Guidelines for the Layout and Contents of Safety Reports ... Safety Standards/Safety_Series_034_1970.pdf(a) a Pre-Construction Safety Report which supports the appli cation for authorization

f h 1S A F E T Y S E R IE S

N o. 34

Guidelines for the Layout and Contents

of Safety Reports for Stationary Nuclear Power Plants

S P O N S O R E D B Y I A E A A N D W H O

I N T E R N A T I O N A L A T O M I C E N E R G Y A G E N C Y V I E N N A , 1970

This publication is no longer valid Please see http://www-ns.iaea.org/standards/


GUIDELINES FOR THE LAYOUT A N D CO NTENTS

OF SAFETY REPORTS FOR STATIONARY NUCLEAR POWER PLANTS



SAFETY SERIES No. 34

G U I D E L I N E S F O R T H E L A Y O U T A N D C O N T E N T S

O F S A F E T Y R E P O R T S F O R S T A T I O N A R Y N U C L E A R P O W E R P L A N T S

REPORT BASED ON A PANEL HELD IN VIENNA

30 JUN E — 4 JULY 1969SPONSORED BY

THE INTERNATIONAL ATOMIC ENERGY AGENCYAND

THE WORLD HEALTH ORGANIZATION

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1970


THIS SAFETY SERIES IS ALSO PUBLISHED IN FRENCH

GUIDELINES FOR THE LAYOUT AND CONTENTS OF SAFETY REPORTS FOR

STATIONARY NUCLEAR POWER PLANTS IAEA, VIENNA, 1970

S T I/P U B /272

P r in te d by th e IA EA in A u s tr ia O c to b e r 1 9 7 0


F O R E W O R D

The purpose of this manual is to suggest guidelines for the layout and contents of the Safety Reports which support the application for the authorization to construct a stationary nuclear power plant. The guidelines should be considered by the applicant as a series of recommendations to be interpreted according to each specific case.

It was prepared as a result of a meeting of experts in Vienna, on 30 June - 4 July 1969, attended by experts from nine countries. Also present at the meeting were representatives from two international organizations.

The guidelines have been prepared by the International Atomic Energy Agency in co-operation with the World Health Organization, and the publication is sponsored by both organizations.



C O N T E N T S

1. PURPOSE.............................................................................. 92. SCOPE................................................................................... 103. PRINCIPLES........................................................................ 10

CHAPTER II — SITE............................................................................. 131. SITE DESCRIPTION........................................................ 132. METEOROLOGY............................................................... 143. HYDROLOGY.................................................................... 154. GEOLOGY AND SEISMOLOGY.................................. 155. ECO LO G Y ........................................................................... 166. SUMMARY........................................................................... 17

CHAPTER III — COMPONENTS AND SYSTEMS........................ 191. SUMMARY DESCRIPTION OF PLANT..................... 192. REACTOR............................................................................ 203. REACTOR COOLING SYSTEM..................... -.............. 234. CONTAINM ENT SYSTEM.............................................. 265. CONTROL AND INSTRUM ENTATION ................... 306. ELECTRICAL SYSTEMS.................................................. 3 37. POWER CONVERSION SYSTEMS............................... 348. FUEL HANDLING AND STORAGE........................... 379. PLANT AUXILIARIES AND MISCELLANEOUS

SERVICES............................................................................ 3910. RADIATION PROTECTION......................................... 4011. RADIOACTIVE WAVE SYSTEMS................................. 41

C H A P TER I — IN T R O D U C T IO N ........................................................... 9


1. IN IT IA TIN G EVENT...................................................... 432. ANALYSES........................................................................... 45

CHAPTER V — OPERATIONAL ASPECTS.................................... 471. OPERATING ORGANIZATION................................. 472. OPERATION DURING CO M M ISSIONING......483. NORMAL OPERA TIO NS................................................ 494. OPERATION DURING ABNORMAL AND

ACCIDENT C O N D ITIO N S............................................ 50

LIST OF PARTICIPANTS.................................................................... 53

C H A P T ER IV — SA FE T Y A N A LYSES...................................................... 43


C H A P TER I — IN T R O D U C T IO N

1. PURPOSEThe purpose of the present document is to suggest guidelines

for the organization and contents of the Safety Reports which support the request for authorization to construct and operate a nuclear power plant incorporating one or more reactors.

Safety Reports represent the principal communication between the applicant and the Regulatory Body, as outlined in the Code of Practice for the Safe Operation of Nuclear Power Plants 1.

It should be understood that these Safety Reports will be a valuable document for the applicant. They should contain, therefore, precise information on the plant and its operating conditions.

The writing of Safety Reports should be considered an opportunity to enhance the safety of the plant and its operating conditions.

Their main purpose is to provide information to permit the assessment of the nuclear safety implications which may arise from the establishment of the plant at the chosen site with due consideration to the health and safety of the general public and the operating personnel.

Safety Reports should include information such as design bases 2 , site and plant characteristics, limits and conditions, conduct of operation and safety analyses3 , in such way that the Regulatory Body may be able to evaluate the safety of the plant.

The applicant should consider the present guidelines as a series of recommendations to be interpreted according to each specific case.

1 IN T E R N A T IO N A L A T O M IC E N E R G Y A G E N C Y , Safe O p e ra t io n o f N u c le a rP o w er P la n ts , Safety S e ries N o .3 1 , IA E A , V ie n n a ( 1 9 6 9 ) 1 25 .2 Design basis. In fo rm a t io n id en tify in g th e specific fu n c tio n s to b e p e rfo rm e dby a major component o r sy s te m in te rm s o f performance objectives to g e th e r w ith specific v a lu e s o r ra n g e o f v a lu e s c h o s e n fo r c o n tro ll in g p a ra m e te rs a s re fe ren ce b o u n d s o r l im itin g c o n d it io n s fo r d e s ig n . S u ch v a lu e s fo rm th e b a s is fo r th e sp ec ifica tio n . ^ Safety analysis. A s tu d y o f th e r e s p o n s e o f th e p la n t to d is tu rb a n c e s in p ro c essv a ria b le s an d to p o s tu la te d m a lfu n c tio n s o r fa ilu res o f e q u ip m e n t a n d th e p o te n t ia l effec ts o f th a t r e s p o n s e , o r lack th e re o f , in re sp e c t o f th e p la n t e n v iro n s b o th o n a n d oflf-site.

9


10 CHAPTER I

2. SCOPEThe present guide is intended to be applicable to stationary

nuclear power plants.

3. PRINCIPLES3.1. Quality and depth of information

Design and operation description as well as the safety analyses should be presented in a precise and concise manner. It should be of such a quality that the contents of the Safety Reports can be easily understood and analysed by the Regulatory Body. The Safety Reports should present thorough and sufficient information on the plant so that for the purposes of safety analyses no other documentation would normally be required. Supplementary information may however be made available.

It has been recommended4 that the Safety Reports be issued in successive and complementary parts, including:

. (a ) a Pre-Construction Safety Report which supports the application for authorization to construct.(b) a Pre-Operation Safety R eport5 which accompanies theapplication for authorization to operate.The Pre-Construction Safety Report should take the form of

preliminary safety analyses of the proposed nuclear power plant. Informal contacts before the pre-construction review stage are encouraged between those planning to build a reactor and the Regulatory Body to develop mutual understanding of the nature of the project and the Regulatory Body’s requirements. The report should be a statement both of safety principles and of the intended design. It should contain sufficient detailed information, specifications and supporting calculations to enable those responsible for evaluating safety

4 IN T E R N A T IO N A L A T O M IC E N E R G Y A G E N C Y , Safe O p e ra t io n o f N u c lea r P o w er P la n ts , Safety S e ries N o .3 1 , IA EA , V ien n a ( 1 9 6 9 ) 1 25 .^ In so m e c o u n tr ie s th e P re -O p e ra tio n Safety R e p o r t c o n s is ts o f th e P re -C o n s tru c tio n Safety R e p o r t a n d s u b s e q u e n t a m e n d m e n ts a n d a d d it io n s .


INTRODUCTION 11

to assess whether the station can be constructed in a manner conducive to safe operation. It should emphasize both the safety features and the possible hazards, having due regard to the site.

The Pre-Operation Safety Report should provide up-to-date and more specific information on the safety topics discussed in the Pre- Construction Safety Report and on any departure from the safety provisions set out in the latter report (agreed by the Regulatory Body); reference should also be made to any major differences in the design. It should justify the detailed design of the plant and prove its safety.

In addition, the Pre-Operation Safety Report should deal in greater detail than the Pre-Construction Safety Report with matters relating to plant operation.

3.2. TimingIt is most desirable to follow as closely and as continuously

as possible the development of the plant from the moment of site selection on through design, construction and commissioning. It is therefore suggested that sections of the Safety Reports be submitted to the Regulatory Body at an early stage and in accordance with an agreed programme; this approach will facilitate a smooth review procedure and prevent delays in construction and commissioning.



C H A P TER I I — S IT E

This section of the Safety Reports should provide information relating to the site, with particular emphasis on factors important to radiation safety, and emphasizing those site characteristics which influence the design and operation of the reactor plant as they relate to the reactor and its environment.

1. SITE DESCRIPTION1.1. Location

This section should contain a description of the site and its location, several maps of successively larger scale are a convenient way to do this. In addition, there should be a general explanation of why this particular site has been selected. This justification will usually include such factors as growth of power demand in the area, proximity to a power transmission centre, and availability of cooling water and land and the characteristics of the site and public health and safety aspects considered in the selection of the site.

1.2. PopulationA presentation of the current distribution of permanent residents

in the surrounding area, from the most recent survey, as a function of distance and direction, should be described in some convenient graphical and tabular form. Similar information for transient and seasonal inhabitants should also be given. Any regular use of public rights of way, paths and beaches should be mentioned.

A forecast of future development in the area should be given on the basis of the best available information. In making this forecast, account should be taken of the influence the plant itself will have on future development.

13


14 CHAPTER II

1.3. Land useThis section should contain information on the present and

predictable uses to which the surrounding land is put. Of importance in this connection are agriculture and forestry in all their various forms. Data on food and milk production and industry in the area will be compiled with special attention to food processing and sensitive industries like photographic materials and papers. Here, also, due regard will be given to the influence of the nuclear power plant in changing and increasing the utilization of land in the area. Location of any local airfields existing or planned, restriction on chimney heights and other precautions necessary should be indicated. Further, consideration should be given to possible influence of other industrial activities on the operation of the nuclear power plant.

1.4. Access considerationsAccess to the plant site needs to be examined. In particular

cases safety considerations may be involved, for example evacuation and accidents to spent fuel shipments in nearby tunnels or on bridges.

2. METEOROLOGYThis section should contain information on those meteorological

conditions that could have an important influence on the consequences of normal, abnormal and accidental releases of radioactive materials. An explanation will be given of the effects which the meteorological considerations have in establishing design bases and operating requirements for the plant.

Data may be drawn from nearby meteorological stations — usually in operation for many years — as well as from micro- meteorological studies made at the site itself. The information to be presented may include the following:2.1. Wind: Speed, direction and duration. Four seasonal wind-roses are often a convenient form of presentation;2.2. Temperature;2.3. Precipitation: Average, maximum and frequency;


2.4. Storms: Types, severity and frequency;2.5. Atmospheric stability: Types, frequency and duration;2.6. Description of the micro-meteorological programme at the site;2.7. Sources o f meteorological data.

3. HYDROLOGYThis section should summarize the information about water

on, under and around the site which has a bearing on safety, and on cooling requirements. Attention will be given to the sources of water, the total volume and to the uses — both present and projected — of water originating in, or flowing through the area. Much of the data for this section will come from existing records, but field studies may be necessary at some sites. The factors to be considered are listed below:3.1. Surface waters: Coastal waters, lakes, rivers, streams, and swamps;3.2. Ground water: Water tables and wells;3.3. High water levels: Floods, seiches, tsunamis and other tidal waves, etc.3.4. Low water levels: Drought, draw-down, etc.;3.5. Field programme;3.6. Sources o f hydrological data.

SITE 15

4. GEOLOGY AND SEISMOLOGYThis section should summarize information from existing records

and from field studies initiated for the purpose. Emphasis should be given to geological abnormalities which call for any unusual design criteria. This information together with that on earthquakes and ground m otion should be sufficient to explain those design requirements that


have been established by seismic considerations. The section will contain the following information, where appropriate.4.1. Regional geology.4.2. Stratigraphy and bed-rock conditions at the site.4.3. Faulting at and near the site.4.4. Seismic history o f the area.4.5. Geological and seismological field programmes.4.6. Sources o f geological and seismological data.

5. ECOLOGY

This section should consider the biological aspects of the site. The effects of chemical and thermal discharges to the environment and the dilution and reconcentration of radioactive materials reaching man through the critical food pathways are the essence of this section. Consideration may be given to the following items, as necessary:

5.1. Radioactive releases to the environment;5.2. Chemical releases to the environment;5.3. Thermal discharges into the environment;5.4. Critical food pathways to man;5.5. Field programme;5.6. Sources of ecological information;5.7. Estimates on the capacity of the environment, as safe recipient of plant effluents.

16 CHAPTER II


SITE 17

Information on the site should be summarized and evaluated in this section in order to:6.1. Estimate typical environmental conditions;6.2. Anticipate extreme environmental conditions;6.3. Explain the influences of the environment on the plant design and operation;6.4. Give basic information for safety analyses.

A tabulated summary should be given of the specific values or range of values for these environmental variables chosen as the design bases, for example wind velocity, minimum and maximum air temperatures, seismic accelerations, and maximum discharge temperature of liquid effluents.

6. S U M M A R Y



C H A P T E R I I I — C O M P O N E N T S A N D SY ST E M S

1. SUMMARY DESCRIPTION OF PLANTThe purpose of this section is to provide a concise description

of the plant, its principal design criteria6 , its design bases, principal operating characteristics and safety implications.

The following are the types of information that should be included in this section:

(a ) Identification of the principal design criteria to which the plant was designed and is to be operated.(b) The most im portant design and operating characteristics of the plant extracted from succeeding sections of the Reports, without details of description, calculation, or discussion.(c) Identification of those features of the plant likely to be 6f special interest because of their relationship to safety together with a summary of the results of any special research and development programmes undertaken to establish the final design.(d) Identification of cross-linking of systems or any shared systems and their effect on the plant.(e ) Identification of prime contractors for the design, construction and operation of the plant.In addition to the above summary, it is expected that an appli

cant will provide in the following sections full evaluation of the plant together with any supporting information to establish the capability of the plant to perform throughout its lifetime under all normal operational modes, including both transient and steady state.

Broadly speaking, the information submitted should show how the principal design criteria are met by:1.1. Identifying the design bases and explaining the reasons therefore.

^ Principal design criteria m e a n s th o s e fu n d a m e n ta l a rc h ite c tu ra l a n d e n g in e e rin g d e s ig n o b jec tiv e s e s ta b l is h e d for the project. A s su ch , th e s e c r ite r ia re p re s e n t t h e b ro a d fram e o f re fe re n ce w ith in w h ic h th e m o re d e ta i le d p la n t d e s ig n e ffo rt is to p ro c e e d a n d a g a in s t w h ich th e e n d p ro je c t w ill b e ju d g ed .

19


20 CHAPTER III

1.2. Describing the plant to show how the design bases have been satisfied and the code7 used, where applicable.1.3. Showing through evaluations that design bases have been met with a reasonable margin for contingencies, that failures, either in equipment or systems, which could be initiating events for accidents have been identified, and that susceptibility to failures of the various parts of the system has been considered.1.4. Providing the bases for limits and conditions upon operation that might be appropriate in the interest of safety.

The development and construction of the plant should be covered by an adequate quality assurance programme. The Safety Reports should identify the structures, systems and components to be covered by the programme and the major organizations participating in it together with their designated functions.

Controls under-the programme would include verification, inspection, testing and documentation over activities affecting the quality of the identified structures, systems and components to an extent consistent with their importance to safety.

The above programme of inspections and testing should be extended throughout the plant lifetime in order to ensure reliable performance of all critical components and systems.

2. REACTORThe nuclear evaluation should include a description of the

methods of calculation employed in arriving at important nuclear parameters, with an estimate of accuracy by comparison with experiments or with the performance of other reactors.2.1. Nuclear characteristics

2.1.1. Cold and hot excess reactivity and shutdown margin, for the clean condition and the maximum reactivity condition

7 In so m e c o u n tr ie s n a tio n a l c o d e s fo r b u ild in g , v e sse ls an d c o m p o n e n ts a re av a ilab le .


COMPONENTS AND SYSTEMS 21

2.1.2. Coefficients of reactivity and their variation with different parameters2.1.3. H ot and cold reactivity worth of individual control rods and groups of rods for planned patterns and core operating modes with estimates of reductions in effectiveness during core lifetime2.1.4. H ot and cold reactivity worth of fuel assemblies and mechanical or chemical shims and of any materials which could have significant reactivity effect2.1.5. Maximum controlled reactivity insertion rates at startup and at operating conditions2.1.6. Gross and local radial and axial power distribution for different planned rod patterns2.1.7. Minimum critical mass2.1.8. N eutron flux distribution and spectrum at core boundaries and at the pressure vessel wall

2.2. Heat transfer and flow characteristics in the core2.2.1. Correlations and physical data employed in determining important characteristics such as heat transfer coefficients and pressure drop2.2.2. Fuel and cladding temperatures, both local and distributed, with an indication of the correlations used2.2.3. Maximum heat flux and indication of the method for determining hot channel factors2.2.4. Coolant flow distribution and orificing2.2.5. Core pressure drop during normal, abnormal and accident conditions2 .2 .6 . Analysis o f tem perature transients


22 CHAPTER III

2.3.1. Core components and internal structuresDescription of the arrangement of core components and internal structures, such as those related to fuel assemblies and control rods. Sufficient information to support the adequacy of the design for both normal and transient conditions and vibration analysis.Description of the proposals for vibration testing and monitoring of the core components and internal structures.2.3.2. FuelDescription and evaluation of fuel assemblies, including information on the arrangement, dimensions, methods of support and fission gas spaces.The requirements for fuel integrity during normal and abnormal operation.A review of the behaviour of the fuel if subject to a loss of coolant accident.A statement of the failure criteria for fuel and cladding.2.3.3. Reactivity control systemInformation on all reactivity sources and control mechanisms is required (absorber rods, booster rods, poison curtains, soluble and burnable poisons, moderator level control, etc.).The basis of the design of the reactivity control system including a description of the control rods and drives.Information to support their adequacy for the lifetime of the plant.An assessment of the sensitivity of the system to mechanical damage with respect to its capability to provide continuous reactivity control.

2.3. Mechanical design


A review of previous experience and /o r development work with similar systems and materials.Surveillance programmes to demonstrate proper functioning at startup throughout their lifetime.2.3.4. Damaged fuel identification and location systemBases of the design of the system including a description.Information on the adequacy and reliability of the system for the lifetime of the plant.Assessment of the possible consequences of failure of the system.Assessment of the sensitivity of the system with respect to estimated fuel damage.The criteria adopted in allowing damaged fuel to remain in the reactor taking into consideration:(a) Difficulty in handling damaged fuel on removal;(b) Unacceptable contamination to the pressure circuit or

damage to the reactor;(c ) Release of fuel power to the environment.


3. REACTOR COOLING SYSTEMThe reactor cooling system covers both the reactor coolant or

coolants and all components that provide a boundary for coolant flow.The coolant, enclosure represents a principal safeguard whose

integrity is important in the protection of public health and safety. Evaluations, together with the necessary supporting material, should be submitted to show that the reactor coolant system is adequate to accomplish its intended objective and to maintain its integrity under conditions imposed by foreseeable reactor behaviour, either normal or abnormal.

The codes, standards and criteria for pressure-retaining components applied in the design, construction, inspection and testing of


24 CHAPTER III

the primary coolant system and other connected systems should be stated.

Design provisions for inspection, maintenance or replacement accessibility of the main components should be described.

3.1. Pressure vessel and/ or tubes and calandria3.1.1. The bases on which the design was established, considering the performance objectives for both normal and transient conditions throughout plant lifetime (temperature, pressure).3.1.2. Sufficient information to show that design and fabrication will meet performance objectives and the requirements of codes, standards and criteria, especially for the critical points of high stresses.3.1.3. Specifications of the materials used for fabrication.3.1.4. Sufficient descriptive information on the material surveillance programme to demonstrate adequate protection throughout the plant life.3.1.5. Procedures and organizations involved in the quality control during manufacture and assembly; methods of inspection and testing applied.3.1.6. Sufficient information to demonstrate the adequacy ofsystems supporting pressure vessel or pressure tubes to copewith loads induced under normal, abnormal and accidentconditions.3.1.7. Descriptive information on penetrations and nozzles demonstrating adequacy at points of high stress; accessibility for inspection, testing and maintenance.3.1.8. Information on the thermal insulation, its possibleinfluence on transient behaviour and on accessibility for inspection purposes.3.1.9. Description of sealing provisions.



3.2. Main coolant system3 .2 .1 . Description of reactor coolant system, information on mechanical design, material, missile protection, isolability of single loops or com ponents, interconnection with other systems.

3 .2 .2 . Information on all major components with indication of principal pressures, tem peratures, flow rates, and coolant volume under normal and transient conditions.

3 .2 .3 . Regulation (flow, pressure, level, etc.).

3 .2 .4 . Coolant clean-up.

3 .2 .5 . Chemical additions (for reactivity control or corrosion effect prevention).

3 .2 .6 . Pressure relief systems.

3.3. Reactor shut-down cooling system/

3 .3 .1 . Description.


3 .3 .3 . Extent of independent multiplicity or duality of function of systems or their components.

3.4. Emergency cooling systems3 .4 .1 . Description of the various systems.




26 CHAPTER III

3 .4 .4 . Functional requirement and performance analysis;including the range of main coolant system ruptures considered, flow distribution, nuclear and chemical effects.

3 .4 .5 . Methods of actuation.

3 .4 .6 . Provisions for facilitating testing and inspection of components.

3 .4 .7 . Integrated operation of the various core emergency cooling systems.

3 .4 .8 . Concept upon which the operation of the system is predicted has been or will be adequately proven. An evaluation of the long-term capability of the emergency core-cooling system to withstand failure of either an active or passive component after a loss of coolant accident.

4 . C O N TA IN M EN T SYSTEM

4.1. Containment philosophy for the system4 .1 .1 . Principles of design.

4 .1 .2 . Description of the types of accident for which any secondary containment barriers are required.

4 .1 .3 . Performance requirements for any secondary containment barriers.

4 .2 . Structure design and compartmentation4 .2 .1 . Description of the major components and associated systems provided to fulfil the required containment function.

4 .2 .2 . Bases upon which the containment system requirements were established and, in particular, identification and explanation for the choice of values of the principal design parameters; viz. the design pressure and the allowable leakage rate.



4 .2 .3 . The sources and amounts of energy and material which might be released into the containment structure, and the postaccident time dependency associated with these releases.

4 .2 .4 . The contribution of any engineered safeguard system in limiting the maximum value of the energy released in the containment structure in the event of an accident.

4 .2 .5 . Code and vessel classification applicable to the design, fabrication, inspection, and testing of the structure.

4 .2 .6 . Capability of the containment system to continue to function in accordance with design specifications when subjected to environmental forces such as winds, floods, and seismic activity associated with the site location, and a description of the analytical methods employed.

4 .2 .7 . Missile protection features and protection provided against combustible, explosive, or reactive materials being released inside the containment structure.

4 .2 .8 . The corrosion protection or material allowances provided.

4 .2 .9 . Insulation.

4 .2 .1 0 . Shielding requirements incorporated.

4 .2 .1 1 . Provisions of vacuum relief.

4 .2 .1 2 . The extent to which the containment system’s effectiveness and functional dependability will be maintained and verified by testing throughout the plant’s operating lifetime and procedures for testing.

4 .2 .1 3 . Compartmentation of containment by internal structure.

4 .3 . Isolation

4 .3 .1 . Criteria applied with respect to the number and location (inside or outside of containment) of independent isolation valves provided for each fluid system penetrating the containment and the basis thereof.


28 CHAPTER III

4 .3 .2 . Summary of the types of isolation valves applied and their open or closed status under normal operating conditions, shutdown or accident situations.

4 .3 .3 . Governing conditions under which containment isolation becomes mandatory.

4 .3 .4 . Modes of actuation, parameters sensed, closure time and sequence of timing for the principal isolation valves to secure containment isolation.

4 .3 .5 . Provisions to ensure operability of isolation valve system under accident environments and measures taken to prevent damage from missiles.

4 .3 .6 . Provisions for monitoring leakage.

4 .3 .7 . Provisions for facilitating testing and maintenance of the components.

4 .4 . Penetrations

4 .4 .1 . N ature, type, number and position of cable penetrations, pipe penetrations, equipment access doors, emergency escape openings, air locks.

4 .4 .2 . Provisions for inspection, testing and maintenance.

4 .5 . Post-accident pressure reduction systems

4 .5 .1 . Description of systems and components.

4 .5 .2 . Performance objectives and capability of the systems.

4 .5 .3 . Methods of actuation.




4 .5 .5 . Concept upon which the operation of the system is predicted has been or will be adequately proven. Experience, tests under accident simulated conditions, or conservative extrapolations from present knowledge are required.

4 .5 .6 . Provisions for inspection, testing and maintenance of the com ponents.

4 .6 . Clean-up systems

4 6 .1 . Description of systems used to reduce radioactive release during norm al, abnormal and accident conditions (filtering, spraying, etc.).

4 .6 .2 . Performance objectives and capability of the systems.

4 .6 .3 . Description of the purification facilities, delineating the extent of system located within the containment structure.

4 .6 .4 . Provisions to exhaust, monitor, and filter the ventilation air and provisions for safe disposal of the effluent to the outside atm osphere in normal, abnormal and accident conditions.

4 .6 .5 . Provisions for inspection, testing and maintenance of components.

4 .7 . Ventilation systems

4 .7 .1 . Description of the systems and operating conditions.

4 .7 .2 . Information on com ponents and their location.

4 .7 .3 . Performance objectives for normal, abnormal and accident conditions (flow rates and temperature, humidity, radiation levels to be maintained) and provision for monitoring.

4 .7 .4 . Conditions during which ventilation of the containment structure is necessary.

4 .7 .5 . Provisions for inspection, testing and maintenance.


30 CHAPTER III

Instrumentation senses the results of the power generating process and makes it possible to actuate appropriate mechanisms to keep the process within safe and economical bounds. The information sensed by the instrumentation is used to perform regulating and protective functions.

Regulating systems take the plant from shut-down to power and then monitor and maintain key plant variables, such as reactor power, flow, pressure, temperature and radioactivity levels within predetermined limits at steady state and during normal plant transients.

Protective systems are utilized to shut down the reactor to protect the core and the integrity of the coolant boundary against effects of abnormalities such as equipment malfunctions, component failures, and operator errors. Protective systems are also provided to actuate containment isolation and other emergency systems in the event of abnormal conditions.

Redundancy and diversity in the regulating and protecting systems should be discussed and the principles of operation and fail-safe nature of design, wherever applicable, should be clearly dealt with.

5. C O N T R O L A N D IN S T R U M E N T A T IO N

5.1. Regulating systems5 .1 .1 . Bases upon which the design of the regulating systems was established with special reference to design requirements dictated by safety considerations.

5 .1 .2 . Description of the systems provided for regulating the main parameters of the plant, with an explanation of their principal characteristics including items such as automatic and manual control schemes, system inputs and interlocking.

5 .1 .3 . Analysis of system stability with reference to typical plant disturbances.

5 .1 .4 . Sufficient information on important components such as sensors, amplifiers, trip elements, actuators, etc. to explain the principal features of their design and to show that the design requirement will be met.



5 .1 .5 . Capability of the systems to meet functional requirements and susceptibility to com ponent failures or faults, that might impair pertinent functions.

5 .1 .6 . Evaluation of digital computer functions, if any, and relationship with other regulating systems.

5.2. Protecting systems5 .2 .1 . Functional requirements to be satisfied by the various protecting systems, with special reference to the reactor protection systems.

5 .2 .2 . Description of the various systems with indication of type, number and range of independent instrument channels.

5 .2 .3 . Relationship between protecting and regulating systems.

5 .2 .4 . Location of critical equipment and operating environment in which this equipment must function.

5 .2 .5 . Provisions for inspection, testing and maintaining of equipment without impairing the safety of the plant.

5 .2 .6 . Sufficient information on important components such as sensors, amplifiers, trip elements, actuators, etc. to explain the principal features of their design and to show how the design requirements will be met.

5 .2 .7 . Capability of the protecting systems to meet their functional requirements, and susceptibility of com ponent failures or faults.

5.3. InstrumentationIn addition to protective and regulating systems a diversity of

instruments is required. These instruments sense a number of plant process variables such as neutron flux, temperature, pressure and flow. Information should be provided about the individual instrument channels.


32 CHAPTER III

5 .3 .1 . Description of the major components of the systems having safety significance.

5 .3 .2 . Functional requirements of protecting systems and their reliability under environmental conditions.

5 .3 .3 . The nature, type and number of instruments used for monitoring any plant parameters relevant to nuclear safety.

5 .3 .4 . Evaluation of effects of loss of power.

5 -3 .5 . A list of critical components in the systems required to operate following a design basis accident, their location and the environments in which these must function.

5 .3 .6 . Provisions for testing, calibrating and monitoring the systems.

5 .4 . Computer systems

5 .4 .1 . Functional requirements to be satisfied by the automatic com puter control system.

5 .4 .2 . Relationship with the protecting and regulating systems.

5 .4 .3 . Location of critical equipment and operating environment in which this equipment must function if protective actions are performed.

5 .4 .4 . Assurance that testing and maintaining the equipment will not impair the safety of the plant.

5 .4 .5 . Sufficient information on important components such as sensors, amplifiers, actuators, etc. to explain the principal features of their design and to show how the design requirements will be met.

5 .4 .6 . Capability of the system to meet functional requirements and susceptibility to component failures or faults, that might impair safety functions.



This section of the Reports should discuss the electrical power distribution system of the plant with emphasis upon those features whereby continuity of power sufficient for safety needs is assured and to evaluate the reliability of sources, distribution systems, equipments, redundancy, independence of the systems under various conditions of operation.

6 .1 . Network interconnections

6 .1 .1 . A schematic single line diagram showing tie-in of the nuclear plant to the grid and the availability of incoming lines as alternate sources for station loads should be included.

6 .1 .2 . An evaluation of the reliability of the arrangement in providing continuity of power should be made, including such things as:

6 .1 .2 .1 . A resume of credible faults capable of causing the loss of more than one source of power.

6 .1 .2 .2 . Effectiveness of the protective features provided to minimize power failures.

6 .1 .2 .3 . Local environmental factors that might create unusual demands upon alternate or emergency power sources.

6 .2 . Station distribution systems

6 .2 .1 . Single line diagrams of a.c. and d.c. systems, as appropriate, for identifying arrangement of station buses, major loads and switching circuitry should be included.

6 .2 .2 . Evaluation of the electrical layout for vulnerability to physical damage of vital circuits as a result of accidents should be made.

6 .2 .3 . Evaluation of the distribution of loads to the various buses in the interest of maintaining power to vital loads.

6. ELECTR ICAL SYSTEM S


34 CHAPTER III

6 .2 .4 . Evaluation of possible consequences of step or ramp electrical load changes either through operator error, equipment malfunction, or environmental disturbances.

6 .3 . Emergency power sources

6 .3 .1 . Line diagrams showing interconnection of locally generated standby and emergency power, a.c. and d .c., and the loads to be fed by such sources.

6 .3 .2 . Brief descriptions of emergency power sources with emphasis on those features provided to protect the supplies from faults, overloads or damage under emergency conditions which could interfere with performance.

6 .3 .3 . Evaluation of loads required to be powered in the interest of safety and the relationship of the emergency load to installed capacity.

6 .3 .4 . Performance history of emergency sources of power andthe possible effects of failure to deliver should an emergencyarise.

7. POW ER CO N V ER SIO N SYSTEMS

7 .1 . This section should have in view at least three main objectives:

( a ) to provide information (technical description) to understand the systems and their influence on the reactor in steady conditions and specially during transients,(b ) to allow an analysis of how the Power Conversion System may alter reactor conditions by changes in tem perature, pressure or reactivity,( c ) to permit an analysis of the radioactivity problems which may be associated with the Power Conversion System.

7 .1 .1 . Steady state characteristics of the various parameters versus load (tem perature, rates of flow, pressure, masses, volumes, etc.) and corresponding limiting values.



Limits in the rates of change of such parameters either imposed by the nature of the considered equipment or by any external influence (for example electrical load changes). Influence of these limits on the reactor and possible provisions to cope with them (e .g . steam discharge).

7 .1 .2 . Pressure relief devices.

Thermal power dumps or discharges.

Backup systems.

Analyses of the associated control systems.

• Influence on the reactor of malfunction of components (pumps, valves, reheaters, etc.).

7 .1 .3 . Physical limits of the systems and means of their isolation from radioactive contaminants or control of such.

Provisions to control radioactive releases under normal and abnormal conditions.

7 .2 . The components of the systems to be reviewed are those which have a bearing on the considerations developed above. The m ore important com ponents which may require special consideration are listed below:

7 .2 .1 . Turbine or turbines

Speed and load control. Overspeed trip. Speed excursion during load transients.

Steam reliefs or discharges.

7 .2 .2 . Condenser

Incondensable gas removal monitoring and control.

Capability of treating dump-steam.

Backup s t e r n s associated with above (pum ps, ejectors).


CHAPTER III

Heat rejection from condenser cooling system.

Reference should also be made to the cooling water (circulating water) system or cooling towers inasmuch as these influence the proper functioning of the condenser and other heat exchangers cooled by the system (e.g . rate of flow at river intake, siphon). An indication of the temperature of the outlet circulating water should be given.

7 .2 .3 . Condensate heating system

Influence of abnormal conditions on the reactivity of the core.

7 .2 .4 . Condensate purification system

Effects on the core, if any.

7 .2 .5 . Feed water storage, pumping and supply to the steam generator

Implication on reactivity.

Under conditions of temperature transients or in case of malfunctioning of components (pumps, valves, regulators).

Isolation of the feed water lines.

Vibration of piping or components and their consequences.

7 .2 .6 . Steam generators (boilers) or separators

Alignment in respect of rest of the plant.

Consequences of failure.

7 .2 .7 . Steam piping

Vibrations.

Relief-valves.

Pressure control and relief.



8. FU EL H A N D LIN G AND STORAGE

A description should be given of the fuel charging operations. An estimate of the frequency of fuel changing should be given, this will be related to core physics studies and reactivity requirements. The fuel handling for on-load fuelling should be dealt with in detail.

A record of the amount and movement of all fissile and fertile material should be kept on each reactor site. Information should also be given on the mode, rate of delivery and shipment of fuel (fresh an d /o r irradiated).

8 .1 . New or unirradiated fuel

Description and evaluation should be given of:

8 .1 .1 . Storage location and arrangement, including physical security and health physics and criticality considerations.

8 .1 .2 . Means of fire prevention, detection and control.

8 .1 .3 . Facilities for handling fuel assemblies from receipt until transfer to the fuelling equipment.

8 .1 .4 . Facilities for inspection.

8 .2 . Fuelling equipment

This includes equipment to handle fuel at the reactor and until it is transferred to storage. Description and evaluation should be given of:

8 .2 .1 . Mechanical design, including shielding considerations and arrangement of standby equipment.

8 .2 .2 . The procedures for removing, repositioning or replacing fuel.

8 .2 .3 . Controls required, their location, function, operation.

8 .2 .4 . H eat removal provisions and their interaction, if any, with the reactor cooling system;


38 CHAPTER III

8 .2 .5 . Special facilities for handling and storing damaged fuel.

8 .2 .6 . Effects of delays in discharging irradiated fuel — temperatures reached and time scales.

8 .2 .7 . Provisions for emergency cooling in event of delays.

8 .3 . Spent or irradiated fuel

Description and evaluation should be given of:

8 .3 .1 . Storage location and arrangement, including physical security and criticality considerations.

8 .3 .2 . Means of fire prevention, detection and control.

8 .3 .3 . Facilities for area ventilation, including filtering equipment and monitoring for radioactivity.

8 .3 .4 . Cooling arrangements.

8 .3 .5 . Any special facilities for handling and storing damaged fuel.

8 .3 .6 . Inspection facilities.

8 .3 .7 . Additional facilities provided and spare storage capacity to meet the requirement, if any, of emergency unloading of fuel.

8 .4 . Other components

A description and evaluation should be given of facilities and procedures which may be required to handle (for repair, inspection or replacement) core com ponents, reactivity mechanisms such as control or shut-off elements, and other radioactive equipment or sources.

8 .5 . Shipping facilities

A description and evaluation should be given of the facilities and procedures required for loading and handling, including monitor



ing and decontamination, of casks and other containers8 . Arrangements for shipment or off-site disposal of fuel and other components should be described.

9 . PLA N T A U X ILIA R IES AND M ISCELLANEO US SERVICES

9 .1 . The Safety Reports should give appropriate information on those auxiliary systems and services which may have a direct or an indirect bearing on nuclear safety.

9-2 . Such items should be reviewed with due consideration to the consequences on the plant in the event of malfunction of their com ponents. Backup systems should be described and their capability demonstrated.

9 .3 . Some of the systems and services to be considered here are directly associated with proper operation of equipment (o r with cooling objectives (heat-sinks) or both). When heat-sinks are considered, the associated system becomes important whether core cooling is involved directly or not. Such systems should therefore be given full attention in the Safety Reports; among them, the following should be included:

(a ) Raw water system;(b ) Intermediate cooling system;(c ) Service water systems;(d ) Compressed air systems;(e ) Fire fighting equipment.

9 .4 . Some other systems or auxiliaries may be associated with conduct of.operation and their malfunctioning might prevent appropriate actions in case of need.

F o r c o n s id e ra tio n o f d e s ig n o f s h ip p in g c o n ta in e rs an d tra n s p o r ta t io n facilit ie s see IN T E R N A T IO N A L A T O M IC E N E R G Y A G E N C Y , R e g u la tio n s fo r th e Safe T ra n s p o r t o f R a d io ac tiv e M a te r ia ls (1 9 6 7 E d n ) , Safety S eries N o .6 , IA EA , V ie n n a ( 1 9 6 7 ) 1 17 .


4 0 CHAPTER III

Such systems should also be reviewed in the Safety Reports since their behaviour may complicate safety problems. The following should be included:

( a ) Communicating systems (public-address, telephone, radiotelephone, television, acoustic signal, visual signal);(b ) Internal transportation (elevators, conveyors).

9 .5 . Provisions of emergency equipment both on and off-site should be described.

9 .6 . Provisions for prevention of accidental dropping of heavy objects in the plant and evaluation of the arrangements for dealing with exigencies of such events should be detailed.

10. R A D IA TIO N P R O T EC T IO N

This section should contain information regarding the radiation protection instrumentation features of design and operation provided primarily for the protection of operating personnel and the general public.

1 0 .1 . Shielding

1 0 .1 .1 . Assumed radiation sources.

1 0 .1 .2 . Specification of radiation exposure limits to be applied.

1 0 .1 .3 . Locations, thickness and materials of the shields.

1 0 .1 .4 . Radiation levels during normal operation, shut-down and in event of accicents.

1 0 .1 .5 . Basic methods of calculation.

1 0 .1 .6 . Provisions to ensure continuous effectiveness and integrity.

1 0 .1 .7 . Plans for radiation surveys.



10 .2 . Area radiation monitoring systems

1 0 .2 .1 . Location, types and sensitivity of radiation detectors.

1 0 .2 .2 . Operating characteristics and set levels for alarms.

1 0 .3 . Health physics

1 0 .3 .1 . Personnel monitoring system and the frequency of determination of radiation exposures.

1 0 .3 .2 . Personnel protective equipment.

1 0 .3 .3 . Different active zones and control of access, change room s, decontamination facilities.

1 0 .3 .4 . The facilities for the health physics service.

1 0 .3 .5 . Arrangements for medical examination and assessment of internally deposited radioactivity.

10 .4 . Plans for environmental radiation surveys

11. R A D IO A CTIV E WASTE SYSTEMS

The Safety Reports should detail the design bases and evaluate the radioactive waste systems for normal and abnormal conditions of operation and for accident conditions highlighting the designed containment capability of the systems.

11.1. The estimation of maximum and average arisings of volume rates of solid, liquid and gaseous particulate wastes, identification of their origin, com position of nuclides, activity levels and chemical forms.

11.2. Description of methods for treatment and disposal for each type of waste and identification of limiting factors, such as limits for releases or release rates.


42 CHAPTER III

11. 3. The collection and processing equipment for solid wastes, storage and disposal.

11.4. The collection and processing equipment of liquid wastes, storage, points of release to the environment.

11. 5. The collection and processing equipment o f gaseous wastes, storage, points of release to the atmosphere.

11.6. Dose estimates within and at the station boundary— a convenient method of presenting the information would be a site layout giving the radiation contours.

11. 7. Process control and monitoring of releases.

11. 8. Evaluation of possible failures in the systems.

11.9. Environmental monitoring.


C H A P T E R IV — SA F E T Y A N A LYSES

Safety of a reactor plant as achieved through design is commonly shown by studies made of the response of the plant to disturbances in process variables and to postulated malfunctions or failures of equipment, human errors. Such Safety Analyses provide a significant contribution to the design specifications for components and systems and subsequently are important in showing that a design consistent with public and personnel safety has been achieved. These analyses are a focal point of safety review of a reactor plant.

While such analyses generally lean heavily upon calculational models, they are none the less expected to be based sufficiently upon experimental evidence an d /o r substantiated previous design experience to provide confidence in predicted results.

In previous sections of this guide, it has been stated that the individual system and com ponent designs should be evaluated for effects of anticipated disturbances and for susceptibility to component malfunction or failures. In this section, it is expected that the consequences of those failures or abnormal situations will be developed and the capacity built into the plant to control or accomm odate such situations or the limitations of expected performance revealed.

1. IN IT IA T IN G EVENT

The postulated initiating events should be selected on the basis of one or both of the following:

1.1. An examination of the response of the plant to malfunctions which may result in severe consequences.

1.2. The susceptibility of components to malfunctions or failures. It is recognized that situations analysed may range from a fairly comm on disturbance such as load loss from a line fault to the remote possibility of loss of integrity of a major component.

4 3


44 CHAPTER IV

The abnormal situations analysed should start from consideration of deviant events such as9 :

1.2.1. Anticipated variations in the reactivity of the reactor, to be compensated by buildup and burnout of xenon poisoning, fuel burnup, soluble or burnable poisons, on-load refuelling, fuel followers, temperature, moderator and void coefficients.1.2.2. Reactivity excursion from such causes as uncontrolled rod withdrawals, control rod failure, unintended coolant insertion, loss of coolant, changes in core geometry.1.2.3. Electrical load disturbances such as loss of load or turbine trip.1.2.4. Failure of the regulating instrumentation, causing, for example, an unacceptable mis-match between power produced and removed.1.2.5. Situations created by malfunction of components such as loss of one or more coolant circulators, or fuel channel blockage.1.2.6. Possibilities for equipment failures or loss of component integrity including the whole spectrum of loss of coolant accidents.1.2.7. Spent fuel handling movement.1.2.8. Inadvertent criticality.1.2.9. External causes such as storms, missiles or explosions, earthquakes.1.2.10. Malfunction of auxiliary systems.1.2.11. Extreme environmental variations.For the most part it would be expected that the abnormal events

analysed would have been identified from the systematic design evalua-

9 T h is l is t s h o u ld b e rev iew ed a s illu s tra tiv e o n ly .


SAFETY ANALYSES 45

tionio of the reactor systems and components. However, postulated additional events may also be a useful device in assessing the performance capability of the safety systems. In any case, information provided with respect to the initiating event should make clear the perspective within which the assumed event should be viewed and why the particular parameters of the abnormality have been chosen.

2. ANALYSESAnalyses should show the extent to which protective systems

must function. Where standby systems provide redundant or backup protective functions, the analyses should show not only the effects if expected performance is achieved, but also effects if a degraded performance is experienced. The analyses should take into account the possible effects on the environment of radioactive discharges under normal and abnormal plant operations. This section should include the following:

2.1. The methods, assumptions, conditions, employed in estimating the course of events and the consequences.2.2. The physical and mathematical model employed denoting significant simplification or approximations introduced to render the analyses tractable.2.3. Information sufficient to evaluate the adequacy of any digital computer programme or analogue simulation used in the analyses and the extent or range of variables investigated.2.4. The considerations of uncertainties in calculations! methods, in equipm ent performance, such as instrumentation response characteristics, or other indeterminate effects taken into account in the evaluation of the results.

Design evaluation. A s tu d y o f th e fu n c tio n a l a n d p h y sica l fe a tu re s o f th e m a jo r r e a c to r p la n t sy s te m s a n d c o m p o n e n ts to d e te rm in e :

(a ) W h e th e r th e d e s ig n can m e e t o r h a s m e t p e rfo rm a n c e o b jec tiv e s with an adequate margin o f safe ty in c lu d in g th e a d h e re n c e to a p p lic ab le c o d es ;( b ) T h e id e n ti ty a n d su sce p tib ility o f failures, either in equipment or control over process variable j w h ich c o u ld b e in itia tin g e v en ts fo r a cc id e n ts .


46 CHAPTER IV

2.5. References to published data or research and development investigations in substantiation of the assumed or calculated conditions.2.6. The results and consequences derived from each analysis and the margin of protection provided by either a backup or protective system which is depended upon to limit the extent or magnitude of the consequences.2.7. The time-dependent variation of the more important variables such as temperatures, pressure, release rate of the fission and other radiotoxic nuclear products within the containment system, their escape to the environment and dose calculations.2.8. The extent o f systems interdependency in contributing directly or indirectly to containment function such as the contribution of:

(a) Emergency core cooling systems;(b) Radiotoxic material removal system;(c ) Post-accident heat removal system.


C H A P TER V — O P E R A T IO N A L A SP E C T S

This section of the Safety Reports should provide information describing the way operation of the plant will be conducted.

It involves information on the competence and responsibility of operating personnel to perform their duties safely during normal, abnormal and accident conditions of the plant. It should also include any special operational provisions related to commissioning.

1. OPERATING ORGANIZATIONIt is very important that all personnel engaged in a nuclear

power plant should have the necessary, qualifications and experience appropriate to their duties. The Safety Reports should therefore give enough information on the various aspects of the problems, some of which are reviewed hereunder.1.1. Job description

Functional description should be provided and related to startup, regular operation and emergency periods.1.2. Qualifications and experiences

The preoperational Safety Reports should provide details of the educational qualifications and experience of the personnel who will actually carry out operational duties in the plant.1.3. Training

The applicant should describe his training programmes emphasizing the topics to be covered for various aspects of operation. On-site and off-site training should be mentioned if any.1.4. Competency assessment

In addition to fulfilling any requirements which may be laid down by the Regulatory Body, the management should periodically

47


48 CHAPTER V

check and satisfy itself that individual staff members are maintaining the required standards and competence appropriate to their duties.1.5. Management and administration

In order to ensure that they are well known and fully understood, the duties and responsibilities of staff members at all levels in the operating organization should be clearly set out in writing. The authority vested in each staff member should be sufficient to ensure the effective discharge of his responsibilities. Organization charts delineating the lines of responsibility and authority for both operation and safety, during start-up, regular operation and emergencies should be included.

2. OPERATION DU RIN G COMMISSIONINGThis subsection of the Safety Reports should provide the informa

tion required to permit a safety evaluation of the plant operation during commissioning. It is during this phase that the final operator of the plant will complete his familiarizing with the systems, components and their interplay and the achievement of the design intentions will be demonstrated by acceptance and performance tests of all systems and components.

For an evaluation of the commissioning, the following information should be provided in the Safety Reports.2.1. Staff requirem ents fo r commissioning

If the responsibility for the plant does not rest with the plant operator, capabilities and responsibilities of the commissioning staff should be given in the same detail as for the final operator.2.2. Review process

The results obtained will be checked and compared with the design intentions after completions of each commissioning phase before the beginning of the next phase. Details of the review process should be given.

The fulfilment of requirements regarding limits and conditions can take place step by step as the commissioning programme pro


OPERATIONAL ASPECTS 49

ceeds, and a clear statement of the prerequisites for each phase of commissioning should be approved by the Regulatory Body.2.3. Documentation

Type, detail, timing and distribution of the documentation, containing the results of the commissioning, corrective measures, recommendations etc., should be given.2.4. Preoperational tests

The sequence, purpose and limitations of the planned acceptance and performance tests should be outlined. Additional equipment, safety precautions and their reason should be described. Reference should be made to a detailed programme of the planned preoperational tests.2.5. Low power tests

The sequence, purpose and limitations of the planned tests of the physical and technical characteristics of the plant after reactor fuelling, including initial criticality should be oudined. Additional equipment, safety precautions and their reason shall be described. Reference should be made to a detailed programme of the planned low power tests.2.6. Power tests

The sequence, purpose and limitations of the planned tests and operations conducted in several steps with increasing power levels in order to check the overall plant behaviour under transient conditions should be outlined. Additional equipment, safety precautions and their reasons should be given. Reference should be made to a detailed programme of the planned power tests and operations.

3. NORMAL OPERATIONS3.1. Assurance should be given that comprehensive procedures will be prepared for all routine operations, and for correcting or otherwise dealing with expected deviations from normality. Routine operations will include regular maintenance, tests and inspections at appropriate


50 CHAPTER V

intervals, and radiological surveys in the plant and its environment. Before the operation of the reactor a set of proposed limits and conditions should be submitted to the Regulatory Body for review and approval. An official copy of the approved limits and conditions 11 must be held by the operating organization. These approved limits and conditions may form a part of the operating license.3.2. The extent and formal of the logs, records and reports to be kept should be described, e.g. periodic summary reports on matters relating to plant safety and reports on modifications should be supplied by the operating organization to the Regulatory Body12 . In addition, the Regulatory Body may specify certain categories of occurrences for which it will require separate reports. The scope, timing and distribution of reports should be agreed on by the Regulatory Body and the operating organization.3.3. An indication should be given of mechanisms for appropriate review of operating limits and procedures to evaluate their effectiveness and appropriateness.

4. OPERATION DU RING ABNORMAL AND ACCIDENTC O N D ITIO N SThis subsection should describe the provisions designed to cope

with or mitigate the consequences of abnormal and accident situations. The Safety Reports may also contain operational procedures and descriptions of the measures to be taken. The final operator should detail the plan of action. Some abnormal situations may have some consequences on the environment and require some external support from, or co-operation with, an external body (fire-squads, civil services). The Safety Reports should identify such cases, outline possible new lines of responsibility and authority and give reference to specific and appropriate procedures.

* * In so m e c o u n tr ie s p ro v is io n s g o v e rn in g lim its a n d c o n d it io n s a n d commission- in g p ro c e d u re s a re n o t se t o u t in specific d o c u m e n ts ; th e re le v an t in fo rm a tio n sh o u ld , h o w e v e r, b e e as ily id en tif ia b le a n d re ad ily av ailab le .12 In s o m e c o u n tr ie s i t is c o n s id e re d a d v isab le th a t su ch re p o r ts s h o u ld b e s u p p lie d a t in te rv a ls ra n g in g fro m ev ery s ix m o n th s to every th re e y ea rs , d e p e n d in g o n th e ty p e o f o p e ra t io n .


OPERATIONAL ASPECTS 51

The Safety Reports should also give assurance that adequate equipment and facilities, which might be needed to meet emergency conditions, have been provided.

The Safety Reports should give some indication as to how the above provisions will be periodically checked, reviewed and evaluated.



L I S T O F P A R T I C I P A N T S

Chairman GANGULY, A.K.

Panel Membersb e r An e k , J.

BIRKHOFER, A.

CLEMENT, B.

D iN U N N O , J.

DOYEN, R.

GAUSDEN, R.

IANSITI, E.

Health Physics Division,Bhabha Atomic Research Centre, Bombay, India

Czechoslovak Atomic Energy Commission, Prague, Czechoslovak Socialist RepublicLabor fur Reaktorregelung und Anlagensicherung, Reaktorstation,Garching, Federal Republic of GermanyDepartement des etudes de piles,CEN Saclay,Gif-sur-Yvette, FranceUSAEC Scientific Representative,American Embassy,Paris 8, FranceS. A. Centre st Sud,Brussels, BelgiumInspectorate of Nuclear Installations, Ministry of Power,London S.W., United KingdomDirector, Division of Safety and Control, Comitato Na2ionale per l ’Energia Nucleare, Rome, Italy

53


MOLLOY, T.J.

VESTERGAARD, R.

Observers FRANZEN, L.F.

JANSSON, E.

TOCCAFOND1, G. VKLONA, 1;.

Representatives SHALMON, E.

VINCK, W.

WHIPPLE, G.H.

Scientific Secretary McCULLEN, J.D .

Atomic Energy of Canada Ltd., Ontario, CanadaAktiebolaget Atomenergi, Stockholm, Sweden

Institut fiir Reaktorsicherheit, Koln-Ehrenfeld,Federal Republic of GermanyNuclear Inspector,Swedish Atomic Energy Board, Stockholm, SwedenENEL, Rome, ItalyENEL, Rome, Italy

World Health Organization,Environmental Pollution,Geneva, SwitzerlandCommission des Communaut6s Europeennes, Brussels, BelgiumWorld Health Organization,Geneva, Switzerland

Division of Health, Safety and Waste Management,IAEA,Vienna, Austria

54


HOW TO ORDER IAEA PUBLICATIONS■ An exclusive sales agent for IAEA publications, to whom all orders

and inquiries should be addressed, has been appointed in the following country:

UNITED STATES OF AMERICA UNIPUB, P.O. Box 433 , M urray Hill S ta tio n , N ew Y ork, N .Y. 10016

In the following countries IAEA publications may be purchased from the sales agents or booksellers listed or through your major local booksellers. Payment can be made in local currency or with UNESCO coupons.

ARGENTINA

AUSTRALIABELGIUMCANADA

C.S.S.R.

FRANCE

HUNGARY

INDIAISRAEL

ITALY

JAPANNETHERLANDS

PAKISTANPOLAND

ROMANIA SOUTH AFRICA

SPAIN

SWEDEN U N ITE D KINGDOM

U .S.S.R . Y U G OSLAV IA

C om isi6n N acional d e E nergia A t6m ica , A venida del L ibertador 8250 , B uenos A iresH un ter Publications, 58 A G ipps S tree t, C ollingw ood, V ictoria 3066Service du C ourrier de I'UNESCO, 112 , Rue du T rone, B-1050 BrusselsIn fo rm ation C anada, 171 S later S tree t, O ttaw a , O n t. K1A 0 S9S.N .T .L ., Spaiend 51 , CS-1 10 00 PragueAlfa, Publishers, H urbanovo ndm estie 6, CS-800 00 BratislavaO ffice In te rna tion a l de D ocu m en ta tion e t L ibrairie, 48 , rue G ay-Lussac,F-75005 ParisK ultu ra, H ungarian T rading C om pany for B ooks and N ew spapers,P.O. Box 149 , H-1011 B udapest 62O xford Book and S ta tio nery C om p., 17 , Park S tree t, C alcu tta 1 6 ;O xford Book and S ta tio nery C om p., Scindia House, New Delhi-110001 H eiliger and Co., 3 , N athan S trauss S tr., Jerusalem L ibreria Scien tifica, D ott. d e Biasio Lucio "a e io u ” .Via Meravigli 16, 1*20123 MilanM aruzen C om pany , L td ., P.O .Box 5050 , 100*31 T o k y o In ternationalM arinus N ijhoff N .V ., Lange V oo rho u t 9- 11 , P.O. Box 269 , T he HagueM irza Book A gency, 65 , T he Mall, P.O .B ox 729 , L ahore-3A rs Polona, C en trala H andlu Zagranicznego, K rakow skie P rzedm iescie 7 ,W arsawC artim ex, 3*5 13 D ecem brie S tree t, P.O .Box 134- 135 , Bucarest Van S cha ik 's B ooksto re, P.O .Box 724 , P reto ria U niversitas B ooks (Pty) L td ., P.O .Box 1557 , P reto ria Diaz de S an tos, Lagasca 9 5 , M adrid-6 Calle F rancisco N avacerrada, 8, M adrid-28C.E. F ritzes Kungl. H ovbokhandel, F redsgatan 2 , S -103 07 S tockho lm H er M ajesty 's S ta tio nery O ffice, P.O . Box 569 , L ondon S E 1 9 NH M ezhdunarodnaya Kniga, Sm olenskaya-Sennaya 32 -3 4 , M oscow G -200 Jugoslovenska Knjiga, T erazije 2 7 , Y U -11000 Belgrade

Orders from countries where sales agents have not yet been appointed and requests for information should be addressed directly to:Division of Publications

^ International Atomic Energy Agency & Karntner Ring 11, P.O.Box 590, A-1011 Vienna, Austria


IN T E R N A T IO N A L S U B JE C T G R O U P : IIA T O M IC E N E R G Y A G E N C Y Nuclear Safety and Environm ental Protection/Nuclear SafetyV IE N N A , 1970 P R IC E : U S $ 2 .5 0


Guidelines for the Layout and Contents of Safety Reports ... Safety Standards/Safety_Series_034_1970.pdf(a) a Pre-Construction Safety Report which supports the appli cation for authorization

Documents