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Safety Design Criteria (SDC) Development for Generation-IV Sodium-cooled Fast

Reactor System

Ryodai Nakai and Tanju Sofu

GIF SDC Task Force

GIF Symposium 2015/ICONE23

Chiba

19 May 2015

GIF Symposium 2015/ICONE23, Chiba, Japan, 19 May 2015

Contents

• Introduction

– Background

– GIF’s Safety Goals & Basis for Safety Approach

• Safety Design Criteria (SDC) for Gen IV SFR

– SDC Phase 1 Report

– Status of International Reviews on the SDC Report

• Safety Design Guidelines (SDGs) Development

– Objectives and Current Status

• Concluding Remarks


Background

• Safety Design Criteria (SDC) development for Generation-IV systems

was proposed at GIF Policy Group meeting in October 2010

– SFR system was selected as the initial application since it

represents one of the more mature next generation nuclear

energy concepts

» Several prototypes being pursued by GIF member states

• Task Force (TF) started work in 2011 and completed SDC in 2013

– Establish reference criteria for safety design of structures,

systems and components

– Achieve harmonization of safety approaches among GIF

member states

» Realization of enhanced safety designs common to Gen-IV

SFRs

» Preparation for upcoming licensing efforts


Hierarchy of Safety Standards

Safety Design Guideline

Domestic Codes & Standards


GIF’s Safety & Reliability Goals SR-1: Excel in Operational Safety and Reliability

Safety and reliability during normal operation, and likely kinds of operational events that set forced outage rate

SR-2: Very low likelihood & degree of reactor core damage

Minimizing frequency of initiating events, and design features for controlling & mitigating any initiating events w/o causing core damage

SR-3: Eliminate the need for offsite emergency response

Safety architecture to manage & mitigate severe plant conditions, for making small the possibility of releases of radiation

• Defence-in-depth

• A combination of deterministic and risk-informed safety

approach

• Safety to be built-in to the design, not added-on

• Emphasis on utilization of inherent and passive safety features

GIF’s Basic Safety Approach


SFR Design Tracks under GIF

JSFR

[Large Loop]

ESFR

[Large Pool]

KALIMER

[Pool]

SMFR

[Small Modular]


Safety Advantages of SFRs • Low pressure primary and intermediate coolant system

– Guard vessel and guard pipes to “maintain” coolant inventory

– No LOCA concern, no ECCS, no risk for control-rod ejection

• Liquid-metal coolant with excellent natural circulation

characteristics and a wide margin (~400 degC) to boiling

• Inherent safety with “net” negative reactivity feedback during

accidents that lead to elevated core/coolant temperatures

• Dedicated systems for decay heat removal to an ultimate heat sink

– Large difference between core outlet and inlet temperatures to

facilitate reliance on passive systems

• Low pressure (~0.5 bar) design pressure for containment (mostly

against heat from sodium fires)

• Much simpler operation and accident management (long grace

period for corrective action)


Challenges with SFRs

• High temperature (>500 degC core outlet temperature) and

high core power density

• Liquid sodium coolant that reacts with air, water and

concrete

– These reactions have to be prevented and/or mitigated

to avoid their effect on SSCs important to safety

• Fast reactor cores are not in their most reactive

configuration

– Relocation of core materials may lead to a hypothetical

core disruptive accident (HCDA)

• For large cores, sodium void worth can be positive

• Opaque sodium coolant could pose in-service inspection

and maintenance challenges


SFR Safety Principals

• Like LWRs, SFR safety is first based on utilization of multiple

redundant engineered protection systems to lower the probability

of accident occurrence and to limit its consequences:

– independent scram systems,

– multiple coolant pumps and heat transport loops, and

– multiple barriers to release of radioactive materials.

• SFR safety analyses traditionally focus on ATWS during which the

reactor scram system is assumed to fail.

– Because HCDAs could potentially result in re-criticalities.

– The safety design features that enhance inherent negative

reactivity feedback and passive decay heat removal

capabilities provide additional measures to prevent/mitigate

HCDAs even during these very low probability accidents.


Defence-in-depth (DiD) & Plant States

based on IAEA INSAG-12 & SSR-2/1


Basic Scheme to outline the SDC High level safety fundamentals, and safety design goals

GIF’s Goals for safety & reliability

Basis for safety approach for design & assessment

Requirements in SFR System Research Plan

1) Particular issues for SFR 2) Reference of SDC Structure

3) Lessons learned from Fukushima Dai-ichi NPPs accident

IAEA SSR 2/1 Management of safety in design

Principal technical requirement

General Plant design

Design of specific plant system

Characteristic of Sodium-cooled Fast Reactor Reactivity (void) … Sodium fire & Sodium-water reaction…

Consideration on Severe Accident Re-criticality during Core Disruptive Accident

High Temperature & Low pressure system Creep property, Leak-Before-Break… No LOCA and no need of ECCS…

Enhanced Safety Approach Passive system for shutdown & cooling

Common cause failure by external event Loss of power for longer period

Decay heat removal, Fuel pool cooling Containment function on spent fuel in the pool Preparing multiple AMs, e.t.c.

GIF SFR SDC

Table-Of-Contents of “SDC Phase 1 Report”*

1. INTRODUCTION

1.1 Background and Objectives

1.2 Principles of the SDC formulation

2. SAFETY APPROACH TO THE SFR

AS A GENERATION-IV REACTOR SYSTEM

2.1 GIF Safety Goals and Basic Safety Approach

2.2 Fundamental Orientations on Safety

2.3 Safety approach of the Generation-IV SFR systems

3. MANAGEMENT OF SAFETY IN DESIGN Criteria 1-3

4. PRINCIPAL TECHNICAL CRITERIA Criteria 4-12

5. GENERAL PLANT DESIGN

5.1 Design Basis Criteria 13-28

5.2 Design for Safe Operation over the Lifetime of the Plant Cri.29-31

5.3 Human Factors Criterion 32

5.4 Other Design Considerations Criteria 33-41

5.5 Safety Analysis Criterion 42

*SDC-TF/2013/01, May 1, 2013

6. DESIGN OF SPECIFIC PLANT SYSTEMS

6.1 Overall Plant System Criterion 42bis

6.2 Reactor Core and Associated Features Criteria 43-46

6.3 Reactor Coolant Systems Criteria 47-53

6.4 Containment Structure and Containment System Cri.54-58

6.5 Instrumentation and Control Systems Criteria 59-67

6.6 Emergency Power Supply Criterion 68

6.7 Supporting Systems and Auxiliary Systems Cri.69-76bis

6.8 Other Power Conversion Systems Criterion 77

6.9 Treatment of Radioactive Effluents and Radioactive Waste Cri.78-79

6.10 Fuel Handling and Storage Systems Criterion 80

6.11 Radiation Protection Criteria 81-82

GLOSSARY

APPENDIX:

(A) Definitions of Boundaries of SFR systems

(B) Guide to Design Extension Conditions

(C) Guide to Practical Elimination of accident situations

(D) Guide to Utilisation of Passive/Inherent Features

(E) Approach to Extreme External Events


• GIF SFR “SDC Phase 1 Report”

– Review requests for the SDC Report

» For “Review by external organizations” and

» For “Enhancing interaction with regulatory bodies”

– Sent the report (ca. July 2013) to

» International organizations

IAEA, MDEP, OECD/NEA/CNRA

» Regulatory authorities at national level

China (NNSA), Euratom (ENSREG), France (ASN),

Japan (NRA), Republic of Korea (NSSC),

Russia (Rostechnadzor), USA (NRC)

Status of International reviews on SDC


• NRC (USA)

– Comprehensive & detailed review, with proposals (Jan. 2014)

– GIF SDC Task Force prepared the resolutions to incorporate

• NNSA (China)

– Review results (Oct. 2013 & Jan. 2014)

– GIF SDC TF resolution replied (Aug. 2014)

• IRSN (France)

– Comments on interim version at the GIF-IAEA Safety Workshop

(Feb. 2012), resolutions already included in Phase I report.

• IAEA

– General and technically specific reviews (April 2014)

– GIF SDC Task Force prepared the resolutions to incorporate

Status of international reviews on “SDC”

External feedbacks have been or are being incorporated


Safety Design Guidelines (SDG) Development

• Main objective

– to support practical application of SDC in design process

for improving safety in specific topical areas

» including use of inherent/passive safety features

» design measures for prevention and mitigation of

severe accidents.

– Initial topical areas are considered:

» Particular importance since a fast reactor core is

typically not in its most reactive configuration

» Quantification of key criteria for safety improvement


Schematic View of SDG Development Schedule

2013 2014 2015 2016

Terms of Reference

Table of Contents

Identification of

discussion points

SDGs on Key

Structures,

Systems and

Components

SDG on

Safety Approach

Identification of

discussion points

Table of Contents Guidelines for Reactor Core Discussion points (e.g. fuel performance in DBA, DEC,

Passive or inherent reactivity features)

Guidelines for Reactor Coolant System Discussion points (e.g. sodium chemical reactions, passive

or alternative cooling features)

Functional Requirements

on SSC

Set of design conditions (e.g. postulated events, design

parameters & constraints…)

Guidelines for Containment Vessel Discussion points (e.g. severe accident conditions and

measures, Accident management)

Draft Report

Loss of heat removal

issue Accident conditions to be

practically eliminated

Reactivity issue Prevention & Mitigation of

severe accidents Final Report

Draft Report Final Report


General Approach to Normal Operation, AOOs, and DBAs

General Approach to

Design Extension Conditions

• Normal Operation- Stable operation, with

controlling reactivity, temperature, flow…

• AOOs/DBAs- Shutdown the reactor and

maintain decay heat removal sufficient to keep

reactor core and system temperatures within

the applicable design limits.

• Prevention of Core Damage

Accident sequences typically caused by

failure of one or more systems

related to safety

Postulated initiating events more severe than those in DBA

• Mitigation of Core Damage

Mitigation of consequences of postulated accidents where significant core

damage may occur, with the objective of maintaining the containment

function to limit radioactive releases.

DiD

Level 1: Normal Operation

Level 2: AOO

Level 3: DBA

Level 4: DEC

Prevention of

Core Damage

Mitigation of Core

Damage

Level 5: Offsite Emergency Response


Exploiting SFR Characteristics to Enhance Safety

• Passive/Inherent safety for DEC

– On reactivity

» Inherent reactivity feedback to reduce the power as

core temperatures rise or

» Passive mechanism are applicable for shutdown

systems, such as SASS, HSR, and GEM

– On decay heat removal

» Natural circulation of single phase sodium coolant

» can be placed in different locations for enhancing

diversity


Exploiting SFR Characteristics to Enhance Safety

• In-Vessel Retention

– In the course of core degradation during unprotected

transients, measures should be provided to prevent

prompt criticality

– Reactor coolant boundary should maintain the

boundary function against pressure load including fuel-

coolant interaction

– Measures should be provided for ensuring long term

cooling of core materials inside the reactor vessel

under sub-critical condition


Practical Elimination of Accident Situations:

• Severe accidents with mechanical energy release higher

than the containment capability

– Power excursions for intact core situations

» Large gas flow through the core

» Large-scale core compaction

» Collapse of the core support structures

• Situations leading to the failure of the containment with

risk of fuel damage

– Complete loss of decay heat removal function that leads to

core damage and failure of primary coolant boundary

– Core uncovering due to sodium inventory loss

• Fuel degradation in fuel storage or during when the

containment may not be functional due to maintenance

– Core damage during maintenance

– Spent fuel melting in the storage


Design requirements on reactivity characteristics

Normal Operation, AOO, DBA, DEC w/o Core Damage

» Shall require inherent reactor power stability

» Reactor Shutdown System shall prevent sodium

boiling and maintain core coolable geometry

Design Extension Condition with Core Damage

» Shall prevent excessive insertion of reactivity by

coolant boiling, cladding and fuel relocation after

core damage


Quantification of requirement

on reactivity characteristics

• For Normal operation, AOO and DBA

» Power reactivity coefficient < 0 (Negative)

» Reactor shutdown capability with inherent feedback

> Postulated reactivity insertion

• For Design Extension Condition

» Before core damage: same as the requirement for DBA,

– Achieved by passive measures or inherent features

» After core damage:

– Total reactor core reactivity < 1$ (below prompt criticality)

» Sodium void worth can be positive as far as the above

conditions are satisfied.


Concluding Remarks • The “Safety Design Criteria Phase 1 Report”

– Issued by the GIF on May 2013

– Disseminated for international review to:

» International organizations

» National Regulatory Bodies

– Important feedbacks have been or are being incorporated:

» e.g. IAEA, IRSN, USNRC, NNSA …

• The “Safety Design Guidelines” development in Phase II

– Started from Sept. 2013

– Two Safety Design Guidelines (SDG):

» Safety Approach and Design Conditions SDG in final

drafting stage

» Key Structures, Systems and Components SDG


Thank you

for your attention !!

Status* Status* IAEA SSR 2/1 GIF SFR SDC Status* Status*Requirement # paragraph # Criterion # paragraph # M/A/D/U Requirement # paragraph # Criterion # paragraph # M/A/D/U Requirement # paragraph # Criterion # paragraph # M/A/D/U Requirement # paragraph # Criterion # paragraph # M/A/D/U

20 20 M 6. DESIGN OF SPECIFIC PLANT SYSTEMS INSTRUMENTATION AND CONTROL SYSTEMS

1 1 U 5.27 5.27 M OVERALL PLANT SYSTEM 60 60 U

3.1 3.1 M 5.28 5.28 U 42bis A 61 61 U

2 2 U 5.29 5.29 M REACTOR CORE AND ASSOCIATED FEATURES 6.32 6.32 M

3.2 3.2 U 5.30 5.30 U 43 43 M 6.33 6.33 U

3.3 3.3 M 5.31 5.31 M 6.1 6.1 M 62 62 U

3.4 3.4 M 5.32 5.32 M 6.2-6.3 6.2-6.3 U 6.34-6.36 6.34-6.36 U

3 3 U 21 21 U 44 44 M 63 63 U

3.5-3.6 3.5-3.6 U 5.33 5.33 U 6.3bis A 6.37 6.37 M

22 22 U 6.3ter A 64 64 U

4 4 M 5.34 5.34 M 6.3quater A 6.38 6.38 U

4.1-4.2 4.1-4.2 U 5.35-5.36 5.35-5.36 U 45 45 U 65 65 U

5 5 U 23 23 U 6.4 6.4 M 6.39-6.40 6.39-6.40 U

4.3-4.4 4.3-4.4 U 5.37-5.38 5.37-5.38 U 6.5 6.5 M 66 66 M

6 6 U 24 24 U 6.6 6.6 M 6.41 6.41 U

4.5-4.8 4.5-4.8 U 25 25 U 6.6bis A 67 67 U

7 7 M 5.39-5.40 5.39-5.40 U 46 46 M 6.42 6.42 U

4.9 4.9 M 26 26 U 6.7-6.8 6.7-6.8 U EMERGENCY POWER SUPPLY

4.10 4.10 U 5.41 5.41 U 6.9 6.9 M 68 68 U

4.11 4.11 M 27 27 U 6.10-6.12 6.10-6.12 U 6.43 6.43 M

4.12-4.13 4.12-4.13 U 5.42-5.43 5.42-5.43 U REACTOR COOLANT SYSTEMS 6.44 6.44 M

8 8 U 28 28 U 47 47 U 6.45 6.45 U

9 9 U 5.44 5.44 M 6.13 6.13 M SUPPORTING SYSTEMS AND AUXILIARY SYSTEMS

4.14-4.16 4.14-4.16 U DESIGN FOR SAFE OPERATION OVER THE LIFETIME OF THE PLANT 6.14 6.14 M 69 69 U

10 10 U 29 29 U 6.14bis A 70 70 U

4.17-4.18 4.17-4.18 U 5.45-5.47 5.45-5.47 U 6.14ter A 6.46 6.46 U

11 11 U 30 30 U 6.15 6.15 M 71 71 U

4.19 4.19 U 5.48-5.50 5.48-5.50 U 6.15bis A 6.47 6.47 U

12 12 M 31 31 M 6.15ter A 72 72 M

4.20 4.20 M 5.51-5.52 5.51-5.52 U 6.16 6.16 M 73 73 U

13 13 U HUMAN FACTORS 6.16bis A 6.48 6.48 M

5.1 5.1 M 32 32 U 6.16ter A 6.49 6.49 U

5.2 5.2 U 5.53-5.62 5.53-5.62 U 6.16quater A 74 74 M

5. GENERAL PLANT DESIGN OTHER DESIGN CONSIDERATIONS 6.16quinquies A 6.50 6.50 M

DESIGN BASIS 33 33 U 48 48 M 6.51-6.54 6.51-6.54 U

14 14 U 5.63 5.63 M 49 49 M 6.54bis A

5.3 5.3 U 34 34 M 50 50 M 6.54ter A

15 15 U 35 35 M 6.17 6.17 M 75 75 U

5.4 5.4 U 36 36 U 6.17bis A 76 76 U

16 16 U 5.64-5.65 5.64-5.65 U 51 51 M 6.55 6.55 U

5.5 5.5 M 37 37 U 52 D [incl. in #51] 76bis A

5.6-5.9 5.6-5.9 U 5.66-5.67 5.66-5.67 U 6.18 6.18 M OTHER POWER CONVERSION SYSTEMS

5.10 5.10 M 6.19 6.19 M 77 77 M

5.11-5.15 5.11-5.15 U 38 38 U 6.19bis A 6.56 6.56 M

17 17 U 5.68 5.68 U 53 53 M 6.57 6.57 M

5.16 5.16 M 39 39 U CONTAINMENT STRUCTURE AND CONTAINMENT SYSTEM 6.58 6.58 U

5.17 5.17 M 40 40 U 54 54 U TREATMENT OF RADIOACTIVE EFFLUENTS AND RADIOACTIVE WASTE

5.18 5.15bis M 5.69-5.70 5.69-5.70 U 55 55 U 78 78 M

5.18 A 41 41 U 6.20 6.20 M 6.59-6.60 6.59-6.60 U

5.19 5.19 M SAFETY ANALYSIS 6.21 6.21 M 79 79 M

5.20 5.20 M 42 42 U 56 56 M 6.61-6.63 6.61-6.63 U

5.21 5.21 M 5.71-5.74 5.71-5.74 U 6.22 6.22 M FUEL HANDLING AND STORAGE SYSTEMS

5.22 5.22 U 5.75 5.75 M 6.23 6.23 M

18 18 M 5.76 5.76 U 6.24 6.24 M 80 80 U

5.23 5.23 M 57 57 U 6.64-6.65 6.64-6.65 U

19 19 U 6.25-6.26 6.25-6.26 U 6.66 6.66 M

5.24-5.25 5.24-5.25 U 58 58 U 6.67 6.67 M

5.26 5.26 M 6.27 6.27 M 6.68 6.68 M

6.28 6.28 M 6.68bis A

*M: Modified A: Added D: Deleted U: Unchanged 6.29 6.29 M RADIATION PROTECTION

6.30 D 81 81 U

59 59 U 6.69 6.69 M

6.31 6.31 U 6.70-6.76 6.70-6.76 U

6.31bis A 82 82 M

6.77-6.84 6.77-6.84 U

3. MANAGEMENT OF SAFETY IN DESIGN

4. PRINCIPAL TECHNICAL CRITERIA

IAEA SSR 2/1 GIF SFR SDC IAEA SSR 2/1 GIF SFR SDC IAEA SSR 2/1 GIF SFR SDC

*M: Modified A: Added D: Deleted U: Unchanged

SDC Criteria (total 83): Modified 20, Added 2, Deleted 1, Un-changed 60

[Added: Overall Plant System & Sodium heating systems / Deleted: ECCS]

Difference between “GIF SDC Criteria” and “IAEA SSR 2/1 Requirements”

Status* Status* IAEA SSR 2/1 GIF SFR SDC Status* Status*Requirement # paragraph # Criterion # paragraph # M/A/D/U Requirement # paragraph # Criterion # paragraph # M/A/D/U Requirement # paragraph # Criterion # paragraph # M/A/D/U Requirement # paragraph # Criterion # paragraph # M/A/D/U

20 20 M 6. DESIGN OF SPECIFIC PLANT SYSTEMS INSTRUMENTATION AND CONTROL SYSTEMS

1 1 U 5.27 5.27 M OVERALL PLANT SYSTEM 60 60 U

3.1 3.1 M 5.28 5.28 U 42bis A 61 61 U

2 2 U 5.29 5.29 M REACTOR CORE AND ASSOCIATED FEATURES 6.32 6.32 M

3.2 3.2 U 5.30 5.30 U 43 43 M 6.33 6.33 U

3.3 3.3 M 5.31 5.31 M 6.1 6.1 M 62 62 U

3.4 3.4 M 5.32 5.32 M 6.2-6.3 6.2-6.3 U 6.34-6.36 6.34-6.36 U

3 3 U 21 21 U 44 44 M 63 63 U

3.5-3.6 3.5-3.6 U 5.33 5.33 U 6.3bis A 6.37 6.37 M

22 22 U 6.3ter A 64 64 U

4 4 M 5.34 5.34 M 6.3quater A 6.38 6.38 U

4.1-4.2 4.1-4.2 U 5.35-5.36 5.35-5.36 U 45 45 U 65 65 U

5 5 U 23 23 U 6.4 6.4 M 6.39-6.40 6.39-6.40 U

4.3-4.4 4.3-4.4 U 5.37-5.38 5.37-5.38 U 6.5 6.5 M 66 66 M

6 6 U 24 24 U 6.6 6.6 M 6.41 6.41 U

4.5-4.8 4.5-4.8 U 25 25 U 6.6bis A 67 67 U

7 7 M 5.39-5.40 5.39-5.40 U 46 46 M 6.42 6.42 U

4.9 4.9 M 26 26 U 6.7-6.8 6.7-6.8 U EMERGENCY POWER SUPPLY

4.10 4.10 U 5.41 5.41 U 6.9 6.9 M 68 68 U

4.11 4.11 M 27 27 U 6.10-6.12 6.10-6.12 U 6.43 6.43 M

4.12-4.13 4.12-4.13 U 5.42-5.43 5.42-5.43 U REACTOR COOLANT SYSTEMS 6.44 6.44 M

8 8 U 28 28 U 47 47 U 6.45 6.45 U

9 9 U 5.44 5.44 M 6.13 6.13 M SUPPORTING SYSTEMS AND AUXILIARY SYSTEMS

4.14-4.16 4.14-4.16 U DESIGN FOR SAFE OPERATION OVER THE LIFETIME OF THE PLANT 6.14 6.14 M 69 69 U

10 10 U 29 29 U 6.14bis A 70 70 U

4.17-4.18 4.17-4.18 U 5.45-5.47 5.45-5.47 U 6.14ter A 6.46 6.46 U

11 11 U 30 30 U 6.15 6.15 M 71 71 U

4.19 4.19 U 5.48-5.50 5.48-5.50 U 6.15bis A 6.47 6.47 U

12 12 M 31 31 M 6.15ter A 72 72 M

4.20 4.20 M 5.51-5.52 5.51-5.52 U 6.16 6.16 M 73 73 U

13 13 U HUMAN FACTORS 6.16bis A 6.48 6.48 M

5.1 5.1 M 32 32 U 6.16ter A 6.49 6.49 U

5.2 5.2 U 5.53-5.62 5.53-5.62 U 6.16quater A 74 74 M

5. GENERAL PLANT DESIGN OTHER DESIGN CONSIDERATIONS 6.16quinquies A 6.50 6.50 M

DESIGN BASIS 33 33 U 48 48 M 6.51-6.54 6.51-6.54 U

14 14 U 5.63 5.63 M 49 49 M 6.54bis A

5.3 5.3 U 34 34 M 50 50 M 6.54ter A

15 15 U 35 35 M 6.17 6.17 M 75 75 U

5.4 5.4 U 36 36 U 6.17bis A 76 76 U

16 16 U 5.64-5.65 5.64-5.65 U 51 51 M 6.55 6.55 U

5.5 5.5 M 37 37 U 52 D [incl. in #51] 76bis A

5.6-5.9 5.6-5.9 U 5.66-5.67 5.66-5.67 U 6.18 6.18 M OTHER POWER CONVERSION SYSTEMS

5.10 5.10 M 6.19 6.19 M 77 77 M

5.11-5.15 5.11-5.15 U 38 38 U 6.19bis A 6.56 6.56 M

17 17 U 5.68 5.68 U 53 53 M 6.57 6.57 M

5.16 5.16 M 39 39 U CONTAINMENT STRUCTURE AND CONTAINMENT SYSTEM 6.58 6.58 U

5.17 5.17 M 40 40 U 54 54 U TREATMENT OF RADIOACTIVE EFFLUENTS AND RADIOACTIVE WASTE

5.18 5.15bis M 5.69-5.70 5.69-5.70 U 55 55 U 78 78 M

5.18 A 41 41 U 6.20 6.20 M 6.59-6.60 6.59-6.60 U

5.19 5.19 M SAFETY ANALYSIS 6.21 6.21 M 79 79 M

5.20 5.20 M 42 42 U 56 56 M 6.61-6.63 6.61-6.63 U

5.21 5.21 M 5.71-5.74 5.71-5.74 U 6.22 6.22 M FUEL HANDLING AND STORAGE SYSTEMS

5.22 5.22 U 5.75 5.75 M 6.23 6.23 M

18 18 M 5.76 5.76 U 6.24 6.24 M 80 80 U

5.23 5.23 M 57 57 U 6.64-6.65 6.64-6.65 U

19 19 U 6.25-6.26 6.25-6.26 U 6.66 6.66 M

5.24-5.25 5.24-5.25 U 58 58 U 6.67 6.67 M

5.26 5.26 M 6.27 6.27 M 6.68 6.68 M

6.28 6.28 M 6.68bis A

*M: Modified A: Added D: Deleted U: Unchanged 6.29 6.29 M RADIATION PROTECTION

6.30 D 81 81 U

59 59 U 6.69 6.69 M

6.31 6.31 U 6.70-6.76 6.70-6.76 U

6.31bis A 82 82 M

6.77-6.84 6.77-6.84 U

3. MANAGEMENT OF SAFETY IN DESIGN

4. PRINCIPAL TECHNICAL CRITERIA

IAEA SSR 2/1 GIF SFR SDC IAEA SSR 2/1 GIF SFR SDC IAEA SSR 2/1 GIF SFR SDC

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