Table Of ContentNUREG-1368
I I I IIIII IIII III I
Preapplication Safety Evaluation
Report for the Power Reactor
Innovative Small Module (PRISM)
Liquid-Metal Reactor
Final Report
III|111 '1 " I IIIIII I III
Manuscript Completed: January 1994
Date Published: February 1994
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
2°,
DISTRIBUTION OF THI6 DOCUMENT IS UNLIMITED
f
ABSTRACT
This preapplication safety evaluation report (PSER) Accordingly, in thereviewofthePRISM design, the staff
presents the results of the preapplication design review for proposes touse andbuildon applicable existing regulations
thepowerreactor innovative small module(PRISM)liquid- and guidelines for safety developed for application to
metal (sodium)-cooled reactor, Nuclear Regulatory LWRs, to develop additional criteria when necessary to
Commission (NRC) Project No. 674. The PRISM address the unique characteristics of these designs, and to
conceptual design was submitted by the U.S. Department require that they be assessed for enhanced safety.
of Energy (DOE) in accordance with the NRC's Additionally, thestaff created further criteria following the
"Statement of Policy for the Regulation of Advanced guidance provided by the Commission in the Staff
Nuclear Power Plants" (51 Federal Re_ister 24643). This Requirements Memorandum dated July 30, 1993, that dealt
policy provides for the early Commission review and with key policy issues for the advanced reactors. In the
interaction with designers and licensees. The PRISM application of the existing regulations and guidelines, the
reactor design proposed by DOE is for a small, modular, staff, in some cases, has had to interpret the guidance
pool-type, liquid-metal (sodium)-cooled reactor. The developed for LWRs for application to the PRISM
standard plant design for the PRISM consists of three concept and for issues under review. In making such
identical power blocks with atotal electrical outputrating interpretations, the staff has directed its approach toward
of 1395 MWe. Each power block comprises three reactor maintaining limits and criteria at least equivalent to those
modules, each with an individual thermal rating of 471 of LWRs for quality design, construction, and operation,
MWt. Each module islocated in its own below-grade silo and for the release of radiation, maintaining defense in
and is connected to its own intermediate heat transport depth, providing for conservatisms to account for plant-
system and steam generator system. The reactors utilize specific uncertainties in the designs, and maintaining
a metallic-type fuel, a ternary alloy of U-Pu-Zr. The consistency with the guidance underdevelopment of future
design includes passive reactorshutdown and passive decay LWRs for the treatment of severe accidents.
heat removal features.
The PSER isthe NRC staff's preliminary evaluation of the
The review approach and criteria used by the staff are safety features in the PRISM design, including the
directed toward meeting the guidance in the Commission's projected research and development programs required to
Advanced Reactor Policy which states that advanced support the design and the proposed testing needs.
reactors must, as a minimum, provide at least the same Because the NRC staff review waS based on a conceptual
degree of protection for the public and the environment design, the PSER did not result in an approval of the
that is required for current-generation LWRs. The staff design. Instead it identified certain key safety issues,
has interpreted current-generation LWRs to be those provided some guidance on applicable licensing criteria,
evolutionary designs currently under review as standard assessed the adequacy of the preapplicant's research and
plantdesigns, such as the advanced boiling water reactor, development programs, and concluded that no obvious
Further, the policy states that the Commission expects impediments to licensing the PRISM design had been
advanced designs to provide enhanced margins of safety, identified.
iii NUREG-1368
CONTENTS
Page
ABSTRACT ................................................... iii
EXECUTIVE SUMMARY ........................................ xxiii
PREFACE .................................................. xxvii
ACKNOWLEDGEMENTS ........................................ xxvii
ABBREVIATIONS .............................................. xxix
I INTRODUCTION AND SUMMARY ................................. I-I
I. I Introduction .............................................. I-I
1.2 PRISM Approach and Objectives ................................. I-I
1.3 General Plant Description ...................................... 1-2
1.4 Comparison With Other Liquid-Metal Reactors ......................... 1-6
1.5 Program Status Overview and Research and Development (R&D) ............ 1-14
1.6 Scope of the Review ........................................ 1-17
1.7 Review Approach and Criteria ................................... 1-18
I
1.8 References .............................................. 1-19
2 SITE LOCATION AND DESCRIPTION ............................... 2-1
2.1 Site Characteristics .......................................... 2-1
2.2 Siting Parameters ........................................... 2-1
2.3 Conclusions .............................................. 2-1
2.4 References ............................................... 2-1
3 REVIEW APPROACH AND CRITERIA ............................... 3-1
3.1 Review Criteria ............................................ 3-1
3.1.1 General Approach and Criteria .............................. 3-2
3.1.2 Proposed Licensing Criteria From SECY-93-092 (Ref. 3.1) ............ 3-3
3.2 Conformance With General Design Criteria .......................... 3-13
3.2.1 Review Criteria ...................................... 3-14
3.2.2 Evaluation .......................................... 3-15
3.2.3 Comparison to the GDC in 10 CFR Part 50 ..................... 3-26
3.2.4 Additional GDC Proposed for the PRISM Design Not in 10 CFR Part 50 . . . 3-55
3.2.5 10 CFR Part 52 ...................................... 3-58
3.2.6 Commission's Advanced Reactor Policy Statement ................. 3-58
3.2.7 Review of Clinch River Breeder Reactor Plant (CRBRP) ............. 3-58
3.2.8 Industry Standard ANSI/ANS-54.1 .......................... 3-59
3.2.9 Advanced Reactor Design Policy Issues ........................ 3-59
3.2.10 Conclusions ......................................... 3-59
3.3 Seismic Design ........................................... 3-59
3.3.1 Seismic Input, Plant System Analysis, and System Seismic Analysis ...... 3-59
3.3.2 Seismic Instrumentation ................................. 3-60
3.3.3 Seismic Base Isolation System .............................. 3-62
v NUREG-1368
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3.4 Seismic Category I Structures .................................. 3-71
3.4.1 Design Description and Safety Objectives ....................... 3-71
3.4.2 Scope of Review ...................................... 3-71
3.4.3 Design Criteria ....................................... 3-72
3.4.4 Research and Development ............................... 3-72
3.4.5 Safety Issues ........................................ 3-72
3.4.6 Evaluation .......................................... 3-72
3.4.7 Conclusions ......................................... 3-73
3.5 Mechanical Systems and Components ............................. 3-73
3.5.1 Design Description and Safety Objectives.. ..................... 3-73
3.5.2 Scope of Review ....................................... 3-73
3.5.3 Design Criteria ....................................... 3-73
3.5.4 Research and Development ............................... 3-74
3.5.5 Safety Issues ........................................ 3-74
3.5.6 Evaluation .......................................... 3-74
3.5.7 Conclusions ......................................... 3-75
3.6 References .............................................. 3-75
4 REACTOR SYSTEM ........................................... 4-1
4.1 Summary Description ........................................ 4-1
4.2 Fuel System .............................................. 4-1
4.2.1 Design Description and Safety Objectives ........................ 4-I
4.2.2 Scope of Review ....................................... 4-2
4.2.3 Design Criteria ........................................ 4-2
4.2.4 Research and Development ................................ 4-3
4.2.5 Safety Issues ........................................ 4-11
4.2.6 Evaluation .......................................... 4-12
4.2.7 Conclusions ......................................... 4-13
4.3 Nuclear Design ........................................... 4-15
4.3.1 Design Description and Safety Objectives ....................... 4-15
4.3.2 Scope of Review ...................................... 4-16
4.3.3 Review Criteria ...................................... 4-16
4.3.4 Research and Development ............................... 4-16
4.3.5 Safety Issues- Identification and Evaluation ..................... 4-17
4.3.6 Conclusions ......................................... 4-18
4.4 Thermal and Hydraulic Design .................................. 4-19
4.4.1 Design Description and Safety Objectives ....................... 4-19
4.4.2 Scope of Review ...................................... 4-20
4.4.3 Review Criteria ...................................... 4-20
4.4.4 Safety Issues ........................................ 4-20
4.4.5 Evaluation .......................................... 4-20
4.4.6 Research and Development ............................... 4-22
4.4._ Conclusions ......................................... 4-22
4.5 Active Reactivity Control and Shutdown System ....................... 4-23
4.5.1 Design Description and Safety Objectives ....................... 4-23
4.5.2 Scope of Review ...................................... 4-23
4.5.3 Review Criteria ...................................... 4-23
NUREG-1368 vi
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4.5.4 Research and Development ............................... 4-23
4.5.5 Safety Issues ......................................... 4-28
4.5.6 Conclusions', ......................................... 4-28
4.6 Passive Safety System Design .................................. 4-29
4.6.1 Design Description and Safety Objectives ....................... 4-29
4.6.2 Scope of Review ...................................... 4-30
4.6.3 Review Criteria ...................................... 4-30
4.'6.4 Research and Development ............................... 4-30
4.6.5 Safety Issues and Evaluation ............................... 4-30
4.6.6 Conclusions ......................................... 4-32
4.7 References .............................................. 4-32
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS .............. 5-1
5.1 Summary Description ........................................ 5-1
5.2 Reactor Vessel and Closure Head ................................. 5-6
5.2.1 Design Description and Safety Objectives ........................ 5-6
5.2.2 Scope of Review ....................................... 5-6
5.2.3 Review Criteria ....................................... 5-7
5.2.4 Safety Issues ......................................... 5-7
5.2.5 Evaluation ........................................... 5-8
5.2.6 Research and Development ............................... 5-10
5.2.7 Conclusions ......................................... 5-10
5.3 Reactor Internal Structures .................................... 5-11
5.3.1 Design Description and Safety Objectives ....................... 5-11
5.3.2 Scope of Review ...................................... 5-15
5.3.3 Review Criteria ...................................... 5-15
5.3.4 Safety Issues ........................................ 5-15
5.3.5 Evaluation .......................................... 5-16
5.3.6 Research and Development ............................... 5-18
5.3.7 Conclusions ......................................... 5-18
5.4 Primary Heat Transport System ................................. Y-19
5.4.1 Design Description and Safety Objectives ....................... 5-19
5.4.2 Scope of Review ...................................... 5-21
5.4.3 Review Criteria ...................................... 5-21
5.4.4 Safety Issues ........................................ 5-21
5.4.5 Evaluation .......................................... 5-24
5.4.6 Research and Development ............................... 5-25
5.4.7 Conclusions ......................................... 5-26
5.5 Intermediate Heat Transport System .............................. 5-26
5.5.1 Design Description and Safety Objectives ....................... 5-26
5.5.2 Scope of Review ..................................... 5-28
5.5.3 Design Criteria ....................................... 5-28
5.5.4 Safety Issues ........................................ 5-28
5.5.5 Evaluation .......................................... 5-28
5.5.6 Research and Development ............................... 5-29
5.5.7 Conclusions ......................................... 5-29
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5.6 Steam Generator System ..................................... 5-29
5.6.1 Design Description and Safety Objectives ....................... 5-29
5.6.2 Scope of Review ...................................... 5-32
5.6.3 Design Criteria ....................................... 5-34
5.6.4 Research and Development ............................... 5-34
5.6.5 Safety Issues ........................................ 5-34
5.6.6 Evaluation .......................................... 5-34
5.6.7 Conclusions ......................................... 5-37
5.7 Residual Heat Removal Systems ................................. 5-37
5.7.1 Design Description and Safety Objectives ....................... 5-37
5.7.2 Scope of Review ...................................... 5-39
5.7.3 Review Criteria ...................................... 5-39
5.7.4 Discussion of Safety Issues ............................... 5-40
5.7.5 Research and Development ............................... 5-42
5.7.6 Conclusions ......................................... 5-42
5.8 References .............................................. 5-42
6 ENGINEERED SAFETY FEATURES ................................ 6-1
6.1 Design Description and Safety Objectives ............................ 6-1
6.1.1 Containment System ..................................... 6-1
6.1.2 Containment During Maintenance ............................ 6-3
6.1.3 Source Term and Site Suitability Analysis ....................... 6-3
6.1.4 Containment Performance ................................. 6-5
6.1.5 Guard Pipes .......................................... 6-6
6.1.6 Leak Testing and Inspections ............................... 6-6
6.2 Scope of the Review ......................................... 6-6
6.3 Design Criteria ............................................ 6-7
6.4 Research and Development ..................................... 6-7
6.5 Safety Issues .............................................. 6-8
6.5.1 Response to Challenges ................................... 6-8
6.5.2 Containment Atmosphere and Combustible Gas Control ............... 6-8
6.5.3 Containment During Maintenance ............................ 6-9
6.5.4 Material Characteristics .................................. 6-9
6.5.5 Source Term ......................................... 6-9
6.5.6 Containment Isolation .................................... 6-9
6.5.7 Containment Penetrations ................................. 6-9
6.6 Evaluation .............................................. 6-10
6.7 Conclusion ............................................... 6-14
6.8 References .............................................. 6-14
7 INSTRUMENTATION AND CONTROL .............................. 7-1
7.1 Reactor Protection .......................................... 7-1
7.1.1 Design Description and Safety Objectives ........................ 7-1
7.1.2 Scope of Review ....................................... 7-3
7.1.3 Design Criteria ........................................ 7-3
7.1.4 Research and Development (R&D) ............................ 7-5
7.1.5 Safety Issues ......................................... 7-5
NUREG-1368 viii
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7.1.6 Evaluatio.n.......................................... 7-7
7.1.7 Conclusions.......................................... 7-7
7.2Safety-RelatIendstrumentati.o.n................................. 7-7
7.2.1 Design Description and Safety Objectives ........................ 7-7
7.2.2 Scope of Review ...................................... 7-11
7.2.3 Design Criteria ....................................... 7-11
7.2.4 Research and Development ............................... 7-11
7.2.5 Safety Issues ........................................ 7-12
7.2.6 Evaluation .......................................... 7-12
7.2.7 Conclusions ......................................... 7-14
7.3 Other Instrumentation and Monitoring Systems ........................ 7-14
7.3.1 Design Description and Safety Objectives ....................... 7-14
7.3.2 Scope of Review ...................................... 7-14
7.3.3 Review Criteria ................. . .................... 7-14
7.3.4 Research and Development ............................... 7-15
7.3.5 Safety Issues ........................................ 7-15
7.3.6 Evaluation .......................................... 7-15
7.3.7 Conclusions ......................................... 7-16
7.4 References .............................................. 7-16
8 ELECTRICAL SYSTEMS ........................................ 8-1
8.1 Overall Electrical System ...................................... 8-1
8.2 Safety-Related Electrical Power System ............................. 8-1
8.2.1 Design Description and Safety Objectives ........................ 8-1
8.2.2 Scope of the Review .................................... 8-1
8.2.3 Design Criteria ........................................ 8-2
8.2.4 Research and Development ................................ 8-2
8.2.5 Safety Issues ......................................... 8-3
8.2.6 Evaluation ........................................... 8-3
8.2.7 Conclusion .......................................... 8-4
8.3 Electromagnetic Pump Power System .............................. 8-4
8.3.1 Design Description and Safety Objectives ........................ 8-4
8.3.2 Scope of Review ....................................... 8-7
8.3.3 Design Criteria ........................................ 8-7
8.3.4 Research and Development ................................ 8-8
8.3.5 Safety Issues ......................................... 8-8
8.3.6 Evaluation ........................................... 8-9
8.3.7 Conclusion ......................................... 8-10
8.4 References .............................................. 8-10
9 AUXILIARY SYSTEMS ......................................... 9-1
9.1 Summary Descriptions .................... .................... 9-1
9.1.1 Fuel Handling and Storage System ............................ 9-1
9.1.2 Water System ......................................... 9-1
9.1.3 Process Auxiliaries ..................................... 9-1
9.1.4 Heating, Ventilation, and Air Conditioning System .................. 9-1
9.1.5 Auxiliary Liquid-Metal System .............................. 9-1
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9.1.6 SodiumPiping and Equipment Heating and Insulation System ........... 9-2
9.1.7 Other Auxiliary Systems .................................. 9-2
9.2 Fuel Handling and Storage System (Reactor Refueling System) ............... 9-2
9.2.1 Design Description and Safety Objectives ........................ 9-2
9.2.2 ' Scope of Staff Review ................................... 9-2
9.2.3 Design Criteria ........................................ 9-3
9.2.4 Research and Development Plans ............................. 9-3
9.2.5 Discussion of Safety Issues ................................ 9-3
9.2.6 Evaluation ........................................... 9-3
9.2.7 Conclusions .......................................... 9-4
9.3 Water System ............................................. 9-5
9.3.1 Design Description and Safety Objectives ........................ 9-5
9.3.2 Scope of Staff Review ................................... 9-5
9.3.3 Design Criteria ........................................ 9-5
9.3.4 Research and Development Plans ............................. 9-5
9.3.5 Discussion of Safety Issues ................................ 9-5
9.3.6 Evaluation ........................................... 9-5
9.3.7 Conclusions .......................................... 9-5
9.4 Process Auxiliaries .......................................... 9-5
9.4.1 Design Description and Safety Objectives ........................ 9-5
9.4.2 Scope of Staff Review ................................... 9-6
9.4.3 Design Criteria ........................................ 9-7
9.4.4 Research and Development Plans ............................. 9-8
9.4.5 Discussion of Safety Issues ................................ 9-8
9.4.6 Evaluation ........................................... 9-8
9.4.7 Conclusior .......................................... 9-8
9.5 Heating, Ventila _n, and Air Conditioning System ...................... 9-8
9.5.1 Design Description and Safety Objectives ........................ 9-8
9.5.2 Scope of Staff Review ................................... 9-8
9.5.3 Review Criteria ....................................... 9-8
9.5.4 Design Criteria ........................................ 9-9
9.5.5 Discussion of Safety Issues ................................ 9-9
9.5.6 Evaluation ........................................... 9-9
9.5.7 Research and Development Plans ......... . ................... 9-9
9.5.8 Conclusions .......................................... %9
9.6 Auxiliary Liquid-Metal System .................................. 9-9
9.6.1 Design Description and Safety Objectives ........................ 9-9
9.6.2 Scope of Staff Review .................................. 9-10
9.6.3 Review Criteria ...................................... 9-10
9.6.4 Design Criteria ....................................... 9-10
9.6.5 Discussion of Safety Issues ............................... 9-10
9.6.6 Evaluation .......................................... 9-10
9.6.7 Research and Development Plans ........................... 9-10
9.6.8 Conclusions ......................................... 9-10
9.7 Sodium Piping and Equipment Heating and Insulation System ............... 9-10
9.7.1 Design Description and Safety Objectives.. : .................... 9-10
9.7.2 Scope of Staff Review ................................. 9-11
NUREG-1368 x
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9.7.3 Design Criteria ....................................... 9-11
9.7.4 Discussion of Safety Issues ............................... 9-11
9.7.5 Evaluation .......................................... 9-12
9.7.6 Research and Development Plans ........................... 9-12
9.7.7 Conclusions ......................................... 9-12
9.8 Plant Fire Protection System ................................... 9-12
9.8.1 Design Description and Safety Objectives ....................... 9-12
9.8.2 Scope of Staff Review .................................. 9-12
9.8.3 Design Criteria ....................................... 9-13
9.8.4 Discussion of Safety Issues ............................... 9-13
9.8.5 Evaluation .......................................... 9-13
9.8.6 Conclusions ......................................... 9-14
9.9 References .............................................. 9-14
10 STEAM AND POWER CONVERSION SYSTEMS ..................... 10-1
10.1 Design Description and Safety Objectives .......................... 10-1
10.1.1 Summary Description ................................... 10-1
10.1.2 Turbine Generators .................................... 10-1
10.1.3 Main and Auxiliary Steam Systems .......................... 10-1
10.1.4 Feedwater and Condensate System ........................... 10-3
10.1.5 Water Chemistry Control ................................ 10-3
10.2 Scope of the Review ....................................... 10-3
10.3 Design Criteria .......................................... 10-3
10.4 Research and Development ................................... 10-3
10.5 Safety Issues ......................... ................... 10-4
10.6 Evaluation ............................................. 10-4
10.7 Conclusions ............................................ 10-5
10.8 References ............................................. 10-5
11 RADIOACTIVE WASTE MANAGEMENT .......................... 11-1
11.1 Source Terms ........................................... 11-1
11.2 Liquid Waste Management Systems .............................. 11-1
11.3 Gaseous Waste Management Systems ............................. 11-1
11.4 Solid Waste Management System ............................... 11-2
11.5 Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems 11-2
11.6 References ............................................. 11-2
12 RADIATION PROTECTION ................................... 12-1
12.1 Ensuring That Occupational Radiation Exposures Are ALARA ............. 12-1
12.1.1 Policy Considerations ................................... 12-1
12.1.2 Design Considerations .............................. 12-1
12.1.3 Operational Considerations ................................ 12-1
12.2 Radiation Sources ......................................... 12-2
12.3 Radiation Protection Design Features ............................. 12-2
12.3.1 Facility Design Features ................................. 12-2
12.3.2 Shielding .......................................... 12-2
xi NUREG-1368
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12.3.3 Ventilation System ................. . ................... 12-3
12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation .... 12-3
12.4 Dose Assessment ......................................... 12-3
12.5 Operational Radiation Protection Program .......................... 12-3
12.6 References ............................................. 12-4
13 CONDUCT OF OPERA'ITON ................................... 13-1
13.1 Radiological Emergency Planning ............................... 13-1
13.1.1 Preapplicant's Proposed Planning ............................ 13-1
13.1.2 Scope of the Review ................................... 13-2
13.1.3 Discussion .......................................... 13-2
13.1.4 Evaluation .......................................... 13-3
13.1.5 Conclusion . . ....................................... 13-3
13.2 Plant Operations ......................................... 13-4
13.2.1 General ................................... ........ 13-4
13.2.2 Operating Modes 13-4
13.2.3 Control Room and Remote Shutdown Facility .................... 13-5
13.2.4 Role of the Operator and Staffing ........................... 13-6
13.2.5 Multi-Module Control .................................. 13-7
13.2.6 Conclusion ......................................... 13-7
13.3 Safeguards and Security ..................................... 13-7
13.3.1 Design Description and Safety Issues ......................... 13-7
13.3.2 Scope of the Review ................................... 13-10
13.3.3 Conclusions ......................................... 13-10
13.4 References ............................................. 13-11
14 SAFETY TEST PROGRAM ...................................... 14-1
14.1 Design Description and Safety Objectives ..... _.................... 14-1
14.1. ! Safety Test P,rogram ................................... 14-1
14.2 Scope of Review ......................................... 14-4
14.3 Review Criteria .......................................... 14-4
14.3.1 10 CFR Part 52 ...................................... 14-4
14.3.2 Commission's Advanced Reactor Policy Statement ................. 14-5
14.3.3 Commission Paper SECY-91-074 ............................ 14-5
14.3.4 Standard Review Plan Section 14.2 .......................... 14-6
14.4 Research and Development ................................... 14-7
14.5 Safety Issues ............................................ 14-8
14.5.1 Risks Associated With the Prototype Tests ...................... 14-8
14.5.2 Lack of Detail on the Safety Test and R&D Programs ............... 14-8
14.5.3 Justification of the Single-Module Prototype ..................... 14-9
14.5.4 Justification of the Systems for the Prototype .................... 14-9
14.5.5 Additional Development or Testing in the R&D Program ............. 14-9
14.6 Evaluation ............................................. 14-9
14.6.1 Comparison to 10CFR 52.47(b)(2) .......................... 14-9
14.6.2 Comparison to the Commission's Advancex' Reactor Policy ............ 14-12
14.6.3 Comparison to Commission Paper SECY-91-074 .................. 14-12
14.6.4 Comparison to SRP Section 14.2 and RG 1.68 ................... 14-12
NUREG- 1368 xii
