ism uth

enceinte ¿ *:.dur é a c t e u r '

A * * ' • ' ^ ¿

FILTRE P O U R F A IS C E A U D E N E U T R O N S THERMIQUES

V id e d ’ isolement

A r r i v é e N? liquide

D épart N 2 g a z

iJ. ¿‘..j

Bore !

PlombA spirationcontact entre les fenêtres « M

FILTRE POUR FAIS CEAU DE N E U T R O N S FROIDS

APPAREIL DE NEUTROGRAPHIE I N DU S T RIE L LE * ME LU S 1NE* 196 9 *I C e n t r e d étu d e s nucléaires dé G R E N O B L E ( F f a n ç t ï |

Pig. 10 Montage de la source chaude dans sa cuve à eau lourde a SILOETTEInstallation of the hot source in a heavy-water vessel in SILOETTE

Pig. 11 Dispositifs expérimentaux autour du coeur de SILOE

Experimental devices around the SILOE core

Fig, 12 Chateau de transport des parties inférieures irradiées des dispositifs expérimantaux standards

Lead container for transporting the lower irradiated parts of standard irradiation devices

TITLE: Utilisation of a Research Reactor as preparation for the introductionof Nuclear Power.

by A.C. Wood

Bandung, Indonesia, 2-6th August 1971

Category: Engineering

(b) Engineering work in support of a nuclear power programme.

ABSTRACTIn order to successfully introduce nuclear power a developing country needs

a group of project oriented experienced nuclear engineers. Because the national

institutions are usually scientifically oriented and the conventional power station

engineers are not familiar with nuclear problems it is desirable to form a

composite team which is given a project as a prelude to the construction of the

nuclear power station.

A number of project types are described briefly in the paper ranging from

the construction of a small prototype power station to an engineering loop. An

example is given of a project concerning a modification to the Australian

research reactor HIFAR. Emphasis is laid on the need for a project to result in

new hardware or a change to existing hardware if the experience obtained is to be

relevant to the construction and operation of a nuclear power station.

This meeting is concerned with Engineering programmes as applied to the utilisation

of research reactors. All the developing countries represented here possess research

reactors and most countries see themselves either in the short term or the long termA

as users of nuclear power.

In order to make the best decisions in relation to the choice of reactor, choice

of site, quality assurance during construction and later, the problems of operation of

nuclear power plants, it is necessary to make use of expert engineering judgement at

each stage. Because most countries would prefer to draw on their nationals for this

expertise rather than rely on foreign sources, there is a strong incentive to use

existing research reactors as a vehicle for the training of engineers. It is our

objective here to propose and discuss possible Engineering programmes which might

further this end. My personal view on how this may best be accomplished is offered.

In introducing the panel on "Engineering Programmes in Research Reactors" sponsored

by the IAEA in Vienna in July 1970, Kolbasov and Gonzales - Montes observed that "most

national institutes begin their nuclear activity with strong emphasis on the scientific

aspects of nuclear energy programmes". The panel however was asked "to focus its

attention on seeking ways of helping developing countries to train the engineering

specialists needed

(1) To act as good customers and operators in the case where the

intent is to buy turnkey plants.

(2) To develop the countries' capacity to launch appropriate facets

of their domestic nuclear ability".

One might ask whether the latter two objectives can be accomplished at all by

undertaking engineering programmes in a research reactor at a scientific institution.

In those developed countries where power reactors are designed and constructed

locally this question does not arise because there are enough engineers continually

employed on design, construction and operation of reactors as well as in the support­

ing industries for adequate project oriented expertise to be available to a potential

reactor purchaser. J.A.L. Robertson of Canada noted in his comments on the papers

presented to the IAEA panel of July 1970 that "the best person to assess a power

reactor is someone experienced enough to design the reactor itself". In the

developed countries the research engineers and scientists at the national institutions

provide in depth support to the designers, constructors and operators. A very high

level of skill can be brought to bear in the national scientific laboratories on a

specific problem when the aid of these specialists is enlisted.

On the other hand in a developing country without an existing nuclear power

programme one would not expect to find an experienced group of project oriented

nuclear engineers. There is likely to be a group of engineers with some experience

in the construction and operation of conventional power stations and another group

of engineer-scientists at the research institution with little experience in the

problems of design, construction and operation of large reactors, but quite a

detailed knowledge of the scientific aspects of heat transfer, materials compatibility,

reactor theory and so on.

There is no cheap solution to the problem of acquiring this essential experienced

group of project oriented engineers. Some overseas experience is essential because one

must have first hand knowledge of the engineering details of the reactors and the type

and magnitude of the problems regularly encountered. However overseas experience on

an attachment basis is not sufficient because one tends to be insulated from the

rigours of the actual decision making process. At all levels of engineering there is

no substitute for experience in the line of command where decisions have to be made to

a schedule and where the decisions have considerable financial and safety implications.

One suggestion is to simulate on a smaller scale a project which will confront

people with the type of problems they would be expected to solve if they were expert

advisers in their respective fields. Ideally one woúld recruit both conventional

power station engineers and reactor engineers from research institutions, and from these

form a team with the specific responsibility of undertaking a small nuclear project.

A suitable engineering project for such a team would be the design and construction of

a small prototype power generating reactor but such a project would be expensive and

might be beyond the means of most developing countries. A less ambitious plan would be to

design and construct a research reactor, because the project would offer experience in

all the engineering aspects of a power reactor except for the turbo-generating stage.

I believe Argentina has successfully conducted such a project. All the countries

represented here have at least one research reactor of foreign design and in many

cases of foreign construction. If the justification exists to construct another

research reactor, every opportunity should be taken to maximise local participation

at all levels and to minimise the degree of foreign assistance.

Finally, modification of an existing reactor to provide for power uprating, the

use of a new type of fuel element or the replacement of the safety circuitry could

provide extensive Engineering training relevant to power reactors. The opportunity

to do this might arise from the need to provide higher fluxes for experimental

purposes or for more flexible and efficient operation. A competent engineering group

can thereby broaden their training in reactor engineering and in turn save large sums

of money through making wise decisions when nuclear power is introduced. Such training

will not be of great use unless the project involves provision of new hardware or

significant changes to existing hardware in spite of a cost incentive to stop short of

this phase. Programmes which yield no new hardware or changes to existing hardware but

instead result in an interesting scientific paper (such as many of those proposed by

the IAEA panel of July 1970) are useful only in demonstrating the existence of

technical specialists who are available for consultation. While this is a necessary

requirement for a local nuclear power industry, it is not a sufficient requirement.

The first line of engineering expertise necessary for countries interested in becoming

"good customers and operators" must consist of project oriented engineers.

To demonstrate that worthwhile engineering projects can be formulated about

modifications to existing reactors, this paper is devoted to describing an example

concerning a modification to the Australian research reactor HIFAR.

This particular modification stimulated a complete review of the safety of HIFAR.

A sum of $45,000 was spent on improvements to plant as a result of the study. A

number of engineers and scientists received valuable experience of the type appropriate

to project nuclear engineering.

HIFAR is a 10 MW heavy water moderated and cooled reactor using fully enriched

uranium fuel and is almost identical in design to the English DIDO reactor. The

reactor vessel operates at atmospheric pressure and is constructed of half inch

aluminium plate. It contains several re-entrant tubes in the horizontal plane and one

horizontal through tube. These tubes are used for neutron beam experiments and

self-service isotope production. Vertical experiments may be loaded into experimental

tubes located outside the core in the heavy water reflector or inside the core within

hollow fuel elements.

The original fuel elements known as Mark II were of the standard MTR box con­

figuration containing 10 parallel and slightly curved plates. Each fuel element

contained 110g of 235jj metal in aluminium alloy. The Mark II type of fuel element

was replaced by the Mark III, when fast flux irradiation facilities were required.

The main feature of the Mark III fuel element was a 2" diameter unfuelled central

tube to contain the experiments, surrounded by a 4" diameter unfuelled outer tube.

The 10 fuel plates were arranged in a helical pattern in the annulus and the 235u

fuel loading was increased to 150g in each fuel element.

To increase the power of this type of reactor from 10 to 20 MW a first

requirement was modification of the fuel element. The Mark III fuel element had

insufficient heat transfer area, giving an excessively high heat flux for a twofold

increase in power level. In addition the pressure drop across the element was

excessive and any substantial improvement in coolant flow rate through the element

would require an excessive pressure in the plenum area in the lower vessel region.

Turbulence was also observed on the free heavy water surface from currents arising

from the exit coolant flow from the fuel elements<, The control rods are of a

blade type which fall between rows of fuel elements after the manner of railway

signal arms and the currents unnecessarily extended the control rod drop times

following a reactor scram by impeding their free fall.

The UK engineers designed a new fuel element designated Mark IV using

concentric tube geometry. It superseded the Mark III fuel element and was intended

to offer improvements in the areas mentioned above. It retained the outer 4"

diameter unfuelled tube and consisted of 4 concentric fuelled tubes with a total

of 507. increase in heat transfer area. The inlet nozzle was redesigned to a

venturi shape, resulting in a considerable reduction in pressure drop across the

element. The outlet porting was also considerably modified with a consequent

reduction in heavy water surface turbulence, which reduced the control rod drop

times by 257. compared with those of the Mark III element. The construction of

the new fuel element was much less rigid and some structural failures occurred

during its development. Nevertheless successive modifications had resulted in

a high level of reliability for completing a four month operational life which

was expected to be its normal irradiation period (slides 1 and 2).

Careful loading of the new Mark IV fuel element was necessary because failure

to load an inner thimble would result in by-passing coolant around the fuel plates.

This problem did not arise with the old Mark III type element because the inner

unfuelled tube was an integral part of the element.

Another feature of the geometry of the Mark IV fuel element is the poor

conduction heat transfer path from the fuelled tubes to the outer tube. This

has two important effects on safety, viz

(1) An irradiated element held in stagnant air would reach a

higher temperature from decay heating than would an

equivalent Mark III element. Fuel is unloaded from the

reactor by using a 20 ton shielded transport flask. Care

has to be exercised to avoid fuel melting during the fuel

unload operation, either by ensuring that air cooling on the

fuel element transport flask is guaranteed or that fuel

unloading is deferred until air cooling is no longer

required to prevent fuel meltdown.

(2) In the original design concept a seal was provided between

the fuel element nozzle and the plenum chamber at the bottom

of the vessel to prevent a fuel meltdown in the event of a

primary circuit rupture. Even if a pipe ruptured, the reactor

vessel could only drain to the level of the outlet ports of

the fuel elements. Sufficient heat could then be conducted

along the fuel plates to the outer surface of the fuel elements

and thence to the water in the reactor vessel to prevent

melting (Slide 3)» The Mark III fuel element design reduced

the efficiency of this heat removal process by removing

one-half of the heat conduction path and the problem was

further aggravated in the new Mark IV fuel element design*,

We were not able to convince ourselves that catastrophic

failure of the reactor pipework was incredible although we

believed that it was extremely unlikely. Our accident

analysis then had to take into account a full core meltdown

and to show that sufficient engineered safety features

existed to minimise the off-site exposure to acceptable levels.

The escape mechanism and the escape route of the fission products from the

primary circuit into the reactor building and then into the atmosphere were

defined for a hypothetical accident as follows. After the coolant had drained

from the interior of the fuel elements through the pipe rupture, the fuel plates

would melt within a matter of minutes. Molten fuel would then slump to the bottom

of each fuel element where much of it would be in contact with the unfuelled

outer tubes. This heat would boil off the water remaining in the vessel and the

steam generated would contain fission products, particularly I. The steam would

vent through the ruptured pipe into the heavy water plant room and thence into

the building.

The reactor sealed building offers the main line of defence against fission

products released from the ruptured circuit escaping into the environment.

Ideally a reactor building should be held at slight negative pressure during

this phase of a fission break accident but this is very difficult to achieve

as the air pumped from the building to maintain the negative pressure has to be

discharged somewhere and this air contains fission products. HIFAR has a fairly

airtight sealed containment building of mild steel, capable of withstanding an

over-pressure of 1.5 p.s.i. with a leak rate of about 17. per day. When sealed

it is particularly sensitive to inleakage of compressed air from air operated

instruments and sources of heat which would raise the internal pressure. The

main normal sources of heat in the building are

(1) Decay heat of the fission products contained in the

irradiated fuel.

(2) Heat stored in the coolant-moderator and reactor shield

at their operating temperatures.

(3) Heat dissipated from electrical instruments.

(4) High pressure hot water ducted to the space conditioners

for air temperature control within the building.

The main heat sinks in the building are

(1) The mass of structural steel in the crane bridge and the

steel floor joists and plating.

(2) The space conditioner system referred to earlier which is

supplied with chilled water from refrigeration plants

outside. (Each space conditioner consists of a box

containing a fan which draws air over two sets of coils.

Hot water circulates through one set of coils and chilled

water circulates through the other. The flow rates of

hot and chilled water are automatically regulated to the

demand for heating or cooling as sensed by a thermometer).

The rate of heat injection to the building was calculated assuming that the heat

was transported in the manner described earlier. We conducted experiments in

which heat was released into the building to determine what fraction appeared as

sensible heat to cause overpressure and what fraction was absorbed by the steel­

work. The experiments were undertaken with the space conditioner system in

operation and also in the "failed" condition. The space conditioner system was

shown to be very effective in reducing the overpressure, but on the other hand

it was shown that without the space conditioner system, the building pressure

would rise to approximately 2 p.s.i.g. and there is no effective mechanism to

reduce this pressure. The thermal insulation on the interior of the building

necessary to make the air conditioning effective works against heat removal by

free air convection on the outside. A calculation of the fission product release

resulting from this overpressure showed that the release would be unacceptably

high. Thus it was shown that the space conditioner system must play a major

role in minimising the consequences of this unlikely but credible accident.

We then looked at the design of our space conditioner system to determine

whether the design met the high standard of reliability required for an engineered

safety feature. Slide 4 shows the original circuit which on examination reveals

that although there is adequate redundancy in that we have six space conditioners

and three refrigeration-compressor units, these units are virtually not independent

and consequently a failure of any one of several critical items could render the

whole system ineffective. A typical critical item would be a thermostat or even

the pipework itself, which could fail, discharging all the chilled water.

In the modified system shown in the next slide we have three completely

independent refrigeration units, each one served by two space conditioners.

Provision has been made for third units to be installed on each of the three

systems. So far it has not been found necessary to do this. Tests under simulated

accident conditions have shown that two space conditioners and one refrigeration-

compressor unit are well matched for capacity and effectively reduce the overpressure.

In fact it was shown that three space conditioners under accident conditions could

overload a single refrigeration-compressor unit and cause it to trip itself out

on high refrigerant pressure. If the third space conditioner were ever to be

installed it would be necessary to provide a load sharing and limiting device.

Another aspect investigated by the test under simulated accident conditions was

the reliability of the electric motors and the heat exchanger coils in an

atmosphere of saturated steam. Electric motors were found to be unaffected by

prolonged operation (one week) and that condensate lying between the fins of the

heat exchanger coils did not markedly affect their efficiency.

We now believe that we have three reliable systems and that one of these

systems can be down for maintenance at any time. Either one of the remaining two

has sufficient capacity to reduce the heat load to the extent that 131x release

in an accident will be within prescribed limits.

The space conditioner systems though essential, form only part, of the

containment building line of defence. It has been found necessary to explore

the reliability of the system to seal the ventilation trunking if an accident

should occur. The mechanisms to seal the ventilation trunking, its sensors,

its activators and its monitoring, should all contain redundancy and independence

necessary for a well engineered safety feature. Other aspects to consider are

fixed penetrations (power cables, communications, water, effluent, rabbit facilities

and other experimental tubes) and airlocks for personnel and vehicle access.

Regular pressure testing of the sealed building is necessary to establish

confidence that the acceptable leak rate of 17. volume per day exists for most of

the time. Our experience has been that even static seals deteriorate and a high

level of supervision is required on maintenance and modifications to building

penetrations to ensure that seals are still intact when work has been done in the

vicinity. This type of supervision can easily be overlooked, resulting in a

reduction in the effectiveness of engineered safety features.

The example of the HTFAR modification quoted here is typical of the type of

project which can be undertaken in training of engineers. Design and construction

of a high pressure water loop similar to the one under construction by the AAEC

would also be a suitable project. Even if the small reactors owned by most of

the developing countries could not accommodate such a loop many larger foreign

reactors have vacant space which can be hired out. In design and construction

it has many of the characteristics of a reactor and also it would permit

engineering experiments to be undertaken. The loop could be locally designed,

constructed and operated out of pile. When thoroughly tested it could be

dismantled and shipped to the foreign reactor centre. A small team could be sent

overseas to assist in its re-assembly and operation. The final experimental results

together with in-pile components could then be returned for evaluation.

None of the proposed solutions for the development of engineering ability are

inexpensive. This is because project oriented engineers are not going to become

proficient through undertaking low cost applied physics experiments in small

research reactors. If the intention is to undertake pure scientific work with the

object of extending man's knowledge, then many low cost experiments could probably

be proposed. On the other hand if the developing country is concerned about its

capacity to write a specification for a nuclear power plant, to make a sensible evaluation of a wide variety of claims by vendors when tenders are evaluated, to

ensure that the quality specified is in fact provided and finally to operate the

plant in an efficient manner it should be understood that the acquisition of

experience is expensive. It requires participation of key personnel in overseas

large reactor design and construction projects and it requires local projects

where engineers are given a decision making role in smaller construction,

modification, and operation phases of the use of nuclear reactors.

205

MAßk. 3 PULL LLÍMÍNJ.

S ta lin g Surfoc*' utocfor Kraswt.

4M m MVJO ■»yy.f »y ,» « w i*«MW O» M 4

>•*•55*2H s H

3 E EAUÎMLUI ATIIIIC BK1SY CIHHIUIOI

T = S T

o*rc

SLIDE 1 QZCZNT FULL QLÍMÍNT 7YP£5 USED IN UffAÛ.

SCMS *<Hf fuH SfUMM dCtrttM. |fi*n 9 7-ft._i---- \nr----

206

1 * 1 » ----1 ««■»». I U>cT~....OONOTSCAI*

MAÛIC 2 FULL ¡LLLMLNT.

MARI' 5 FULL LUM LNT MhtUi 4 FULL £L£M£NT.

c o w w y w w gg »— n » tm m m m i ai— miw aIM« MAVW*. VMM 8 TM| MCMtnr 0* IM Al____a WXVN OH tK W«piHil«1M ft**»« «JP jiJ !

~n«T MAWW I MOO I tCMAMIAISTIALIAM ATOMIC EffiUY C6HHISSI0N oarc

SLIDE 2RACnONSt

uNua oTHKvn ttAtm C/BOSS SECTIONS OfH I FAQ. FU£L £L£M£NTS.

iCMI ?Utl 3*1*1 If. Ctrksk“

207

208

SIX 3 9 S 4 SPACE CONDITIONERS IN R.S.& EACH-AIR FLOW 2,000 C.F.M.APPROX. COOLING CAPACITY 8Q000 Ä U ^ w .

THREE FREON 22 CIRCUITSEACH t&O.OOO STu/hr.

WEATHERMAKER IN AL5C PLANT ROOM AIR R O W -4 ,5 0 0 C.F.M.COOUN6 CAiP^crrY 646,000 &.T.U/fr.

HIFAR SPACE CONDITIONINGe x i s t i n g c h il l e d w a t e r c ir c u i t s

SLIDE 4 N? 3 PLANT

209

SYSTEM N? Z

C H IU E D W ATER WkAKEUP FROM SEPARATE HEADTANKS EXTERNAL t o R .S .B ------------ ÿ

THREE PAIRS OF THERMOSTATS IN R.S.B.

Ï ® ©

— L C*3—

N? g PLANTTHERMOSTATS CONTROLLING COMPRESSORS

f + T t P T l f + W 6

SYSTEM N? 3

-crS~:- M -

43Q 7

— u

~©

WtATH ER MAKER IN AIK. PLANT ROOW AIRFLOW - 4,500 C.F.M.COOL1N6 CAPACITY ■ 246,000 B-T.u/hr.

L « M-{X J— i

N? 3 PLANT

HI FAR SPACE CONDITIONINGPROPOSED CHILLED WATER CIRCUITS.

SLIDE 5

RELATIVE PC6ITI0NS OF SPACE CONDITIONER UNITS

INSIDE P .S .B . --------- »

Leonard J. Koch .Senior Engineer

“ÍCArgonne National Laboratory Argonne, Illinois, U. S. A.

ABSTRACT

Research reactors provide an effective focal point for a broad spectrum of nuclear engineering activities. Many of these activities can provide excellent reactor engineering prepara­tion for the introduction of nuclear power systems. Effective engineering programs include the participation of engineers in all the activities related to the research reactor program.These activities should include: maintenance, improvement and modification of the reactor and experimental facilities; engineering development of "non-engineering" reactor exper­iments; and a broad program of engineering experiments. A program of this type will assure the participation of the engi­neering staff at the research reactor facility (irrespective of the number of engineers in the group) in virtually all of the activities at the facility. This will result in the best utili­zation of the total technical staff and the best preparation of the engineering staff for future reactor engineering responsi­bility. It should also result in the most effective utilization of the research reactor and the most productive research program.

INTRODUCTION

Research reactor operations and programs play an important role in nuclear research and development. This is particularly true in the early or formative phases of a country's nuclear program. A research reactor provides a unique focus for the various participants in the program and plays an important role in coordinating their activities. Research reactors, as their name implies, are used to perform research-oriented experiments and this focus tends to emphasize

Work performed under the auspices of the U. S. Atomic Energy Commission.

scientific research rather than engineering development. It is the purpose of this paper to discuss methods of increasing the emphasis of research reactor engineering programs, particularly as these programs may contribute to pre­paring the members of an organization's staff for future participation in nuclear power programs.

It is customary to conduct the engineering development program at a re­search reactor in much the same manner as the physical or biological sciences research programs. As a result, the main emphasis tends to be placed on de­veloping engineering experiments to be performed in research reactors. However, because of the limited capability of small research reactors, it is often difficult to devise meaningful engineering experiments which can produce solutions to major engineering problems. Proposed engineering experiments must be carefully evaluated to ensure that they will produce the desired results. Such evaluations should consider the optimum utilization of engineering resources, particularly where they are limited. The available engineering resources should be allocated to: maintenance, improvement, and modification of the reactor and experimental facilities; engineering development of "non-engineering" reactor experiments; and a broad program of engineering activities and experiments which emphasize the effective and efficient utilization of the reactor. These steps will result in direct participation of engineers in the reactor program; the engineering staff will make larger contributions as participants or collaborators in the various ex­perimental programs and will simultaneously become more knowledgeable engi­neers with broader experience. This will result because the engineer must not only understand the reactor and the engineering problems associated with its use, but he must also understand the experiments to the extent that he can trans­late the experiment requirements into engineering requirements for the reactor and the experiment.

In this paper, I will discuss engineering activities related to research reactor operations and programs within the broad scope defined above. I will attempt to emphasize the application of this perspective to the CP-5 and JANUS research reactors at Argonne National Laboratory. In describing the activities related to these reactors, I hope to describe a philosophy which can be applied to the variety of research reactors in the countries of the participants at this meeting. I hope also that these experiences will suggest additional programs which may be conducted on your reactors and which will stimulate increased participation by your engineers.

Most engineering experiments in research or experimental reactors are performed to solve specific problems involving nuclear reactor technology. In most instances, it is necessary (or at least very desirable) to simulate the op­erating environment which will exist in a new reactor under development. This requirement has resulted in the development of large, high powered, sophisticated test reactors because most of the engineering problems are related to power re­actors. In many respects, these engineering test reactors have become as com­plex as the power reactors which they are supporting. Also, many of them have been designed to provide solutions for very specific problems. As a result, special purpose test reactors with very specialized operating conditions and characteristics have evolved. A rather common characteristic of these reactors is that they are expensive and they require the application of advanced technol­ogy and manufacturing capabilities.

The United States Atomic Energy Commission (US-AEC) has supported the construction and operation of a large number of engineering test reactors which have been concerned primarily with the solution of engineering problems related to water-cooled power reactors. A similar program is now underway to develop experimental reactor facilities to solve engineering problems related to the sodium- cooled fast breeder power reactors.

1. The primary operating emphasis of EBR-n, a 62. 5-MWt experi­mental power reactor station, has been shifted to experimental fuel and material irradiation and a large facility for examination of irradiated fuel is being added to EBR-II,

2. The SEFOR reactor, a 20-MWt sodium-cooled fast test reactor is operating to demonstrate and measure the Doppler coefficient in oxide-fueled fast reactors.

3. The Fast Test Reactor, a 400-MWt sodium-cooled fast test reactor incorporating closed sodium loops and open instru­mented test positions is under construction.

These facilities and programs have a strong engineering orientation and emphasis. They require large staffs of engineers to develop the facilities, to define the experiments and to analyze the results. The results produced by these programs may be of interest to the participants at this conference, but they contribute very little to the subject of this conference except to the extent that they serve to place one aspect of engineering experiments into some per­spective.

Engineering experiments in small research reactors tend to fall in the fol­lowing catagories:

1. The development or verification of basic engineering data.

2. Testing and evaluation of concepts or components.

3. The development and testing of basic experimental or operational techniques.

Some of these programs may have as their objective the solution of prob­lems applicable to higher flux, higher power reactor systems. These problems can be solved in a low flux reactor if they are amenable to extrapolation to high flux conditions or if the problems involve low flux conditions and the small research reactor can provide a reasonable simulation. The following are pos­sible examples of engineering experiments in each catagory.

Engineering Data

1. Determine irradiation effects on relatively sensitive materials which normally will be used in relatively low flux regions of a reactor such as: electronic components, semiconductors, plas­tics, electrical insulation, thermal insulation, lubricants, coat­ings, etc.

2. Determine irradiation effects on properties that are relatively well known and understood in the absence of radiation, such as: effect of irradiation on nucleate boiling in water or super­heat in sodium, effects of irradiation on materials conductivity, corrosion, or physical properties.

Testing Concepts or Components

1. Develop concepts for reactor or test facilities, such as: producing specific irradiation environments of neutron and gamma for seed mutations, biological irradiations, etc.

2. Test radiation monitoring instrumentation, self-powered detectors, and low level instrumentation for power reactors.

3. Test effectiveness of shield materials and combinations.

Testing Experimental or Operational Techniques

1. Develop and evaluate low power testing techniques for application to high power reactors, su ch as: low power flux mapping to determine power generation distribution,

danger coefficient measurement techniques to determine material worths, and subcritical monitoring techniques to improve safety during fuel handling.

2. Develop methods of reactor control, including automatic systems with feedback and simulated malfunctions.

3. Develop methods of identifying and controlling the distri­bution of fission products released from fuel elements.

REACTOR MAINTENANCE, IMPROVEMENT AND MODIFICATION

Maintaining a research reactor so that it will provide efficient and effec­tive service to the experimenters is an important function. It should include improvements and modifications which will upgrade the facility and permit the performance of new and better experiments. In the extreme, these activities may include major modifications and reconstruction of large segments of the reactor. Argonne National Laboratory recently completed two such major modi­fications to the CP-5 and JANUS research reactors.

CP-5 had operated approximately 15 years as the basic research reactor at Argonne for physical research. During that period, obsolesence and deterioration had occurred in many systems and components and, as a result, the reactor was taken out of service at the beginning of 1969 for more than 1-1/2 years for a complete "rehabilitation." Major rearrangements of facilities and systems were made and to the extent practicable, they were modernized. For example, the instrumentation system was essentially replaced in total and updated to mod­ern standards. The primary D2O system was rearranged and the reliability of the emergency cooling system was improved to include earthquake and tornado considerations. Services and facilities used by experiments were improved and, to a large extent, separated from those used for reactor operation. The entire (approximately eight-man) engineering staff of the facility was involved in these modifications and approximately $2 million was expended in completing this undertaking. During this same period, the experimenters modified and updated their experimental equipment and a detailed review was conducted of the inter­action between each experiment and the reactor. Although we have no active plans for the design or construction of a new research reactor, there is no doubt that this experience has greatly increased the capability of our staff to design, build, and operate research reactors of increased complexity and sophistication. CP-5 has been back in operation at 5 MW for almost a year and the improvements have been very effective.

During essentially this same period, the JANUS reactor was shut down for major modifications. This is a much smaller reactor (200 kw) for biological re­search. The modification of this facility was quite different than that for CP-5 and involved only the experimental facilities; the reactor was essentially un­changed. Modifications were made to improve the irradiation characteristics for biological experimentation and involved the redesign of the high flux irradiation room, the neutron shutters, attenuator and converter. Figure 1 is a photograph of a model of JANUS which shows a section through the facility. Figure 2 is a draw­ing which identifies the major components. The modifications involved the very close collaboration of engineers, physicists, and biologists. It required sophis­ticated neutron physics and shielding analysis as well as dosimetry to produce controllable and measurable doses to the animals and specimens.

Although JANUS must be characterized as a "special purpose" research re­actor, some of the experience with it is particularly applicable to this meeting because: (1) it is a 200-kw reactor which places it well within the power range of most small research reactors, and (2) it is used for biological research and some of the information obtained can be applied to biological research with other reactors.

The JANUS modification program was directed at two basic objectives:

1. To provide a flux of fission neutrons at a uniform intensity over a large volume (a room approximately 7' x 15' x 6 - 1 / 2 ’).

2. To reduce unwanted background radiation (low energy neutrons and gammas) to the lowest practicable levels.

Fission neutrons are produced in a curved converter plate approximately 147" long and 39" high. The converter contains approximately 34 kg of highly enriched uranium in thin stainless-steel-clad plates. The neutron flux from the reactor incident on the converter plate varies in intensity by a factor of about 6 from the center to the edge of the converter. A "graded " attenuator plate is positioned between the reactor and the converter to "flatten" the neu­tron flux to the converter. The attenuator consists of a curved aluminum plate (approximately the same size as the converter) to which are fastened 1-inch- square plates of boral. The boral plates are non-uniformly spaced to vary the neutron absorption and provide uniform flux upon the converter plate. Figure 3 is a photograph of the attenuator plate. The spacing of the boral squares was developed with a computer program which produced a punched tape used directly in the machine for drilling the irregularly-spaced locating holes for the squares.

This procedure avoided the necessity for a tedious process of dimensioning and locating during manufacture of the attenuator.

The high flux irradiation room required extensive modification to reduce background radiation to acceptable levels. The room was originally constructed with concrete floor, walls and ceiling. The neutron spectrum in the room was softened by the moderating effect of the hydrogen in the concrete. Neutrons captured in the concrete produced prompt gamma radiation which complicated the interpretation of the experiments. Activation of the concrete made it impracti­cable to enter the room for at least ten minutes after an irradiation. To correct this condition, the entire room was lined with a 4" thickness of lead. A 4 " layer of borated hardboard was placed behind the lead on the floor and walls, while an 8 " thick layer of special concrete made with bauxite and boron carbide was placed behind the lead on the ceiling. The ceiling construction presented a challenging structural problem since approximately 15 tons of lead is supported by a steel and aluminum framework, as shown in Fig. 4. Aluminum studs were cast into the lead bricks and aluminum was used extensively in the support struc­ture to minimize capture gamma radiation.

The JANUS modifications were very extensive and the reactor was shutdown for approximately 1-1/2 years. The extent of the task was influenced very much by the size of the experimental irradiation area—the size of the face and the vol­ume of the room. It was designed to accommodate the simultaneous irradiation of at least 500 mice at one time (see Fig. 5), and an extensive long-term irradi­ation program is now in progress.

The specific details of the JANUS modifications are, of course, applicable to that specific reactor facility design. However, the design principles and procedures are probably applicable to other reactor facilities in which radiation biology experiments are performed. Also this experience demonstrates the im­portance (in fact, the necessity) of achieving effective interaction between the different disciplines involved and the central responsibility that the engineer must fulfill in such an undertaking.

ENGINEERING OF REACTOR EXPERIMENTS

Virtually all experiments performed in and with research reactors involve some interaction with the reactor and/or require the use of extensive apparatus. Either of these requirements provides the opportunity for active participation by engineers. The engineer can be particularly effective if he participates as a collaborator because this encourages his involvement in the planning, conduct,

and results of the experiment. This is not a simple relationship to establish and it differs considerably depending upon the personal characteristics of the people involved. It is an important relationship to achieve, however, and it is very important to achieve on relatively small projects.

The following are a few examples of classes of experiments in which engi­neers should participate. Also, they should participate as members of the experimental "team" involved in planning, evaluation, decision making, and reporting. Referring again to our experience with the JANUS modification, the program proceeded rather slowly and ineffectively until a project team was or­ganized. It was a relatively small group (less than six people) but they inter­acted and combined their skills very effectively. This kind of interaction can be applied to the following types of programs (even when only two people are involved) :

1. Cross-section measurements—time of flight, neutron "choppers," etc. These can be relatively sophisticated experiments involving complex mechanisms, precision structures, and sophisticated instrumentation. The engineer has expertise in these fields and their inter­actions.

2. Activation analysis is becoming a very important diag­nostic tool. Reactor neutrons will probably continue to be the most convenient and flexible radiation source for these analyses. A larger variety of special irradiation capabilities will be required to accommodate the large variety of samples which will be involved. The ecol­ogical programs at Argonne have generated a large in­crease in water samples and marine life samples. New low-temperature irradiation facilities are being developed for CP-5 to irradiate these samples in the frozen state.These facilities and the equipment and procedures for handling the samples are being developed by the CP-5 engineering group.

3. There is a continuing need for the production of isotopes.This need is dependent somewhat upon the availability of "packaged" isotopes when needed. In the more iso­lated regions of the world where they may be less avail­

able, it would seem that even the more common isotopes should be produced locally. Of course, it is always necessary to produce the short-lived isotopes locally.Development of the facilities and procedures for producing, packaging, and handling these materials should be included in the engineering program at research reactors. A very important part of such a program involves the preparation of new isotopes or new forms of the isotope for new applications.Here again, effective dialogue is required between the poten­tial user of the isotopes and the producer of the isotopes to ensure that the user is aware of what can be produced and the producer is aware of what the user may need.

4. The development of beam facilities and "tailored" fluxes can generate difficult engineering problems that require the par­ticipation of the engineering staff. These projects tradition­ally require a very high level of collaboration and cooperation between scientist and engineer because of the variety of options and constraints to be considered.

5. Virtually all experiments involve some instrumentation and circuitry and most of them require very sophisticated systems.They provide a fertile field for the instrumentation engineer and participation in such tasks provide invaluable experience for future application to instrumentation and control of reactors.

Radiobiology experiments in small research reactors provide an excellent example of the opportunities available for the engineer to contribute to reactor experiments. Figure 6 shows an animal exposure arrangement described by Ainsworth, et al.* adapted to a Triga Mark-F reactor. For neutron irradiation, a void tank was interposed between the reactor and the exposure tube, as shown in view A, to provide an air path for the neutrons from the reactor to the animal exposure volume. The exposure tube was shielded with 2" of lead and 1 /4" of boral. For gamma irradiation, the void tank was removed to increase the ratio of gamma-ray to neutron dose as shown in view B. The mice were placed in­side a cannister (see Fig. 7). Within the cannister the mice were placed on a retaining board with dosimeters as shown in Fig. 8. The loaded cannister _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

E. J. Ainsworth, G. F. Leong, K. Kendall, and E. L. Alpen, "The Lethal Effects of Pulsed Neutron or Gamma Irradiation in Mice," Radiation Research 21, 75-85 (1964).

assembly was loaded into the exposure tube with a building crane. Figure 9 is a photograph of such a facility in the reactor tank while Fig. 10 is a photo­graph of the external portions of the facility. As can be seen, it is a rela­tively simple facility and illustrates how this kind of capability can be added to a conventional research reactor.

It should be noted that much of the radiobiological experimental research in the U. S. and elsewhere is conducted in facilities of this type. JANUS is an excellent facility and provides a unique capability for this type of research, but excellent work can be performed (on a significantly smaller scale) in much simpler and less expensive facilities. The "engineering program" at virtually all research reactor facilities could include the development of facilities of this kind for their reactors.

CONCLUSIONS

It is not the purpose of this paper to promote the construction of specific facilities or promote specific programs. I have used experimental radiobiology as an example because it lends itself nicely to my principle theme that engineers can enhance the experimental capability of research reactors and can contribute to the experimental program. Hopefully, the examples cited here will stimulate and encourage such activity. I do not believe that such activities by engineers at research reactors detracts from their capability or interest in performing engi­neering experiments in the reactor. On the contrary, the engineers will generate better engineering experiments because they have a better understanding of the reactor and have been exposed to an environment in which they are continually attempting to perform new or improved experiments in and with the reactor.

Finally, the success of these programs is very dependent upon the cooper­ation and coordination of all participants. This should be encouraged in every way possible including the organizational structure and the procedural require­ments for the operation. At CP-5, where there are a large number of experiment­ers, we have found that an experimenter's committee and an experimental review procedure are effective in encouraging communication and discussion between the experimenters and the engineers and operators, Achieving effective dialogue should be an objective of the program along with achieving efficient operations and meaningful experimental results.

221

Fig. 1. JANUS M o d e l - -R e a c t o r and High F lu x R o o m

SHUTTERACTUATOR

ACTUATOR SUPPORT STRUCTURE

REACTOR WORK ROOM FLOOR

EYEBROW

2“ THICK LEAD WALL

7" THICK LEAD WALL­SHUTTER

7" THICK LEAD PEDEST/

10" THICK LEAD LEDGE

CONVERTER PLATE 147 ^ " x 47V|6 x IÜ4

ATTENUATOR PLATES

SHUTTER PEDESTAL

7" THICK CONCRETE PEDESTAL

CONTAINMENT TANKS

ELEVATION

223

F ig . 3. JANUS G raded Attenuator

224

F ig . 4. JANUS Irradiation R o o m Ceiling

22

6

ALUMINUM TUBE , 10* 0.0. *9$" 1.0.

2 INCH LE AO PUIS I INCH BORAL

ANIMAL EXPOSURE VOLUME

Fig. 6 . Animal Exposure Arrangem ent--A for Neutrons, B for Gamma

F ig . 8. M ouse Retain ing B oa rd and D o s im e te r P la ce m e n t

229

F ig . 9. Photo of Irradiat ion F a c i l i ty in R e a c t o r

RESEARCH REACTOR UTILIZATION

ENGINEERING WORK IN SUPPORT OF A NUCLEAR POWER PROGRAMME

by

. S.K. Mehta and S.R. Sastry

Reactor Engineering Division

Bhabha Atomic Research Centre, Trombay,

Bombay, India.

A B S T R A C T

The Department of Atomic Energy of India has committed itself to

a definite programme of building a series of nuclear power plants in India.

This programme calls for a supporting development programme. This report

highlights the facilities available at Trombay for carrying our in*reactor

engineering experiments. The utilization of these facilities so far and

the programmes on hand are outlined.

Introduction

The Department of Atomic Energy of India has committed itself to a

definite programme of building a series of nuclear power plants in India.

The power reactors under construction and proposed to be constructed in

the near future are heavy water moderated, natural uranium f u e M and

pressurized heavy water cooled, i.e. the CANDU-PHW type. The power reactor

programme to be carried out would necessitate maximum indigenous effort.

Such an effort calls for a good research and development programme. This paper discusses the role of the utilization of the research reactors at

Bhabha Atomic Research Centre for engineering studies in support of the

overall power reactor programme.

Research Reactors at Trombay

There are three research reactors currently available at Trombay.

They are:

a) 40 MWt CIRUS Reactor (similar to the original NRX of the AECL,Canada),

b) 1 MWt Apsara Reactor

c) Zero energy ZERLINA Reactor.

Various engineering facilities available in these reactors are elaborated

belowî

a) 40 MWt CIRUS Reactor

This is a heavy water moderated natural uranium metal fueled reactor

with light water in the closed primary heat transport system. The reactor

was built in technical collaboration with the AECL, Canada. Apart from the

thermal columns and the self serve holes there ares

i) 25 horizontal holes of various sizes available for experiments.

These holes extend from the outer face of the biological shielding

.. to the calandria. These horizontal holes are being mainly used for

physics experiments.

ii) Six vertical holes in the calandria, each having a diameter of

approximately 10 cms, which may be used to accommodate test

sections to conduct in-pile engineering experiments. The positions

of the holes in the lattice are shown in figure-1.

iii) One Central experimental hole approximately 13*75 cms i*1 diameter

and called the central thimble for use where high neutron flux is

required and where a relatively large sample or experimental assembly

can be accommodated.

b) 1 MWt Apsara Reactor

Apsara is a swimming pool type reactor with enriched uranium in the form

of MTR type fuel plate assembly. The core is suspended from a movable trolley

and can take various positions in the pool, thus facilitating various types

of experiments to be carried out. There are eight beam holes provided for

conducting a variety of experiments pertaining to nuclear physics, radiation

damage, and biological studies.

In addition there is a shielding corner where the properties of various

shielding materials can be studied.

c) ZERLINA •

ZERLINA is a zero energy pile for lattice studies and has all the

flexibilities to study the lattice parameters of various cores or assemblies.

Engineering Programmes in the Research Reactors

The existing research reactor facilities at Trombay can be used mainly

for the studies in the field of:

i) Fuel Development,

ii) Material testing,

iii) Coolant Chemistry and

iv) Studies in Apsara and ZERLINA

i) Fuel Development

a) Facilities

Mainly the six vertical holes and the central thimble available in the CIRUS reactor, as shown in figure-1, are of immense use to this programme. In

position N-I9» at present, we are installing a "Pressurized Water Loop".

Table-1 gives the details of this loop while figure-2 gives a simplified flow

sheet. Appendix-1 gives a brief description of the loop.

The main loop was designed and supplied by the Atomic Energy Canada

Limited; while the auxiliary facilities were engineered by our organization.

A detailed hazard evaluation of the loop was carried out by us. The hazards

evaluated include the effects ofi

a) equipment failures

b) control failures

c) sequential power systems failures

d) loss of coolant flow

e) loss of coolant

f) in reactor failure of the pressure tube

g) failure of the jacket cooling and the safety of thereactor calandria tube.

The analyses indicated that with no jacket cooling (i.e. loss of

cooling water), the loop operating at the rated conditions of temperature,

and the reactor moderator dumped, the reactor calandria tube temperature is

likely to differentially rise to a stage where it might damage it. As a

safeguard, provision has been made to automatically bring down the loop coolant

temperature to a safe level in a short time. Also the jacket cooling water

system has been connected to the reactor emergency cooling system.

It is obvious that the reactor and the loop by themselves are not

enough to carry out a fuel development programme. The programme in addition

calls for

1) Facilities for the fabrication of test fuel. elements of different enrichments. At present

we are contemplating putonium enrichment for test irradiations.

2) Post irradiation handling and transport equipment and facilities.

3) Hot cells for metallurgical and radio-chemical examination.

For obvious reasons we have chosen to use plutonium enrichment for

test irradiations. We have commissioned our facilities for the fabrication

of mixed oxide fuel pins with stainless steel cladding. The standardization

of procedures for fabricating the pins to the required specifications either

by selective assembly procedures or otherwise has been accomplished. We are

also in the process of equipping the facilities for fabricating Zircaloy clad

fuel pins.

The suspension assemblies, for irradiation of the test specimens in

the loop have been designed and fabricated. The procedure for irradiation and

the loop operation have been decided.

The post irradiation handling equipment calls for procedures and

equipment for the withdrawal of irradiated test pins from the loop, facilities

for post irradiation cooling prior to transportation and transport facilities.

Such facilities for handling irradiated assemblies from the pressurzied water

loop have been designed and fabricated.

For carrying out the post irradiation examination we are commissioning

the hot metallurgical cells. These will be available soon. There are six

hot cells for carrying out the meta,llurgical examinations providing a total

floor area of about 50 square meters. These are provided with the necessary

support facilities. Procedures for loading, unloading, material flow and

testing operations have been worked out. The examinations that can be carried

out are shown in Table-2.

The above facilities in short cater to the needs of loop irradiations.

In addition we have worked out a conceptual design to use some of the vertical

10 cms holes in the CIRUS reactor to carry out capsule irradiations. These

locations are well suited for such irradiation set ups as,, the reactor coolant

itself can be used as the coolant for the specimens and the instrument leads

can be taken out conveniently from the top. The experimental set ups would

be similar to the basket facilities in ETR and MTR. The facilities in the

CIRUS reactor are planned to be used mainly for irradiation of test fuel

specimens to study swelling of fuel elements (both restrained and unrestrained)

fission gas release from the fuel etc., for irradiation of structural materials

at various temperatures and the study of realtor coolants. For the instru­

mentation during irradiation, particularly for fuel and clad temperatures, and

the fission gas pressure, it is most desirable to study the detailed behavious

of test specimens under irradiation. The instrumentation of the test specimens

is being planned at a later date, even though this xvould increase the recurring

cost of the experiments.

b) Fuel Development Programme:

Using the above facilities we have chalked out a consolidated fuel

development programme to cater to our immediate needs. Though we have a

working design available with us for the CANDU type reactors we have to keep

up with the changing needs and technology and to equip ourselves to answer the

outcome of operational experience.

Our immediate programme includes loop irradiations (on low enrichment,

up to 4$, mixed UC^- PuC^ samples). The enrichment is added to accelerate

the tests and to achieve the required ratings. The irradiation programme

should be geared to give sufficient data to formulate theoretical models for

design codes. The irradiations should cover standard operating as well as off

standard conditions. The total fuel development programme engulfs the factors

as shown in fugre-3 while Table-3 shows the immediate development programme,

utilizing the "Pressurized Water Loop" in the CIRUS reactor.

To supplement the in-reactor irradiation programme, we also have an

out-of-pile testing programme. We are also developing the fuel design computer

codes. Some of these codes are operative and the rest are being made. The

main points of study requiring a deep understanding in the fuel behaviour are:

1

2

3

4

5

67

89

10

11

1213

14

15

Fuel relocation

Fuel slumping

Plutonium migration or seggregation

Fuel plasticity

Swelling and grain growth

Long terra creep of fuel pins

Dimensional stability

Fuel structure, crack formation and propagation

Void migration

Fission gas release and migration

Fuel ratchetting

Gas plenum temperatures and methods of reducing it

Ridge formation

Long term corrosion and

Effect of fluence on Zircaloy-2 mechanical properties.

ii) Material Testing:

We mainly intend to study the behaviour of clad material, Zircaloy-2,

being used in the CANDU type povjer reactors. Zircaloy-2 so far, has been tirell

established as a cladding material for pressurized and boiling water cooled

reactors. It is still essential to study and understand fully the "behaviour

of zircaloy-2 under various conditions of coolant chemistry, heat flux,

temperature and irradiation. This calls for study of aspects like texture,

hydride orientation, effect of heat flux, temperature and other parameters

especially from the point of view of fracture mechanics. The inpile irradiation

study of clad materials should include:

a) the effect of fuel clad clearances

b) fuel clad interaction

c) the ridge formation

d) the inpile corrosion

e) effect of the water chemistry

f) the effect of irradiation on mechanical properties(creep, fatigue, nil - ductility transition)

Zr - 2 1/2'fo 1Tb alloys hold promise as the pressure tube material. Following data on the in reactor "behaviour is desirable.

a) mechanical properties

b) creep data

c) corrosion properties

The Nuclear Fuel Complex in India will be producing Zircaloy-2 ,

pressure tube. The plant can be utilized for the production of Zr M Nb

pressure tube.

Even though irradiation of the pressure tube (or any structural

material) in CIRUS to a desired level will take a long time, it would be

possible to check the material behaviour after a reasonable irradiation.

However, the Fast Breeder Test Reactor, under design in India, will provide a

better facility for in-reactor material testing of reactor structural materials.

iii) Coolant Chemistry

The water chemistry conditions in a water (pressurized or boiling)

cooled reactor can affect the corrosion behaviour of various components as

well as the behaviour of carrying the activity from the core to the other

components. The study of the behaviour of crud formation on the fuel pins

is of great significance. The coolant chemistry studies programme would

generally include î

a) Methods of water chemistry control,

236

b) Development of techniques for accurate and reliable analysis of samples

c) Studies on the kinetics of reactions of various system component materials including fuel clad under various water chemistry conditions.

Also attempts may be made to develop on line instrumentation for conductivity and pH. Purification systems suitable for high temperature operation could be of great interest.

Various out of pile experiments will have to proceed before inpile experiments on water chemistry are conducted. At present we are studying the methods of water chemistry control and corrosion of zircaloy-2, monel etc. in two out-of-pile water loops. Both loops are designed for 100 bars and 280°C. One of the loops is mainly used for the out of pile acceptance test of the Rajasthan Atomic Power Station fuel bundles and the other for conducting boiling heat transfer experiments.

The inpile experiments are planned to be conducted in the Pressurized Water Loop in CIRUS, described earlier.

iv) Studies in Apsara and ZERLIMA ReactorsThe Apsara reactor was used for various studies on reactor instru—

mentation* irradiation studies on chemical compounds and other non-engineering studies. Por instance studies on the radiation effects on organic coolants (terphenyls) were carried out using one of the beam holes in the reactor.The studies were carried out by irradiating the samples in the temperature range of 200 - 400°C. This was achieved by using a high temperature >irradiation assembly with heaters. Necessary driving mechanisms, temperature control features and cooling arrangements.were provided in the assembly (2,3)«

In the field of reactor instrumentation considerable work has been . carried out by the Electronics Group on ionization chambers, burst fuel element detection, and solid state logic employing Apsara as a tool (2, 4» 5» 6, & 7)»

Void coefficient Measurements:

The design of a heavy water moderated boiling water cooled reactor requires a lot of experimental data on the void coefficient of reactivity.These studies can be carried out in the Zero energy ZERLINA reactor by simulating the voids and their distribution in the unit cell.

Poison Injection System

The present 200 MWe CAIflXJ type reactors adopt dumping of moderator as a safety measure in Case of a failure of the reactor regulating system. In the large stations of 500 MWe capacity it is difficult to achieve the required dumping rates. An emergency poison injection system is used to achieve the objective. Studies on the effectiveness of such an emergency poison injection system are planned to be carried out in the Zero energy reactor ZERLINA. The methods of injection, the rates of injection and their effectiveness are to be studied in detail.

Testing of materials and components under low radiation fields

In the CANDU type reactors, the components of fuelling machines and some of other equipment are housed in areas of low radiation fields. The suitability of the various materials and components such as hoses, seals, drive mechanisms etc. need to be established over their life time. The ■behaviour of such materials can be studied utilizing the Apsara and the CIRUS reactors.

Beferenoe»»

1, Canada India Reactor, AECL 1443

2, Utilisation of a Research Reactor t

10 years of Apsara

3, K. ÏJàrayana Rao et.al«

"Studies on the Pyrolytic and Badiolytic Stabilities of Organio Coolants”AEET/ÖD/20

4« Satyanarayana £, and Bao S.V.R."Ionization Chambers for Reactor Control*1

Paper 37o.CN 22/59» IAEA Conference onHuclear Electronics , Bombay, Hot. 22 - 26 (1965)

5» Prabhakar, B.S. et. al

"Design of an Electrostatic PrecipitatorMonitoring System for Ruptured Fuel Element Detection"»

Third U.K. International Conference on the Peaceful uses of Atonde Energy, Geneva, Aug« 31 - Sept« 9 (1964)

6« Kaaargod S,V*, and Bao K,R.,ttSemi-conductor logic for Reaotor Safety and Interlock Systems"

Report AJEET/SD/SQ/y8 (1963)

7* Kasargod S.V.,

"Solid State Logio for Reaotor Safety System”

Paper Fo. CU/22/40, IAEAConference on Nuclear Electronicst Bombay

Hwrember 22 - 26 (1965)

APPENDIX 1

The pressurized water loop^Fig.^is mainly intended to be used for testing various reactor fuels and to a lesser extent for the evaluation of mechanical components, instrumentation and studies on coolant.chemistry. The loop is installed to investigate the following aspects which affect the fuel elements operating in a power reactor.

a) To study the dimensional and structural stability of various fuel elements.

b) To develop corrosion resistant cladding materials for fuel elements.

c) To investigate the effect of dissolved gases in high pressure, high temperature water on the corrosion of various metals for use in various power reactors.

d) To develop pressure tube and other structural materials used in a power reactor.

The loop is being installed in one of the 10 cms experimental positions (N-19) in the CIRUS reactor. The loop is an installation in which the coolant is recirculated at a preset temperature and pressure over an experimental fuel stringer or any other testing material held in the in-reactor pressure tube or any where in the loop outside the reactor.

In the pressurized water loop water at high temperature (max.292°C)

and high pressure (max. 137 "bars) is circulated over the test specimens by­using two glandless motor pumps. The loop system includes high pressuré

circulating pumps, a surge tank for pressurizing the system (which also

accommodates the swells and shrinkages in the loop), a heater for controlling

the water inlet temperature to the test section, a delayed neutron monitor,

an in-reactor test section, a control valve and a cooler. The loop is

designed to circulate a maximum of 400 ltrs./min. of demineralised water at

a maximum pressure and temperature of 137 bars and 292°C respectively over

an experimental fuel stringer. The coolant is maintained at a pH of 9-10 to

minimize the release of corrosion products. The piping in the main loop system

is made of stainless steel type 347»

The heat from the main loop is transferred from the high pressure water

to dowtherm in the secondary circuit through the loop cooler. The loop has

a heat removal capacity of 400 Kw. The heat from the low pressure dowtherm

circuit in turn is dumped into a third circuit containing the low pressure water

which passes through the dowtherm cooler. Ultimately cooling is achieved "by using a spray pond in the third circuit.

The loop is also provided with auxiliary circuits consisting of

purification system, catch tank system, decontamination system, make up

water system, loop room cooling and ventilation system, purification and

sampling system.

The loop is provided with sufficient instrumentation and control

with triplication of the instrumentation wherever necessary to enable a

safe operation of the loop and to ensure the safety of the reactor. Through

the control system the reactor can be tripped under abnormal conditions to

ensure the safety of the loop and the reactor.

t a b l e - 1

C I R U S - L - 5 PRESSURISED WATER LOOP

MAXIMUM UNPERTÜBED THERMAL 13 2 : 5*5 X10 h /cm/sec.FLUX CAT 40 MW OPERATION OF

THE REACTOR) TEST SECTION 1. D. * 5.79 cmsPRIMARY LOOP COOLANT : DEMINERALISED WATERPH : 9 - 1 0OXYGEN CONTENT • LESS THAN 0*1 i»pmCOOLANT FLOW DIRECTION : UPWARDS

COOLANT FLOW RATE: NORMAL : 350 Ltrs/mlMAXIMUM : 400 Lti-s/m*

COOLANT TEST SECTION INLET 0TEMPERATURE s 270 C MAX.

COOLANT TEST SECTION OUTLET TEMPERATURE : 292° C MAX.

LOOP DESIGN PRESSURE * 17 1 Bams

LOOP OPERATING PRESSURE : 13 7 Bars

A P MAX. ACROSS TEST SECTION : 30*5 itvWi

MAX. HEAT REMOVAL CAPACITY OF THE LOOP : 400 Kw.

242

T A B L E - 2

POST IRRADIATION EXAMINATIONS

A. NON-DESTRUCTIVE INSPECTION

<*) VISUAL INSPECTION BY PERISCOPE

AND TELES C O P E.

Q ULTRASONIC INSPECTION,

c) DIMENSIONAL MEASUREMENTS

LENGTHS UPTO 48* WITH

t/64 QUICK READINGS.

DIAMETERS UPTO 2* WITH AN

ACCURACY OF ± 0 001;

B. DESTRUCTIVE EXAMINATION

a ) FISSION GAS COLLECTION.

b) DECANNING THE FU EL AND

EXAMINATION OF FUEL.

c) COLLECTION OF FUEL SAMPLES

TO DETERMINE BURN-UP ETC.

<L) METALLOGRAPHY.

MICROHARDNESS TESTING

DENSITY MEASUREMENT AND

ELECTRICAL RESISTIVITY

MEASUREMENT.

243

FUEL DEVELOPMENT PROGRAMME

EXPT. d e sc r ip tio n ano PIN tim e OF IRR. DATA TO BE POST IRR. EXAM. RESULTS EXPECTED ORN a PURPOSE INSTRUMENTATION N REACTOR COLLECTED R E M A R K S

MWD* DURWG IRR.

M AT UOxPM CANDU SIZE — . . 1000 OIMENSIONAL CHECK, MAMLY TO ESTABLISH ALL THE50 CvftLONC ONE PM EXAMINATION OF CLAD, TECHNIQUES OF POST «RADIATION

0-521 Cm O IA ) PLUG, WELDS ETC, EXAMINATION.

<$CALORMETRtC FISSION GAS RELEASE

DATA PUNCTURE TEST,

METALLOGRAPHY TOOEOOE CENTRAL

TEtffERATUPE, P » -

V) LOOP WATER DISTRIBUTION-ANALYSIS

CHEMISTRY FOR BURN UP-

2. SAME AS ABOVE WITH TWO « ) ONE PIN SURFACE 4 0 0 0 SAME AS ABOVE SAME AS ABOVE, PERFORMANCE OF SURFACE TEMP.PINS-W ITH ONE W STRU- TEMR ♦ PIN SURFACE THERMOCOUPLES-CONFIDENCE IN

MENTEO PIN $ FLUX MONITORS TEMP. EXTRA-POLATtON IN REASONABLE

(FOILS) LIMITS.

3. U0»-4% Pa 0 » SAME SIZE - d o - 4 0 0 0 SAME AS ABOVE SAME AS ABOVE + EFFECT OF HEAT RATING AND BURN

AS ABOVE DETAILED ANALYSIS OF UP ON METALLURGICAL STRUCTURE,

«9THREE PINS WITH VARY­ X« AND Kr + GRAIN GRAIN GROWTH AND SWELLING

ING INITIAL FUEL CLAD GROWTH AND SWELLING

GAPS. STUDY.

THER

MAL

C

OLU

MN

SCHEMATIC LA TTIC E DIAGRAM FOR C1R.

M »<eN « a o - n n t n « K c o i o ~ n n < t s « A ------------ -------------- — --------------N G t f l i t M N N I M O f O K M C . T .

©

FIG. 1

C E N TR A L THIM BLE

PRESSURISED WATER

L O O P .

IQ CM. VERTICAL HOLES

245

L 5 - PRESSURIZED WATER LOOP SIMPLIFIED FLOW SHEET

■f lG;2L

246

FUEL

CDNC.

DESGN

r-p-SftJGLE PIN «¡RADIATIONS-|_

CAPSULE IRRADIATIONS LOOP IRRADIATIONS

- JA7ROO CLUSTER IRR.LOOP IRRADIATIONS

*— t* BOD CLUSTER IRA . __ ] -------------

• SHORT RUPTURE TESTS

IRR. UNDER EXTREME

OPERATING CONDITIONS

• OPERATION WITH CONTROL

MELTING

~£\*LUAT?ON FOR SEFETY

UNEAR POWER RATINGS

* ( « CONST. OPERATIONAL

- ( « ) SHORT TRANSIENT PEAKS

- REQUIRED BURN UP RATINGS

• ADDITIONAL ASPECTS TO BE INVESTIGATED

— MECH. INTEGRITY

—AGGRE VATlON OF TRANS£NTS DUE TO SLUMPING

. — CATASTROPHIC SWELLING

- DISPERSAL t REACTIVITY PROPERTIES OF FUEL

RELEASED TO COOLANT

r— (?) VARIOUS POROSITIES ft DISHINGS

— <D VARIOUS FUEL CLAD CLEARENCES

— © COLLAPSIBLE VtSELF STANDING CLAD

— @ CRUO FORMATION ON CLAD SURFACES DUE

TO VARIOUS WATER CHEMISTRIES

EVALUATION--flST-

p— © INPftX THERMAL CONDUCTIVITY MEASUREMENT»

— Q MPtLE ÖURNOUT TESTS

— (D SHORT TIME IRRAOIAHON OF PINS IN LIQUID

OEFtOENT REGIONS

STEAM * WATER REACTIONS WITH ZRY AT WGH

TEMPERATURES MAGNITUDE AND DURATION

OF RESULTING PRESSURE PULSES

ACTIVITY CARRY TO TURBINE WITH NORMAL

ANO RUPTURED P« FUEL ELEMENTS

r— <§> FUEL RELOCATION

— <g) SLUMPING

— © Pu MIGRATION OR SEGGREGATION

— FUEL PLASTICITY DUE TO THE

PRESENCE OF P»

SWELLING AND GRAIN GROWTH

— 0 LONG TERM CREEP

-Q DIMENSIONAL STABILITY

__ (g) FUEL STRUCTURE CROCK FOR

MATION AND p r o p a g a tio n

— VOI D m ig r a tio n

__Q) OPERATING TEMPERATURES

— ($ ftSSON GAS RELEASE ft MK5RATION

— FUEL RATCHETTMG

— O GAS PLENUM TEMPERATURES

ANO METHODS OF REDUCING IT

— ® RIDGE FORMATION

__ Q LONG TERM CORROSSION

— ® EFFECT OF FLUENCE ON ZRY

MECH. PROPERTIES.

DATA OBTAINED FOR

OUT OF PILE

EXPERIMENTS AND

DESIGN CODES

AS A FEED-BACK

FIG.3 IN PILE DEVELOPMENT

to£>■-J

HOSE COHN.

DECONTAMINAN! PURGE PUMP

L 5 -P -1 0

----!SAMPLE {

FAST FILL PUMP jrL S -P -5 ^

DISPOSAL AREA

* /f f t PROCESS AIR LINES------------------M AK E-UP S Y S TE M L IN E S-------------------CATCH TANK S Y S T E M Llf lES

-frfa------- VALVE WITH EXTENSION HANDLE

S C A -£ -/ / .T ¿ .

LRçln ~ F-S05Cm$2F.

■ewus SERVICE w a te r

L-5 P R E S S U R IZ E D WATE.R LD DP S IM P L IF IE D FLOW S H E E T

ù k N . \X.Hk 'D.\a PP'D ;

\r>M.S. S.

FIGURE 4

POWER UPGRADING OP THE TRIGA MARK II REACTOR

PROM 25O kW TO 1000 kW

by

Soetarjo Soepadi,Ijos Subki, and Karsono Linggoatmodjo

Bandung Reactor Centre Indonesia

Abstract

The current utilization of the TRIGA Mark II reactor and its future programme are described» Upgrading the reactor power to 1000 kW, will increase its capacity to supply isotopes for various applications and significantly support the research and engineering programmes related to the reactor.

The major modifications to the reactor as well as the phases of the upgrading are also described. The modifications, installations, criticality and commissioning tests will be executed by the local staff.

1. INTRODUCTION

The Triga Mark II research reactor at the Bandung Reactor

Centre, National Atomic Energy Agency, has been in operation for

more than 6 years. Since its first criticality and commissioning

in October 1964, the reactor has been available for operation up to

25O kW maximum power.

The reactor has been utilized for isotope production, research

and training purposes. The current applications of the reactor

have been briefly described in reference /l/.

The present reactor operating power level will not be sufficient

to supply the foreseen isotope demand for various applications. More­

over, the present power and neutron flux could not properly support

the neutron inelastic scattering research as well as the engineering

programme related to the reactor.

Feasibility studies were conducted and it was concluded to

upgrade the reactor to 1000 kW with no appreciable changes to the

reactor systems.

II. MAJOR CHANGES FOR UPGRADING

The programme to upgrade the reactor from 250 kW to 1000 kw is

aimed at obtaining a higher flux to accomodate new research using

neutron beams including engineering research, but with the following

constraints in mind, that the cost of upgrading as well as the reactor

operating costs would be minimal.

This requires at least the following changes to be undertaken.

Core and core components

The present fuel of the TRIGA Mark II reactor is aluminium

cladded UZrH^ q fuel, which is limited to a steady-state operating

power of 25O kW. The reason for this is that UZrHn _ has a phase1 *u

transition at approximately 530 C. In order to avoid any dimensional

changes of the fuel which may occur at this phase transition tempera­

ture, the reactor is limited to operation at a maximum power level

of 25O kW.

For reactor operations above 250 kW, a new stainless steel

cladded UZriL , core would be required. This fuel has a singleo

phase up to temperatures above 1000 C. As a result, the operation

is not limited by a phase transition such as the UZrH^ q fuel.

By retaining the present F-ring grid plate with 85 fuel positions,

the operating flexibility as well as the irradiation positions would

be more limited as compared to the use of a G-ring plate. Therefore,

we will install a new larger G-ring plate,/'wl®^ will be able to accomo­

date 121 fuel positions. The use of this grid has the following ad­

vantages: it permits the reactor to be run fou an extended period of

time and additional fuel elements can be added as required without

necessitating the removal of fuel elements, the larger fuel-to-fuel

distance in the central positions (B-ring) permits better cooling,

this hexagonal central section in addition to two groups of 3

elements are removable allowing larger samples to be inserted in

in-core positions.

The use of 4 fuelled follower control rods (FFCR) will permit

a smaller compact core arrengement, and assist further in extending

the core lifetime.

It is also necessary to replace the existing reflector with

a reflector having a larger inside diameter so as to permit the

installation of the larger size grid plate. This new reflector has

been designed to permit the utilization of the existing rotary

specimen rack. *

Cooling systems

The 1000 kW reactor will still use natural convection cooling,

but operation at powers higher than 1000 kW will need forced cooling

to piieclude incipient boiling in the core.

The new primary cooling system will use 4” diameter aluminium

piping with a 15 hp stainless steel pump to drive a flow rate of the

order of 350 gpm, while the secondary system will use 6 M steel piping

with two 20 hp pumps in series, and a new 1000 kW cooling tower which

will also be installed. The system will keep the bulk coolant tem­

perature below 45° C.

Control system

As has been mentioned earlier the use of fuelled follower control

rods will be quite advantageous. It will then onty require an additional

control rod drive and the corresponding position indicator on the con­

sole.

Other modification to the console will include the adjustment

of the linear, logarithmic and percent power indicator to the 1000 kW

full scale.

Adequate preparation and procedures have been laid down to

assist the personnel in the execution of the upgrading programme.

The outline of the operations will be described.

Removal of TRIGA Fuel

The irradiated fuel in the tank will be removed to the bulk

shielding facility (temporary storage) using the shielded fuel

transfer cask. Special care should be exercised to eliminate any

possible over-exposure to the personnel during the fuel removal and

to prevent any possible criticality in the temporary storage.

Dismantling of basic reactor components

The basic reactor components should be subsequently dismantled,

these consist of: central channel assembly, control rod drives, pneu­

matic system, rotary specimen rack, central channel plate and central

thimble.

The reactor pool water will then be drained through the waste

treatment tank, a careful survey will be exercised since radiation

still exists due to activation of reactor components. Then the

following procedures will be executed: dismantling and lifting of

the piercing beam port!s bellow assembly and reflector. All active

components will be placed in the storage pit.

Dismantling of the cooling system

The preparatory work will proceed with the dismantling of the

primary cooling system, the purification system, and the secondary

cooling system with its existing cooling tower.

The upgrading phases will now be started with a thorough in­

spection of the reactor tank?surface, welds, beamports, and the

thermal column to check any damage or crack.

Core assembly

The core reflector assembly will be assembled outside the

reactor tank, this work will cover installation of the ion chamber

guide tubes and securing the top and bottom grid plates. The re­

flector assembly will then be lowered into the tank and set on

the tank bottom. The top grid plate will be levelled, and the re­

flector assembly port aligned with the piercing beamport. The core

assembly will then be completed after the execution of the bellow

assembly leak test.

Cooling systems

The piping for the primary as well as secondary systems will be

installed without any welding, since the piping will be coupled

together using flexible connections. The basic diagrams of the pri­

mary and secondary systems are shown in Figs. 2 and 3 . The installa­

tion of the cooling systems will be completed with watertightness and

flow rate tests.

Control systems

As has been mentioned earlier there are no major changes in the

control systems except for the followings use is made of 4 fuelled

follower control rods instead of the 3 existing control rodsj the

linear, logarithmic and per cent power channels will be adjusted to

the 1000 kW full scale and an additional position indicator will be

placed in the control console.

Pre-operational and operational tests

Prior to the criticality experiments all the control channels,

safety systems, interlocks, and control rod drives will be

thoroughly checked out. The neutron source for reactor start-up

is a 3.56 Gi Am-Be source.

After the criticality experiments the reactor will then be

loaded with 82 stainless steel clad fuels to achieve an excess re­

activity of around $ 7»0

Control rods calibration will then proceed to obtain the

calibration cruves of the rods and determine the available excess

reactivity.

Power calibration will be followed step by step at power tests

up to 1000 kW, the instruments linearity, and the reactor tank

water temperature rise will be observed; and health physics sur­

veys around the reactor deck, the reactor bridge, the piping system

and the demineralizer should be carried out. The particulate air

monitor should also be on line throughout the tests to detect any

possible fuel leak.

Since the reactor will be capable of operation at 1000 kW,

more attention should be given to reactor safety aspects to limit

any possible hazard to the surrounding and to the personnel.

Accordingly, new operating procedures and regulations are being

adopted.

17. SUMMARY

The power upgrading of the TRIGA Mark II reactor at the Bandung

Reactor Centre from 2^0 kW to 1000 kW has started with the prepara­

tion of procedures and necessary tools to expedite the execution

of the programme. Prior to this activity close communication with

Gulf Energy & Environmental Systems (GE & ES) has been conducted in

order to obtain technical information and to procure the necessary com­

ponents .

The upgrading programme will be executed by the local staff

and it will be completed in approximately six weeks time.

The upgraded reactor will not pose any special safety problems,

major probable accidents and their consequences have been evaluated

in the Reactor Safeguards Analysis Report.

References

/l/ Ijos Subki et al.,

"Six Year Operating Experience with the TRIGA Mark II

Reactor at .Bandung Reactor Centre"

Proceedings of the IAEA Study Group Meeting on Research

Reactor Utilization convened in 1971 in Bandung.

Irradiation Space

Source

IternateSoureeLocation

Irradiation¡pace

PneumaSystem (Rabbit)In - core Terminus o Standard Fuel

Element

Graphite Dummy Element

G-ring grid plate

Control Rod

TEMPERATURE

s «PILL

PUMP

X J

TEMPERATURE - ®p i ™ i

.WATER SURFACE SKIMMER

l-Dí XVENT —

sHOLES

DIFFUSERCONE

r.1- RES» j SUPPLY

V I CONDUCTI

©20 a m ,l-l n r

VITI CELL ]

u 'f i r L ------- !— iyjEMINERALlZERI

T *> I CONDUCTIVITY CELL |I • — ----1- ■■■ -II____ J (i- iRESINDISCHARGE

REACTOR

350 gpm

I

-JÄJ-

FLOW METER

vept ff.ll

i

CUSTOMER* IFURNISHED ,

-fCTi

! STORAGE TANK | I (OPTIONAL)

110° F

AHEAT EXCHANGER

20_£ CUSTOMERFURNISHED

I. SECONDARY ,

COOLING , 80 Pj__SYSTEM I

90* F

J--------1 I f-

350 GPM ^ 4 IN

ríaFI*. 2

L E S E N D --- ---

THROTTLING VALVE

s h u t - Off v a l v e

^ o r i f i c e

^ PRESSURE GAGE

LBQBKD 1. Fan Cooling Tower 2» Heat Exohanger 3. Pump 4« Valve

n * .

Secondary Cooling System

byLibrado D. Ibe

Acting Commissioner Philippine Atomic Energy Commission

andGuillermo C. Corpus

Head, Reactor Operations Department Philippine Atomic Research Center

ABSTRACT

The Philippine research reactor PRR-1 became critical in 1963. After one year of testing and low-power operation, the reactor was .brought to full power of one megawatt. The PRR-1 is utilized for radioisotope production, sample irradiations, physics experi­ments and training of personnel. Some operational difficulties en­countered are also discussed.

INTRODUCTION

The Philippine research reactor PRR-1 is an open-pool type

facility designed for an initial operation up to one megawatt thermal

power level. It is operated by the Philippine Atomic Energy

Commission. The reactor facility is located within a 9-hectare lot

inside the campus of the University of the Philippines in Diliman,

Quezon City.

The PRR-1 reactor core is made up of 30 MTR type full ele­

ments arranged in a 5 x 6 array and completely surrounded on the

sides by graphite reflector elements. Four 10 5 /8 inches wide

boral coarse control blades virtually divide the core into three

blocks - a central block of 3 x 6 array and two side blocks each, of 1 x 6

array. For fine and automatic control, a regulating rod is provided.

Radiation baskets, of the game s iz e and shape fuel and reflector ele­

ments, are also provided to permit irradiations in the reactor core

and reflector regions. Other in-core experimental facilities include

two 2 -inch vertical dry-pipes.

The reactor assembly is submerged in an aluminum-lined

concrete pool and supported by a suspension frame attached to a

movable bridge mounted on top of the pool parapet« The pool is

divided into three sections - a high power section where reactor

operation up to one megawatt is possible, a transition section and a

low-power section where the reactor could be operated up to 100 KW

only. Draining of the low-power or high-power perol section, could

be accomplished with the aid of a portable bulkhead, gate and suitable

pumps.

Most of the fixed experimental facilities are located in the high

power pool section. These include six radial beam tubes, two pneu­

matic tube systems and a graphite-filled thermal column.

Initial Operation of PRR-1

The construction of the reactor building and reactor pool was

started in late 19él. When the structure was almost complete, the

facility major equipment were brought inside and installed at their

respective places. The assembly of the reactor core components

took place immediately thereafter. On August 26, 1963, the PRR-1

became critical for the first time.

During the whole period of construction, radiation surveys

to establish background levels were conducted at various stations

around the reactor site up to a radius of 10 m iles. Radiation moni­

toring was continued even during reactor operation at regular inter­

vals. The results showed absolutely no increase in the environmental

radioactivity levels in all areas outside the reactor facility compound.

The first year of reactor operation after attainment of initial

criticality was mostly utilized in the instrument checkout, equipment

performance tests experimental determination and verification of

reactor parameters, core flux mapping, and reactor operator train­

ing. During this period, reactor operation was kept raised at

practically zero power level and later on to not more than 100 KW.

Utilization of PRR-1

In August 11, 1964 the reactor power level was brought to

one megawatt. Since then, the PRR-1 has been operated at different

power levels up to one megawatt to meet the demands and requests

for (a) production of radioisotopes, (b) irradiation of samples,

(c) physics experiments and (d) training courses.

Two of the experimental beam ports are presently tied up

with the two neutron crystal spectrometers used for the physics

experiments. These spectrometers were initially utilized in the

India-Philippines-Agency (IPA) project on neutron spectrometry.

This project lasted for five years and was terminated in late 1969.

The PRR-1 is further utilized in the training of local scien­

tists and technologists in reactor engineering. These trainees

consist mainly of university faculty and technical personnel of

electric power utilities. A specific course on reactor operation

was recently conducted for the engineers of the Manila Electric261

Company (MERALCO). MERALCO, the largest private electric

utility in the country, has shown active interest in putting a nuclear

power plant in the very near future.

On« important investigation that was made possible with the

use of PRR-1 is the verification of the U-235 contents of the 20 new

fuel elements for the second-core received from the fabricator* s

shop in the United States. By placing a new fuel element at a

specified position in the reactor core of given configuration and

measuring the change in resulting reactivity, the corresponding

U-235 content of said element was calculated. With this method,

the experimental values obtained for all the new fuel elements

were found to he very close to the figures supplied by the fabrica­

tor, the discrepancies ranging from 0. 2 to 2. 5 per cent. These

results made us conclude that no appreciable impurities with

"poison** characteristics have been introduced in the fabrication

of said fuel elements.

Operating Problems Met

The performance and utilization of PRR-1 has so far been very

satisfactory, with no incidence of any nuclear accident. Its almost

perfect operation is marred only by occasional unscheduled shut­

downs due to equipment failure, instrumentation difficulties, and

power outages. Some of the operational difficulties encountered

are:

1. Rattling and/or failure of the heat exchanger tubes;

2. Vibration of control blades;

3. Abrasion of detector cord;

4. Failure of the compensated ionization chambers;

5. Dropping of control blades without an instrument scram;and

6 . Power measurements.

Failure of Heat Exchanger Tubes

The heat exchanger used in the reactor cooling system is a

1-Z shell-and-tube type designed to cool 170© gpm of primary water o °

from 114 F to 110 F. It contains 266 3 /4 - inch aluminum tubes bent

in a U-shape form with the ends welded to a single tube sheet which

is an integral part of the primary water manifold. The tube bundles

are supported throughout their lengths by the tube sheet and 4 equally-

spaced cross baffles. A 120-ft head centrifugal pump circulates

1700 gpm water in the primary side and another pump forces 1250

gpm on the secondary side outside the tube bundle. During the

testing of the heat exchanger for the calibration of the primary

and secondary flow instruments, the following observations were

noted:

a. With no flow on the primary side, the equipment started

to rattle at a secondary flow of about 500 gpm. The

rattling intensity increased as the flow rate was in­

creased.

b. Although the secondary flow rate was kept steady, the

sound came on and off with no apparent regularity. The

bursts of sound were generally short and abrupt. Longer

and louder bursts were heard at secondary flow rate of

about 1300 gpm.

c . With the secondary shell drained completely and prim ary

•water allowed to flow into the tubes, no rattling noise

was heard.

d. With flow in the primary and secondary sides, the same

observations in (a) and (b) above w ere again noted.

The cause of rattling was pinpointed to the tube bank itself. Upon

inspection, the U-tubes were found to be loosely supported at the

baffles and could vibrate. In addition, the tube bundle itself seemed

to have been factory-installed 180 out of alignment with respect

to the inlet and outlet secondary water connections. On the sus­

picion that said factory defect could be the principal contributory

cause of the vibrations, the heat exchanger was reassembled witho

the bundle rotated 180 # After one-and-a half years of operation,

the tube bundle has to be pulled out again due to intermixing of

secondary and prim ary water. Six (6) U-tubes on the side closest

to the secondary inlet opening were found leaking, establishing

the failure caused by repeated pounding of the tubes on the baffle

holes. These tubes were subsequently plugged. To minimize

further damage to the other tubes, a rubber band was wrapped

around the bundle to minimize their movement in the baffle holes.

As there was no access to the inner tubes, the rubber band could

only protect the peripheral tubes. Operation of the heat exchang­

er was thereafter relatively more quiet, but 3 years later a tube

situated 3 layers inside the bundle failed again due to vibration.

It appears that the rubber banding of the outer tubes helped to

some extent in minimizing tube failures.264

Vibration of Control Blades

Another vibration problem encountered is that of the four

control blades whose movements are guided by an aluminum

shroud. Each blade is 3 /8 n thick, 10 5 /8 " wide x 54 l / 8 " long.

The shroud is made 1 / 4 inch wider to give a clear 1 / 8 inch space

on either face of the blade. Trouble starts when reactor control

is on automatic mode at power levels in excess of 100 KW. Under this

condition of operation, the primary coolant flow of 1 7 0 0 gpm causes

the four blades to vibrate within the shroud and results in power

fluctuation. As initially set, the servo system, corrects the power

thru the movement of the regulating blade at power variations ex­

ceeding 2%. This condition caused too much strain on the regulating

blade drive motor which had to keep on changing its direction of

rotation as many as 50 times a minute. The motor temperature ex­

ceeded allowable lim its and the blade drive actuating relay suffered

frequent breakdowns. As it was not deemed advisable to reduce the

free space between the control blades and the shrouds since the

emergency dropping of the blades by gravity may be affected, the

remedial measure resorted to was to adjust the automatic response

of the servo system to ^ 4% of operating power. For irradiations

requiring very close degree of neutron flux control, regulating

blade operation is done in the manual mode.

Sealing of Pool Bulkhead Gate

The reactor pool has been designed and constructed so that

an aluminum bulkhead gate with a pheripheral rubber seal could

be installed between any two adjacent pool sections for dewatering

265

a section of the pool. Before a section could be dewatered, the

rubber seal must first be made to press tightly against the gate

seat after which a 20-gpm portable centrifugal pump gradually

pumps the water out. Establishment of the water seal for evacua­

tion of the high power pool section is easily accomplished by utili­

zing an 8-inch prim ary coolant pipe to discharge by gravity flow

the high power pool water into a 15,000-gallon retention tank. The

dewatering of the transition and low-power pool sections, however,

is not as easy. There is no big drain pipe in these sections which

could be utilized to discharge water fast enough and establish the

necessary seating head. In order to effect the sealing of the bulk­

head gate, the water level in the pool is first lowered by about one

foot, then a gate valve in the 8-inch return pipe from the pool to a

3000-gallon holdup tank is closed to prevent pool water from flow­

ing back into this tank. The 1700 - gpm primary coolant pump is

switched on for about 10 seconds to suck water from the holdup

tank and discharge into the high power pool section. The large

volume of water that goes into this pool section provides the dif­

ferential head to seal the gate. A portable pump pool section to be

drained is subsequently operated to complete the dewatering opera­

tion. To avoid collapse of the holdup tank due to negative pressure,

all drain and vent pipes connected to it are kept open and serve as

passages for the on rushing replacement air. Although the scheme

has worked successfully several times in the past, the operation

is still not totally satisfactory in view of the strain, on the holdup

tank. Another method whereby a portable tank in the section to266

be dewatered will be pulled up by the existing overhead crane for

the establishment of the required sealing head is under considera­

tion* In addition, the 20-gpm pump is planned to be replaced by

a bigger semi-portable dewatering pump.

Abrasion of Detector Cords

In the PRR-1, there are two startup channels with movable

neutron detectors for a wider coverage of startup power. A s power

increases the detectors are raised towards the region of low neutron

flux. The BF 3 startup detector is moved by remote control from

the control room while position of the other detector (fission chamber)

is adjusted manually. After five (5) years of operation, the BF 3

channel failed due to abrasion of the flexible cord connecting the

detector to the external terminals at the top end of the core suspension

frame. The cord insulation was damaged by the constant rubbing of

the cord with the detector drive stem and against the housing. To

correct this defect, the detector cord was made to pass thru a small

stainless steel pipe and both detector and pipe caused to move as a

unit. A hole on the top of the suspension frame guides the pipe in

proper position. With this arrangement rubbing of the cord against

the drive stem or housing was eliminated.

Failure of CIG*s

Two of the compensated ionization chambers that send power

signals to the two safety channels and the log N channel have already

failed due to water leaks into the chamber terminals. Apparently,

the 25 feet head of pool water acting on the submerged cable and

plastic seals is high enough to force water into the chamber. When

the first CIC was damaged, the installation was modified by screw­

ing the chamber onto a long dry pipe. This pipe extends beyond the

pool water surface and the cables were ran through this pipe.

The field-repaired CIC lasted for about two years and then

failed again. It was decided to return it to the manufacturer for

factory inspection and repair.

Early this year, another CIC also failed, giving symptoms

similar to that of the first CIC. In order to ascertain the real

cause of trouble and because of the urgency of the situation, it was

decided to open the chamber in the field and try to make the neces­

sary repairs, if possible. Our findings show the chamber failure

to have been caused by the cracking of the ceramic insulator at

the cable connectors. A substitute sleeve assembly is being fab­

ricated at present and, as soon as reconnections are made, field

welding of the aluminum chamber housing will be done. In the

meantime, the reactor operation up to 500 KW is allowed with the

use of the remaining two CIC*s in the safety channels and the BF 3

detector utilized for period indications.

Dropping of Control Blades

We experienced some incidents where all four (4) control

blades would just drop without instrument scram, despite the

cleanliness of the magnet armature in contact with the electro­

magnet. The cause readily escaped detection due to its transient

nature and there is no visible indication in the annunciation panel

except the "blade disengaged” light. At one instance, the same .

event happened and, in addition to the "blade disengaged" light,

the'fcoolant gate" light also went on. This made us to suspect

that the 24-volt D. C. circuit serving the coolant flap gates was

at times momentarily noisy and would cause the dropping of the

blades without a scram. These flap gates, installed one each

on the inlet and outlet primary coolant pipes, are connected to

float switches. They are normally half-closed but are fully

opened when the coolant flow is about 1700 gpm. The setting of

the float switch could be unstable especially since the movement

from half-closed to fully open flap gate position is extremely

small. To remedy the situation, the flap gates were re-adjusted.

Power Measurement

Due to the semi«enclosed nature of the primary cooling

system of the reactor, no reliable power measurement could be

made from temperature changes of the coolant. Besides, the re»

actor power is too low for thermal power measurements. In order

to provide a neutron-flux-independent power instrument, the gamma

sensitive detectors mounted on the thermal column and in the pri­

mary cooling pipe were calibrated against reactor power. At full

power operations, reading of these remote area monitors were

found to be constant. Hence the instruments are used regularly

to check the neutron flux power.

CONCLUSION

The current experience with PRR-1 is believed not unique

to the Philippines but may be typical for a developing country.

269

The reactor facility not only provided with a modern tool for

research but also afforded opportunities for the advancement of

science and technology in the country. Perhaps the greatest

accomplishment of PRR»-1 is in the conditioning of the public

mind regarding the safety of an operating reactor. The PRR-1,

once feared, is now practically accepted by our people. This

experience could influence to some extent the eventual establish­

ment of nuclear power plants in the country.

SIX YEARS OPERATING EXPERIENCE WITH THE

TRIGA MARK II REACTOR AT THE BANDUNG REACTOR CENTRE

by

I. Subki and K. Linggoatmodjo

Bandung Reactor Centre Bandung - Indonesia

Abstract

The utilization, operation and maintenance of the TRIGA Mark II reactor at the Bandung Reactor Centre is described. The reactor is utilized for isotope production and its research applica­tion is increasing steadily every year. The main reactor opera­tion problems come from instrument failure or from the rotary specimen rack. No safety hazard to the surrounding population and personnel has occurred since its first criticality in 1964*

The total integrated reactor power up to the present is about 2 x 10 kW-hours.

I. INTRODUCTION

J The TRIGA Mark II reactor at the Bandung Reactor Centre,

National Atomic Energy Agency, has been operating regularly since

1965. This reactor uses aluminium cladded cylindrical fuel, con­

sisting of 8 Wt io uranium, 20 $ enriched in U-235» homogeneously

mixed with zirconium hydride as moderator. The MARK II reactor is

of the above ground and fixed core type, with four beamports, one

thermal column and one bulk shielding facility, available for various

experiments.

The first criticality was achieved in October 1964» and no

major troubles were encountered during the execution of the tests.

After the calibrations and the power tests were completed, the reactor

has been available for operation up to 250 kW maximum power level.

Up to the present (l5 June 1971) the reactor has logged a total

integrated power of about two million kW-hours or 83.5 MW-days. '

,-1,'he Physics Division has been using the tangential beamport for

neutron diffraction studies with a crystal spectrometer, which in­

cludes crystallography, magnetism and alloysystems. This project

has picked up a great momentum since the arrival of our physicists

from their training abroad in 1969• Future upgrading of the reactor

power to 1000 kW will enable the execution of meaningful research in

inelastic scattering using a Beryllium Detector System.

The Reactor Physics Division is working on the measurements of

reactor data, in-pile dosimetry, flux mapping and establishment of

power and control rod calibration methods, all of which have been

quite useful in the safe operation of the reactor. Work on reactor

noise in the frequency and time domain is in progress.

The reactor has been used by the Chemistry Division for research

in hot atom chemistry and activation analysis. Hundreds of

samples of crude oil. tin ores, fertilizers, hairs of kwashiorkor

children^, and other mineral ores have been irradiated in the

reactor for analysis.

A group of researchers from the Biology Division uses one

beamport for the irradiation of drosophylla and various insect

pests. Their work is directed towards integral pest control.

Another group of researchers uses the bulk shielding facility

for thermal neutron irradiation of seeds and seedlings, and the

Standard TRIGA Irradiation Facility (STIF) in the thermal column

for fast neutron irradiation in the mutation breeding project.

The use of the reactor for training purposes has increased

every year. The Reactor Operator's training has been conducted

four times, most of the trainees consist of members of the local

staff and students of the Bandung Institute of Technology. A

course in radiochemistry and neutron activation analysis was con­

ducted in 1970, staff members from various companies and local

staff as well as chemistry students from the Institute attended

this course. A course in reactor technology is regularly conducted

to upgrade the operation and reactor physics staff; it also serves

as a part of the graduate study at the Institute of Technology.

By far the most intensive use of the reactor is in the production

of short lived radioisotopes to supply various research needs and

which find its applications in industry, hydrology and nuclear

medicine. At present 26 different radioisotopes are produced regu­

larly.

The reactor utilization for research pruposes has been increasing

annually, nevertheless it is still far from saturation. Utilization

of the reactor by universities and other research institutes is still

meagre. Furthermore, it should be noted that diversification of re­

search and other applications of the reactor is limited by the present

available neutron flux.

III. OPERATING EXPERIENCE

Operating data

The reactor has been operated at steady state power only, mostly

at 25O kW. The pulsing operation will be included in the future up­

grading programme. The reactor core performance has been quite satis­

factory, no fuel element has been found to leak during the 6 years of

operation, fuel bowing and elongation pose no problem, this among

others might be due to the fact that we are not operating in the

pulsing mode.

At the present time, the reactor is utilized for isotope pro­

duction and research work from Monday through Friday, while it is

reserved on each Saturday for weekly check and nuclear reactor ex­

periments for graduate students. As a normal practice, the reactor

is operated continuously 72 hours per week.

This schedule is supported by the operating crew on a 3 shift

basis, each shift consisting of one supervisor and 2 operators.

Table I shows operating data and reactor utilization from 1965

to I97O. In I97O, it can be seen that reactor utilization time for

isotope production is 10Q $ while for research and training it is

approximately 50 Ía ami 25 i<> respectively.

Table I

Operating data from 1965 to 1970

Year 1965 19 6 6 1967 1 9 6 8 1969 1970

Total operating time 242 427 826 IO99 2965 3252(hours)

Total energy generated 4 8 , 5 8 6 91,766 182,927 2 6 9 , 0 6 5 540,350 577,775(kW hours)

Burn-up U-235 (grams) 2.5 4-5 . 9-6 13.98 28.0 30.0Utilization time fors

(hours)- isotope production 242 427 826 1099 2466 3252- research - 107 207 330 1245 1 6 2 6- training 50 64 124 265 741 813

Core physics data

Reactor operation began to be very intensive in 1968/1969 when the demand for isotopes increased significantly. In 1969 the reactor

could not operate continuously at 250 kW for more than 11 hours,

the main reason was the loss of reactivity due to fuel burn-up,

eventhough Xe-poisoning and temperature effects contributed to this

reactivity loss. It was then decided to procure new fuel from

Gulf General Atomic.

In September 1970 the reactor was reloaded in the central position

with 4 new stainless-steel clad, high hydride fuels. The core excess

reactivity of the reloaded core was measured to be $ 3.90. With the

present measured reactivity loss of 2.50 cents per MW-day, we pre­

dicted that the reactor would be able to operate continuously at

25O kW for 72 days. This turned out to be true.

The control rod worth is calibrated routinely by the inhour as

well as by the rod drop techniques. During the six years of operation

the worth of each control rod did not change more than 10

Power calibration is done routinely by the heat generation measure­

ment method to check instrument drift in the Í» power channel as well as

in the linear channel. This calibration was frequently done due to

the frequent unloading of the Rotary Specimen Rack (R.S.R.). Under

the unloaded condition (the R.S.R. diassembled) the chamber's sensi­

tivity is reduced significantly due to the modified flux distribution

in the core. Under these circumstances operating the reactor at 80 ’fo

as indicated by the meter will mean that the actual reactor power

would be 4OO kW instead of 200 kW. Thus in this case power calibra­

tion is a necessity to insure safe operation of the reactor.

IV. MAINTENANCE PROBLEMS •

Instrument troubles

As far as core performance is concerned the reactor has been

operating quite satisfactorily. The reactor's availability is

dependent only upon the performance of the instruments, auxiliary

systems and the irradiation devices.

The control instrumentation has caused a lot of operating delays

and reactor shutdowns. In brief, the causes may be classified as:

imperfect design, lack of locally available spare parts and environ­

mental conditions.

The dual-pen Westronics recorder for linear and logarithmic

power indication has failed frequently. Bad contact in the slide

wire caused oscillations and frequently spurious scrams. Routine

cleaning of thfe slide wire could only relieve the symptoms. We have

replaced this slide wire with a locally made substitute. The

synchrovèrters and electrometers should be replaced periodically,

but they are not locally available. Other source of problems are

the input signals that this instrument feeds to the period meter and

to the servo system. We also think that the recorder scram setting is

not necessary since it does not insure safety, it has also been veri­

fied that the period channel scram has a much faster response to

start-up accidents. ■

Troubles in the rod position indicators have also been noticed,

this was mainly due to failures of the interstage transfôrmer

(Triad TY55^) and transistors (2N652).

One compensated ion chamber (Westinghouse) was replaced with a

spare chamber after operating for more than 5 years. This chamber

feeds the log N recorder.

The rod drive assembly for the safety rod was replaced due to a

serious damage in the pinion-rack system. Ho major trouble has ocurred

after the replacement.

Minor troubles such as cable leaks, worn-out switches, etc.

have occurred in the period and log count rate circuits. Such

troubles also cause start-up delays because they are sometimes diffi­

cult to detect.

The neutron source is related to the instrument's sensitivity.

The Po-Be neutron source which was installed in 1964» does not supply

sufficient neutrons for start-up. However, since our operation

schedule is rather intensive, we could start-up the reactor safely

with the available photo-neutrons, provided we know beforehand the

critical positions of the control rods.

Auxiliary system troubles

No major difficulties have been encountered with the primary

as well as secondary cooling system. However, the Weinmann pump in

the primary circuit has been replaced, due to a small leak through

a graphite gland seal. The primary cooling water is maintained at

pH values of 5*5 - 6 . 5 and the conductivity value is about

1 micromhos.

Routine maintenance work on the secondary cooling water system

(city water) is done every six months to prevent algae and mud de­

position in the heat exchanger. For this purpose we have used a

cleaning solvent called VEC0M-200 made by Vecom International, Belgium.

The results have been satisfactory. Fig. 1 shows the performance

of the cooling system when the deposit exists in the heat exchanger

(upper curve), the curve shows a faster increase in bulk coolant

temperature as well as a higher equilibrium temperature; the lower

curve shows a better performance of the exchanger. Fig. 2 shows the

performance of the heat exchanger (after cleaning) at peak power level

of 25O kW, the temperature of the coolant is kept below 50° C.

Irradiation device

The rotary specimen rack (Lazy Susan) began to fail in the second

year of operation. The main problems we have encountered are the dry

■ban-bearing, the loose drive shaft joints and the locking shaft.

The first Lazy Susan was disassembled four times due to the dry

ball-bearing and the drive chain. Galling of the lubricant under high

temperature and irradiation might be the cause of the trouble. Every

time the Lazy Susan was taken out of the core, it was cleaned with

acetone and thereafter reinstalled into the reactor. In 1968 galling

again occurred, and unfortunately, the lower joint of the drive shaft

broke. The Lazy Susan was irreparable because the breach occurred

in an inacessible area. For almost one year we used bent aluminium

pipes for sample irradiations and isotope production until a new

Lazy Susan was available in 1969» This second Lazy Susan was dis­

assembled in February 1971 due to failure in the locking shaft. At

that time the shaft could not be inserted into its locking position,

careful examination showed that the "oilite" bushing which guides

the shaft, underwent mechanical deformation. This was easily reparable,

without undue radiation exposure to the maintenance personnel.

Since 1967 we have adopted a maintenance procedure for the Lazy

Susan using light oil NERO-358 (Standard Oil Company) every six months

or whenever the Lazy Susan gets sticky. Both Lazy Susans are not

equipped with a motor drive.

V. RADIATION SAFETY ASPECTS

Eventhough the TRIGA reactor is an inherently safe system as

has been experienced by many operators, the radiation safety aspects

have been given serious attention.

This health physics programme covers: Personnel Monitoring

where every staff member is equipped with a beta-gamma film dosimeter,

Air Monitoring in the Reactor Area required to maintain constant

surveillance of particulare air activity around the operating reactor,

Area Monitoring consisting of GM-counters in the reactor building and

Sampling of Vegetation etc. to check any contamination due to reactor

operation.

Up to the present there has been no unusual occurrence which lead

to health physics implications.

The TRIGA Mark II reactor at the Bandung Reactor Centre has been

in operation for more than 6 years. Utilization of the reactor for

isotope production and research and training purposes has increased

every year demanding higher availability of the reactor.

The reactor core performance has been quite satisfactory. Much

of the operational troubles have come mainly from the control

instrument's failures and frequent Rotary Specimen Rack bindings.

Nevertheless since 1 9 6c the reactor has been in operation on a quite

intensive basis, this last point is a credit to the well coordinated

effort of the operating and maintenance personnel as well as

experimenters.

BULK COOLAKT TEMPERATURE RISE

Reaotor powert 250 feW

Pig.2 .

BULK COOLANT TEMPERATURE RISE

(H X Cleaned with YECOM - 200)

Troubles with the Rotary Specimen Rack Assembly

by: Tôn-Thât-Coh and Ngô-Dinh-Long

Dalat Nuclear Research Center

Dalat, Vietnam

ABSTRACT

This report deals with the many difficulties encountered

in operating the Rotary Specimen Rack (RSR) of the Triga Mark

II Reactor installed in the Dalat Nuclear Research Center which

resulted in the decision to replace it with a new RSR. Diagnosis

of troubles, the vain attempt to repair it, and the steps

followed to install a new RSR are described.

Introduction

This modest report limits itself to one phase of the main­

tenance of a research reactor: the TRIGA MARK II. Figure 1

shows a vertical section of this low cost, high flux and inherently

safe reactor. This solid-homogeneous, tank-type reactor is

manufactured by the General Atomic Division of the General

Dynamics Corporation, and is used, as its name TRIGA implies,

for Training, Research, and Isotope production. It is particularly

popular among small research institutes and universities. In

Asia and the Far East we can see TRIGA reactors installed and

exploited in South Korea, Indonesia, Japan and Vietnam. Because

of this reason we think that our present report on "Maintenance

Problems" could be of interest to members from other nuclear

research centers and to the participants of this study group

meeting.

The content of our report concerns the problems encountered

in the Rotary Specimen Rack (a radioisotope production facility)

which is a ring-shaped}seal-welded aluminum housing surrounding

the reactor core. Inside there is a rotating rack, which is

commonly known as the "lazy susan", and which provides 40 tubular

recpetacles for specimen containers. The lazy susan is supported

on bearings and its motion is driven from the top of the reactor

through a drive shaft to a sprocket-and-chain drive in the RSR

housing. This is the main component of the isotope production

facility. There are two accesses to reach the RSR from the

reactor platform. One is the specimen-loading-and-removal tube

assembly which is curved to avoid radiation streaming from the

core. The other is the straight tube-and-shaft assembly which

consists of a driving shaft and a locking shaft. Shielding is

provided in this assembly by 5 feet of polystyrene enclosed

within the tubing. The two tube assemblies are located l80°

from each other. Through the drive shaft, the lazy susan can

be driven either manually or by a motor. In our case, the

lazy susan has been since the beginning operated manually. It

seems that manual operation, coupled with some imperfections

in the design of the lazy susan's assembly, had been the main

cause of its breakdown.

Troubles and DiagnosisLate in 1964, during a one-week continuous reactor opera­

tion, we found that the rotation of the lazy susan was getting

harder. As more torque was applied, the roll pin at the mid­

tank junction of the drive shaft apparently broke. After

evacuating the water to half level and upon dismounting the

weatherhead joint, we found no roll pin at the coupling, instead

broken bits of a l/l6 in. drill tip were observed. It appeared

that the mechanic from General Atomic had not inserted the roll

pin and the hole was not even drilled through. The coupling was

subsequently securely, fixed.

Until early in 1966, when the lazy susan came again to a real

state of inertia, repeated tries to rotate it back and forth a

small angle resulted in the driving shaft getting loose. We did

not have the opportunity to diagnose this fact until March 1966. Upon dismantling the upper coupling we found that we could raise

the lower drive shaft, which meant that the roll pin on the lower

coupling was indeed broken. The shaft was then completely

removed and closely examined. The dose rate measured at 10 cm

from the lower end of the shaft was about 15 mr/hr.Subsequent discussions resulted in the following courses

of action:

- For the time being, we could go ahead using the only hole

available in the RSR for limited isotope production.

- Consultations to be carried out with General Atomic and

other institutions which own TRIGA reactors on the radiation

hazards condition involved upon removing the RSR, and on the

repair operation.

Attempt to Repair the Rotary Specimen Rack

Subsequent correspondence through May-July 1966 brought back

valuable reports from Prof. Jauho, Director of the Reactor Labora­

tory, Technical University of Helsinki, Finland; from the Nuclear

Reactor Laboratory, University of Illinois, USA; and from the

General Atomic Co. It appeared that similar troubles with the

RSR had been experienced in a number of TRIGA reactors in several

institutions, especially in Finland where the TRIGA reactor had

drive shaft failures on three occasions. Their sharing of

experience in the repair of their RSR proved to be helpful to

us. Mr. A. P. Graff, Manager, Triga Reactor Program, General

Atomic Company, also acknowledged some weaknesses in the design

of the lazy susan system, and sent us a new design "dowel pin

conversion". The dowel pins would replace the roll pins at the

couplings to strengthen the drive shaft mechanism.

With the arrival, late in 1966, of the dowel pins and the

safety sleeves, we decided in December to take the RSR out of the

reactor tank for repair.

The RSR was then taken down to the ground floor and positioned

at a specially prepared place surrounded with lead bricks for

shielding purposes. The radiation level monitored at 1 ft. from

around the RSR without shielding was about 0.5 - 1 r/hr. Part

of this high dose rate was due to a small Co-60 source which was

still trapped in the RSR. We started work on removing the lower

weatherhead joint of the drive shaft tube assembly. We could

therefore try to turn on the short part of the drive shaft which

protruded from the RSR. Impossibility to turn this shaft more

than a few degrees confirmed the suspicion that the rack got

stuck inside the housing. We then proceeded to flush the RSR

with petroleum either in order to dissolve any sticky lubricating

oil which had frozen the proper motion of the ring bearings.

No improvement, even after we had left the petroleum ether

inside for three days. Repeated attempts to turn the shaft

with stronger torsional forces resulted in turning it loose,

without having the rack moved.

At this áage there was little hope for us to repair it,

since the work is very difficult as reported by Prof. Jauho.

One must cut a small opening at the side of the aluminum

housing to replace the broken roll pin from the sprocket, and

we had no facility for aluminum welding. Furthermore, it is

recommended, according to the manual GA-3510» that the RSR be

replaced with a new unit if it becomes inoperative because of

internal mechanical difficulties. We did nevertheless cut a

small hole to remove the Co-60 source and then decided the following:

- A new rotary specimen rack was definitely needed. And how

to finance it was also a problem.

- For the time being we could use the fishing method for

radioisotope production: irradiation of chemical targets placed

in watertight aluminum containers positioned close to the core.

Difficulties in searching for financial resources to pay

for the RSR, coupled with disruption of activities at the Dalat

Nuclear Research Center by the Viet-Cong Tet offensive in 1968,

and loss of personnel due to the military draft — without men­

tioning in passing a certain degree of indifference shown by the

government regarding nuclear researches — all this somehow

explained the state of operation of the reactor until now.

Installation of the new Rotary Specimen Rack

Early in 1971 "the new RSR was finally delivered in Saigon

and then transported to Dalat. And in April we proceeded to

install it in the reactor core. The work took three full days —

day and night — and was achieved by a staff of seven people

following the steps described below:

Step 1 :

- Test run to check the reactor working condition.

- Shuffling of fuel and graphite elements: interchange

of positions between the six fuel elements from the innermost

ring in the core, and the graphite elements from the outermost

ring.

- Test run in these conditions to check subcriticality

with all three control rods completely up (out of the core).

Step 2:

- Dismounting of the three rod drives.

- Control rods raised, disconnected from upper extension

rods, then reinserted into the core for safety.

- Central thimble and rabbit tube removed and hung against

the tank wall.

Step 3 ?

- Lower part of the drive shaft and the locking shaft

fixed to the RSR. Aluminum sleeves are fixed around the couplings

to insure that the dowel pins are always held in place.

- Tube assemblies (for lazy susan drive and sample inser­

tion and removal) fixed, and epoxy applied around the weather-

head joints.

Step 4 :

- Water evacuated from the tank to half level (part to

fill up the bulk shielding tank, part must be drained outside).

The water volume evacuated is about eight cubic meters.

- Radiation levels measured at reactor platform and at

water level in the tank are less than 0.5 mr/hr.

- Water tightness test of RSR and tube assemblies, with

soap film applied around weatherhead joints: OK with 15 psi

air pressure.

steP 5*