safety design criteria india - iaea

Safety Design Criteria of Indian Sodium Cooled Fast Reactors

P. Pillai, P.Chellapandi, S.C.Chetal*, P.R.Vasudeva Rao

IGCAR, Kalpakkam, India* Former Director, IGCAR

Third Joint GIF – IAEA Workshop onSafety Design Criteria for Sodium-Cooled Fast Reactors

26 – 27 February 2013 IAEA, Vienna

Section I : About PFBR, Safety Criteria and

Safety Review

Section II : Feedback from PFBR safety review

Section III : Safety Design Criteria for Future Indian SFRs

Outline

Section I

About PFBR safety systems in brief, Safety Criteria and Safety Review

Indian SFR Program

10 0907 08

06

04 05

03

02 01

11

12

13

72

5

Ø11950

FBTR• 40 MWt (13.5 MWe) • Loop type reactor • PuC – UC• Design: CEA, France • Since 1985…

PFBR• 1250 MWt (500 MWe)• Pool Type• UO2-PuO2• Indigenous design and

construction

CFBR• 500 MWe• Pool Type• UO2-PuO2• Six units (3 twin units) • Launching from 2023

Future FBR• 1000 MWe• Pool Type• Metallic fuel • Series

construction• Beyond 2025

Credible

confidence in

fuel cycle

Human resources

Human resources

25 years ofsuccessfuloperation

State-of-art

concepts

Innovative concepts

MFTR� Industrial scale subassembly size

� Experience with fuel behavior, Consolidation of

fuel performance & data, Closed fuel cycle aspects

FBTR:

Station Operation Review Committee (SORC)

AERB - Safety Review Committee

SARCOP – SAfety Review Committee for Operating Plants

PFBR:

Three tier safety review

Internal Safety Committee

Project Design Safety Committee

Apex Committee for Project Safety Review

AERB – Project Clearance

Safety Review Structure for Indian Fast Reactors

PFBR Reactor Assembly

01 Main Vessel

02 Core Support Structure

03 Core Catcher

04 Grid Plate

05 Core

06 Inner Vessel

07 Roof Slab

08 Large Rotating Plug

09 Small Rotating Plug

10 Control Plug

11 CSRDM / DSRDM

12 Transfer Arm

13 Intermediate Heat Exchanger

14 Primary Sodium Pump

15 Safety Vessel

16 Reactor Vault

• Two diverse shutdown systems

• Control & Safety Rod (CSR) - 9 rods (65%

enriched B4C in Central portion + 20 cm

natural B4C at top & bottom)

• Diverse Safety Rod (DSR) - 3 rods of 65%

enriched B4C

• Diversity in design

• 9 CSR + 3 DSR worth – 13300 pcm

• CSR + DSR - minimum SD margin 5000

pcm

CSR : (n–1) criterion for cold shut down

DSR : (n-1) criterion for cold shut down

• CSR + DSR with loading error

1 FSA replacing absorber SA – 1840 pcm

2 absorber SA removal – 3820 pcm

• Reliability - 10-6/r-y

PFBR Shutdown System

PFBR Decay Heat Removal Systems

• Safety Grade Decay Heat Removal - 4

independent circuits - 8 MW capacity each.

• Diversity in design of heat exchangers.

• Natural circulation on both intermediate

sodium and air sides.

• Redundancy provided for dampers. 2

dampers provided at inlet and outlet of

each AHX.

• Provision of pony motor to run at 15% for

primary sodium pump.

• No pony motor or Class-III power for

secondary sodium pumps.

• Automatic operation with no credit for a

period of 30 minutes following a PIE.

• Reliability - 10-7/r-y

PFBR Containment

• RCB design pressure - 25 kPa based on sodium release and consequent pressure rise due to CDA.

• RCB leak rate 0.1% v/h at design pressure for analysis

• Unlined reinforced concrete construction

Protection Against Sodium Fire

Protection Against Sodium Water Reaction

PFBR SCRAM Parameters

* - for core subassemblies

Design Basis Events

Category Definition

1 Normal Operation

2 Upset Condition (expected once or more during operation) *

Abnormal behaviour due to minor faults

Do not require any major repair or regulatory inspection

( > 10 - 2 / reactor year )

3 Emergency Condition (not likely to occur – single occurrence

considered in design)*

Requires mandatory inspection

( < 10 - 2 / reactor year & > 10 - 4 / reactor year )

4 Faulted Condition (not likely to occur – configuration remains

coolable)*

Plant Restart not essential & Public health and safety to be

ensured ( < 10 - 4 / reactor year & > 10 - 6 / reactor year )

Cumulative Damage Fraction Apportionment for Reactor Core Components

EVENT

CATEGORY

PHYSICAL LIMITS APPORTIONED

CUMULATIVE

DAMAGE

FRACTION

1 Clad integrity maintained for steady

state loads for a peak burnup of 100

GWd/t

0.25

2 Clad integrity ensured to sustain for 10

events

0.25

3 Clad integrity ensured for a single

event

0.25

4 Coolable geometry is maintained

(no clad melting)

Clad Integrity not essential

--

A fraction of 0.25 CDF reserved for spent fuel storage and handling

PARAMETERSEVENT CATEGORIES

1 2 3 4

Frequency f > 1 10-2 < f < 1 10-4 < f < 10-2 10-6 < f < 10-4

Functional

Shutdown Yes Yes Yes Yes

Decay heat removal Yes Yes Yes Yes

Containment Yes Yes Yes Yes

Restart YesYes, after

fault clearance

Yes, after inspection, repair and

requalification

Not necessary

Temperatures, K

Cold pool@ 670 813 873 913

Hot pool 820 873 898 923

Average SA coolant hotspotNo Bulk Coolant Boiling

No burnout in local hotspots

Clad hotspot

Driver fuel SA

973and

CDF ≤ 0.25

974 – 1023 for 75 min &

1023 - 1073for 15 min

for all cate.2events andCDF ≤ 0.25

974-1073for 15 min &

1073–1123 for 6 min &

1123–1173 for 2 min andCDF ≤ 0.25

1473

Storage SA823 873 923

1223& CDF < 0.25

Design Design Safety Limits & Criteria for DBELimits & Criteria for DBE

Fuel hotspot No melting No melting No melting

Partial melting to the extent that there is no clad failure associated with this melting

Structural

As per RCC-MR Level A Level A Level C Level D

Radiation

Plant Personnel 30 mSv/y/p, 100 mSv/p cumulative for a block of 5 y 250 mSv / p

Public at site boundary 0.1 mSv / y (for PFBR – 10% of Kalpakkam Complex)100 mSv /

event *

Public evacuation No No No No

Events with probability of occurrence < 10-6 / r-y - BDBE* Same 100 mSv as that specified for CDA (BDBE), Temperature in K

Design Design Safety Limits and Criteria for DBE …Limits and Criteria for DBE …

PFBR Design & Commissioning Review

� Peer design and safety review by national andinternational experts & agencies – e.gNOVATOME and OKBM

� Extensive safety review – ISC, PDSC (115 ISCmeetings, 102 PDSC meetings)

� 18 Specialists’ Groups formed by PDSC fordetailed review of Preliminary Safety analysisReport – Chapter wise – Comprehensiveaddressal of all safety issues in detail

� Project and equipment erection clearance fromAERB and other statutory clearances fromGovernment departments

� Specialists’ Groups to prepare and reviewInitial Test Program, Commissioning Sequence& Procedures and Tech. Specs for Operation

� Safety series 50-C-D Rev D, 1988

� Re-issued as IAEA NS-R-1, 2000

� NS-R-1 mainly applicable for water reactors

� Atomic Energy Regulatory Board Code forming the basic safety criteria for PFBR (DAE-SRC sub-committee in 1986, report in 1987, SARCOP approval 1989, Review by Novatome, issued in April 1990)

� Major provisions of IAEA NS-R-1 also have been incorporated in AERB code

Safety Design Criteria Adopted for PFBR

• Compliance of PFBR design to IAEA standard NS-R-1 2000 reviewed byAERB by a Specialists’ Group with extensive discussions with IGCAR in2006 – As a part of consenting process for the project

Total 61 clauses, 50 complied with, 1 not complied with, 1 not

applicable for PFBR, 9 conditional compliance (for analysis, information)

� Reactor core not in most reactive configuration

� Possibility of either positive or small negative coolant void reactivity

� High burnup & High decay heat content

� Liquid metal sodium coolant which is chemically active & Absence ofmoderator

� Shutdown systems with limited diverse features

� Absence of pressurised coolant system and primary boundary

� Components subjected to very high neutron dose with associated largeradiation damage

� Large size thin shell structures subjected to high temperature

� Additional failure modes due to high temperature effects like creep &fatigue in addition to the conventional structural failure modes

� Scheme of fuel handling such as offline as against online

� Manual reactor control against automatic control due to large reactivityworth availability

� Sodium leaks and fires

� Large thermal capacity in pool type fast reactors

� Presence of large and massive structures above main vessel

� Different types of initiating events with different unfolding safetyscenarios

Major Differences w.r.t Thermal Reactors (not exhaustive)

Section II

Feedback from PFBR Design Safety Review

• Shutdown margin - 5000 pcm for all absorber rods

• Allowable reactivity addition due to SSE owing to relative movement of fueland absorber SA - presently no limit is specified and 0.5 $ is adopted indesign

• Design margin for the case of fuel SA immersed in water (0.95 Keff )

• Diverse Safety Rod drop time – 1 s specified - no provision to measure droptime

• Blanket subassembly temp monitoring – EOL LHR limited - design andeconomic implications

• Design Safety Limits for fuel – Partial melting of fuel

• Subassembly melt propagation – to surrounding SA – detection scenario(from thermocouple of adjacent SA or DND from the affected SA)

• Safety margins in the DBE (e.g category 4 primary pipe rupture eventNo coolant boiling for first SCRAM parameter, local boiling predicted forsecond SCRAM parameter; Review and rationalisation of hot spot factors)

Safety Design Criteria Safety Design Criteria -- Feedback ExperienceFeedback Experience

• Comprehensive safety analysis for DBE : Sub-grouping of events depending on

their extent of severity and Analysis of enveloping events of such sub-groups

• SCRAM Parameters and allowable limits - For every PIE, two scram parameters

envisaged in general (First scram parameter would limit the severity of the

event within the limits corresponding to the event category, 2nd scram

parameter limits within the criteria for the next higher category)

• DBE with single scram parameter

for example: Individual SA cooling faults for which only one scram parameter –

ie temp is available (SA flow blockage) – DND comes after failure

• Exhaustiveness of DBE considered in the design

• Analysis of one PSP seizure event - Seizure time 1.0 s assumed

• Location of decontamination facility in RCB - Use of water inside RCB

• Non-provision of isolation valves on secondary sodium lines penetrating RCBand SGDHR lines – need to have exception to have design flexibility

• Automation of fire fighting system for sodium fires

• Airlock - Both doors open condition not permitted - rare exceptions may beneeded

Safety Design Criteria Safety Design Criteria -- Feedback Experience …Feedback Experience …

� Shutdown system reliability levels

� DHR reliability - both forced circulation and/or natural circulation

� Sodium – Water reaction exclusively not addressed in safety criteria

� Spent fuel handling – provision to detect whether decay heat exceeds the

limit

� Applicable dose criteria for back-up control room in safety criteria

� Under Sodium Viewing and feasibility of fault detection under sodium

• PI (Pump / IHX) flask - criteria for handling- Exclusive criteria needed for

design of such huge component handling and rare operation.

• Site boundary limits for sodium aerosols - Need for specification of limits in

safety criteria

• Control Room habitability - Dose limits during BDBA

• Radiation zoning over reactor assembly roof slab - Rationalised approach for

the design of Complimentary shielding.


• Availability of two independent parameters for SCRAM and Diverse means of

protection function

• Containment isolation and isolation signals – Gamma activity and pressure –

diverse and faster means - Two signals

• Minimum Neutron counts for core monitoring during shutdown and fuel

handling

• Core Thermocouple – 2/2 logic used - Future Safety criteria to reflect the

philosophy ?

• Core flow – not measured directly in core – Measured at the pump location

• Mix-up of codes in design

• Sodium fire protection and sodium–water reaction – Exclusive criteria

• Containment heat removal system – Need

• Efforts for increasing sodium void negative reactivity

• Addition/augmentation of diversity to shutdown systems


Important Issues with Extensive Important Issues with Extensive Discussions with Discussions with Regulatory AgencyRegulatory Agency

• Fuel temperature limits during DBE – Acceptance of partialmelting and its extent

• Blanket subassembly coolant outlet temperature monitoring(periodic flow measurement introduced)

• Core support structure redundancy

• Neutron flux monitoring

• Acceptability of manual mode of reactor operation

• Decay heat removal system – Active Vs Passive, adequacy,diversity etc

• Reliability analysis of shutdown and decay heat removalsystems

• SCRAM parameters

• Core disruptive accident analysis – Energy release

• Containment design pressure

Post Fukushima Safety Margin AssessmentPost Fukushima Safety Margin Assessment

• Primary Containment Capacity Against CDA as a Design Extension Condition

• Seismic Capacity Margin of Reactor Components Beyond SSE

• Seismic Capacity Margin of Safety Related Civil Structures Beyond SSE

• Simultaneous Double Ended Guillotine Rupture of Two Primary Pipes

• Sequential Leakage of Main and Safety Vessels

• Multiple Failures of SGDHR Circuits as a Design Extension Condition

• Effects of Extended SBO

• Reactor Vault and Top Shield Cooling systems

• Fresh and Spent Subassembly Storage Bay

• H2 Production and Mitigation

• Possibility of H2 Generation Due to Water Leak in BSC Circuit

• Possibility of Hydrogen Generation in Steam Generator Building

• Need for Hydrogen Recombiners in Reactor Containment Building

• Issues of Flooding of Power Island

Section III

Safety Design Criteria for Future Indian SFRs

Future SFRs – Improved Safety Features & Directions

• Eight primary pipes – enhanced safety

• In-vessel sodium purification

• Additional passive features in existing shutdown

systems – Temp sensitive magnetic switch, stroke

limiting device

• Third shutdown system – liquid poison based

• Improved decay heat removal capability with higher

capacity of natural circulation

• Advanced ISI-R techniques

• Enhanced post accident heat removal systems

• Core catcher for whole core

• Retaining the existing safety features (e.g. blockage

adaptor in SA)

• Enhanced safety margins through detailed modelling

and numerical and experimental investigations

• Rationalisation of events

• Sharing of non-safety facilities for twin units

• R&D on safety related areas

Need for Comprehensive and Specific Safety Criteria for

Sodium Cooled Fast Reactors

� India has ambitious fast breeder reactor program – FR with closed

fuel cycle program is essential

� Need was felt for new and improved safety criteria based on the

international developments and our own experience from PFBR

during the safety review deliberations.

� Review of IAEA NS-R-1 standard (DS-414 >> SSR 2-1) – India is a

member

� AERB, Indian regulatory agency is planning to issue general safety

criteria applicable for all types of reactors (including LWR imports)

� Compared to PFBR design safety criteria (1990), new criteria aim

for higher level of safety – probably less than Gen-IV but higher or

equal to GEN-III.

Constitution of Task Force for Safety Criteria Drafting

� To assist AERB, a Task force consisting of 12members from various disciplines in IGCARformed in August 2006

� Specific criteria for pool type FBR of 500 MWeas twin units was prepared by task force

� General criteria also included as first fivechapters

� Formatting similar to IAEA NS-R-1

� Experience of PFBR safety review includedmainly as guidelines

Approach followed

� Safety criteria -broadly classified into two categories.

� First category covers the general safety requirements which areapplicable to all types of reactors and generically applicable. Thesecond category covers requirements which are reactor systemspecific.

� General safety principles, defence in depth requirements, safetyfunctions, radiological dose limits to plant personnel & public, siteboundary radiological requirements, emergency preparedness etc.fall in the first category (AERB)

� Nuclear steam supply systems design requirements form thesecond category (IGCAR)

� Due consideration to feedback experience from PFBR

� Provisions to incorporate evolving innovative designs

� International experience

� Revision of IAEA standard

Background Reports referred

� PFBR safety criteria, AERB (1990)

� Safety of Nuclear Power plants: Design Requirements SafetyStandards Series No. NS-R-1, IAEA (2000)

� LMFBR safety criteria and guidelines report EUR-12669 EN(2000)

� Safety of Nuclear Power Plants: Design; Draft SafetyRequirements - DS414, Revision of NS-R-1, IAEA, Draft-14 (2009)

� Draft report “Principles, Goals and Criteria for Safety inDesign of AHWR” (AERB), 2008

� Design of Pressurised Heavy Water Based Nuclear PowerPlants, AERB Safety Code No. AERB/NPP-PHWR/SC/D, DraftReport (AERB), 2009.

� Minutes of meetings of ISC, PDSC & SG and other committeesunder PDSC and response from design and constructionagency

Review Process

� Internal review in IGCAR – Emphasis on FBRspecific criteria (Chapters 6 to 21) – CompletedNov. 2007

� Preliminary review in AERB – Emphasis ongeneral criteria (Chapters 1 to 5) .

� Decision to include guidelines where applicablefor all chapters (given in italics at end of eachchapter)

� Completed in Nov. 2008

� Review by 9 member committee chaired byFormer Chairman-PDSC and Chairman, AERB -2009

� Foot notes added where applicable onadvanced design features which may beconsidered for enhanced safety.

Contents of Report

1. Introduction

2. Safety Objectives and Concepts

3. Requirements for Management of Safety

4. Principal Technical Requirements

5. Requirements of Plant Design

6. Core

7. Reactor Assembly

8. Reactor Shutdown System

9. Heat Transport System

10. Sodium water Reactions

11. Core Component Handling and Storage

12. Plant Layout

13. Electric Power Supply

14. Air Conditioning and Ventilation System

15. In-service Inspection

16. Sodium Fire Protection

17. Instrumentation and Control

18. Plant Protection System

19. Containment

20. Steam Supply System, Auxiliary Systems And Services

21. Radiological Protection

Abbreviations

Appendix: List Of Design Basis Events and their Classifications

Glossary

Safety Criteria clauses are exhaustive:

Only, some selected safety criteria and guidelines in differentchapters are presented …..

Design extension for BDBA

• Most of the BDBA are practically eliminated by design; A set from the remaining

event/event sequences are selected for implementation of practicable mitigation

measures – (e.g. core catcher, parameters of containment design)

• Credible scenario with potential for significant releases imposing highest loads on

containment (at least one such scenario to be considered; e.g. whole core

accident initiated by unprotected loss of flow)

Safety Criteria – Few Aspects (not exhaustive)

Core & Shutdown Systems

• Combined failure frequency of two SDS shall be < 10-7 / reactor year

• Minimum shutdown margin (SDM) taking into account worst combination of

scenario shall be more than 1 $ at cold shutdown state for all PIE

• In case of SSE, the external reactivity inserted due to the relative movement of

fuel and absorber rod sub-assemblies shall be below the prescribed limits, within

the capability of shut down system to avoid crossing Design Safety Limits

• Reactivity inserted due to relative SA movements in SSE should be < 0.5 $.

• In estimating the minimum SDM, the reactivity inserted by any PIE should be

accounted. Also, account of the decay of Np-239 or Pa-233 should be considered

Safety Criteria – Few Aspects …

Reactor Assembly

• The main vessel design should take into account the expected end-of-life

properties affected by erosion, creep, fatigue, the chemical environment, the

radiation environment and ageing.

• Design provisions should be made for the top shield design to minimize the

deposition of sodium aerosols in the narrow gaps for smooth functioning of the

mechanisms.

• Consideration should be given for roof slab and rotatable plugs made of solid

metal which enables machining and consequently reduced gaps.

Heat Transport Systems

• The failure frequency of decay heat removal system shall be < 10-7 / reactor year.

• For ultimate heat sink as air, the design shall take into account the effect of

cyclonic and severe weather conditions

• Consideration may be given for in-vessel purification of sodium

• Consideration may be given for use of oil free bearings and seals in the sodium

pumps.

Plant Layout

• Control room and backup control room shall be habitable during the postulated

whole core accident. The maximum cumulative dose to personnel shall be less

than 50 mSv during the whole event.

• If possibility of differential motion exist between reactor containment building

(RCB) and fuel handling building (FHB) due to SSE, its consequences shall be

considered in design

• RCB, FHB and building housing sodium systems with piping originating from

RCB should be on common raft to minimize the consequences of earth quake

from risk of sodium fire and damages to machines handling subassemblies

between RCB and fuel building.


Probability Safety Analysis (PSA)

• Level-1 PSA - reliability of shutdown system and decay heat removal systems and

consequent core damage frequency

• Level-2 PSA - Probability of large off-site release

• Level-3 PSA - Requirement of emergency measures

Containment isolation

• The containment structure shall be designed and constructed so that it is

possible to perform a pressure test at a specified pressure to demonstrate its

structural integrity before operation of the plant and over the plant’s lifetime.

• At least two adequately diverse parameters shall be monitored (in value or

logic) for initiation of containment isolation signal

• In the event of design basis events (DBE) and severe accidents the

temperature of the containment shall be within allowed limits.


I&C aspects

• Control rod positions for different insertion levels shall be measured and

indicated. The in and out conditions of safety rods shall be measured and

indicated.

• During shutdown state, the core shall be monitored. The minimum count rate in

the neutron detectors arising from core neutrons shall be more than prescribed

limits in the shut down state.

• The minimum count rate in neutron detectors monitoring reactor shut down state

should be greater than 2.0 for neutron multiplication factor greater than 0.9.

Sodium Leaks

• Sodium leaks shall be detected by diverse instrumentation systems.

• Sodium leak detection shall be provided which can reliably detect at an early

stage and then locate sodium leakages to allow initiation of corrective action (eg:

drainage of loops)

• A Na leak rate of the order of 100 gm/h should be detected in 20 hours in air

filled rooms and within 250 hours in nitrogen filled rooms. (Ref- ASME Section

XI, Division -3 )


Steam Supply System and Auxiliary Systems

• The design of the steam supply system provides for appropriately rated and

qualified steam isolation valves capable of closing under the specified

conditions.

• Steam and feed water systems are of adequate capacity and are designed to

prevent anticipated operational occurrences escalating into accident

conditions.

• The turbine generators are provided with appropriate protection such as over

speed and vibration, and measures are taken to minimize the potential effects

of turbine disintegration on items important to safety

• The design basis of any compressed air system that provides a service to an

item important to safety shall specify the quality and cleanliness of the air to be

provided and the design measures to be provided to ensure that the required

reliability is consistent with the associated component or system.


Category Frequency information range DiD level Consequences

Category-1

Normal operation and

operational transients.

Normal operation and all

operational transients

1 L1

Category-2

Events of moderate

frequency

Expected to occur once in life

time of reactor

2 L2

Category-3

Events of low frequency

Expected to occur atleast once

in one generation of reactors

3 L3

Category-4

Multiple failures and rare

events

Accidents which are likely to

occur less frequently than

those in category 3

4 L4

Category-5

Hypothetical events

Not expected to occur but yet

accounted in emergency

preparedness

5 L5

Preferred approach of DBE classification is Defence in Depth based but is frequency informed and considering the radiological

consequences

Categorization of DBEs

Release to public during normal operation is L1 (1 mSv/yr)L1 < L2 < L3 < L4 < L5

Sl no. Event description

1. Failure of core support structure

2. Ejection of all control rods from the core

3. Sodium water reaction in SG + Failure of SG depressurization valve to open

4. Simultaneous failure of both main and safety vessels

5. Total loss of all DHR systems on demand

6. Rupture discs fails to open, on demand, following a large sodium water reaction

7.Sodium – concrete reaction due to leakage from safety vessel after a leak from

main vessel

8. Failure of both RCB ventilation dampers to isolate RCB on demand

9. Dropping of loads (example: IHX, PSP, handling flask)

10.Transient Over Power (TOP) initiated by uncontrolled withdrawal of a CSR from

operating position to fully out followed by failure of shut down system

List of BDBA (DiD Level-5)

Events mentioned below are considered hypothetical due to various design provisions made in terms of redundancy, diversity and sufficient

safety margin in component & systems design. These are practically eliminated. However, emergency control centre and plans of on-site and

off-site emergency response are a must

Major Safety Domains and R&D Issues to be Focussed

• How much margin should be adopted for external events like seismic

and flooding ?

• Identification of design extension conditions and establishing plant

capability to withstand damage and spell out margins

• Fundamental safety studies : core re-criticality

• Phenomena modelling capability – core transients, core meltdown

and PAHR

• Numerical analysis and Safety experiments – seismic margins, gas

entrainment, mechanical consequences of CDA, PAHR, science

behind sodium fires, aerosol behaviour and mitigation, fuel melt and

melt through structures, molten fuel coolant interactions

• Design of passive safety systems for reactor control and decay heat

removal

• Design measures to assure there is no radiological impact on public

for any event both internal & external

• Na void worth – in large size metallic fuelled core (MFTR & MFBR)

• Important feedback has been gained through the design and safety review of PFBR.

• The safety criteria document prepared by AERB & IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level

• A common approach with regard to safety, among countries pursuing fast reactor program, is desirable.

• Sharing knowledge and experimental facilities on collaborative basis

• Evolution of strong safety criteria – fundamental to assure safety

SummarySummary

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