safety aspects in nuclear fuel cycle sixth iarp conference

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SAFETY ASPECTS IN

NUCLEAR FUEL CYCLE

SIXTH IARP CONFERENCE

BHABHA ATOMIC RESEARCH CENTRE

BOMBAY

March 7-9 , 1979

PROGRAMME & SYNOPSES

INDIAN ASSOCIATION FOR RADIATION PROTECTIONBOMBAY 400 085

INDIAN ASSOCIATION FOR RADIATION PROTECTION, Bombay - 400 085i (An Affiliate of International Radiation Protection Association)

President Emeritus : A.R. Gopal-Ayengar

Board of Trustees

U. Madhvanath P. H. PatelV. R. Shah T. Subbaratnam

Executive Committee

President : A. K. GangulyVice-President : K.C. VohraSecretary : K. R. DasTreasurer : V. K. JainMembers : M.S.S. Murthy

P.S. NagarajanA. R. ReddyS.D. SomanG. Venkataraman

Executive Committee ( Delhi Branch)

President : A. NagaratnamSecretary : A.R. ReddyTreasurer : K. SanthanamMembers : P. N. Arumugham

M.M. GuptaRarolal SharmaTejasvi Sharma ' |

SAFETY ASPECTSIN

NUCLEAR FUEL CYCLE

VI IARP CONFERENCEBHABHA ATOMIC RESEARCH CENTRE

BOMBAYMARCH 7 - 9 , 1979

INDIAN ASSOCIATION FOR RADIATION PROTECTIONBOMBAY 400 085

ACKNOWLEDGEMENT

The Association gratefully acknowledges the ftnancial

support given by the following organisations :

Research and Development Organisation,Ministry of Defence.

M/s Indian Industrial Instruments Company,Bombay.

M/s Vace Engineering, Bombay.

M/s Walchandnagar Industries, Walchandnagar.

Department of Atomic Energy, Govt. of India.

M/s Corrosion Control Services (Bombay) Pvt. Ltd.

We are thankful to Director, Bhabha Atomic Research

Centre for providing all the facilities to hold this Conference

in B. A. R. C. We are thankful to Dr. V. A. Kamath and his

staff of the Library and Information Services for bringing out

the Conference booklet.

INDIAN ASSOCIATION FOR RADIATION PROTECTION

Brief History, Objectives aad Activities

In 1968 a group of health physicists in Bhabha AtomicResearch Centre initiated a proposal to set up a professionalassociation to promote safety in the manifold uses of ionisingradiations in the country. After extensive discussions, thedraft constitution was adopted in 1969 and the IndianAssociation for Radiation Protection (IARP) was registered asa public trust, under the Bombay Public Trusts Act.

The objectives of the Association are

*to bring about proper awareness of the hazards fromionising radiation amongst their users in particular andthe public in general;

*to promote adoption of appropriate means for avoidingor reducing radiation exposures in the applications ofionising radiations and nuclear technology in the countrysuch as power generation, industry, medicine,agriculture, scientific research etc., therebymaximising the benefits which are derived out ofthese applications while minimising the risks; and

• to facilitate contacts and exchange of informationamongst specialists in radiation protection and relateddisciplines in the country and with their counterpartsin other countries.

The Association has a total membership of about 300specialists from different parts of the country. It is managedby an Executive Committee headed by Dr. A. K. Ganguly, awell-known authority in the field of radiation protection.Dr. A. R. Gopal-Ayengar, the internationally renownedradiation biologist, is the President Emeritus of the Association.The Association has on its roll numerous scientists, medicaldoctors and engineers who are well known in India and abroadfor scientific contributions and technical accomplishments.

The Association has successfully organised the followingmeetings:

First Annual Conference on Radiation Protection,Bombay, March 1973.

First Topical Meeting: Safety Standards for Siting ofNuclear Installations, Bombay, June 1974. y'

First Asian Regional Congress on Radiation Protection,Bombay, December 1974.

Second Topical Meeting: Accidential Criticaiity, Bombay,June 1975.

Third Annual Conference on Radiation Protection,Vinjyan Bhavan, New Delhi, January 1976.

Third Topical Meeting: Protection from ModernInd""trial & Environmental Risks: Lessons fromNuclear Industry, Bombay,. June 1976 (Jointly withSociety for Clean Environment and EnvironmentalMutagen Society of India).

National Seminar on Diagnostic Radiology Sc Radiotherapywith particular reference to Radiation Protection inMedical Institutions & Hospitals, Bombay, November1976 (Jointly with Bhabha Atomic Research Centreand with the financial support of WHO Regional Officefor South East Asia).

Fourth Annual Conference on Radiation Protection,Cancer Institute, Adyar, Madras and Reactor ResearchCentre, Kalpakkam, March 1977.

Fifth Conference on Radiation Protection. CancerInstitute, Gwalior, February 1978.

The Association publishes a quarterly journal, "Bulletinof Radiation Protection", which is distributed free to themembers.

The Indian Council of Medical Research has approvedthe Association as a Scientific Research Association for thepurpose of section 35(3)(ii) of the Income Tax Act 1961. TheAssociation has been given tax exemption by the CentralBoard of Direct Taxes.

The Association was admitted in May 1970 as anassociate society of International Radiation ProtectionAssociation (IRPA). Members of IARP are also members of

the International association. Some members of the associationare actively associated in the work of a number of committeesof IRPA. Members of IARP are eligible to obtain copies ofHEALTH PHYSICS (a professional journal published by IRPA'sassociate society in USA and Canada) at a special concessionalsubscription rate.

For further particulars, including membership of theAssociation, please write to:

K.R. OAS(Secretary, IARP),Division of Radiological Protection,Modular Laboratories,Bhabha Atomic Research Centre,Bombay - 400 085.

P R O G R A M M E

SAFETY ASPECTSIN

NUCLEAR FUEL CYCLEBhabha Atomic Research Centre,

Bombay, March 7 - 9, 1979

P R O G R A M M E

WednesdayMarch 7. 1979

0830 hrs REGISTRATION

INAUGURAL SESSION

0930 hrs Welcome

0935 hrs

0945 hrs Inauguration

1000 hrs Keynote address"Risk Estimates forFuel Cycles"

1025 hrs Vote of thanks

CENTRAL COMPLEXAUDITORIUM

1030 hrs Coffee

1100 hrs Invited Talk:Negligence of RadiationExposures - Its Medico-Legal Implications

Shri K.R. DasSecretary, IARP.

Shri S. FareeduddinDirector, BARC.

Dr. M. R. SrinivasanDirector, PPED

Dr. K.G. VohraVice.-President, IARP

Dr. K.C. PillaiConvenor,Programmes Committee

Dr. R. H. DasturBombay

Session L Safety Aspectsof Mining kMilling

1200 hrs Evaluation of Activity-Distribution of RadonDaughter-products in Airand its possible Applica-tion in a MineAtmosphere

(IARP-10/19)

1215 hrs Radon and Thoron in theAtmosphere and Thoronin the Breath of ThsriumWorkers

(IARP-10/17)

1230 hrs Radiation Exposure inMonazite Industry

(IARP-10/13)

1245 hrs Lunch Break

M. C. Subbaramu

Y.S. MayyaP. Katrappa

A.C. Paul

Session II Safety Aspects D-Blockof Fissile AuditoriumMaterialProduction andFabrication

1400 hrs Safety Assessment andand Experience in theFabrication of U-PuMixed Carbide-Nitrideand CarbonitrideCompounds at TrombayLaboratory

(IARP-10/29)

1415 hrs Occupational Radiation S.Exposure in the Nuclear B.Fuel Complex S.

(IARP-10/32)

Sudarshan K. MehtaP. M. B. PillaiK.N. KuttyJacob JohnN. SwaminathanS. B. Wfitamwar

ViswanathanSurya RaoMehdi AU

1430 hrs Determination of U-233content in SealedContainers and WastePackages by GammaCounting

(IARP -10/36)

1435 hrs Development of anExploding Wire FerosolOperator fox MetalOxide Aerosols

(IARP-10/38)

1500 hr6 Tea Break

M. V. DingankarP. P. JoshiL. V. Kulkarni

J. C.A. A.

KapoorKhan

Session III

1515 hrs

15 30 hrs

1545 hxs

1600 hrs

Safety Aspectsof ReactorOoerations

A Review of R.adiation M. G. KolhalkarExposure to Occupational S. K. SharmaWorkers at CIRUS

(IARP-10/43)

Investigations into aCase of High TritiumExposure at RajasthanAtomic Power Station

(IARP-10/9)

Internal ExposureProfile of OccupationalWorkers of a BWE.Type Atomic PowerStation

(IARP-10/6)

Organic Binding, ofTritium and itsSignificance inExposures in HeavyWater Moderated Reactor

(IARP-10/15)

T. Subbaratnam

A. G. HegdeI.S. Bhat

Kamala Rudran

1615

1630

1645

ThursdayMarch 8,

hrs

h r s

h r s

1979

Recovery of LiquidScintillator from TritiumMonitoring at ReactorSites

(IARP-10/31)

Improvements in theActivated CharcoalFilter for BetterConfinement of Air-borneRadioiodine

(IARP-10/37)

Close of Session

S.R.S.D.

K.G.A.G.D.S..A. A.

SachanSoman

GandhiKishoreOeshingkarKhan

Session IV Development inPersonnelDosimetry

1000 hrs An Intercomparison Study S. Somasundaramfor Assessment of LungBurdens of Low EnergyX-ray Emitters

(IARP-10/39)

1015 hrs

1030 hrs

Estimation of Radium-226 by Emanometry

(IARP-10/2)

Comparison of DiffusionChambers and the WireSereen Samples forEstimation of UnattachedRadon-daughters

(IARP-10/3)

R. C. SharmaP. KotrappaT. SurendranT.K. HaridasanS. P. GargD. P. BhartiN.S. Pimpale

M. RaghavayyaM.A.R. IyengarP. M. Markose

A. K. KhanG. K. SrivastavaM. Raghavayya

1045 hrs Computed Fast Neutron K.S.Kertna and TL Responseof BeO

(IARP-10/14)

Sharada

1100 hrs

1115 hrs

CaSO4:D7 EmbeddedTeflon TLO Tape forPersonnel Monitoring

(IARP-10/26)

Coffee Break

1130 hrs Effect of IntenseNeutron Flux in TL.XSPersonnel MonitoringCards

(IARP-10/24)

1145 hrs Influence of Detectorsize and Air-gap onthe Sensitivity ofBonner Spheres

(IARP-10/7)

1200 hrs Angular Dependenceto Fast' Neutrons ofPersonnel MonitoringKodak NTA Film-packfor Neutrons Energiesttpto 14 MeV

(IARP-10/8)

1215 hrs A Versatile Instrumentfor Counting Lumines-cence Photons fromChemicals.or TLDPhosphors

(IARP-10/25)

Bhuvan ChandraA.S. PradhanA. R. LakshmananK.C. PopliR.C. Bhatt

R. R. ViswakarmaR.G. KhadakeP. Gangadharan

M.F. DhairyavanP.S. NagarajanG. Venkataraman

P.S. Nagarajan

S. KannanN.H. RisbudP. Gangadharan

Session V Safety Aspectsof FuelReprocessing

1400 hrs Safety Considerations ina Fuel ReprocessingPlant

(TARP-10/28)

1415 hrs Development of a fissilematerial transportcontainer

(IARP-10/33)

H. R. SiddiqueS.V. KumarS. Krishnamony

T. S. Lax minarayananP. HaraprakashK. VijayanV. P. BaiakriahnanS. RajagopalasP. F.S. Jacob

1430 hrs Decommissioning ofTrombay Fuel Reproce-ssing Plant -Radiological Health andSafety Aspects

(IARP-10/42)

1445 hrs Nondestructive isotopicassay of Plutonium

(IARP-10/10)

1500 hrs Tea Break

1515 hrs Minimum risk basedsolution of wastefueldumping site

(IARP-10/45)

1530 hrs Radiation Stability ofAmberlyst-15, amacroporous strongacid cation exchangeresin

(IARP-10/41)

1545 hrs Further evaluation ofcyclone for themeasurement of aerosolparameters

(IARP-10/18)

T. K. TeyyunniB. M. SidhwaM. N. Nadkavni

M. R. IyerH. EberleH. Ottmer

D. D. SharmaD. ChatterjeeK. Sri Ram

P. C. Mayan KuttyN.S. PillaiS.S. ShindeM.N. Nadkarni

V. B. MenonP. Kotrappa

1600 hrs Development of an P. R. Joahiindigeneous high volume P. Kotrappaair sampler N. N. Shetty

(IARP-10/16) B. RaghunathP. S. Sivasubramaniam

1615 hrs Session Closes

FridayMarch 9. 1979

1000 hrs Invited talk:

1100 hrs Coffee Break

Session VI Dose Records,Standards k.Legislation

1115 hrs A proposal to reduce the P. S. Iyermaximum permissible U. Madhvanathdose to meet the require-ment of high standard ofsafety in radiationindustry

(IARP-10/21)

1130 hrs Operational limit of Giridhar Jhauranium ore dust M. Raghavayyaapplicable' to uraniumcomplex at Jaduguda

(1ARP-10/4)

1145 hrs Analysis of occupational K.S. Shenoypersonnel exposure in P. H. PatelNuclear Fuel Cycle S.J. Supeoperations in India

(IARP-10/23)

1200 hrs

1215 hrs

1230 hrs

1400 hrs

1415 hrs

1430 hrs

1445 hrs

Occupational life-timedose expectancy in theNuclear Fuel CycleWorkers in India

(IARP-10/22)

R. SadagopanG. SadagopanS.J. Supe

P. M. MarkoseK. P. Eappen

Sai»f? in Fuel S. E. BritoReprocessing - Aregulatory view

(IARP-10/27)

Lunch Break

Session VII EnvironmentalRadioactivity

t

Some aspects ofaccumulation of naturalradionuclides in aquaticproducts and vegetationaround a uraniumcomplex

(IARP-10/5)

Radium-2 2 8 in theMonazite industry

(IARP-10/12)

Studies of naturalradioactivity inKalpakkamenvironment

(IARP-10/34)

• Some aspects ofaccumulation ofradionuclides incoastal sediments

(IARP-10/35)

A. C. PaulK.C. Pillai

M.A.R. IyengarM. P. Raj anS. Ganapathy

N-N. DeyVasanthi MalkarElizabeth MathewK.C. Pillai

1500 hrs Tea Break

1515 hrs

1530 hrs

1545 hrs

Comprehensiveenvironmentalsurveillance andimpact-experiences atRajasthan AtomicPower Station

(IARP-10/11)

Radioactivity releaseto the environment bythermal power stationsusing coal as a fuel

(IARP-10/20)

R. P. GurgT.A. SebastianB. DubeK.G. VarugheseA. R. Lakshmanan

U.C. MishraB.Y. LalitT.V. Ramachandran

Technologically enhanced V.S. Londheradiation exposure from K. C. PilJaisome non-nuclear S.D. Soniansources

(IARP-10/40)

1600 hrs A continuous effluentstream samples

(IARP-10/1)

1615 hrs Organometallicinteraction ofmanganese withhumic acid

(IARP-10/44)

1630 hrs Closing Session

S. VenkataramanG. K. SrivastavaS.C. SahaP.R. Kamath

A. N. AggarwalM. V. M. Desai

1

S Y N O P S E S

1 !

I

1

IARP/10/19

Evaluation of Activity Distribution of Radon Daughter ProductFin Air sad its Possible Applications in a Mine Atmosphere

M. C. Subba RamaAir Monitoring SectionBhabha Atomic Research CentreBombay - 400 085

The working level concept based on potentialej^-energyand used in uranium mines for inhalation riBk evaluation isnot derived from adequate biological knowledge. If thehazard is associated with the presence of the fraction ofradon daughter products not attached to aerosols, the potentialo^-energy does not represent the risk where as radonconcentration has been found to be quite representative 1. Thedeposition of the unattached fraction in the bronchial regionhas been associated with development of cancer in uraniumminers2. The attached activity can also deposit in thebronchial region if the aerosols carrying the radon daughteratoms grow in the human lung during inhalation3. It is thusnecessary to measure both the unattached and the attachedactivity distributions in the mine air for a. proper evaluationof inhalation hazard.

This paper describes, the results of the measurementof unattached fraction using a diffusion sampler calibrated inthe laboratory in ar. atmosphere of unattached radon daughterproducts. This ha; been achieved by avoiding the inter-ference by the aerosols that are formed in air spontaneouslyby ionising radiations.

Attachment studies using radiolytic aerosols have beencarried out by measuring the unattached fraction by thediffusion sampler and Millipede filter combination. The un- >,attached fraction has been found to vary inversely as the jaerosol concentration. The attachment method can "be used !to measure the mean aerosol size.

The distribution of the attached activity has beenmeasured using an Andersen sampler. The interference by

the unattached fraction has been taken- into consideration incalculating the distribution of the activity on aerosols. Ithas been found that the activity mern aerodynamic diameterchanges with time due to growth by coagulation in anenclosed atmosphere. Radiolytic aerosols which grow vithrelative humidity are used for the studies. The resultsshow that the bulk of the attached activity is associatedwith aerosol size-range which when grown in the humanrespiratory tract would deposit in the bronchial region ofthe lung.

The use of diffusion sampler-Millipore filtercombination and the Andersen Impactor in uranium minesseems to be adequate for obtaining information necessaryfor the evaluation of inhalation risk to which the workersare exposed.

References:

1. Pradel, J. "Radon Protection in Uranium Mines",Noble gas Symposium, Las Vegas (1973).

2. Chamberlain, A. C. and Dyson, E. D. "The Dose toTrachea and Bronchi from the Decay Products of RadonlThoron", The Brit. J. of Radiol, 29, 317 (1956).

3. George, A. C. "Deposition of Radon Daughters in HumansExposed to Uranium Mine Atmospheres", HASL - 68 - 11(1968).

IARP/10/17

Radon and Thoron in the Atmosphere and Thoron in theBreath of Thorium Workers

Y.S. Mayya and P. KotrappaHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Introduction:

Radon and thoron are present in the atmosphere atvery low concentration levels. Also, the individuals possess-ing internally deposited amounts of thorium and uraniumrelease small quantities of thoron and radon in their breath.To measure these levels, the conventionally used double-filter apparatus was modified both to increase the sensitivityand to improve the calibration procedure.

Description of the method:

In the conventional form, the double-filter unitconsists of two filters in series separated by a cylindricaldecay volume of about 1 litre to permit the decay of radonand thoron in transit. The inlet filter collects all the parti-culates and lets pure radon or thoron along with the air toenter the decay volume. The downstream filter collects apart of the decay products generated in the decay volumeduring transit. The remaining part gets deposited on thecylindrical wall by diffusion. The alpha activity on the endfilter is measured for a certain interval of time after theend of sampling. To deduce the concentration C from thealpha disintegrations D(Tj, T ) in the interval from T, toT2 reckoned from the end of sampling duration t , thefollowing formula is used,

. 0.45 D (Tfr T2)

F f Z V

V = volume of the cylinder in litres

Z = A function of, decay constants of radon or thorondaughters, t8, Tj b T2 .

A

Ff = The fraction of the decay products depositingon the end filter

Ff has to be empirically calculated. Theoritical equationsgiving the values are approximate due the uncertainly in thediffiusion coefficients of decay products. Also one assumesa uniform parent gas concentration in the derivation of F£and thoron being short lived, does not conform to thisassumption. Also, for one litre tube, the sensitivity is low.

In the present apparatus the wall loss is reduced toless than 5% at relatively email flow rates, using ammoniumchloride aerosol as the scavenger of the decay products.Air stream passes at the centre while the aerosol is let nearthe periphery of the tube; thus stopping the decay productsfrom depositing on the wall. Also, since the entire activitygenerated in decay volume is transferred on the end filter,the method becomes absolute. The present decay volume isabout 150 litres and can detect 80 pCi/m for thoron and6. 0 pCi/m3 of radon.

Atmospheric measurements were done near BARCHospital both in the open and in the room air. The followingmethod is employed for the thorium-workers. The subjectinhales thoron free air from a bag through a respirator withinlet - outlet valves and exhales into the double-filter unit;the pulsating flow from the breath is converted to unidirect-ional flow in the apparatus by using a total suction largerthan the breathing rate and sucking the extra air needed fromanother bag free of thoron. The measurement is done for30 minutes. The concentration measured in pCi/litre isconverted into the emanation rate of thoron by simplymultiplying the former by the tofal flow rate.

Results and Discussions:

Reproduceable results were obtained in both the typesof measurements. Radon values ranged from 50 to 100pCi/m while thoron values were found to be between 100 to300 pCi/m3 in the environment.

There was a marked difference between the thoronemanation rates of thorium and non-thorium workers. Whilethe thoron emanations for thorium workers ranged from 60 to

130 pCi/min, that for non-thorium workers was around8 pCi/min. Experiments are being carried out in. Alwayeto correlate the thoron emanation rate with the actual bodyburden of thorium.

IARF/10/13

Radiation Exposure in Monazite Industry

A. C. PaulHealth Physics DivisiciBhabha Atomic Research CentreBombay - 400 085

The monazite present in the beach sands of Kerala andTamil Nadu is separated in the mineral separation plants atManavalakurichi and Chavara operated by M/ s Indian RareEarths Ltd. The sand later on is subjected to chemical processingto separate thorium from rare-earths at the rare-earths plant"situated at Udyogamandal. The physical and chemical processingof the sand involves radiation hazards due to the presence ofthorium, uranium and their daughter products in monazite.

In the mineral separation plant at Manavalakurichi wherenearly 90% of the monazite is produced the problem of occupatio-nal exposure is confined to 10% of the operating staff who arcdirectly engaged in the manual handling of the monazite concentra-tes obtained from the electro magnetic separators,, Externalexposure resulting from this operation is estimated to be in therange of 1 to 4 rem. yr . The man-Rem. t of monazite in thisplant works out to be 0. 02. The low fraction of respirable dustreduces inhalation hazards. Some of the other topics identifiedhere are the environmental levels and ingestion pathway throughfood items.

Chemical processing of monazite results in the fractiona-tion of radio-nuclides in different streams. The major activitiesof Th-232 and Ra-228 get concentrated in the hydroxide and chlo-ride fractions respectively. Persons who cone directly intocontact with thorium concentrates and solid wastes receive themaximum external exposure. The average exposure per personworks out to be nearly 1 remyr"^.

Institution of Health Physics practices since lq63 hasresulted in gradual decline of the radiation exposures due toimprovements in working conditions, rotation of personnel andincreased awareness of radiation safety. However, of late the

7

build-up of activity in plant and premises, resulting fromdeactivation and an accumulation of solid wastes have alteredthe situation. 'Table 1 which gives the Man-Rem exposures fordifferent years high- lights the trend.

Table 1

MAN-REM EXPOSURE IN THE CHEMICAL.PROCESSING OF MONAZITE

Year . Man-Rem.f1

1965 0.063

1967 0.047

1969 . 0.049

1971 K 028

1973 0.029

H9T5 0.027

1976 0,055

1977 0.070

The internal contamination as assessed by urinaryexcretion is found to be 0. 3. tolerance on an average with respectto Th-nat and Ra-228.

The paper also discusses the impact of the operationson environmental radiation levels and measures adopted toreduce the same*

1ARP/10/29

Safety Assessment and Experience in the Fabrication ofUranium-Plutonium Mixed Carbide, Nitride and CarbonitrideCompounds at Trorobay Laboratories

S. Janardhanan, S.K. Mehta, P. M. B. Pillai, K. N. Kutty,Jacob John, N. Swaroinathan and S. B. WatamwarHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Interest in carbide, nitride and carbonitride actiniaecompounds have merited greater attention in several nuclearlaboratories in view of their use as potential fuel foradvanced fast breeders and metal-cooled power reactors, onaccount of their unique properties amenable to high tempera-ture nilclear operation. At Trombay, recently a flow sheetwas developed in Radiometallurgy Division for preparation ofpowders based on the carbothermic reduction of oxides.

Metallurgical operations involved handling of finegraphite, uranium and plutonium oxide powder which bringsin its wake a host of radiation and contamination problemsand functional safety requirements of concern- to the healthphysics profession.

Primary risks identified in the manufacture route are:

(1) Radiation. Safety Exposure problems from gammaand neutrons during mixing, pressing,, powder characterisationjobs as well as sintering. Critical masses for the synthe-sised compounds are several kg weights and for the smallbatch sizes handled this did not arise.

(2) Equipment Safety Design requirements forplanetory ball mills, karber hydraulic press with controlsand leak tightness to ensure alpha seal and inhibit accidentconditions are essential. Mills have gasketted lids, time andspeed controls, interlock systems for operation and presshas self sealing coupling to maintain alpha seal and shieldin the die systems.

(3) Fire safety and handling care Pyrophoricity dueto the extreme reactivity of the powder with water vapourand air call for inert gas boxes for handling, storage etc.

(4) Furnace safety The synthesis results in hightemperature operation of furnaces for prolonged periods withattendant risks and product handling care needed. Furnacesafety includes mainly emergency cooling arrangements,current controls and trip circuits etc. in untowardsituations.

(5) Product handling during inspection X-ray diffrac-tion studies called for careful mounting of samples and alphatight enclosures which are described.

Data showed that \ kg batches could be handled withmaximum chest and wrist exposure of 40 m re'm and 145 mrem for carbides, nitrides and 115 m rem and 60 m remfor carbonitrides respectively. About 50 urine sampleanalyses were carried out and showed no case exceeding0.05 tol (Pu) and 40 whole body counts done over the twoyear period also pointed to no exposure. With 60 gm batchsize, gamma and neutron dose on box ports were 4-5 mR/hrand less than 1 mR/hr respectively. Air monitoring indicat-ed an ambient background of 0.1 tol for Pu during hightemperature work and no release through stack. The efflue-nts amounting to 1.8 x 10° litres per year from furnacecoolants, decontamination wastes with an aggregate activityof 3.5 mCi (alpha) and 5.4 mCi (beta-gamma) as well asmore than 400 solid waste packets of nearly one kg.wt. eachand surface dose 5 mR/hr, indicated no significant Purelease to environment.

A few minor incidents requiring corrective steps andof health physics significance are also reported. Equipmentdesign safety and pyrophoricity are matters of major concernin work with these compounds.

/o

IARP/10/32

Occupational Radiation Exposure in the Nuclear Fuel Complex

S. Viswanathan, B. Surya Rao and S. Mehdi AliHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

The Nuclear Fuel Complex, Hyderabad, the majorfabricator of zircaloy clad uranium fuels, houses variousproduction plants for the purification of natural and slightlyenriched uranium and fabrication into fuel assemblies.About 400 persons working on Tound-the-clock shifts in theplants are covered by centralised film badge service fromBhabha Atomic Research Centre for the evaluation ofexternal exposure. For internal contamination due totransportable uranium, regular urine analysis has beencarried out and uranium body burden is evaluated. Estimationof the lung dose has been carried out by using the aerosoldata.

The paper presents the radiation exposure datacollected during the past five years. The total annual dosefor different plants and the average annual dose per personare tabulated in table-1. Internal contamination levels dueto transportable uranium expressed in terms of maximumpermissible body burden (MPBB) is given in table-2.. Airborne uranium activity during the various operations areplotted on log -probability paper and median values are taken.Activity Median Aerodynamic Diameters (AMAD) of theaerosol are found out experimentally using the \" HASLcyclone assembly. Using the above data, and also thereported values of the effective half lives of the uranium,lung dose estimates have been carried out for fifty years'exposures based on ICRP new lung model. , j

IThe results show that both average annual external i

exposure and the uranium body burden are found to be low. • •The estimated average lung dose due to non-transportableuranium is worked out and found to be far below themaximum permissible dose recommended by ICRP (15 reins).

k

EXTERNAL EXPOSURE OF RADIATION WORKERS (CUMULATIVE DOSE IN MREM5) - TABLE-!

Y E A R

1973

1974

1975

1976

1977

Y E A R .

NANCE IN

Laaa than

URANIUM(Production

OXIDE PLANTof Natural UOg }

Total dot a for Avaraga doiatha Iaitltutioa par paraon' (mraia) (mram) .

14*34

6654

4125

9135

9233

161

66

51

74

•1

ggTJBRJNAL CONTAMINATION

TOLERANCE

0.05

0. «5 to 0.1

0.1 to 0.2

0.2 to O.S

Abova OjS

1973

UOP EUOP FffB

4 41 7

S 10 4

3 4 3

1 -

• • •'

ENRICHED URANIUM OXIDE PLANT(Production of Enrichad UOj)

Total doaa fortha Institution

(raram)

!44«

3607

1975

4?00

7540

DUE TO URANIUM

1974

UOP EUOP FFP

20 •• 24 35

5 1ft 6

1 • 1

2 11 1

* m m

Avaraga deaapar parton

(mram)

^22

33

17

40

79

(BIO ASSAY RESULTS)

5975

UOP EUOP FFP

51 140 S3

• ia 6

2 3 2 •

i 7 1

. . .

FUEL FABRICATION PLANT(Fual Fabrication)

Total doaa lortha Inattttiiion

(mi am)

7«44

••59

•240

7955

2749

• TABLED

1976

UOP EUOP

40 U

9 34

2 11

4 S

• •

Avaraga doaapar parien

(mram)

„ts

61

S3

17

1977

FFP UOP EUOP FFX

11 23 40 7

S 3 12 1

1 6 1 1

3 7

. . . .1 tolaranca (MPBB) corraeponda to ascratlen of SO ug. of nat. uranium in ona litra urlnai

1ARF/1O/36

Determination of U-233 Contest in Sealed Containers &Waste Packages by Gamma Counting.

M.V. Dingankar, P.P. JoshiHealth Physics Division, B.A.R. C, andL V. KolkarniNeutron Physics Division, B.A.R.C.

1.. Introduction:

Uranium-233 is a fispile material which is expectedto play an important role in our Nuclear energy programmein the context of the utilisation of our Thorium resources.Minute amounts (ppm levels) of U-232 are invariably presentin the uranium-233, separated from reactor irradiatedThorium-232. Bi-212 and Tl-208 are two daughter productsin the decay chain of U-232, which are hard gamma emitters.The radioactivity of these daughter products increases withtime following chemical separation of uranium, attainingmore than 50% of the saturation value within a couple ofyears. Although this is an undesirable feature of the ThoriumCycle, it can be used to monitor and detect U-233 content insealed containers or waste packages. At the PurnimaFacility a gamma counting technique using a 5 cm x 5 cmNal (Tl) scintillation detector, has been developed.

2. Materials and Methods.;

A conventional counting set up consisting of a Nal (Tl)scintillation detector, pre-amplifier, amplifier, single channelanalyser, sealer and a High Voltage Supply Unit, is used forgamma counting of the samples.

The test sample consisted of contaminated BeO brickssealed in PVC bags. The single channel analyser was set tocount gammas of Tl-208, which have energy of 2. 614 Mev.The efficiency of the counting set up for Tl-208 gammas wasdetermined using a Th-232 sample of known history. Theamount of Tl-208 in the Th-232 sample can be calculatedfrom the mass of Th-232 and its age since separation. Thetest samples were counted at various distances from the

detector. The Th-232 standard sample and the testsamples showed identical variation of count rate with distan-ce. By counting both the samples at a suitable fixed distan-ce, the relative strengths of the T1-20S content of tha testand standard sample, can be found. To calculate the U-233content of the test sample; the ppm content of U-232 ia itand time since separation, of U-233 must be knows. Activitybuild up data of Tl-208 from U-232 for various times afterseparation are taken from the literature available. U) Thisenables us to convert the Tl-208 content into the U-232content. Knowing the ppm content value of U-232 in theU-233, the U-233 content of the sample can be obtained.

3. Results and Discussions:

All the test samples counted, showed statisticallysignificant count rates in the 2. 614 Mev energy channel c*ithe counting set up. The background count rate in thischannel was very low. Thus there was a clear indicationof the presence of U-233 in the test samples. The amountsof U-233 estimated for the test samples were in the rangeof 5 mgms. to 245 mgms. The U-233 contained 3 ppm ofU-232 and its age was 7 years.

This technique can be used to monitor the presenceof U-233 in milligramme amounts in inaccessible places orto locate hold up of U-233 solutions in pipes or vessels.

Reference

(1) Raman N., IyerKLR.. 'Activity build-up of artificiallymade uranium isotopes.1 B.A. R. C. Report. B.A.R. C. /1-331. (1975)

IARP/10/38

Development of an Exploding Wire Aerosol Generator forMetal-O-dde Aerosols

J. C. Kapoor and A. A. KhanAir Cleaning Engineering Research SectionBhabha Atomic Research CentreBombav - 400 085

Studies of aerodynamic properties like effective density,shape factor, mass median diameter etc. of agglomerates ofcertain metal oxide aerosols is of importance in the.developmentof emergency air cleaning system tor fast breeder reactors.In hypothetical core disruptive accidents in fast breeders,large energy would be available in the core for evaporation inm sec. duration and specific energy have been estimated to beabout 4. 0 kilo joules. Simulation of these conditions of energydissipation in core materials in laboratory can be achieved byusing i) exploding 'wire technique and ii) pulsed laser evaporation.In this paper we describe the exploding wire technique whichhave been developed for aerosol generation for these conditions.

This system consists of a hank of four 0. 5. /*F capacitorswhich are charged to 10 kilovolts, providing a maximum storedenergy of 100 joules. This capacitor bank is discharged throughthe material of the desired aerosols using a SCR-fired spark-gap* -Amount of the material evaporated is of the order of a fewhundred milligrams per shot. The specific energy can be changedby charging the capacitors to a lower voltage or by using adifferent amount of wire material. The system is operated incritically damped mode where maximum of the stored energyis discharged da the first peak.

Some results of the iron-oxide aerosols indicate that theprimary particle size is' of the order of 0. 20 microns, for aspecific energy of 4. 0 kilo joules /gm. This primary particle

Isize decreases with the increasing specific energy. For size »analysis the aerosols were collected on electros microscopegrids directly inside the explosion chamber and analysed by usingelectron micrographs.

IARP/10/43

A Review of Radiation Exposures to OccupationalWorkers at CIRUS

M.G. KolhatkarHealth Physics Division andS.K. SharmaReactor Operations DivisionBhabha Atomic Research CentreBombay - 400 085

Cirus, a 40 MW (Th), natural uranium fuelled, heavywater moderated and light water cooled research reactor islocated at Bhabha Atomic Research Centre, Bombay and isin operation since July I960. About 600 personnel who workin the reactor complex include staff from operations, mainte-nance, health physics and other disciplines. Film badgeswere used for monitoring of personnel exposures till 1978when a change over to thermo-lumniscent dosimetry waseffected. Spot assessment of exposures during work in highradiation areas is done by direct reading pocket dosimeters.

Bulk of the total exposures are shared by operations(52%) and maintenance personnel (36%). Regular assessmentfor body burdens due to 58Co, 6oCo, 137Cs a n d 131J i s

carried out and this has remained less tha.i one percent ofthe permissible levels. Exposures due to tritium intake havebeen insignificant inspite of large amount of work carried outon equipment and piping carrying tritiated heavy water. Thiscan be attributed to the general awareness amongst staff,inculcated through training, and strict procedures followedwhile carrying out such work.

All significant exposures are investigated in detail bya competent committe.e and the findings reported to higherauthorities. As a first step, the genuiness or otherwise ofthe reported exposure is established. So far, in about 6%of the cases the reported exposures have been found to benot genuine. The major causes for this have been (i) conta-mination of the film while in use (ii) storage of the film ata contaminated place during the non-use period and (iii)inadvertent placement of the film in high radiation areas.

In case of genuine exposures, the* causes are identi-fied and remedial actions taken to prevent recurrence.Analysis indicates that a majority of significant exposureshave occured during underwater handling of irradiated assem-blies in the spent fuel storage building and while working atthe bottom of irradiated fuel assemblies in the reactorbuilding.

In the spent fuel storage building, exposures resultfrom unavoidable long hours of work in a relatively lowradiation background of 25 to 40 mR/hr. Disposal of radio-active resins which are used for polishing the bay-water wasanother source of high radiation exposures. Maintenancework on the resin beds also results in high exposures mainlydue to highly congested working conditions leading to difficultapproach and longer durations for relatively simpler work.

Exposures while working below irradiated fuelassemblies are mainly from handling of fuel rods with rupturedcladding. Adherence of fission product laden water to theprotective clothing is another source of exposure.

Based on investigations and the experience over theyears, various steps have been taken to bring down radiationexposures in Cirus. These measures, coupled with reductionin fuel clad-failure incidences have brought down the totalexposures gradually and the present annual exposures areless by a factor of three as compared to the exposures duringthe period 1966 to 1970.

•7

IARP/KV9

Investigation!! into a Case of High Tritium Exposure at theRajasthan Atomic Power Station

T. SubbaratnamHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

A case of high exposure to tritium resulting in «. contami-nation in urine of 2800 uCi/C was investigated. The case cameto the nctice of the health physics unit when a routine sampleof the individual registered 30? uCi/C on the first day, 859uCi/Con the second day and 17180 uCi/C on the 6th day. Repeatsamples taken on the subsequent days showed only values rangingbetween 29 and 135 uCi/C. Since the values noted did not followany logical pattern of elimination for the radionuclide, personalsamples were obtained and found to register values averaging2600 uCi/C. Fts double checking the results, samples of spit,breathing air and blood were analysed. Chromosomal aberationtests were done to support the conjecture that tho very high valuesof uriae concentration observed earlier to the incident were notgenuine.

The second step was to establish and understand the cir-cumstances under which the exposure was incurred. Plantconditions during the period when the exposure was estimatedto have oc cured as well as the preceding fortnight during whichthe individual showed high tritium concentrations in urine werecarefully analysed. Results were double checked at site as wellas at the* laboratories at BARC. By computing the intake quantaby. oral intake due to a splash of heavy water (as was claimedby the individual), by skin absorption and by breathing air atpostulated high concentration of tritium and by the process ofreasoning, it was established that the intake was due to intakeof tritiated heavy water by deliberate action by the individual.The dose commitment to the individual due to this intake was . <estimated to be 9. 4 reins. !

The above incidence highlights the role of theplant health physicist in not only enforcing rigid methodsof control but also in establishing the bonafides of eachcase of over-exposure so as to safeguard not only theinterests of the individual but also that of the organization.

IARP/10/6

Internal Exposure Profile of Occupational Workers of a BWRType Atomic Power Station

A. G. Hegde and L S. BhatEnvironmental Survey LaboratoryTAPS Colony, TAPS Post, Thana Dist.Maharashtra State 401 504 -

The radiation workers in a nuclear power station receiveroutinely internal exposure well below permissible levels frombody deposited radionuclidea. In boiling water reactors (BWR)it has been observed that the major internal contaminants aregamma emitters 134CSj 137CBf 58Go> 60Co, 131^ 103+106Rllj

95Zr. These can be easily and accurately measured by in vivocounting with whole body monitors.

The occupational staff of the Tarapur Atomic PowerStation (TAPS) are examined for whole body activity once everyyear as routine. The measurements are carried out at SSLwith the shadow shield type of whole body counter having 10 cmdiameter Nal (TI) detector connected to a multichannel analyser.The major "nuclidea in the TAPS workers are observed to be°°Co and 134Cs+ Cs. During refuelling and carrying out other

131 ^ 3special maintenance jobs, other nuclides like*^"Ba Ad 95z r a a v e been found on whole body examination for •body burden. -

The internal exposures are expressed as nCi/mao. Theaverage nCi/man for a radionuclide is evaluated by dividingtotal internal exposure in nCi of all the workers involved by thenumber of workers. Average internal exposure due to ^ C ochanged from an initial value of 1. 7 nCi/man in 1970 to 102. 0nCi/man in 1973 and came down to 25. 5 nCi/man in 1978 (MPBBfor 60Co is 10000 nCi). Similarly 134+l37cs increased from 2. 5nCi/man in 1970 to 72. 3 nCi/man in 1?2 and then decreased to11. 3 nCi/man in 1978 (MPBB for 137cs is 30000 nCi). Thus theinternal exposures have been found to be only 1% or even lessthan the MPBB values.

The paper discusses the internal exposure profile ofTAPS radiation workers during the last nine years in relation-

to job assignments. The staff exposures are classified underfour different groups namely i) Maintenance ii) Operationsiii) Technical Services and iv) non-Technical. This study-revealed that Maintenance workers group had highest incidenceof internal exposure among these. The special section ofmaintenance workers who have the higher internal exposurecompared to others in the same group has also been identified.

IARP/10/15

Organic Binding of Tritium and its Significance in Exposuresin a Heavy Water Moderated Reactor.

Kamala RudranEnvironmental Studies SectionHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Introduction

OUT present radiation protection recommendations likeMFCa, MPC^, derived investigation levels and organ of concernfor 3H are all based on a 10-12 day biological halflife fortritium in body. The present summary deals with some of theobservations made by the author when a few cases of exposureto tritiated water by inhalation were studied for possible longerhalf lif es.

Three cases of exposure to tritium in a heavy watermoderated reactor wore studied for elimination of tritiumin urine for periods upto 500 days. Two components wereobserved in all the three cases. Biological halflives andthe proportionation between the components were calculatedby least square fitting of the results obtained. The valuesof TD and fractions eliminated with each T^ are given inTable I

It was considered that the second component couldprobably arise from tissue binding of tritium in vivo whichon metabolic breakdown may give rise to tritiated waterand get eliminated in urine. However, some of the tissuebound tritium should than be eliminated in organic compoundsin urine. To check this, a series of urine samples fromexposed persons were analysed for organic bound tritium.The procedure adopted was to evaporate the urine to drynessin partial vacuum, remove easily labile H by repeatedwetting and drying, burn known weighty of dry urine inoxygen flask W, collect the liberated water in liquid

scintillates: and court for H. Organic matter content in thesample was astimated as the loss in weight an ashing at450°C. Concentration per gram of urinary organic matterwas aboai 1th to l t n of that of urinary watsr.

10 30

Discussion

Table I shows that in all the three cases studied,radioactive tritium is eliminated with two biological half-lives, 3-10 days and more than 100 days and that thefraction eliminated with a longer Tj, is-about 3-13% of thetotal elimination. In addition to a firet short Th of 3-ISdays reported by various authors, an intermediate TV of20-25 days has been reported in five cages (2)» (3)» T*)»and a longer T*D of 280-350 days has been reported in fourcases (3) (4)# Some of the authors have attributed the secondand third component to invivo organic binding of H. Presentstudy gives concrete proof of invivo organic binding since theurinary organic tritium which is not easily exchaig ed withhydrogen of water has been formed in the body, the intakes be-ing tri&ate'd water. This indicates the necessity of lookinginto the bio-chemical nature of the organic binding and itsimportance for radiological protection purposes.

Reference

1. Vaze P. K. et. al. Indian Journal of Pure Applied Physics1976 Vol. 14, No. 11 p 935-936.

2. Snyder W. 5. et al. Physics in Medicine «t Biology,13, 547, 1968.

3. Sanders and Reinig: Proceedings of Diagnosis and treat-ment of Deposited Radionuclides.Symposium, Richland, Washington, May 1967 editorskornberg HA and Norwood W. D.

4. Moghissl et al Health Physics 23, 805, 1972.

TABLE I

Biological HalfUves of 3H afterinhalation of tritiated water

Case Biological Halflives (d) Fractions of 3HNo. T^i T ^ eliminated with

1. 5.84 226.0 0.965 0.035

2. 7.66 130.9 0.876 0.124

3. 9.1 554.8 0.914 0.086

IARP/10/31

Recovery of Liquid Scintillator from Tritium Monitoringat Reactor Sites

S.R. Sachan and S.D. SomanEnvironmental Studies SecticnHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Tritium monitoring around heavy water moderatedreactor site requires a very large number of samples to becounted daily. This involves consumption of good amount ofliquid scintillator (L£). Ins tag el, Tritol and Diozane basedscintillators are commonly employed for aqueous samples.Besides disposal problem, cost involvement of these scinti-llators is very high.

It would be doubly advantageous if (i) cheaper scinti-llator with better or similar counting efficiency are employed(ii) the counted scintillators are repeatedly used after decon-tamination, Toluene based LS though gives good countingefficiency for tritium in organic samples, is unsuitable foraqueous samples since it does not take water. However,attempts are made to solubilise the aqueous samples intoluene LS.

LS prepared with 70-30% mixture of toluene andethanol gives the counting efficiency of 35-16% respectivelywith water leading capacity of 3-13%. The used US can bedecontaminated by addition of excess water. The two phasesseparate out. Bottom aqueout layer containing tritium andalcohol is drained off. The top organic layer is again dilu-ted with alcohol to make up the losses and counted for resi-dual activity. It is observed that recovered IS is almostat background level. The recovered IS is spiked withtritiated water and counted again. This gives about 90-93%counting efficiency with respect to fresh one. Thus it ispossible to recycle the used LS.

IARP/10/J?

Improvements in the Activated Charcoal Filter for BettorConfinement of Airborne Radioiodi&e,

K.G. Gandhi, A.G. Kishore, D. S. Deshingkar, and A. A. KhanAir Cleaning Engineering Research SectionBhabha Atomic Research CentreBombay - 400 085

In view of high standard* of radiation safety, the radio*iedine filters must have high removal efficiency for all chemicalforms of airborne radioiodine especially for molecular iodineand methyliodide. The commercially available activatedcharcoals find limited use as adsorbent in iodine filters becauseof their poor methyl iodide removal efficiency. Commerciallyavailable activated charcoal when tested at 0. 24 sec contact timeshows 99. 9% removal effic ie'ncy for molecular iodine but has82. 8% removal efficiency for methyl iodide is high, ft i s 2%when desorption is carried out at 10 times higher face velocityfor 90 mts.

To improve these draw backs,' the activated charcoal isimpregnated with chemicals like potassium iodide, sodiumhydroxide, triethylene diamine etc. Potassium iodide impregna-ted charcoal shows better removal efficiencies for radio iodineand and methyl iodide. This impegnation is carried out by wettingthe activated charcoal inaKI solution of determined strength.The charcoal is taken out after \ houx ~>and dried. In this processthe alkali content of charcoal decreases. To compensate thisalkali doss this' charcoal is treated with sodium hydroxidesolution,*" Experimental result s reveal that the charcoal thusimpregnated has removal efficiencies 99. 9% and 99. 6% for radio-iodine and methyl iodide respectively. Impregnation with tri-ethylene diamine also improves the performance of activatedcharcoal to the desired level but being a costly chemical it is notused on large scales.

Activated charcoal after impregnation is analysed for theimpregnants quantiativeiy. For alkali and KI determination aweighed quantity of activated charcoal from average sampleof a batch is taken in soxhlet apparatus for extraction with distilledwater. A fraction extvact is then titrated against standard acid

2 6

using conductivity measurement technique KI determination isdone poteniiometrically or by liberating iodine from the extractwith potassium iodate and acid and then extracting the liberatediodine in benzene or carbon tetrachloride. This is thentitrated against standard thiosulfste solution. Gravimetricmethods are no* used because of the interfering of chloride

£8 present in tap water used for impregnation.

The advantages of impregnated activated charcoal filteri'j that it has good removal efficiency and retention for iodineand methyl iodide. Since excess free alkali is present, allchemical forms of radioiodine are trapped including HOI.

Impregnated activated charcoal has less iodine loadingcapacity and useful life as compared tc unimpregnated activatedcharcoal. V-rater content of this activated charcoal aiso increasesbecause of excess alkali present. Also it is very sensitiveto the poisoning elements like chlorine gas, acid vapours andthinaers used for painting purposes.

-4 .

iARP/10/39

An Intercomparison Study on Assessment of Lung Burdens ofLow-Energy X-ray Emitters

S. Somasundaram, R. C. Sharma, P. Kotrappa, T. Surendran,T. K. Haridasan, S. P. Gar§:, D. P. Bhanti and N. S. Pimpale

Health Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Several factors complicate the calibration of externaldetectors for in-vivo assessment of Pu and other transactinides.Many laboratories, using large-area detectors such as proporti-onal counters (1, 2, 3), phoswiches (4, 5, 6), have adopted widelydiffering calibration techniques (7). An-interlaboratory exercise(7) had been conducted through AERE, Harwell amongstEuropean and North American laboratories as participants*- inorder to compare on a common realistic basis the sensitivitiesof equipment and methods employed for measuring long burdensof low-energy photon emitters. While it suggested that phos-wich, ostensibly, is the detector of choice for such work, it alsobrought out that laboratories underestimated a subject's radio-activity contents by a factor of three or more (7).

Following the installation of a fhoswich assembly (4)comprising a 20 cm x 3 cm Nal (TI) primary detector and a 20cm x 5 cm Cst (TI) secondary detector, we performed itscalibration (8).by adopting and perfecting an in-vivo calibrationtechnique. The technique consists in making human volunteersinhale 'mock-Pu' aerosol, namely, polydisperse polystyreneparticles (AMAD = 1 urn; og = 1. 8) labelled with known relativeconcentrations of Pd-103 acid Cr-51; ascertain the onset of long-term lung retention and use that period for making variedmeasurements on subjects with a view to assess first Cr-51by established gamma spectrometry methods and then Pd-103from known Cr/Pd ratio in the inhaled aerosol. This was incontrast to the 5 um monodiserse aerosol (poly sty re particleslabelled with the same radioisotopes) which were inhaled bysubjects who participated in the international intercalibratianexercise mentioned earlier. Our first experiments were per-formed in. 1975 with Indian subjects whose body builds arecomparatively smaller.

A comparison at first glance of the calibration factorsfor Pd-103 in lungs of subjects for a single 20 cm dia phoswichlocated centrally over chest suggested th^ the estimates obtainedfrom our experiments were about two times higher than thoseexpected from extrapolation of data reported by Harwell for asimilar detector (8). Due to the low value of half-value thicknessof tissue for 20 keV x-rays and various anatomical factorsinvolved, perhaps, this was not surprising. Moreover, althoughthe phoawich detectors used at Trombay and Harwell are almostidentical, different electronic logic and instruments are employedfor effecting pulse-shape discriminaticn. As a consequence ofearlier inter-laboratory comprison, the equipment and methodsused at Harwell for calibrating phoswich stood as internationallyacceptable reference. It was therefore desirable to look for theplausible explanations for the discrepancies in calibration factorsfor Pd-103 ia-vivo. In order to do that, it was essential toevaluate the merits of our phoswich detection system, toscrutinise the accuracy of our measurements and to test andvalidate the assumptions made by us in obtaining the final values.To materialise these aims, we embarked upon a programme ofintercomparison experiments with other laboratories. The scopeof this presentation is limited to the organisation logistic andresults from the intercompxrison experiments between Trombayand AERE Harwell.

•x

Volunteer subjects for the intercomparison experimentsinhaled polydisperse polystyrene particles (AMAD = 1 um; o_=2)and were measured at BARC and at AERE, Harwell afterascertaining the end of first fast clearance phase of the inhaledmaterial. Each subject carried carefully prepared radioactivesources of Pd-Cr being representative of the material inhaled.The radioactive contents of Pd-Cr were deposited on thin filterpapers (one during the inhalation and the other from the Pd-Crsolution) which in turn were enclosed in thin polythene covers.The laboratories reported the results of the radioactive contentsof the filter papers, Cr-51 in-vivo of the subject and calculatedcounting efficiencies of the equipment used for Pd-103 in lungs.

The first intercoxnprison experiment between Trombayand Harwell was performed on subject RCS in 1975. Table 1lists the data reported by the two laboratories for the low-energyphoton outputs and Cr-51 (uCi) contents of the filter papersources. It is clear that once we correctly chose the assumptions,the two phoswich systems gave 20 keV photon outputs of the

2-9 - V

sources within 115% and the estimates of Cr-51 in the sourcesagxeed within + 5%. However, the Trombay- estimate of Cr-51in-vivo of subject RCS was about 16% lower than Harwell'sand the source of this underestimation was traced to the unacc-ounted signal loss due to the operation of pulse-shape discrimi-nation circuit for C3I (TI). Once this was corrected, theagreement was within +5% again. Our subsequent measurementsof Cr-51 in-vivo were carried out using a 20 cm x 10 cmNal (TI) detector which gave a similar accuracy. Table 2 liststhe observed counting efficiencies (count/1000 photons) forPd-103 in lungs of subject RCS for two phoswich geometries.The revised values from Trombay refer to those obtained a&ernormalisation to the same photon emission rate was made. Thedata of both laboratories are in good accord.

Following this experinnt, three further intercomparisonwhave been conducted on subjects KSS, FK and DN. A detailedcalculation following our methods was carried out with deliberateuncertainties (I 10%) introduced at each step and it was observedthat in any case the compounded error due to all these fluctuat-ions would not exceed 22%. This has been supported by theresults of all the intercomparisons performed so far. Table 3gives Harwell and Trombay data on the counting efficienciesfor these three subjects. These results taken together withthe earlier Harwell data suggest that perhaps the particle sizeof an inhaled aerosol might not affect the response of an externalx-ray detector significantly. Further experiments are nowunder way to explain the earlier discrepancies.

T*bl« }

ESTIMATES OF LOW ENEMY (M kaV) PHOTON OUT POTS (PHOTOie/mla) AND *»Cr ( Ml) FOR JTIRST W-Cr SAIiPUEB

1 O i Pd ESTIMATES ilCr ESTIMATESSAMPLES

REFERENCE HARWELL TROMBAY RATIO TROMBAY RATIO RErXRBNCS HARWELL TROUBAYEAT4CDATE (H) OLD H/TO CORRECT H/TC DATE H/T

ASSUMPT. ASSUMPT-IONS IONS

(TC)

2. llxl0s

4.24x10*

4.03x10*

9. t l

O.TS

O.tf .

2.46*1V5

4.17x10*

4.43x10*

0.7»

o.st

e.w

16.I2.TS

It . 12.79

1.99

0.27*

RCS SUSPENSION

RCS FILTER NO.3 23.12.71 J.MxlO* 4.24x10* 0.7S 4.17x10* O.St It. 12.79 0.27* 0.293 -W

RCS riLTER NO. 3 22. It. 73(INVERTED)

TABLE 2

20 cm DIA PHOSWICH COUNTING EFFICIENCIES FOR l 0 3 P d IN LUNGS INCOUNT/1000 PHOTONS FOR SUBJECT RCS. DATA ADJUSTED TO 100%PSD EFFICIENCY.

PHOSWICH GEOMETRY HARWELL VALUE TROMBAY VALUE TROMBAY VALUE(OLD) (REVISED)

1. CENTRAL OVER CHESTWITH WINDOWHORIZONTAL

ARMS BESIDE THORAY 2.03 2.79 1.71

ARMS BEHIND HEAD 2. IS 3.44 2. 11

2. GEOMETRY SIMILAR TO(1) BUT ONEPHOSWICH OVEREACH LUNG

ARMS BESIDE THORAY 2.92 S. 20 3.22

TABLE 3

COUNTING EFFICIENCIES FOR l 0 3 P d IN LUNGS IN COUNT/1000 PHOTONS FORTHREE SUBJECTS. DATA ADJUSTED TO 100%'PSD EFFIEICNCY. SINGLEPHOSWICH CENTRALLY OVER CHEST WITH WINDOW HORIZONTAL AND ARMSBY SIDE POSTURE OF THE SUBJECT ARE ASSUMED

SUBJECT HARWELL VALUE ADJUSTED TROMBAY VALUE

KSS

PK

DN

2.44

2.01

3.73

2.36

1.70

3.20

IARP/10/2

Estimation of Radium«226 by Eraanometry

M. Raghavayya, M.A. R. Iyengar and P.M. MarkoseHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

1. Introduction

Conventionally 2 ^ R a is estimated in the laboratoryby co-precipitation with a carrier, barium. The precipitate:is transferred after separation onto a stainless steel planchet,dried, fired and then the activity thereon is determined withan alpha counter. The radium content is derived from thecount rate after applying corrections .for recovery of radiumfrom the aample, counter efficiency,' build up of daughteractivity and so on. The precipitate on °the planchet is usuallybulky because of the carrier and generally spread on it nonuniformly. This introduced errors due to self absorptionof alphas. Secondly, a part of the radon built up in thesample tends to diffuse out of the planchet thereby causingan underestimate of the final result. It is rather difficultto quantify this loss of radon since it is governed by severalparameters. A rough estimate is of the order of three toten percent. In order to overcome these drawbacks, radonemanometry can be used for estimation of radium.

2. Principle

The principle of emanometry is simple. Radium inthe sample is brought into solution.(if it is not in solutionalready) by suitable chemical extraction. After removingwhatever radon is in the solution by aeration, the solutionis allowed to age for controlled build up of radon. Theradon thus generated in the solution is milked after a suitabledelay and estimated. The radium content of the originalsajnple is calculated by extrapolation.

3. Procedure

The solution containing radium is tajken in a bubbler

34

of about 85 ml. capacity. Usually 70 ml. of the solution istaken. The sample is initially aerated by means of a suctionpump for removal of in-built radon.' The bubbler is thensealed and allowed to stand for a few days generally equiva-lent three to fovx half lives of radon. At the end of thisbuild up period, an evacuated scintillation flask of 150 ml.capacity is connected to the outlet of the bubbler. When thestopcocks are carefully opened, air enters the scintillationflask after bubbling through the solution column. The Bintsreddisc provided in the bubbler breaks up the air stream intotiny bubbles thereby causing vigorous effervescence. Thishelps to quantitatively carry the in built radon into thescintillation flask « ong with the air.

3.1. In about 200 minutes the radon daughter activity in thescintillation flask reaches secular equilibrium with radon.Thereafter the scintillation flask is coupled to a radoncounter and the alpha counts obtained during a specifiedinterval, usually 10 minutes (counting can be done for longerduration, if necessary depending on count rate), are noted.The radium content of the solution taken in the bubbler isobtained from the following expressicr> (RAGHAVAYYA, 1976)?

Ra(pCi) =1.883 z 10"3 x C

E * e"** x (1- e "**) x (1-e"*0)

whore C is the gross counts during the countingduration,

£ is the efficiency of the flask (percent),t . is the counting delay (rain.),T is the counting duration (min.)© is the time allowed for radon build up in

the bubbler (min. )^ i s the decay constant of radon

(1.258 x 10"4 min-1)

4. Conclusion

The em anometric method was compared in the laboratorywith the conventional co-precipitation method. The formerwas found to have several advantages over the latter. Firstof all in the case of moat of th» ifflii»^t nmpl«

is necessary except nitration. An aliquot of the effluentsample can be taken in the bubbler straight away. If theradium level in the sample is very low a pre-concentrationstep is all that is necessary before taking it in the bubbler.For other samples the radium has to be brought intosolution. In all the cases precipitation and separation isavoided.

The method under discussion is selective for Rasince only radon is milked out of the solution for furtherestimation. So interference from oiJier isotopes of radiumis not a problem.

From our experience we find that the emanometricmethod is very simple in operation and is accurate.

•References

Raghavayya M., 1976 : A study on the distributionof radon in uranium mines. M.5c. thesis, University ofBombay.

IARP/10/3

Comparison of the Diffusion Chamber and the Wire-ScreenSampler fox Estimation of Unattached Radon-Daughters

A.H. Khan, G.K. Srivastava and M.Health Physics UnitP. O. Jaduguda MinesSinghbhum, Bihar

Raghavayya

1. Introduction

Among the devices available for estimation of theunattached fraction of radon daughters, the diffusion chamberand wire- screen sampler are useful in uranium mines. Whilethe diffusion chamber (KOTRAPPA et al, 1975) is based on anexact theory postulated by MERCER and MERCER (19?0),the wire-screen sampler relies on a semi-emperical proposi-tion (THOMAS and HINCHLLEFE, 1972).

Performance of the two devices has been compared inlaboratory and their relative merits for routine use in mineshave been discusssed.

2. COLLECTION EFFICIENCY

The collection efficiency of the diffusion chamber isderived from considerations of diffusional losses from a fluidflowing radially inward betwesa concentric, parallel, circularplates, and that for the wire screen is derived by consideringthem as an assembly of very short diffusion tubes. Table-1shows the effect of flow rate on the co llection efficiencies ofthe two systems.

Table-1. Comparison of collection efficiencies of diffusionchamber and wire screen sampler.

Flow rate(1/min)

0.100.501. OC5.00

10.00

3. EXPERIMENT

Collection efficiency (%)Diffusion chamber J_._Both plates

99.9993. 8376.2830.2019.27

AND RESULT

I One Flate l"

49.9946.9138.1415.109.63

Lre-screen sampler

100.00100. 00

99.7173.5353.41

The experimental setup consisted of a steel barrelof 200 litre capacity provided with two sampling outlets andan air inlet. Radon from a drum containing pulverised uraniumore was introduced into the barrel. The diffusion chamberand wire-screen sampler, each having a back up filter, wereconnected to the sampling outlets. Air was sampledsimultaneously through both the devices at indentical flowrates for 5 min. The alpha activities on the diffusion chamberplate, wire screen and on the respective backup filters 'wereseparately counted for three optimum counting periods. Theradon daughters deposited on these surfaces wsre calculatedand their unattached fractions were computed.

The total unattached fractions of radon daughtersobtained from the two devices in a series of experimentswere subjected to a least square analysis. The agreementbetween the results was rather good.

4. REFERENCE

KOTRAPPA P., BHANTI D. P., and DHANDYUTHAM R.Health Phys. 29, 155 (1975).

MERCER T. T. and MERCER R.L., Aerosol Sci. 1, 779(1970)

THOMAS J. W. and HUMCHLIFFE, 1972 J. Aer JSOI Sci (1972).

1ARP/10/14

Computed Fast Neutron Kerma. and TL response of BeO

K.S. SharadaDivision of Radiological ProtectionBhabha- Atomic Research CentreBombay - 400 085 --

Fast neutron kerma values for BeO have been calcu-lated in the present work. From the high - LET and low -LET components of the computed kerma, the T. L. responsehas been estimated using the published LET response data forBeO. T&e energy range covered is from 0. 02 to 14 MeV.The kerma calculation has been done for (a) the bigh-LETcomponents viz (i) Be-9 recoil arising from (n, n) eventsr~(ii) He-6 recoil and the He-4 from the (n, alpha) events; and(iii) Li-7 recoil and H-3 from (n, t) events: and (b) the low -LET component arising from the short-lived beta emissionfrom the decay of He-6. The anistropy of (n, n) is conside-red as available in the third edition report of the. Brookhaven National Laboratory no. BNL-400 (1970). For energiesbelow 1.2 MeV only the elastic scattering contributes tokerma. Abisotropy factor is significant for E n y 0.8 MeV..Beyond 2. 5 MeV, isotopic scattering would overestimate the(n, n) kerma by a facto? of 1.6 to 4 as the nsutron energyincreases. .- The (n, alpha) component is around 30% of thetotal kerma. in the range 3 MeV to 10 MeV and, beyond 10MeV, drops to 15%. The (n, t) component starts only be-yond 12 MeV and rises to 5% of the total kerma at 14 MeV,The lone low - LET component from the residual nucleusHe-6 of the (h,, alpha) events rises from E n = 1.2 MeV to

-about 10% of total kerma at 3 MeV and thereafter falls toless than 2% at 14 MeV.

Employing the LET-TL response data from Tochilinet al1 for the low-and high-LET components, the expectedTL response-vs-neutron energy for BeO has been obtained.A mean value of 2.2 was obtained from the experimentaldata for the relative TL efficiency of the high-LET componentcompared to the low-LET beta radiation.

The thus-computed neutron TL response-vs-neutron

energy shows a generally rising trend -with energy with anumber of 'local' hills and valleys. To give an idea of therange of the variation, the relative TL responses are1 (0.02 MeV), 10(0.02 Mev), 50 (1 MeV), 82 (2 MeV),150 (3 MeV), 150 (4 MeV), 150 (5 MeV), 170 (7.5 MeV),200 (9 MeV), 245 (10 MeV), 360 (14 MeV). These relativesensitivities are for the same fluence. In terms of dosethe relative response is 1 (En<£ 1 MeV) and this rises toapproximately to 2 (for E n > 2 MeV) The above data appliesto fairly large thicknesses of BeO powder. As the He-6beta radiation is of high energy (£ a (Max) = 3 MeV), forBeO layer thicknesses of the order of a few mg/cm ,the £ dose in BeO would be numerically equal only to afraction of the beta lterma. This fraction has also beencomputed for the usual thicknesses one would normally comeacross in practice. Together with this fraction, the TLresponse of the BeO-vs-incident neutron energy can beestimated.

Reference

1. E. Tochilin, N. Goldstein and J. T. Lymen,Proc. Sec. Int. Conf. on Luminescence Dosimetry,CONF-68092 pp. 424-437 (1968)

LARP/10/26

CaSO4:Dy Embedded Teflon TLD Tape for PersonnelMonitoring

Bhuwan Chandra, A.S. Pradhan, A. R. Lakshmanan,K. L. Popli and R. C. BhattDivision of Radiological ProtectionBhabba Atomic Research CentreBombay - 400 085

Introduction

CaSO^Oy embedded Teflon TLD discs have foundincreasing use in radiation dosimetry and personnelmonitoring. Based on these discs a TLD badge has beendeveloped and has been found quite useful for personnelmonitoring applications. In order to cater to the need fora large number of dosimeters in our countrywide personnelmonitoring programme, CaSO rDy embedded Teflon tapehas been developed for providing a large number of dosi-meter discs af a known and constant sensitivity.

Experimental

From a 2.5 kg mixture of CaSO^Dy and Teflonpowder in a weight ratio of 1:4, a tape (0. 8 mm thick and16.0 wide) was developed with the help of a local plasticprocessing company.

Discs of 13. 5 him diameter were punched out fromthe tape. Weights of these discs were taken and it wasfound that in a typical batch of discs, 90% of the discswere having their weights in the range of 260-270 mg. andthese were only selected for further study. These discswere divided into five equal batches, out of which fourbatches were heat treated at temperatures 250°C, 300°C,350°C and 400°C for one hour duration each, and the fifthbatch was not given any heat: treatment.

Since shrinkage was observed in the discs obtainedfrom the raw tape after temperature treatment both at35G°C and 400°C two Bets of the tape pieces of length 6 cm

A\

each were cut out from the untreated tape and were giventemperature treatments of 350°C-l hr and 40C°Cri hr resp-ectively. After this treatment, circular discs were punchedout from these tapes.

All these discs were loaded into the aluminium cards(having dimensions 5.2 cm x 3. 2 cm x 0.08 cm) and exposedto 2.5 R of °"Co gamma rays. They were readout on aTLD card reader. For reuse, annealing treatments ofeither 200°C - 16 hrs or 250°C-l hr were employed allthrough. A minimum number of five cards were taken ineach stage of reuse. An exposure-readout-anneal cycle wascontinued for 15 cycles of reuse. :.

Results

The variation in the TL outpit of three independentsets each containing 30 discs which were taken from differentregions of the tape length and pre-treated at 400 C for onehour was less than - 5% (10") at the' above mentionedexposure level. It was found that the untreated tape <£scsshow a heavy deterioration in the TL sensitivity in fifteencycles of reuse, whereas, the treated cards show much lessdeterioration. This is due to the lower mechanical strengthof untreated discs which develop a depression in the centreduring heating and recycling and as a result do not get intoa proper contact with "the heater strip of the TLD reader.The 350°C as well as 400°C heat treatments to the discsprior to their use give them sufficient mechanical strengthfor subsequent use. However, the cards based on .the discspunched out' from the tape treated at 400°C give the bestresults. No significant cuange, in the optical density of 400°Ctreated discs has been observed after 15 cycles of reuse.

Conclusion

1. The tape provides an easier method for productionof large number of CaSO4:Dy Teflon TLD discs.

2. The additional heat treatment to the TLD tape isnecessary for imparting to It strength and flexibility.

3. The 400°C-l hr treatment gives the best results forreusability of the TLD cards which, are .made frojn..the_ IXiD_discs punched out from the tape.

41.

4. About 8000 discs can be obtained from the tapemade from a convenient amount of 2.5 kg mixture of

and Teflon.

43

IARP/10/24

Effects of Intense Neutron Flux on. TLD PersonnelMonitoring Cards

R.R. Vishwakarma, R. G. Khadake and P. GangadharanDivision of Radiological Protection.Bhabha Atomic Research. CentreBombay - 400 085

Introduction ,

Most of the nuclear establishments all over the wor,dare switching over to TLD badges for personnel monitcring.The reliability, the wide dof° range and the high sensitivityof these badges are the main reasons behind this change.During the dose measurement and annealing processes, theTL dosimeteres are subjected to high temperatures. There-fore, even though the cassettes of these badges are genera-lly made out of low Z materials, the TL, dosimeters aremounted on metallic cards (Ref. 1, 2) Thus the use of ametal in a TLD badge as a backing material and hence its -activation particularly due to neutrons in a mixed radiationfield is inevitable. The amount of radioactivity produced ina badge would depend upon the- neutron fluence and the badgematerial. Therefore, the study of neutron activation in aTLD badge and its contribution towards the total dose is ofconsiderable interest. Even if the instantaneous activityproduced in the badge material during irradiation is not veryhigh, due to the continuous exposure of the TL dosimeterto the radiations emitted by the activity induced in thebadge, its contribution towards the total dose soay besignificantly high. Henc , for accurate beta and gaxmna dosesmeasurements, the contribution due to activation should beevaluated and the necessary correction should be ap.plied.

The neutron activation produced in a typical TLDpersonnel monitoring device being used in this departmenthas been measured (Ref. 3). The badge consists of threecircular discs of TL material, each of a diameter 13 mm,made out of a mixture of CaSO^Dy and Teflon in a weightproportion of 1 : 3 (Weight of each disc is 280 mg). Thesediscs are mounted on a chromium, plated aluminium, card.

This card alongwith a paper strip containing the badgeidentification information is loaded inside a plastic cassette.

Measurements

The TLD badge including the plastic cassette and themetallic cards with and without TL discs were exposed tothermal and fast neutrons separately and the induced gammaspectra were measured using 3 x 3 inch Nal (TI) detectorand a 400 channel analyser spectrometer assembly. It wasfound that the radioactivity produced in the cassette materialand the TL discs were negligible as compared to that in thecard. Thus neutron activation measurements were carriedout with TLD cards alone. Two different sets of TLD cardswere irradiated in thermal and fast neutron columns inAPSARA reactor separately. The induced activities due tothermal neutrons is predominently due to ^7 Al (Y1,Y ) ^ A l •reactions Here the product 28 .1 is a short lived isotope c

and decays with a half life of 2. 3 min. Fast neutrons inducesome more energies in the card mainly due to ^ A 1 ( T ^ ^ J ^ N areactions. These induced energies have been recorded byusing the above said spectrometer assembly. Beta gammadose contribution due to thermal and fast neutron activationof TLD cards (at saturation) have been measured by usinga semi automatic TLD reader (Ref. 4) separately and theyare found to be 1/3 rd of the total gamma dose inside thereactor core.

Results and Discussion

The increase in the dose due to neutron activationof the metallic card in the TLD badge as described abovecorresponds to high thermal and fast neutron fluences. Inmost of the noTirial wo rking" conditions in which radioisotopes

. including neutron sources are used, such high neutron fluxesare not expected and the dose contribution from the neutronactivation of the TLD cards are negligible and can be ignored.However, in the vicinity of a neutron generator as •well asin case of a criticality accident the neutron fluence encount-ered is high. In such cases the contribution due to neutronactivation would be quite appreciable and the doses would beover-estimated. If the badge is exposed in a high neutronflux, the activity induced in the card gives rise to a surfacedose rate indicating that the badge: has been exposed to

high neutron flux. Under such conditions it would bepossible to correct the exposure recorded by hhe badge andincidently indicates the unexpected high neutron exposure.

References

1. D. Herrman, W. Kraus, W. Will - 4th Int. Conf .onlumniscent dosimetry Krakow - Poland, Aug. 1974pp 801 - 814

2. D. Grogan, V. Balasubrahmanyam. Ibid pp 841-858

3. "Vohra K.G. et al - Proc. 5th Conf. luminescent dosimetrySao Paulo. Feb-1977

4. A. Sankaran et al - 3rd Annual Conf. on Rad. Protection,Vigyan Bhawan, New Delhi, Jan. 19-21 (1976) pp 22-23.

IARP/LO/7

Influence of Detector Size and Air Gap on the Sensitivityof Bonnor Spheres

. Dhairyawan, P.S. NagaxaJ&n and G. VenkataramanDivision, of Radiological Protection.Bhabha Atomic Research CentreBombay - 400 085

Boaner sphere type of instrument are widely used inneutron dosimetry. The system consists of a polythenemoderating sphere with a centrally placed thermal neutrondetector. Detectors of different types have been used: Lilscintillatora of sizes ranging from 25 mm dia x 2 mm longto 4 mm dia. x 4 mm long, ^^BF, and' ^He counters of dia1" or 2". The diameter of the spherical moderator variesin the range of 2 inch to 12 inches. The Influence ofdetector size or detector itself on the response of thesphere has been neglected all along on the tacit belief thatthe response curves will be parallel when one detector isreplaced by another and what is needed is only a calibrationat a known neutron energy. So far only one calculation isavailable on the influence of the detector size.

This paper presents the results of a Monte CarloCalculation, on the response function of 3" and 10" diameterpolythene spheres where the central detector is a Ldl scinti-llator of varying sizes or BF? counters of diameter 1" and 2"The beat microscopic cross section data available in theliterature was used for all the constituents of the moderatorand detectors. Anisotropy of neutron scattering was takeninto account. 5000 neutron histories were followed for eachof the neutron energies ranging from 10"7 MeV to 14 MeV.

The results shew that for a 10" diameter spherethe response is proportional to the volume of the BF3 dete-ctor whereas for a 3" diameter moderator no such simplerule exists. The epithermal neutrons inside the moderatoralso contribute to the sensitivity of the smaller diameterspheres and therefore the epitliermal response of the centraldetector must be taken, into account while changing from Ldlto BF^ or some other detector.

Al

The effect of Lil Bcintillator size is also studiedfor three different sizes. The other result presented in thepaper is the influence of the air gap which is necessarilypresent when a central detector is introduced. It is seenthat for an air gap of about 0. 7 cm thickness sensitivityfalls by a factor of three for a 5" diameter polythene spherewith a Lil detector of 0.49 cm diameter. The influence ofair gap for two other different sphere sizes are studied,These are the first studies of this type.

4S

IARP/10/8

Angular Dependence of Fast Neutron Personnel MonitoringKodak NTA Film Pack for Neutron Energies up to 14 MeV

P.S. NagarajanDivision of Radiological ProtectionBh&bha. Atomic Research CentreBombay - 400 085

Personnel monitoring of fast neutrons is done usingthe Kodak Nuclear emulsion Type A (NTA) film pack. Thepack has a fine grain nuclear emulsion layer coated on ahydrogeneous film base and wrapped by light-tight paper.Thus the emulsion layer, about 30 micrometre thick, liessandwiched between two hydrogeneous material layers, vizrfilm base, (cellulose acetate) and paper amounting to about70 mg/ciBT on the back surface and paper amounting to25 mg/cm on the front surface. These two layers serveas proton radiators on neutron irradiation. The emulsionrecords recoil protons emerging from these radiators inaddition to those generated within the emulsion itself.Limited experimental results of angular dependence of NTAfilms are available (1). As regards theoretical results,none is available fer angles of incidence other than normalincidence. The present computational work gives the angularresponse of NTA film pack over the fast neutron energyrange of up to 14 MeV.

In the case of normal incidence, only the frontradiator can contribute recoil protons for track formationin the emulsion, since the protons are emitted only in theforward direction. For angles of incidence other than normalincidence both the front and back radiators contribute to pro-tons tracks in the emulsion. For a given incident neutronof energy EQ in. the direction Qm, a recoil proton producedat a depth x from the emulsion-radiator intereface and emittedin the direction (01, fo ) could be recorded as a recognisables^ack in the developed film if the range R of the protonaatisfies the inequality R^{(x/cos 0i )+ R B ] , where RB is thebias range, taken as 0. 3 Jim in this study, since tracks oflength less than Rj$ in the emulsion are not recognisable astracks. The integration for a given E and Q,, is carried

over 0 1 , Yi.an** x» with the condition given by the abovementioned inequality.

Results are obtained for EQ= 1.0, 2.5, 5.0, 7.5, 10,12 and 14 MeV and for each EQ at incident directions CosQs0, 0.2, 0.4, 0.6, 0.8 and 1.0; and for the two radiatorthicknesses used in the NTA film pack. The sum of thenumber of tracks in the emulsion arising from both theradiators and the number of those originating within theemulsion arising itself gives the film response. The responseas (tracks/cm ) per unit incident fluence is thus obtained inabsolute units. The computed results were compared withthe available experimental ones U). The agreement is gooda.nd within acceptable experimental errors. At E Q ^ 1 MeVthe angular dependence is not significant and within about10%, This is because not many recognisable tracks arisingfrom the radiators are produced in the emulsion due to thesmall range of recoils. But at higher Eo 's, the angulardependence is significant. For example the ratio of theresponse for normal incidence to that for parallel incidenceis 1.1, 1.5, 2, l . r . 1.4 and 1.4 at Eo=l, 2.5, 5, 7.5, 10and 14 MeV. Thus, in addition to the previously known andrecognized uncertainties in fast neutron personnel monitoringauch as energy dependence, fading, etc., that arising fromangular dependence of the film should also be recognisedas it is seen to be significant.

Reference

1. Kathren R. L., Prevo C. T. and Block S. HealthPhysics 11, 1067(1965), also Kathren R. L.,Health Physics 13, 1039(1967)

IARP/10/25

A Versatile Instrument for Counting Luminescent Photonsfrom Chemicals or TLD PhoBphors

S. Kannan, V.H. Risbud and P. GangadharanDivision of Radiological ProtectionBhabha Atomic Research CentreBombay - 406 085

The use of Phosoa counting technique in the measure-ment of low levels of luminescence in dosimetry (e.g. Ly-oluminescence, thermaluminescence and radiophotoluminescence)is known to improve the aignal-to-noise ratio by an order ofmagnitude over that obtained with D. C. techniques (1,2).This signal-to-noise ratio improvement is based on the pulseamplitude discrimination and on the stability of the digitalprocessing circuitry. This .schnique does not eliminate thenoise pulses with amplitudes which fall within the window ofthe discriminator circuits. Consequently, the pulses due tothe spontaneous emission from the pho toe at node or due tostray light or system noise with amplitudes in the windowregion of the discriminator set the lower bounds for anymeasurement. This paper describes a Photon Counter/Processor developed in our laboratory, that uses the digitallock-in technique (3) to provide an additional signal recoverycapability.

The system makaB use of a mechanical chopper tointerrupt the signal periodically. When the signal is blackedthe accumulated counts reflect only the noise sources whichlie anywhere between the chopper and the final output. Whenthe input signal is allowed to fall on the photo-cathode, theaccumulated counts are due to true signal, information aswell as to the system noise. In the Photon Counter/Processordeveloped, two separate 8-digit counters register the countsduring the 'on' and 'off' periods and the difference betweenthe contents of the two counters at the end of the preciselycontrolled accumulation time, gives the signal information.In a typical case where the signal and noise pulses displayPoisson statistics, the signal-to-noise ratio 1B proportionalto the square root of the measuring time. Since no outputI£jro-drift limitations exist in the digital technique, it is

51

possible in principle, to measure the signal to accuracylimits determined by the noise in signal itBelf and not to thelimitB that are introduced by limitations in the detector orsignal processing system.

The instrument consists of a fast pulse amplifierwith selectable gains of 10, 50, 100 or 200, a fast pulseheight analyser, a well-regulated, short circuit and over-voltags protected EHT for the PMT, two 8-digit countingchannels A and B, a crystal-controlled timer, an arithmeticlogic circuitry to compute A+B or A-B and a digital-to-analogconverter to provide the analog output for recording purposes.An 8-digit sev.eit-segment LED display can be selected todisplay A, B, A+B, A-B or live counting time. Front panelthumbwheel switches select the total counting time and thechopper sampling time. Chopper timing information isavailable at a rear panel BNC connector for oscilloscopedisplay. A built-in logic circuitry ensures that irrespectiveof the preset time or preset count, equal number of thebackground an-i signal -Samples are accumulated in the respective channels. The pulse height analyser has a 'coincidence-correct1 mode of operation in addition to the 'integral' and'window' modes. In this mode, the pulses exceeding theupper threshold of the discriminator are counted twice and thisprovides a first order correction for the coincidence of twophotons at the phctocathode at high counting rates. Theinstrument also has 'single cycle* and 'test' modes of opera-tion besides the 'chop' mode. In the 'Test' mode, a 1MHztest signal is fed to all active inputs to evaluate the operati-onal readiness of all counters, the arithmetic unit and thedigital readout and any malfunction is visually discernableon the display.

The Photon Counter/Processor is thus a versatilecounting and computing instrument which serves as an adapt-able interface between photo or electron, multipliers and digitalcomputers, printers or similar output devices. This system,in conjunction with suitable thermoluminescent or radiophoto-luminescent #materials, will find wide use in low level radiationmonitoring (e.g. background radiation in radioactive miningand milling and waste management operations)

1. J.B. Lasky, D. W. Pearaon and P.R. Moran USAECReport COO-1105-188 (1973)

2. I . HSchlesingeri A.Avni and Y. Feige-RISQ-Report 249 (W71)3. The ORTEC Physical Sciences Division Catalrgue (1976)

1ARP/10/28

Safety Considerations in a Fuel Reprocessing Plant

H.R. Siddiqui and S. V. KumarFuel Reprocessing Division andS. KrishnamonyHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

The reprocessing of spent nuclear fuel presents awide range of safety problems including not only the problemsgenerally encountered in a conventional chemical plant butalso those arising due to high levels of radiation, criticality,discharge of effluents from the plant to environment etc. Atthe same time there is need to protect the operation staff,the members of general public and the environment from theeffects of the radiation and radioactive discharges. It is thusobvious that a fuel reprocessing plant must incorporate thehighest safety guarantees.

There is a basic functional difference between a re-processing plant and a nuclear reactor; the safety considerat-ions, therefore, are different in many respects. For example,the fuel reprocessing plant does not have high systempressures as a nuclear reactor has, the bulk of the radio-activity due. to fission products is in liquid state unlike in areactor and finally, inspite of the high degree of remotisationemployed in the operation, there is extensive direct humanaccess to the radioactive material in a reprocessing plant.The basic safety approach in the design of reprocessingplants thus has to take into account these specific charact-eristics and differences.

The main process steps in a reprocessing plant thatare of concern from the point of view of safety and protection,both of the personnel and equipment, are mentioned below:

(a) Fuel element receiving and storage requires extremeprecautions in handling Co avoid risks of cask or fuelelements falling outside or even inside the storage pool,efficient removal of decay heat from the storage pool,and careful geometrical layouts of fuel elements in thestorage pool to avoid criticality.

(b) Dissolution: During dissolution of the chopped fuel thegaseous wastes released include iodine, krypton, andxenon which must be removed through an efficient off-gas system. There, are the criticality problems also•during dissolution which have to be taken care of.

(c) The Solvent used is generally tributylphosphate, dilutedto 30% with a light hydrocarbon or odourless kerosene,having a flash point fairly close to the operating temp-erature of the solutions processed. It can be seen thatthe risk of solvent fire is not negligible.

(d) Concentrator which contains uranyl nitrate, nitric acidand entrained tributylphosphate used as a solvent, at atemperature of more than 160 C, can lead to an explo-sion which would not only extensively damage the processvessel concerned and the other vessels in neighbourhoodin the cell, but would also result in loss of valuablenuclear material.

(e) The high active liquid wastes containing fission productsin a concentrated form with high specific activities arestored in large underground stainless-steel tanks withcooling provision, since these wastes generate heat.The loss or failure of cooling system can lead to unduepreseurisation of the storage vessel thus affecting itsintegrity of- containment.

(f) The avoidance of criticality accident is of paramountimportance in a reprocessing plant. Special attentionis paid to undue accumulation of fissile material inany process vessel. The nuclear safety in this regardis achieved by using geometrically ever-safe vesselsand also by exercising control on mass and volume ofthe solution, whereever possible.

La order to provide maximum safety of the plant, even

under extreme conditions, all conceivable modes of malo-perations and accidents are systematically analysed andprecautions taken to limit the consequences of such accidentsto tolerable levels. The basic technological challenge insuch a plant arises from the need to handle with almosttotal containment, highly radioactive solutions which arechemically reactive as also corrosive. The presence ofplutonium in high concentrations in an easily dispersibleform adds an extra dimension to the problem in view of itsradiotoxic and fissile properties. The abnormal occurrencein a reprocessing operation can cover a wide spectrum fromminor leaks to uncontrolled explosive reactions, within theprocess system. This presentation deals with an analysisof predictable occurrences during the various stages of fuelreprocessing.

IARP/10/33

Development of a Fissile Material Transport Container

T.S. Laxminarayanan, B. Haraprakas, K. Vijayan,V. P. Balakrishnan, S. Rajagopalan and P. P.S. JacobFuel Reprocessing DivisionBhabha Atomic Research CentreTrombay, Bombay - 400 085

Fuel reprocessing is an important step in the nuclearfuel cycle where the strategic nuclear material like Pu whichis fissile is separated from fission products and uranium. Toenable further use of this separated fissile material, it hasto be transported in a safe manner from the fuel reprocessingplant to the fuel fabrication facility either directly or throughan intermediate step of storage facility. In view of the in-herent hazards involved in the nature of the material there isan imperative need for ensuring safe transport of the same.

The IAEA regulations (1967-73 editions) spell out thesafety criteria which should be built into the design of con-tainers and it is the responsibility of the consignor tosatisfy the requirements while euch transport is effected. Thecontainers are designed to withstand the effects of severeimpact and of fire in the event of an accident. In additionthey must be designed to prevent any criticality incident.The regulations also call for the design to protect transportworkers and the general public in both normal operationsand accidents.

In order to meet the above requirements, containers.designed for transport of fissile materials, have to survivea series of tests which are simulated to produce damage tothe containers equivalent to that which would be produced ina severe accident. These tests include the following :

(a) A 9 metre drop on to an unyielding target in themost damaging attitude;

(b) A 1 metre drop on to a metal bar 15 cm in dia.and at least 20 cm long;

(c) A fire teat of 800°C for 30 minutes followedby a natural cooling for 3 hours; and

(d) An external pressure test by immersion in 15metre depth of water or water at a pressure of1.5 Kg /cm gauge for 8 hours.

The fire test has to be carried out after, the two drop testson the same container.

This presentation deals with the development andtesting of a transport container to meet the IAEA regulationsforsafe transport of Pu by road. The following work wasdone :

(a) An electrically heated furnace was specificallyfabricated for testing the various designs of. thebird cage and pressure vessels;

(b) A pressure tank was fabricated for simulatingthe 15 tn£tre water immersion test, when thewater can be subjected to the required pressure;

(c) Fire tests were conducted on dummy containershaving instrumentation for estimating thetemperatures reached at various points;

(d) The developed transport container was put throughall the IAEA recommended tests.

After the tests, the container was examined and foundto have only minor damages on the external containers andabsolutely no damage to the sealed inner container.

The above container i- the first in the series fordevelopment of a safe transport container and modificationsare being incorporated on the basl s of data obtained.

57

IARP/10/42

Decommissioning of Trombay Fuel Reprocessing PlantRadiological Health & Safety Aspects

T.K. Theyyunni, B. IvL Sidhwa and M. N. NadkaraiFuel Reprocessing DivisionBhabha Atomic Research Centre, BombayBombay - 400 085

Fuel reprocessing forms an important step in thenuclear fuel cycle. It is in this step, the spent fuel fromthe reactors is converted into liquid, form from their originalsolid matrix. Operation of a plant of this naturenecessitates a high degree of radiological health and safetyin view of radioactivity encountered from fission productsand fissile materials.

The Trombay Fuel Reprocessing Plant, the first oneto be built in India, was taken up for decommissioning aftera number of years of operation with a view to modify andrebuild the plant to meet the increased reprocessing load.

The decommissioning of a fuel reprocessing plantrequires high standard of planning and team work as it is inthis job a number of persons have to work in direct contactwith radiation both during decontamination and decommissip-ning operations.

The decommissioning operation for the plant atTrombay was carried out in the following sequence:

(1) The recoverable nuclear materials were removedby internal rinsing of process equipment andpiping in the cells.

(2) Detailed radiation surveys were carried out inprocess areas housing equipment and piping todetermine the extent of decontamination requiredfor reducing the radiation dose levels.

58

(3) Internal decontamination of process equipmentand piping was carried out using chemicals such.as nitric acid, sodium hydroxide, tartaric acidand oxalic acid - hydrogen peroxide mixtures.The liquid wastes generated at this stage wereconcentrated and stored in storage tanks.

(4) Once the radiation levels of the in-cell equipmentand piping were brought down to values lowenough to permit personnel entry, more detailedinspection of the cells to plan further- decontami-nation and dismantling operations, was carriedout.

(5) Dismantling of equipment and associated pipingof the systems was planned and executed usingvarious types of cutting and handling tools.

(6) Final decontamination of the plant areas wascarried out after completion of dismantlingoperations. The levels of contamination andradiation field achieved were as per specifiedlimits fixed taking into account the requirementsfor carrying out fresh installation work.

In order to ensure radiological safety and limit thepersonnel exposure to the minimum, each area was consideredseparately and detailed planning -was worked out as givenbelow:

(1) Radiological status of the area was carefullystudied with the help of survey reports.

(Z) The nature of the job to be carried out wasclearly spelt out.

(3) Approximate exposure requirement for a particularjob was determined.

(4) The number of persons required for each jobwas determined.

(5) Exposure and time allowed for each individual toperform the job was decided.

(6) The type of job each person has to carry out wasexplained with the help of sketches and mock up.

(7) Material requirement for performing the duty,including health and safety equipment were workedout and arranged.

(8) One Supervisor was assagned to carry out all jobsas planned in each area.

(9) All the radiation workers were provided with,protective clothings including ventilated plasticsuits and air line respirators as specified bythe RMC unit.

(10) Individual doses received were scrupulouslyregulated keeping the permissible limits in viewand the dose records were systematicallymaintained and reviewed as the job progressed.

Concluding Remarks

A number of process cells and affiliated process areashave been successfully decontaminated and decommissionedwithout even a single case of over exposure beyond permissi-ble levels as stipulated in the ICRP Regulations. Therewere also no cases of spread of contamination to inactiveareas or activity release into the environments beyondpermissible limits which amply proves the importanceattached to radiological health and safety aspects while(executing a job of such stupendous nature.

A

IARP/10/10

Non destructive Isotopic Assay of Plutonium*

M.R. IyerHealth Physics DivisionBhabha Atomic Research CentreBombay -. 400 085andK. Eberle and H. OttmarInstitute for Applied Nuclear Phys\csNuclear Research CentreKarlsruhe, FRG

The precision non-destructive analysis of Plutoniumisotopic ratios using high resolution gamma spectroscopyoffers a very useful tool which can replace the more tediousmass spectrometric analysis under certain circumstances,Quantative measurements on bulk Plutonium samples usingneutron coicidence or calorimetry require information on theisotopic composition for the interpretation of the results andgamma spectrometry provides the.answer. The methodstandardised employs characteristic gamma rays from thedifferent Pu isotopes and uses a fundamental parametertechnique without the necessity of any standards.

The iaotopic ratio N (i)/N (k) of two isotopes i and kare related to the ratio of measured peak intensity I (i) andI (k) by the relationship

where A and RE are the specific gamma activity and relative;efficiency of the gamma rays with energies E(i) and E(k) fromisotopes i and k respectively. This needs nuclear data onthe two isotopes which are available in most of the caseswith sufficient accuracy. The relative efficiency functionincludes the effects due to self and matrix attenuation in thesample, source-detector geometry and the intrinsicefficiency of the detector. This is a sample dependent para-meter which is to be generated for the particular sample.

Thus the precision of the analysis is determined by theprecision with which the nuclear data is known and theprecision with which peak areas can be extracted afte?necessary correction for compton contribution.

The paper gives the results obtained using the methodfor various Pu samples and discusses the application andlimitation of the method. Pairs of gamma, rays suitable forthe different ratios are also given. The peak areas havebeen determined using two different computer codes and achannel summation method. The results of isotopic ratiosare compared with destructive analysis results.

Pu-239/Pu-241 ratio:

The results of these studies show that the pairs ofgamma lines between 144 to 208 keV are best suited foranalysis of low burn up Plutonium. In case of high burnup Pu the 330-420 keV region provides a better choice usinga multi group analysis.

Pu-238/Ptt-239 ratio:

This pair offers a clean set of gamma lines at 152keV and 144 keV respectively. The destructive analysisin this case is by using alpha spectrometry and these studiesindicate that gamma spectrometry is more reliable than alphaspectrometry.

Pu-240/Pu-241 ratio:

The measurement of this ratio is the most difficultFor small sample analysis the 160 keV complex can be usedbut this involves the extraction of 160. 3 keV peak area fromthe 160 keV complex. The accuracy of this is thus limitedby the relative intensity of these two lines in the sample.The 640 keV complex provides a better choice particularlyfor thick samples.

+ Work performed under the framework of theIndo-German Collaboration Programme.

IA&P/10/45

Minimum Risk Based Selection of Waste Fuel Dumping Site

D. D, Sharma, D. Chatterjee and K. Sri RamNuclear Engineering and Technology ProgramIndian Institute of TechnologyKanpur 208 016

Radioactive wasteh andling from nuclear power plant tofinal dumping involves several stages and many personnel. Anyaccident or faulty handling at any of the stages poses dire conse-quence to the protection of workers and environment. Wastefuel dumping site is usually far away from the reactors andtransportation of fuel first to reprocessing plant and then todumping site exposes vast terrain and a large population to theradiation hazard in case of a possible accident.

For a CANDU reactor of a capacity of 200 MWe( 700 MWth), with On-line fuelling machine operating expecteddaily radioactive waste has an activity of 1. c MCi after 60 daysof cooling (See Table 1 for details). India has a 200 MWe unitoperating and another of same capacity awaiting commissioningat Kota, two units of 200 MWe each awaiting commissioning atMadras, two units of 200 MWe under construction at Narora andanother similar pair of units proposed for somewhere inGujarat. . There is only one proposed waste fuel dumping sitesomewhere in Andhrapradesh. In view of the above informationand anticipating that in future we may have to also tackle wastefuel from TAPS (installed capacity 400 MWe).. it's obvious thatlarge amount of activity will have to be handled in future. Theproblems that we face now are:

(i) Is it safe enough to have only one waste fuel dumping site,?

(ii) Does the waste fuel dumping site ensure minimum risk overthe entire waste fuel handling cycle.

Both the questions cover a wide spectrum of safetyrelated considerations in radioactive waste fuel management.

Present paper deals with the problem of finding radio-active waste fuel dumping sites and a schedule of transporting

waste fuel from each of the five nuclear reactors so that thetotal risk in waste fuel handling is minimized.

The basic information required is geographical, geolo-gical and demographic details of proposed regions, accidentdata at all stages of handling and consequent risk. To computerisk it is necessary to know probability of radioactive materialleakages following an accident. Since adequate data is notavailable to provide this information, Fault Tree Analysis is ado-pted to obtain the desired failure probability.

TABLE 1

WASTE ACTIVITY FROM A 700 MWth CANDU REACTOR-PER DAY

N. B. For computational convenience fission products are groupedinto 7 groups according tctheir half lives.

M«an halflife

12d

35d

60d

300d

986d

10220d

33945dand more

Isotopes

Ba140. Nd147

Ru103. Ce*4I.Nb95,

Zr9 S .Sr8 9 , Y91

Ru106. Ce^44

. Sb125. Pm147,Eul5-

&9° *Sm157

Total activity discharge per day

Activity atdischarge

4100 KCi

Te 1 2 9 2560 KCi

-1480 KCi

148 KCi

5 14. 7 KCi

2. 25 KCi

3.65 KCi

8308.60 KCi

Activity after £davs of coolin

27. 37 KCi

461 KCi

546 KCi

121 KCi

13.84 KCi

2. 25 KCi

3. 65 KCi

1175.11 KCi

64

IARP/10/41

Radiation Stability of Amberlyst - 15, A Macroporoue,Strong Acid Cation Exchange Re tin

P.C. Mayan Kutty, N.S. Filial, S.S. Shinde andM.N. NadkarniFuel Reprocessing DivisionBhabha Atomic. Research CentreBombay - 400 OSS

Introduction

MacroporouB ion exchange resins are reported to havesuperior mechanical and chemical stabilities to the conventio-nal gel type resins U» 2). Since they have a porousstructure and consequently possess large surface area theyare able to sorb non-polar, non-aqueous solvents unlike theconventional exchangers. Because of these attractiveproperties macroporous resins could be utilized for ionexchange separations both in aqueous as well as organicmedia I3"5) .

Several radicchemical separations have been'carriedout on macroporous resins. But studies ca the radiationstability of this class of exchangers ara rarely found inliterature. It may be interesting to check whether theirsuperior chemical and physical, properties are matched bybetter radiation stability as'well. This report describesthe results of the investigations carried out on the radiationstability of Amberlyst-15, a macroporous, strong add cationexchange resin. This resin was particularly selected be-cause extensive investigations were carried out on thisexchanger in our laboratories for radiochemical separationsinvolving plutonium, uranium,, thorium and fission products(4~?)Samples of Amberlyst-15 were subjected to external radiationfrom a 6 0 C o gamma source as well as internal radiatica by

<oQ - particles from absorbed plutonium. Variations incapacity, gal l ing behaviour and absorption kinetics werestudied as a function of absorbed dose. The results werecompared with those obtained with Dowex - 50 x 8, aconventional gel type cation exchange resin irradiated underidentical conditions.

65

Resalta and Discussion

Changes in capacity end swollen volumes of the resinsat various gamma doses are summarised in the followingtable:

Dose Capacity (%)

(MR) Amberlyst-15 Dowex

100

200

300

400

94.5

89.0

83.5

78.0

92.

8 5 .

7 7 .

70 .

50x8

5

0

5

0

Increase in swoli%

Amberlyst-15

2 . 5

7 . 5

12.8

18.4

n volume(%)

Dowex 5 Ox

13.6

30.9

52. 7

82. 7

Lossess in capacity due to cC-radiation also were foundto be less in the case of Amberlyst-15 than those exhibited byDowex - 50x8. The superior stability of Amberlyst-15 couldbe seen from the swelling data. While Dowex - 50x8 showsenormous increases in swollen volumes after irradiation,Amberlyst-15 exhibits insignificant changes. The resultsindicate that the latter can be used to much advantage in indu-strial scale columns especially in nuclear industries likefuel reprocessing plants.

After <£ - irradiation both the exchangers exhibitedslight increases in the total exchange capabilities, though,here again, Amberlyst-15 registered less variations. Theincrease in total capacity may be ascribed to the partialoxidation of the resins resulting into the formation of weeklyacidic groups.

Kinetics studies showed that irradiated Dowex 50x8samples attain exchange equilibrium much faster than ths un-irradiated sample, which is obviously due to the extensivede-cross-linking as a result of radiation. Irradiated andfresh samples of Amberlyst-15, on the other hand, did notshow noticeable differences in the kinetics of absorption.

References

1. R. Kunin, E. Meitzer and N. Barttick; .J. Am.Chem.Soc, 84, 305 (1962)

2. R. Kunin, et al, I and E. C. Prod. Res. Develop.,1, No. 2, 140 (1962)

3. W.W. Schulz., ARH-SA-58 (1970)

4. M.N. Nadkarni, at. al. 3 BARC 899 (1977)

5. M, N. Nadkarni et al. , Radiochem. Radioaaal. Letters,31 (b), 347 (1977).

6. P.C. Mayan Kutty, et al . , BARC. 928(1977).

7. P.C. Mayan Kutty, et al . , Radiochem. Radioanaly.Letters, (in print).

IARP/10/18

Further Evaluation of Cyclone for the Measurement of AerosolParameters

V. B. Menon and P. KotrajpaHealth Physic* DivisionBhabha Atomic Research CentreBombay - 4C0 085

1. Introduction

£otrappa et al (1973) have described a new approach ofestimating AMAD using the data obtained by a two stage cycloneassembly. Menon et al (1975) have evaluated the performanceof cyclone in comparison with a centripeter for measuring AMADin actual field conditions. They found the measurements in goodagreement. This proiroted further evaluation of cyclone assemblyin comparison with an Anderson cascade iznpactor which is anadvanced size selective sampler, very commonly used in aero-sol research. The results are presented here.

2. Performance evaluation of cyclone and Anderson Immctor

The Anderson Sampler (A. A. Anderson 1958) used forthe measurements was the one with a flow rate of 28. 3 1/min.It had eight stages each of which was made up of aluminiumround jet plate perforated with 400 holes, a spacer gasket and acollection disc. This instrument permitted direct measurementof activity as a function of aerodynamic diameter. The effectivecut off aerodynamic diameter (ECADg0) for each stage wasgiven by the manufacturer. The cyclone sampler consisted oftwo stages, a miniature version of the industrial cyclone, and afilter paper air sampling head. The following relation was usedto calculate the AMAD from the cyclone measurements:

AMAD = 13. 36 exp (0. 253 P)

where AMAD is in micrometers (l to 10 urn) and P is the %penetration.

A number of simultaneous field measurements weredone in an active area of a fuel reprocessing plant during

AMAD

7.

6.

8.

5.

6.

9.

6.

8.

7.

5.

6.

7.

8.

(H2

8

2

8

6

2

8

4

4

9

0

6

2

maintenance work involving chipping aud mopping of the conta-minated surface.

3. Results

The AMAD values obtained both by cyclone and Andersonimpactor are given in the Table.

Cyclone Andersen. ImpactorMeasurement AMAD (/um) AMAD

1

2

3

4

5

6.

7

8

9

10

11

12

13

Mean Value : 7. 24 + 1. 05

4. Discussion

*8.

6.

7.

7.

7.

9.

7.

8.

7.

6.

6.

7.

0

a

i

5

8

5

8

2

8

4

4

10.0

7. 8 + 1.10

As can be seen from the results, there is a goodagreement between the values of AMAD measured by cyclone andAndersen Samplers. Even though a number of samplers areavailable for AMAD measurements, cyclone is of considerable-importance for field use for the following reasons:

1. For a given flow rate, cyclone is considerablysmaller than horizontal elutriators.

2. Cyclone does not: have a fixed orientation.

Horizontal elutriators may be belter suited for samplingfibrous materials but for routine use in radioactive installationscyclone has definitely an edge over the other samplers.

References.

1. P. KOTRAFFA, S.K. DUA, D.P* BHANTI andP. P. JOSHI "HASL. Cyclone as an instrument formeasuring Aerosol parameters for New Lung Model"P roceedings of /RFA Congress, - WashingtonD.C. (1973)

2. V. B. MENON, A. R. SONDARARAJAN, P. P. J£> SHIaad P. KOTRAPPA "Evaluation of Cyclone Assemblyfor field use in the measurement of aerosolparameter, AMAD". Health Phys. 28, 618 (1975).

3. A. A, Anderses J. BacterJLol 76, 471 (1958).

IARP/10/16L

Development of an Indigenous High Volume Air Sampler

P.P. Joshi, P. Kotrappa, N.P.N. Setty, B. Raghunath andP. S. S iva subramanian.Health Physics DivisionBhabha Atomic Research CentreBombay - 400 035

Introduction

One of the standard methods of measuring theconcentrations of airborne particulates, whether radioactive.or non-radioactive, is to pass a known volume of air througha high efficiency filter paper, and analyse the participatescollected on the filter paper. For the measurement of verylow concentrations, the volume of air sampled must be large.It becomes desirable then to have an air mover of high airflow rate. The currently used sampler for this purpose isthe 'High volume air sampler1 manufactured by the StaplexCompany, U.S.A. Large number of such samplers arerequired for health physics work in nuclear installations likereactors, reprocessing plants and isotope production units.Development of a suitable high volume air sampler from indi-genous sources therefore assumes considerable importance.The objectives in developing such a sampler are thefollowing:

i) Cost should be low

ii) Large scale production should be feasible

iii) Its performance should be comparable with that oftEe staplex sampler.

iv) It should be rugged and capable of heavy duty use.

Development

Since the Staplex air sampler uses a turbine and motorassembly, similar to those found in conventional householdvacuum cleaners, it was decided to try out such a system

71

from a vacuum cleaner manufactured in India, by Delstar(Pvt.) Ltd.. and marketted by Spencer (India) Ltd. A lightweight metal housing having base supports and top handle,was fabricated to accommodate the turbine and motorassembly taken out from a vacuum cleaner. A filter paperholding assembly similar in desitn to that of the Staplexsampler was attached to the intake part of the turbine-motorhousing. The exhaust side plate of the housing was perfor-ated as much as possible to minimise the back pressure.The filter paper holding assembly at the intake side wasprovided with a 2 mm diameter hole to allow insertion of-ahypodermic needle probe for measuring the pressure dropacross the filter paper. It was decided to use this pressuredrop as a measure of the air flow rate, once a calibrationof the pressure drop with flow rate is performed using astandard flow measuring system such as the venturi system.

Performance Evaluation:

The performance of a prototype indigeneous highvolume air sampler was evaluated by measuring the airfiow rates when different types of filter papers were loadedin the filter paper holder assembly. Flow rates for thesame types of filter papers were also measured using the

: Staplex high volume air sampler. A standardised Venturitype flow rate measuring system developed by the IndustrialHygiene and Safety Section of Health Physics Division,B.A. R. C. was used. Filter papers used were HY-70, W-41,

, charcoal impregnated GF/A and GF/A type filter papers.At each flow rate measured, the pressure drop across thefilter paper was also measured using a water manometer.Asmooth curve was -drawn through the points when the flowrates were plotted against the corresponding p- essure drop,on a semi-log graph paper.

Both the samplers were put in continuous operationusing the same type of filter paper. The temperature rise ofthe air sampler body was measured for each, after 1 hourof continuous operation.

Results and Discussion

The BARC high volume air sampler was found to givefiow rates -which were 80% to 90% of the fiow rates shown

by Staplex sampler under identical sampling conditions.The temperature rise after 1 hour continuous operation,using HV-70 filter paper which has the highest resistance,was 30°C for the BARC sampler and 40°C for the Staplexsampler. -

The flow rates with various filter papers thoughmarginally less than those obtained with Staplex samplerare adequate for air sampling purposes.

The cost of the unit works out to be 59. 2, 000/-; ofthis vacuum cleaner needs &. 1, 400/- and fabrication of-filter holder and housing for turbine-motor, needs W. 600/-These samplers can be produced in large numbers. Sincesuch samplers are needed in large numbers not only in ourown Atomic Energy Programme, but by the conventionalindustries also for environmental pollution analysis, consi-derable foreign exchange amount will be saved.

IARF/10/21

A Proposal for Reduction of Maximum Permissible Dose toMeet: the Requirements of High Standards of Safety inRadiation Industry

P. S. Iyer and U. MadhavnathDivision of Radiological ProtectionBhabha Atomic Research CentreBombay - 400 085

Introduction

In setting up dose equivalent limits for radiationworkers, the International Commission on RadiologicalProtection (ICRP) has stated that a valid method for judgingthe acceptability of the level of risk in radiation work is tocompare this level with that in other industries with highsafety standards. Such industries are generally consideredto be those in which the average annual mortality fromoccupational hazza.rds does not exceed 10-"* (ICRP 77 a).In this context, an analysis of various radiation facilities bas-ed on personal doses and risk factors will be important inassessing safety standards.

Proposal for Reduction of Maximum Permissible Dose

The ICRP concludes that the mortality risk factorfor radiation-induced cancers is about 10 - _Sv , whichcorresponds to a risk factor of 10"* rem*' (ICRP 77 a).On this basis, the nuclear industry can be considered ashaving high standards .of safety only when the average annualdose per worker does not exceed 1 lem. The ICRP hasproposed, on' the basis of detailed calculations that the max-imum permissible dose (MPD) of 5 rem/yr (if receivedcoit inuously) corresponds to an annual mortality rate of3.4 x 10 (ICRP 77 b). Based on this, the annual dosecorresponding to an annual mortality factor of 10 will be1.5 rein. Signific&ntly, in either case, these doBes aremuch lower than the MPD. The average annual dose (AAD)for various types of radiation workers in different countrieshas been reported (UNSCEAR 77) and is found to be highin many nuclear installations. This implies that although

the MPD is 5 rem/yr, efforts must be made in all nuclearfacilities to ensure that the AAD is kept below 1.5 rem.In moat countries, the AAO for all types of radiation workersmedical, industrial research and atomic energy-takf « togetherwill be leas than 1.5 rem, even though the AAD for sometypes of workers i s h. Her. However, this in itself shouldnot justify higher doses because it would mean that thesignificance of the risk factor would be lost in the applicationof the law of averages. On the basis of the requirementsfor high safety standards, the authors strongly recommendedto the ICRP and other national bodies responsible forradiation protection, that the MPD level be immediatelybrought down to 1.5 rem/yr.

Restriction of Lifetime Dose.

It is not unlikely that many workers at nuclear fuelreprocessing and nuclear* power stations have been receivingannual doses close to the present MPD of 5 rem. Thecumulative dose of such personnel should be considered, and,their future annual doses must be strictly controlled.Assuming an MPD of 1.5 rem and an average working careerof 30 yr. it will be advisable to restrict the cumulativeoccupational do3es of radiation workers to not more than60 rem. This corresponds to a maximum risk of- 3 x 10 /yrfor radiation induced cancer resulting in death for the workerassuming that the effect of a given increment of dose doesnot persist beyond 25 yr. after irradiation (BEIR 72). Sixtyrem is higher than the cumulative dose based on thesuggested annual MPD of 1.5 rem; however, such flexibilitywill have to be allowed during a transitional period, as wehave had the annual MPD of 5 rem for the past 22 yr.It is hoped that shifting the emphasis to the cumulativelifetime MPD limit would enable countries just begining innuclear power programme to deal with exposure problemsearly and later effectively control them when they have,gained experience and have had the opportunity to improvesafety features.

75"

Referenced

BEIR 72

ICRP 77 a

ICRP 77 b

UNSGEAR 77

Report of the Advisory Committee on BiologicalEffects of Ionising Radiations, 1972, NationalResearch. Council, Washington, D. C.

ICRP, 1977, Recommendations of the Inter-national Conamision. on Radiological Protection,ICRP Publication No. Zd (Oxford : PergamonPress)

ICRP, 1977, Problems Involved in Developingan Index of Harm, ICRP Publication No. 27(Oxford :.Pergamon Press)

United Nations Scientific Committee on theEffects of Atomic Radiation, Sources andEffects of Ionising Radiation, Annex E, Dosefrom Occupational Exposure 1977(U.N., New York\

IARP/10/4

Operational Limit for Uranium Or* Dust Applicable toUranium Complex at Jadvguda

Giridhar Jha and M. RaghavayyaHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Mining and milling operations to. produce uraniumconcentrate are dusty by n&ture* 'The ,extent of the parti -culate airborne contamination (excluding radon daughters)in these operations is such that it san usually cause onlychronic exposure. The operational limit as. employed by theICRP (1976) i s based on the total alpha activity of the long-lived nuclides viz. uranium J2 38, 'tho*ium*2 30, radium-226and polonium-210, present in the dust. This limit has beenarrived at under the assumption that the .constituents of thegross long lived alpha activity are in aejcular equilibrium.

From studies conducted- in the " forking areas of ura-nium mines and mill at Jaduguda, it i& found that the con-stituent» cf longlived alpha activity in the airborne dust re-main in equilibrium only during the initial .«perations. Inlater stages the equilibrium -w s found to be disturbed todifferent degrees. Moreover, 'such factors as .pulmqnarydeposition fraction dependent oa 'the activity median aero-dynamic diameter (AMAD) of tfe* dust available for inhalatio*and clearance pattern of the retained activity from the "pulmon&ry region influence the jdose commitment.

The AMAD of aerosols generated during differentstages of mine and mill operations was found to vary fas m2 to 6 um. Variations were observed' not only duringdifferent operations but also during the similar operations.

The toxicity of the inhaled aerosol is believed todepend mainly on its dissolution! characteristics in the.lungfluid. While calculating the (MPC) for different radio -nuclides, ICRP (1959) has consi&erid a clearance half-lifeof 120 days for all insoluble duet. In-vitro study on thedissolution of uranium in respirable uranium ore dust from

i • •

*77

Jaduguda has indicated a binaodal dissolution pattern, withclearance half-times of 0.07 day (accounting for 33 percentof activity) and 371 days.

Hence, for uranium mines and mill operations atJaduguda, the operational limit for gross longlived alphaemitters has been rederived taXwg into account theexperimental findings reported in this paper and thoserecommended by the ICRP from time to time.

IARP/10/23

Analysis of Occupational Personnel Exposures in NuclearFuel Cycle Operations in India

K.S. Shenoy, P. H. Patel and S. J. SupeDivision of Radiological ProtectionBhabha Atomic Research CentreBombay - 400 085

The growing trend in occupational personnelexposures in the field of Nuclear Energy programme is ofgreat concern to the authorities responsible for radiationprotection in the country. In India, the nuclear energyresearch and its applications are largely confined to theDepartment of Atomic Energy, Government of India. Theauthorities have been very cautious regarding health hazardsassociated with the radiation exposures and radioactivecontamination. The responsibility of measurement andcontrol of radiation exposures of the radiation workers in

""Inaia-Tesfs with the Division of Radiological Protection. • TheDivision conducts a centralised personnel monitoringprogramme and maintains lifetime dose records of allradiation workers.

From the personnel exposure records availablewith the Division of Radiological Protection, attempt is madeto analyse the occupational exposures of radiation workersconnected with nuclear fuel cycle operations from the year1971 to 1977. The data presented in Table-1 shows the \phenomenal rise in collective dose per year in variousoperations connected with nuclear fuel cycle, and in basicnuclear research within Department of Atomic Energy.

It would be better to express the values intocollective dose per unit practice1, which will react-tochanges in practice efficiency as well as the protection

. efficiency. However it is too early to get data in this formin bur country.

The average exposure per person per year is atrue indicator of radiation safety level, and the same isgiven in. Table 2 for various operations. The valuesreported by various countries fox* radiation workers indifferent operations are in the range of 0.5 - 2 rad perperson per year*. Except for the persons working inpower reactor catag ory our values are within, the rangereported elsewhere.

Table 1Collective dose (rem)

Fuel Waste Power Research NuclearYear Mining Fabrica- Manage- React- Reactors Research

tion & ment orsReproce-asing

1971

1972

1973

1974

1975

1976

1977

222

198

217

169

309

416

450

604

811

645

542

815

675

415

46

88

74

116

65

368

3.95

411

2282

2605

3595

4922

4936

5125

317

43<J

301

190

276

266

363

94

114

96

103

90

94

111

Year

Table 2Average clode (nrirein) per person per y«-a:

Fuel 1-aljii- Waate Power Research T«aclearMining cation & Manage- Reactor* Reactor* Research

1971

1972

1973

1974

1975

1976

1977

179

141

155

113

192

257

265

881

1143

991

892

1243

1033

429

197

394

295

429

332

884

1010

808

2503

3261

3380

3562

3620

3324

402

532

371

230

385

320

445

97

lid

102

127

110

112

138

References:1. Sources and Effects of Ionizing Radiation UNSCEAR 1977 Report, ANNEX E

Para 43-81.

I

IARP/10/22

Occupational Life Time Dose Expectancy in the Nuclear FuelCycle Workers in India

R, Sadagopan, Geetha Sadagopan and S. J. SupeDivision of Radiological ProtectionBhabha Atomic. Re s.ea re h C entreBombav - 400 085

The nuclear energy programme in India is now twodecades old. The radiation workers in various constituent unitsunder this programme have been regularly monitored withPersonnel Monitoring Services. Data for cumulative exposuresfor them is available since 1961. From this data the life time

.dose expectancy has been evaluated. Further for power reactors,as well as research reactors radiation -workers have been cate-gorised according to their nature of work. The various categoriesare maintenance workers, operators, Nuclear physics researchworkers, Health Physicist. The life time dose expectancy forthese various categories have been evaluated for both the researchand power reactors.

Various methods for assessing life time doses areavailable which, are'based on retrospective analysis of individualor group radiation exposures. For evaluation in this work we haveused the expression given by UNSCEAR REPORT on the dosesfrom occupational exposure Annex E. Accordingly, the basiccumulative life time doses are given by

where Dj is the average annual dose for each of the n yearsfor which dose records are available, D^Q is the predicted dosefor 40 years employment and 'n' i s the number of years for w hi chthe dose records are available and should have preferably valuegreater than five years.

Table \ gives the average life time dose expectancy for theworkers in various units of the nuclear energy program. Forcomparison the values for the workers from Ontario Hydro havealso been quoted from UNSCEAR REPORT. The values indicatethat the life expectancy dose for various units of nuclear energyprogram are of ihe same order as those for Oatari© ~*ydre;

Table 2 presents category wise life expectancy dose ofradiation workers, for one of the research reactors and onepower reactor. From this data it may be noted that the operatorsin the research reactor are expected to receive the maximum doseand the research workers the minimum. In case of power reactoralmost all category of radiation workers are expected to receivemore or less the same life time dose. Because of thi the averagelife time expectancy dose for all the workers of the power reactoris expected to be higher than those from research reactor.

The average yearly dose for radiation workers in some unitsand its variation with the experience of radiation workers havefurther been analysed.

Table 1

Group of Units Life time dose expectancyin xem

Effluent management 27. 66

Fuel processing 7. 45

Fuel Reprocessing 27. ft!

Mining 24.07

Power Reactors

a. RAPS 30.33

b. TAPS 40.35

Research Reactor 16. 97

Research & Development 6.48

Ontario Hydro 39 to 56

Table 2

Life time dose expectancy in remGroup Research Reactor Power Reactor

C C L - 27.44 15

F H U - 51.09

E M U - 9.34

CM - 34.90

EM - 18.86

Operators 37.34

a. Engineers 17.01

b. Workers 71.21

MaintenanceWorkers 31.28MaintenanceEngineers 16. 53

MechanicalMaintenance - 46. 09

SeriveMaintenance - 31.45

Health Physics 24. 53 38. 63

N P D 16.69

R C S 17.19

ReactorPhysics 7.21

84-

IARP/10/2?

Safety in Fuel Reprocessing - A Regulatory View

S.E. BrittoAtomic Energy Regulatory CellBhabha Atomic Research CentreBombay -. 400 085

In many respects, Fuel Reprocessing Plants are seento be the most hazardous of all nuclear installations bes.idestheir 'proliferation' potential. Consequently, very carefulsafety analysis and hazardB evaluation are performed on thedesign and on the plant before authorising start of regularoperation. However, when looking from a.critical regulatory-angle, these evaluations appear to be rather overly biased to-wards requirements of radiological and criticality safety ofthe being designed, as-designed and as-built plant, undertakenwith the health physicists in the role of the Devils Advocateand the process/plant designers defending against possibleintroduction of a multiplicity of restrictive safety constraints.In this context, this paper makes a point of departure anddetails out some basic requirements in design that shouldalso be evaluated in depth as the primary and importantlevel of ensuring safety. These requirements are groupedunder the following given four categories.

Critical Review of Design and Operation;

With various pressures on the process, fabrication,project and operating engineers, it is not possible to avoidoverlooks, mistakes and design deficiencies in a chemicalprocess plant system as complex and sophisticated as thereprocessing of spent fuel. It should be mandatory thereforeto provide a separate and independent regulatory review andquality assurance check of all design and operation to ensuresafety of plant and public. The paper briefly sketches theactivities in this area.

Automatic Control:

In spite of the provision of elaborate and extensiveprocess and safety related instrumentsc at present precious

little of automatic control is incorporated, placing reliancemostly on operators' judgement in normal, abnormal andaccident conditions. The snags inner rent in this methodof operation are highlighted ';« outline various measures thatshould be taken to improve the situation.

Corrosion:

Stainless steels of the suitable and proven types areused as plant construction materials with all-weldedconstruction and extensive radiog raphical checks. And yet,serious corrosion failures occur triggering criticality-wiseand radiologically unsafe situations causing exposures tomaintenance personnel. Measures that are to be taken tominimise corrosion susceptibility of construction materialsare outlined.

Chemical Safety:

It was the third Geneva Conference papers onreprocessing that brought to light the fact that the seriouspreoccupations of the designers and operators with radiologicaland criticality safety has tended to place a relatively lessconcern to chemical hazard potentials resulting in cases ofradiological accidents intiated by chemically unsafe situations.The hazardous conditions in this area are considered tr. re-commend an approach of building-in chemical safetythrough choice of safe schemes even sacrificing a little ofprocess efficiency.

IARP/I0/5

Some Aspects of the Accumulation of Natural Radionuclides inAquatic Products and Vegetation Around a Uranium Complex

P. M. Markose and K. P. EappenE. S. SectionHealth Physics DivisionEnvironmental Survey LaboratoryJaduguda

Release of radionuclides to the environment, though in avery limitted level, has been a constant feature in the millingand nrunig operations of uranium. Accumulation of theseradlotoxins from soil and water by flora and fauna is an inevitableconsequence. An investigation into the natural uptake of radio-cue Udes by aquatic products and vegetation has been carried outin the environment of Jaduguda uranium complsx. Samples offish and vegetation from a contaminated stream indicate accumu-lation of the radionuclides. Ratio of the encentration or radiumand uranium in the edible and non-edible portions of the sampleshas been derived.

Our analysis indicated that the ratio of radium anduranium in bone to flesh of fish samples was generally 5:1 whenthe fish was collected from contaminated stream of mine effluents.Accumulation of radionuclides by grass samples has beenobserved. Evaluation of the concentration of radiotoxins in grassis important because of the fact that they get metabolically incor-porated in plants and may ultimately reach man through animals.

In this paper, an attempt has been made to bring out someaspects of the accumulation of radionuclides by aquatic and land-products in the uranium mining and milling environment. Somecontrolled experiments are also described to demonstrate theaccumulation of radium and uranium by aquatic products.

Radionuclides in fish samples

226 Ra pCi/kg Ratio U(nat) ug/kg RatioSample

Fish fromGara nala

Fish fromcontaminatedstream

i i

Flesh

0.

162.

142.

4

4

6

Bone

978.1

779.9

bone/flesh

6 .0

5.52

Flesh

ND

90.4

61.5

Bone

—

692.1

388.3

bone /flesh

7 . 7

6.3

Radium content in soil and grass

Sample no. Ra in 3oilpCi/g

Ra in grassPCi/g

Ratiograss/ soii x 10"

1.

2 .

3 .

2.64

7.53

116.10

56.89

134.50

2736. 1

21.5

17.9

23.6

IARP/10/12

Radium-2R8 in the Monazite Processing

A.C. Paul and K. C. PillaiHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Monazite is chemically treated, by alkali digestionand acid leaching, to separate thorium and rare-earths atthe Indian Rare Earths plant at Udyogamandal situated on thebanks of Periyar river. The process effluents, generated atvarious stages of the operations, are discharged into theriver after preliminary treatment.

The major radionuclide present in the effluent isfound to be Ra-228, the daughter product of Th-232 inmonazite. Ra-226 which is the daughter product of U-238is nearly 1/10 of the Ra-228 activity. The activity getsdistributed into the soluble and insoluble fractions in theeffluent depending on its acidic/alkaline nature. As per theprevailing conditions in the plant suspended solids accountfor 90% of the activity in the alkaline streams and 50% inthe acidic stream. :

Modifications carried out to improve the settlingefficiency resulted in 4 to 5 fold reduction of the suspendedactivity in alkaline streams. However, the acidic streamswhich carried nearly 80> of the total effluent activity werenot amenable to such treatments. Subsequently, thedeactivation with BaSC>4 which was introduced in the weakchloride stream, brought down effluent activity of the streamas shown in table 1. The solid wastes obtained from theprocess are suitably contained.

Table 1

REDUCTION m EFFLUENT ACTIVITY AFTER DEACTIVATIONOF WEAK CHLORIDE (ACIDIC STREAM).

EfflueatRa-228, pCi. ml"1

before de- I after de-activation [ activation

Supernatant, rare-earthfluoride pption. 53.5 11.2

Supernatant, fare-earthcarbonate pption 42.0 8.0

In spite of the above treatments the individualeffluents continued to show activity levels in excess of thepermissible limits. .The mixed effluent prepared by simulatingthe plant conditions showed activity levels an order ofmagnitude higher than the stipulated limits. Therefore studieswere carried out to evolve a suitable treatment procedure.

The scheme for further deactivation of the effluentwas based on lime treatment which gave encouraging resultsin laboratory and pilot plant trials. Table 2 shows efficiencyof this method. An effluent treatment plant, making use ofthe procedure, is being put*-.up at site.

Table 2

REDUCTION IN EFFLUENT ACTIVITY AFTER LIMETREATMENT

Characteristic s

Activity in rrjced effluent,1

before linr.e i ifter limetreatment i reatment

gross alpha z o ;. 08gross beta 11 .04Ra-22S . Z J. 04

Environmental impact

Background concentration of Ra-228 in Periyarriver water varied from 3 to 10% of (MPC)W. Theconcentrations at the industrial area varied from 10% to50% of (MPC)W between the high and lean flow periods inmonsoon and summer respectively. The activity was alsofound to get accumulated in sediments at ths out fall area.Sediment shift, towards backwaters, was observed in theriver during monsoon due to the turbulent flow in the season.

Fish caught from the industrial area indicateduptake of Ra, especially in the bone. Uptake by paddy,grown in the fields irrigated by the river water downstream,was significant in hay and the transfer to the grain was low.Exposure of population through the intake of contaminatedfood is considered low as the fish catch and paddy harvestingin the area form only a small fraction of the total producein the locality.

. . The effluent treatment, envisaged, will furtherlower the Ra-228 levels in the river environment.

IARP/10/34

Studies on Natural Radioactivity in the KalpakkamEnvironment

M.A. R. Iyengar, M. P. Raj an and S. GanapathyEnvironmental Survey LaboratoryHealth Physics DivisionBhabha Atomic Research CentreKalpakkam - 603 102

Kalpakkam situated 75 knw south of Madras is thesite of Madras Atomic Power Project and the ReactorResearch Centre. Pre-operational radiological surveys ofthe site revealed significant radiation levels (20-400 uR/hr)particularly at the coastal beaches and nearby areas, whichwas traced to the presence of monazite (0.4-0.6%) indiscrete localised pockets (1). In view of the monaziteoccurrence detailed investigations were conducted to studythe.pattern and level of natural radioactivity distribution indifferent environmental materials of relevance to localpopulation exposure.

Demographic surveys identified the existence oftwo main professional groups: (i) fishermen residing onbeach settlements and (ii) farmers living in the land interior.The paper presents the natural radioactivity distribution inthe local environment, in particular in beach sands, landcrops, marine food sources and local diet. The radio-nuclides investigated are 226R.a, 2 2 8Ra, 2 1 0 Pb, 2 1 0 P o .As could be expected the beach sands display varyingconcentrations of activity fluctuating with its monazitecontent. The marine food sources have been studied inconsiderable detail and the paper reports activitydistribution with reference to different species and furtherin different tissues of the same species. The data has beenutilised for arriving at some of the concentration factorsin the marine environment. From the data it is evidentthat the marine organisms accumulate the natural radio-elements to a significant degree with * Po at the maximumlevel. The preferential uptake of 2 1 0Po is in contrast withthe lower uptake of 2 2 6Ra, 2 2 8Ra and 2 1 0 Pb.

The survey has also brought to notice thatcasuarina foliage contain appreciable amounts of 22o£aj &and Po depending upon its growth site i. e. whether inthe region of a hot spot or away from it. The casuarinafoliage is used as a source of fuel by a segment of the localpopulation and this practice could lead to significant exposures.The paper discusses these various aspects in some detail andattempts to identify the population group which is- subjectedto a higher level of exposure from natural radiation sources.

References

Iyer M. R. et. al: 'Radiation survey of the monazite areas ofKalpakkam1. BARC-I/315 (1974)

! I

IARP/10/35

Some Aspects of Accumulation of Radionuclides in CrystalSediments

N.N. Dey, Vananti Matkar, Elisabeth Mathew and K- C- PillaiEnvironmental Studies SectionHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Suspended particulate materials have profoundinfluence on the behaviour and fate of pollutants dischargedinto aquatic environments. Extensive investigations havebeen carried out for more than a decade at Trombay tostudy the influence of suspended silt and sediments on thedistribution of some of the radionuclides released into theBombay harbour bay. Depending on the location, turbulenceresulting from both tides and winds, and seasonal land run-offinto the bay, the suspended load in sea water varies widelyranging from 1 to 2000 mg/l. Besides the physico-chemicalnature of radionuclides released and the changes that mightundergo when these nuclides are introduced into the seawater medium, the clay mineral and organic matterconstituents and particle size of the sediments greatlyinfluence the accumulation of radionuclides by sediments.Laboratory and field studies have shown wide range of Kjjvalues for sediments for radionuclides from sea water.Average values obtained are for 9 0Sr -16, 1 3 7Cs - 8. 7 x lO2,106R u _4>8 x 1Q3f 144C e -8.7 x 104 and 239,240pu - 9 x 10^.These values have been utilized to estimate the extent towhi *h these nuclides can be removed from the bay watersand are given below:

Localization of some of the nuclides like *44Ce and" ' P u Were observed near discharge location and dispersalof these nuclides were mostly through transport of silt fromdischarge area. Different mechanisms for uptake of variousradionuclides by sediments have been examined.

Sedimentation of radionuclides may apparentlyappear to be beneficial; however consequences of suchaccumulation in causing radiation exposure to man. andhigher uptake to benthic organisms need to be assessed.These factors are taken into consideration for evaluationof the environmental impact of radioactive effluent releasesand to assesB the safe recipient capacity of aquaticenvironment.,

Table 1

Removal of radionuclides from sea waterby suspended Bilt/sediment

% removal for different silt loads (me/I)

9 0Sr 0.02 0.16 0.79 1.58 3.10

1 3 7 Cs 0.86 8.00 30.32 46.52 63.51

l 0 6 Ru 4.61 32.73 58.85 82.91 90.67

1 4 4 Ce 46.43 89.66 97.75 98.86 99.42

2 3 9 P u 47.37 90.00 98.18 9&. 90 99.44

LARP/10/11

Comprehensive Environmental Surveillance and ImpactExperience at the Rajasthan Atomic Power Station ^

R. P. Gurg, T.A. Sebastian, B. Dube, K.G. Varugheseand A. R. Lakshmanan.Environns ntal Survey LaboratoriesRAPP, Kota, Rajasthan

RAFS is the first inland Atomic Power Station inIndia using fresh water of Rana Pratap Sagar as 'a secondarycoolant. Being the only water source in the region it alsosupplies drinking water to the population and receives dis-charged waste effluents and domestic waste effluents from theplant and .colonies. The lake receives water from our landflows and tailings from upstream dam and Chambal tributaries.

The surveillance at the station site was started from1973 and attention was kept on water quality of the lake forconsumption by local population and degradation of qualitydue to release from station and colonies.

Water samples were collected from 14 different loca-tions and were examined for .radioactivity content (H-3, 1-131,C B - 1 3 7 , Sr-90). Excepting in Discharge Canal (D. C) water1-131 was less than 1 pCi/1 against MPCW of 2000 pCi/1.Maximum tritium in D. C. water was 773 pCi/mi againstMPCW of 5000 pCi/ml of water. Sediment samples from thearea nearby to RAPS showed the accumulation of Cs fromlake water. The levels are 20-25 times higher tb>\n thegeneral background levels prevailing in the sediments.

Air activity in the environment contributed by 1-131was monitored by thyroid as indicator tissue. There was nodetectable 1-131 activity mast of the time in the thyroid exceptduring the period following Chinese nuclear weapon tests of1974, 1976 & 1977.

Results of radioactivity analysis for wheat, meat,milk, ground water, fish are reported in the paper.

Tritium is an important radionuclide produced inthe reactor and it finds easy escape in the environment.Tritium in the atmosphere, water, leaves and fish arereported in the paper.

The drinking water qu lity is important to bedetermined as there are no public health laboratoriesnearby. The surveillance was extended to the examinationof drinking water and discharge of domestic effluents to thepublic health specifications. The distribution of -heatedwaters discharged to lake has been kept under surveillancefor temp. & Do. Comprehensive surveillance has shownthere is no degradation of environmental quality.

"V

IARP/10/20

Radioactivity Release to the Environment by Thermal PowerStations Using Coal as a Fuel

U.C. Mishra, B.Y. La!it and T.V. RamachandranAir Monitoring SectionBhabha Atomic Research CentreBombay - 400 085

Man has always been exposed to ionising radiationsfrom various sources, one of such source being thermalpower stations using coal as a fuel. By the year 1978-79,about 55% of total electric power produced in India is expectedto be from thermal power stations using predominently(rw80%) coal as a fuel. Nuclear power will constituteBoth of these sources release radioactivity, though of differentnature, to the environnent. In the present paper, these twosources are compared in relation to radiation hazards arisingout of their operations.

Large coal-fired power stations using pulverized coaldischarge gaseous and particulate combustion products intothe atmosphere containing daughter products of uranium andthorium series. India produces, mostly poor quality coal with25-40% ash content. Many places fly-ash control methodsare also not adequate to abate the pollution problem. Thiscontribute more radioactivity to the atmosphere per megawattgenerated by coal-fired plants when' compared to the thermalpower stations in advanced countries.

Coal and-fly-ash samples from four coal-fired thermalstations in India are analysed for their natural radioactivitycontent using 12.5 cm x 10 cm Nal (Tl) crystal coupled to256-channel pulse height analyser. The radioactivity contentsof Indian coal and fly-ash from these power stations wereas follows:

Coal: Ra-226: 0.4 - 1.3 pCi/g; Th-228:0.7 - 1.2 pCi/g.

Fly-ash: Ra-226:1.4-4. 6 pCi/g; Th-228:2.1-4. 8 pCi/g.

The radioactivity released by thermal power stationswas calculated on assumptions of 500 kg. of coal consumptionper MW (e)-hour ash content of coal 25% and 90% fly-ashcontrol. The total 226Ra + 228Ra released by thermalpower plants under study range between 5.4 x 10"4 to8. 9 x 10~4 curies/year per MW(e) power generated. Basedon calculations by Eisen.jud and Petrow (2) this release isradiobiologically equivalent to 216-316 curies of Kr peryear or 0.216-0.356 Ci 131I/year both per MW(e) power.

On comparing this data with the emissions of thesetwo radioactivities from nuclear power stations of BoilingWater Reactor and CANDU type in use in Indian PowerProgramme, it is found that nuclear power stations ofsimilar size rd ease less radio-biologically significant radio-activity in the atmosphere when compared to coal fuelthermal power stations.

References:

1. Shah K. T. & B. K. Gandhi, Air Pollution by ThermalPower Stations and Methods for their Control Techniques

. . in Proceedings of Seminar on Air Pollution ControlI Techniques, Bombay 11-12 September 1973, Central

/ Labour Institute, Bombay 81.

! 2. Eisenbud, M. & H. G. Petrow, Radioactivity in the/ Atmospheric Effluents of Power Plants that use-Fossil

/ Fuels, Science, 144, 17-4-64, (1964), 288.

IARP/10/40

Technologically Enhanced Radiation Exposure from someNon-nuclear Sources

V.S. Londhe, K. C. Pillai and S. D.Health Physics DivisionBh&bha Atomic Research CentreBombay - 400 085

Soman

Environmental impact of nuclear energy operationshave been studied extensively from the beginning of this stewtechnological development. In recognition of the effects oflow level radiations and consequent to technological innovationson waste management practices in the nuclear industry it isnow possible to restrict radioactive effluent discharges fromnuclear installations such that it contributes only a negligiblefraction of the background radiation dose a person receives.In this context, technologically enhanced radiation exposurefrom non-nuclear sources assumes significance.

Rock-phosphates, a basic material used inFertilizer industry for production of Phosphoric acid andsuper phosphate fertilizer have been examined for its radio-activity content. Radioactivity content were obtained by beta,gamma measurements and radiochemical estimation of Th,U and 2 2 6Ra. 2 z 6 Ra and Gross beta activity obtained insome of the rock phosphate samples are given in Table 1.The waste materials produced from these sources, phospho-gypsum, were also examined for activity content.

Coal used in some thermal stations were alsoexamined for their ash and radioactivity content. Coal ashand phospho-gypsum are used as building materials alongwith cement. Table 2 gives the relative concentrations of2 2°Ra and gross beta activity in cement, coal ash andphospho-gypsum. Personnel radiation exposure that canarise from such sources are estimated and are discussed.

Table 1

Radioactivity ia Rock Phosphates(pCi/g)

No

2 2 6 Ra

Betaactivity

30,

127

4

. 8

29.

104.

1

0

n39.7 -24.4

il6.0 117.0

Table 2

Radioactivity of Cemeet and some waste materials

Sample

Cement

Coal

Phospho-Gypsum

ST. NO.

1.

2.

1.

2.

1.

1.

2.

4.

7.

13.

226Ra

05 t 0.

4 + 0.

22 t 0.

75 t 1.

6 + 1 . 5

35

97

2

Beta activityPCi/g

5.1 i 1.4

7.9 t 1-7

19.8 1 4.0

--

• \' i

to)

SARP/10/1

A Continuous Effluent Stream Sampler

S. Venkataraman, G. K. Srivastava, S. C. Saha andP.R. KamathHealth Physics UnitJaduguda MinesSinghbhum, Bihar

Environmental pollution due to any industry can betruly assessed only if a continuous watch is maintained overthe liquid and gaseous discharges. With this in view asimple, continuous effluent sampler has been developed whichworks on the age old principle of the water wheel, derivingthe motive force from the effluent street itself.

In construction, the sampler, is essentially a paddJewheel mounted on ball bearings for frictionless r'-Hon. Thewheel is mounted in such a way in the discharge ain, thatthe effluent flow keeps the wheel in motion continuously. Asmall vial mounted on the rim of the wheel collects an ali-quot of the effluent during each revolution and drops into asuitable container placed nearby.

In such a system, it can be shown that the samplecollected faring any period, say a day, is representativeof the total quantity of effluents flowing along the drain. AndConversely, from the volume of the sample collected, totalquantity of effluents discharged can be related by:

Q = A ( B . Y - + C)3 (1)

V.T

where Q is the total quantity of effluents discharged

W is the volume of sample collected

V is the volume of the vial

T is the duration of sampling

(02-

A, B h C are constants, which can be determinedexperimentally.

A continuous sampler of the above type, has been inuse in Jaduguda for some time. Before installing the device,grab samples of the outgoing effluents were being collected,which often either underestimated or overestimated thepollution potential. With the installation of the sampler,the sampling drawbacks have been greatly obviated.

The representative samples are generally analysed,in this for uranium and manganese, both in the dissolvedstate and as associated with the suspensions, The paperpresents a typical set of results obtained,' and also comparesthe data with grab sample analysis results.

IARP/10/44

Organometallic Interaction of Manganese with Humic Acid

P. N. Aggarwal and M. V. M. DesaiHealth Physics DivisionBhabha Atomic Research CentreBombay - 400 085

Various metal ions in microquantities aredischarged as effluent, into sea water, from nuclear reactorsand industries. On their interaction with organic matter,these metal ions can be converted to a. soluble compound oran insoluble species. Organic matter in natural environmentis generally present in sufficient quantity to form organo-metallic species with various incoming metal ions. Of theorganic matter present in the marine environment humicacid is a very important component. Humic acid reactswith various metals at competitive rates and mobilise orimmobilise thrm. Humic acid present in sea water willinteract with a metal ion and take it away with it to farocean, thus mobilising the metal ions. On the other handhumic acid present in the sediment will also react with themetal and form an organometallic complex, that is, it willrendsss the metal in a form, non-replaceable on ionexchangers. This humic acid being adhered to clayimmobilises the metal as well. However if this metal canreplace all the clay particles from humic acid, the resultwill be immobilization of metal. With time, of course,this humic acid on interaction with other organic matter mayrelease this metal.

The present studies were carried out to find outthe maximum amount of Manganese, the humic acid componentof- organic matter from sediment of Bombay Shelf can bind.Humic acid was extracted from sediment by a solution 0. 2 Nin sodium carbonate and sodium hydroxide. It was thenpurified by electrodialysis till there was no significantchange in conductance of side compartments of electro-dialysis cell. Humic acid was then allowed to interact witha fixed amount of Manganese (Z0 ugm Mn labelled with 5 4

under various controlled conditions of pH, maintained byNaOH, for one week. After the reaction, the soluble

fraction was passed through Dowex-50 cation exchanger,50-100 mesh size, conditioned to reaction conditions, tofind out the amount of Mn rendered non-replaceable byhumic acid.

The reaction was carried out in pH range of6.5 to 12. 5. It is seen that maximum uptake occurs atpH 11.8 to 11.9. Maintaining this pH values, the variationof conversion of Mn to non-replaceable form shows that themaximum uptake is about 11.0 mg/Mn per gm of humicacid (approx. 8 mg Mn/kg of sediment).

AUTHOR INDEX

(Figures refer to page numbers)

Aggarwal, A. N.

Balakrishnan, V. P.Bhanti, r . P.Bhat, I.S.Bhatt, R.C.Bhuvan ChandraBritto, S.E.

Chatter jee, D.

Desai, M.V. M.De shingkar, D. S.Deye N. N.Dhairyavarii M. P.Dingarkar, M.V.Dube, B.

Eappen, K. P.Eberle, H.

-Ganapathy, S.Gandhi, K. G.Gangadharan, P.Garg, S.P.Gurg, R.

Hara PrakashHaridasan, T.K.Hegde, A. G.

Iyengar, M. A. R.Iyer, M. R.Iyer, P.S.

Jacob, P .P .S .Jacob JohnJanardhanan, S.Jha GiridharJoshi, P. P.

103

552719404084

62

1032593461295

8660

912543,502795

552719

33,916073

5588

7612,70

Kamala RudranKamath, P.R.Kannan, S.Kapoor, J. C.Khadake, R. G.Khan, A.A.Khan, A. H.Kishore, A. C.Kothatkar. M.G.Kotrappa

Krishnaxnony, S.Kulkarni, L. V.Kumar, S.V.Kutty, K.N.

Lakshmanan, A.R.Lakshmanan, A. R.Lalit, B.Y.Laxminarayanan, T. S.Londhe, V.S.

Madhvanath, U.Marko se, P. M.Mathew ElizabethMatkar VasantiMayan Kutty, P.C.Mayya, Y.S.Mehdi AliMehta Sudarshan, K.Menon, V.B.Mishra, U.C.

Nadkarni, M. N.Nagarajan, P.S.

Ottmer, H.

Patel, P.H.Paul, A.C.

21101

50144314,253625153.27:

67,70521252

8

4095975599

7333, 86939364

310

86797

57,6443,48

60

786,88

Piliäfi, K.C.

Pillai, N.S.Pillai, P. M. B.Pimpale, N.S.Popli. K. L,.Pradhan, A. S.

Raghavayya,. M.

Raghvnath, B.Rajagopalan, S.Rajan, M. A.Ramachandran, T.V.Risbud, V. H.

Sac hart, S.R.Sadagopan GeeihaSadagopan, R.Saha, S. C.Sebastian, T. P.Sharada, K. S.Sharma, D. D.S har ma, R. C.Sharma, S. K.Shenoy, K.S.

88, 93.9964

8274040

33, 36767055919750

2 481

101953862241578

Shetty, N. P. N.Stünde, S.S.Siddique, H. R.Sidhwa, B. M.Sivasubramaniam, P.S.Soman, S. D.Somasundaram, S.Sri Ram, K.Srivastava, G. K.Subbaramu, M. C.Subbaratnam, T.Supe, S.J.Surendran, T.Surya Rao, B.Swaminathan, N.

Theyyunni, T. K.

Varughese, K. G.Venkataraman, G.Venkataraman, S.Vijayan, K.Viawakarma, R. R.Viswanathan, S.

Watamwar. S. B.

706452577024,99276236, 101

11778.812710

8

57

954 3

101554 310

8

'• \

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K. C. PillaiP.vS . ChauhanK.S.V. NambiA. RamaswamyA.R. Reddy

ADMISSIONS

R. C. SharmaD. R. BongirwarS. JayaramanP.S. Nag ara jan

ADMINISTRATIVE SERVICES

Kamala RudranElizabeth MathewVijaya R. PullatA. V. Santhamma

FINANCE

V.K. JainK. R. DasP.S. Nagarajan

BULLETIN

T.S. lyengarS. KrishnamonyL. V. KrislmanV.B. Menon.I-. S. MishraT.S. MuraleedharanM.S.S. MurthyG. MuthukrishnanA. NagaratnamM. RaghavayyaB. M.L. Sah