Standard Review PlanN UREG-0800 (formerly issued as NUREG-75/087) Standard Review Plan for the...
33
N UREG-0800 (formerly issued as NUREG-75/087) Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1987
Transcript of Standard Review PlanN UREG-0800 (formerly issued as NUREG-75/087) Standard Review Plan for the...
N UREG-0800(formerly issued asNUREG-75/087)
Standard Review Planfor the Review ofSafety Analysis Reportsfor Nuclear Power Plants
LWR Edition
U.S. Nuclear RegulatoryCommissionOffice of Nuclear Reactor Regulation
June 1987
NUREG-0800(formedy issued asNUREG-75/087)
Standard Review Planfor the Review ofSafety Analysis Reportsfor Nuclear Power Plants
LWR Edition(This June 1987 update includes all revisionsissued between July 1981 and June 1987.)
U.S. Nuclear RegulatoryCommission
Office of Nuclear Reactor Regulation
June 1987
%,,I asot,
* 0M,
0a-*¢
INTRODUCTION
The Standard Review Plan (SRP) is prepared for the guidance of staff reviewersin the Office of Nuclear Reactor Regulation in performing safety reviews ofapplications to construct or operate nuclear power plants. The principalpurpose of the SRP is to assure .the quality and uniformity of staff reviewsand to present a well-defined base from which to evaluate proposed changes inthe scope and requirements of reviews. It is also a purpose of the SRP tomake information about regulatory matters widely available and to improvecommunication and understanding of the staff review process by interestedmembers of the public and the nuclear power industry.
The safety review is primarily based on the information provided by an applicantin a Safety Analysis Report (SAR). Section 50.34 of 10 CFR 50 of the Commission'sregulations requires that each application for a construction permit for anuclear facility shall include a Preliminary Safety Analysis Report (PSAR) andthat each application for a license to operate such a facility shall include aFinal Safety Analysis Report (FSAR). The SAR must be sufficiently detailed topermit the staff to determine whether the plant can be built and operatedwithout undue risk to the health and safety of the public. Prior to submissionof an SAR, an applicant should have designed and analyzed the plant in sufficientdetail to conclude that it can be built and operated safely. The SAR is theprincipal document in which the applicant provides the information needed tounderstand the basis upon which this conclusion'has been reached.
Section 50.34 specifies, in general terms, the information to be supplied in aSAR. The specific information required by the staff for an evaluation of anapplication is identified in Regulatory Guide 1.70, "Standard Format andContent of Safety Analysis Reports for Nuclear Power Plants - LWR Edition."The SRP sections are keyed to the Standard Format, and the SRP sections arenumbered according to the section numbers in the Standard Format. Reviewplans have not been prepared for SAR sections that consist of background ordesign data which are included for information or for use in the review ofother SAR sections.
The Standard Review Plan is written so as to cover a variety of site conditionsand plant designs. Each section is written to provide the complete procedureand all'acceptance criteria for all of the areas of review pertinent to thatsection. However, for any given application, the staff reviewers may selectand emphasize particular aspects of each SRP section as is appropriate for theapplication. In some cases, the major portion of the review of a plant featuremay be done on a generic basis with the designer of that feature rather thanin the context of reviews of particular applications from utilities. In othercases a plant feature may be sufficiently similar to that of a previous plantso that a de novo review of the feature is not needed. For these and othersimilar reasons, the staff may not carry out in detail all of the review stepslisted in each SRP section in the review of every application.
The individual SRP sections address, in detail, who performs the review, thematters that are reviewed, the basis for review, how the review is accomplished,and-the conclusions that are sought. The safety review is performed by 25primary branches. One of the objectives of the SRP is to assign the reviewresponsibilities to the various branches and to define the sometimes complexinterfaces between them. Each SRP section identifies the branch that has theprimary review responsibility for that section. In some review areas theprimary branch may require support, and the branches that are assigned thesesecondary review responsibilities are also identified for each SRP section.
Each SRP is organized into four subsections as follows:
I. Areas of Review
This subsection describes the scope of review, i.e., what is being reviewed bythe branch having primary review responsibility. This subsection contains adescription of the systems, components, analyses, data, or other informationthat is reviewed as part of the particular Safety Analysis Report section inquestion. It also contains a discussion of the information needed or thereview expected from other branches to permit the primary review branch tocomplete its review.
II. Acceptance Criteria
This subsection contains a statement of the purpose of the review, an identifica-jtion of which NRC requirements are applicable, and the technical basis fordetermining the acceptability of the design or the programs within the scopeof the area of review of the SRP section. The technical bases consist ofspecific criteria such as NRC Regulatory Guides, General Design Criteria,Codes and Standards, Branch Technical Positions, and other criteria.
The technical bases for some sections of the SRP are provided in Branch TechnicalPositions or Appendices which are included in the SRP. These documents typicallyset forth the solutions and approaches determined to be acceptable in the pastby the staff in dealing with a specific safety problem or safety-relateddesign area. These solutions and approaches are codified in this form so thatstaff reviewers can take uniform and well-understood positions as the samesafety problems arise in future cases. Some Branch Technical Positions andAppendices may be converted into Regulatory Guides if it appears that thisstep would aid the review process. Like Regulatory Guides, the Branch Techni-cal Positions and Appendices represent solutions and approaches that areacceptable to the staff, but they are not required as the only possible solu-tions and approaches. However, applicants should recognize that, as in thecase of Regulatory Guides, substantial time and effort on the part of thestaff has gone into the development of the Branch Technical Positions andAppendices and that a corresponding amount of time and effort will probably berequired to review and accept new or different solutions and approaches.Thus, applicants proposing solutions and approaches to safety problems orsafety-related design areas other than those described in the Branch TechnicalPositions and Appendices must expect longer review times and more extensivequestioning in these areas. The staff is willing to consider proposals forother solutions and approaches on a generic basis, apart from a specificlicense application, to avoid the impact of the additional review time onindividual cases.
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III. Review Procedures
This subsection discusses how the review is accomplished. The section isgenerally a step-by-step procedure that the reviewer goes through to providereasonable verification that the applicable safety criteri~a have been met.
IV. Evaluation Findings
This subsection presents the type of conclusion that is sought for the particularreview area. For each section, a conclusion of this type is included in thestaff's Safety Evaluation Report in which the staff publishes the results oftheir review. -The SER also contains a description of the review includingsuch subjects as which aspects of the review were selected or emphasized;which matters were modified by the applicant, require additional information,will be resolved in the future, or remain unresolved; where the plant's designor the applicant's programs deviate from the criteria stated in the SRP; andthe bases for any deviations from the SRP or exemptions from the regulations.
V. References
This subsection lists the references used in the review process.
The SRP and the Standard Format are directed toward water-cooled reactor powerplants. Staff reviewers will adapt the SRP for use in the reviews of otherreactor types where applicable.
The Standard Review Plans result from many years of experience by the staff inestablishing and using regulatory requirements in evaluating the safety ofnuclear power plants and in reviewing Safety Analysis Reports. A great dealof progress has been made in the methods of review and in the development ofregulations, guides, and standards since the early years of review. ThisStandard Review Plan may be considered a part of a continuing regulatorystandards development activity that not only documents current methods ofreview but also provides the base of orderly modifications of the reviewprocess in the future.
In 1981, the Standard Review Plan was revised in entirety and published asNUREG-0800. The revision program had three major objectives, i.e., to morecompletely identify the NRC requirements that are germane to each reviewtopic, to more fully describe how the review effort determines satisfaction ofthe requirement, and to incorporate the large number of new and revisedregulatory positions (primarily TMI-related) that had already been established.To accomplish this and to conform to the revised NRR organization, some SRPsections were added, deleted, split, and/or combined.
The SRP will be revised and updated periodically as the need arises to clarifythe content or correct errors and to incorporate modifications approved by theDirector of the Office of Nuclear Reactor Regulation. A revision number andpublication date are printed at a lower corner of each page of each SRP section.Since individual sections have been, and will continue to be, revised asneeded, the revision numbers and dates will not be the same for all sections.The Table of Contents indicates the revision numbers of the currently effectivesections. As necessary, corresponding changes to the Standard Format will
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also be made. Comments and suggestions for improvement will be considered andshould be sent to the Director, Office of Nuclear Reactor Regulation, U.S.Nuclear Regulatory Commission, Washington, DC 20555. Notices of errors oromissions should also be sent to the same address.
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7590-01
U.S. NUCLEAR REGULATORY COMMISSION
NUREG-0800
"STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS"
NOTICE OF ISSUANCE AND AVAILABILITY
REVISED TABLE OF CONTENTS
The U.S. Nuclear Regulatory Commission (NRC) has published a revision to
the "Table of Contents" of NUREG-0800, "Standard Review Plan for the Review
of Safety Analysis Reports for Nuclear Power Plants," LWR Edition (SRP).
The table of contents, Revision 5 incorporates all Standard Review Plan
Sections that have been revised and issued since NUREG-0800 was issued in
July 1981. All changes resulting from incorporating the revised SRP
Sections and a few editorial changes are identified by a line in the margin
of the revised Table.
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A copy of the revised Table is expected to be available in the Public
Document Room within 2 weeks. Copies of the revised SRP Sections or of the
complete Standard Review Plan, NUREG-0800, Accession No. PD-81-920199, are
available for purchase from the National Technical Information Service,
5285 Port Royal Road, Springfield, Virginia 22161; telephone (703) 487-4650.
Dated at Bethesda, Maryland this 26 day of December 1984.
FOR T NUCLEAR REGUL9RY COMMISSION
' Edson G. Case, Acting DirectorOffice of Nuclear Reactor Regulation
2
STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS
TABLE OF CONTENTS
Applicable IssuedSRP No. Revision Year/Month
INTRODUCTION . ....................................... --- 75/111 81/7
Table of Contents .................................. --- 75/111 79/12 79/33 80/54 81/75 84/12
Compilation of Branch Technical Positions ........... 0 81/7
CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.8 Interfaces for Standard Design 0 0 78/121 81/7
CHAPTER 2 SITE CHARACTERISTICS
2.1.1 Site Location and Description --- --- 75/111 78/72 81/7
2.1.2 Exclusion Area Authority andControl ..................... --- 75/11
1 78/122 81/7
2.1.3 Population Distribution . --- 75/111 78/122 81/7
2.2.1-2.2.2 Identification of Potential Hazardsin Site Vicinity 1 --- 75/11
1 78/72 81/7
2.2.3 Evaluation of Potential Accidents ... --- 75/111 78/122 81/7
2.3.1 Regional Climatology . --- 75/112 78/42 81/7
2.3.2 Local Meteorology . --- 75/111 78/42 81/7
2.3.3 Onsite Meteorological MeasurementsPrograms ..............-- 75/11
1 78/52 81/7
Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
SRP No.
2.3.4
2.3.5
2.4.1
2.4.2
2.4.3
2.4.4
2.4.5
2.4.6
2.4.7
2.4.8
2.4.9
2.4.10
2.4.11
2.4.12
Appendix A .....................
Short-Term Diffusion Estimates ForAccidental Atmospheric Releases.....
Long-Term Diffusion Estimates .......
Hydrologic Description ..............
Appendix A .....................
Floods ..............................
Probable Maximum Flood (PMF) onStreams and Rivers ................
Potential Dam Failures ..............
Probable Maximum Surge and SeicheFlooding ..........................
Probable Maximum Tsunami Flooding
Ice Effects .........................
Cooling Water Canals andReservoirs ........................
Channel Diversions ..................
Flood Protection Requirements .......
Cooling Water Supply ................
Groundwater .........................
ii
ApplicableRevision
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IssuedYear/Month
75/1178/581/7
75/1181/7
75/1178/583/7
75/1178/68V17
75/1178/68V17
75/1178/6s8n
12
12
12
12
12
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1
2
12
2
12
Rev.
75/1178/6817
75/1178/6s1n
75/1178/681/7
75/1178/68V17
75/1178/581/7
75/1178/681/7
75/1178/6817
75/1178/581/7
75/1178/5s1n
75/178n81n
- December 1984
TABLE OF CONTENTS (Continued)
ApplicableSRP Ho. Revision YE
BTP HMB/GSB 1 ....... ........... ---1
BTP HGEB 1 .......... ........... 2
2.4.13 Accidental Releases of LiquidEffluents in Ground and SurfaceWaters .*-------------------- 1
22
2.4.14 Technical Specifications andEmergency OperationRequirements. ---
12
2.5.1 Basic Geologic and SeismicInformation ............ ........... ---
12
2.5.2 Vibratory Ground Motion ...... ....... ---1
2.5.3 Surface Faulting ......... ........... ---12
2.5.4 Stability of Subsurface Materialsand Foundations ........ ........... ---
12
2.5.5 Stability of Slopes ....... .......... ---12
CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS. EQUIPMENT, AND SYSTE]
3.2.1 Seismic Classification ...... ........ ---
3.2.2 System Quality GroupClassification ......... ........... ---
Appendix A (FormerlyBTP RSB 3-1) . .................
Appendix B (FormerlyBTP RSB 3-2) .................
Appendix C ........ ........... 01
Appendix D ........ ........... 01
3.3.1 Wind Loadings ........... ............ ---12
3.3.2 Tornado Loadings ......... ........... ---12
Issued?ar/Month
75/1178/781/7
75/1178/681/7
75/1178/681/7
75/1178/1181/7
75/1181/7
75/1178/118117
75/1178/1181/7
75/1178/11817
4S
75/118117
75/1181/7
75/1181/7
75/1181/7
None81/7
None81/7
75/1178/881/7
75/1178/881/7
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TABLE OF CONTENTS (Continued)
SRP No.
3.4.1
3.4.2
3.5.1.1
3.5.1.2
3.5.1.3
3.5.1.4
3.5.1.5
3.5.1.6
3.5.2
3.5.3
3.6.1
3.6.2
Flood Protection ....................
Analysis Procedures .................
Internally Generated Missiles(Outside Containment) .............
Internally Generated Missiles(Inside Containment) ..............
Turbine Missiles ....................
Missiles Generated by NaturalPhenomena .........................
BTP MAB 3-2 ....................
BTP ASB 3-2 ....................
Site Proximity Missiles (ExceptAircraft) .........................
Aircraft Hazards ....................
Structures, Systems, and Componentsto be Protected from ExternallyGenerated Missiles ................
Barrier Design Procedures ...........
Appendix A ........................
Plant Design for Protection AgainstPo'stulated Piping Failures inFluid Systems OutsideContainment .......................
BTP ASB-3-1 ....................
Determination of Rupture Locationsand Dynamic Effects Associatedwith the Postulated Rupture ofPiping ............................
BTP MEB-3-1 ....................
iv
Applicable IssuedRevision Year/Month
--- 75/111 78/32 81/7
--- 75/111 None2 81/7
75/111 78/42 81/7
--- 7/111 78/82 81/7
--- 75/111 78/72 81/7
75/111 78/72 81/7
--- 75/111 None
2 81/7
--- 75/111 81/7
--- 75/111 None2 81/7
75/111 78/32 81/7
___ 75/111 81/7
0 81/7
--- 75/111 81/7
___ 75/111 81/7
--- 75/111 81/7
___ 75/111 81/7
Rev. 5 - December 1984
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TABLE OF CONTENTS (Continued)
SRP No.
3.7.1
3.7.2
3.7.3
3.7.4
3.8.1
3.8.2
3.8.3
3.8.4
3.8.5
3.9.1
3.9.2
3.9.3
3.9.4
3.9.5
Seismic Design Parameters ...........
Seismic System Analysis .............
Seismic Subsystem Analysis ..........
Seismic Instrumentation .............
Concrete Containment ................
Appendix ..........................
Steel Containment ...................
Concrete and Steel InternalStructures of Steel or ConcreteContainments ......................
Other Seismic Category IStructures ........................
Appendix A ........................Appendix B ........................Appendix C ........................Appendix D ........................
Foundations .........................
Special Topics for MechanicalComponents ........................
Dynamic Testing and Analysis ofSystems, Components, andEquipment .........................
ASME Code Class 1, 2, and 3Components, Component Supports,and Core Support Structures .......
Appendix A ........................
Control Rod Drive Systems ...........
Reactor Pressure Vessel Internals
Inservice Testing of Pumps andValves ............................
ApplicableRevision
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75/1181/7
75/118117
75/1181/7
75/1181/7
75/1181/781/7
75/1181/7
I
75/1181/7
75/1181/781/781/781/781/7
75/1181/7
75/1178/481/7
75/1178/881/7
75/1181/7
81/784/4
75/1181/784/4
75/1178/481/7
75/1178/481/7
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3.9.6
v Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
Applicable IssuedSRP No. Revision Year/Month
3.10 Seismic Qualification of Category IInstrumentation and ElectricalEquipment .--- 75/11
1 78/42 81/7
3.11 Environmental Design of Mechanicaland Electrical Equipment .......... --- 75/11
1 78/72 81/7
CHAPTER 4 REACTOR
4.2 Fuel System Design .................. --- 75/111 78/92 81/7
Appendix A ............. 0........... 81/7
4.3 Nuclear Design ...................... --- 75/111 78/42 81/7
BTP CPB 4.3-1 .................. --- 75/111 78/42 81/7
4.4 Thermal and Hydraulic Design ........ --- 75/111 81/7
Appendix ....................... --- 75/111 81/7
4.5.1 Control Rod Drive StructuralMaterials . . --- 75/11
1 78/12 81/7
4.5.2 Reactor Internal and CoreSupport Materials. --- 75/11
1 78/12 81/7
4.6 Functional Design of Control RodDrive System .--- 75/11
1 81/7
CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS
5.2.1.1 Compliance with the Codes and StandardRule, 10 CFR § 5O.55a ...... ....... --- 75/11
1 78/12 81/7
5.2.1.2 Applicable Code Cases ....... ........ --- 75/111 78/12 81/7
5.2.2 Overpressure Protection ...... ....... --- 75/n1 81/7
BTP RSB 5-2 ............ ........... 0 81/7
5.2.3 Reactor Coolant Pressure BoundaryMaterials .--- 75/1
1 78/42 81/7
vi Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
Applicable IssuedSRP No. Revision Year/Month
BTP MTEB 5-7 ......... ........... --- 75/111 78/42 81/7
5.2.4 Reactor Coolant Pressure BoundaryInservice Inspection and Testing ... 75/11
1 81/7
5.2.5 Reactor Coolant Pressure BoundaryLeakage Detection .-.................. - 75/11
1 81/7
5.3.1 Reactor Vessel Materials ............. 75/111 8V7
5.3.2 Pressure-Temperature Limits 7 1.......... - 75V111 81/7
BTP KTEB 5-2 ......... ........... --- 75/111 81/7
5.3.3 Reactor Vessel Integrity ...... 75/....... 11 751 81/7
5.4 Preface .............................. --- 75/111 81/7
5.4.1.1 Pump Flywheel Integrity (PWR) ........ --- 75/111 81/7
5.4.2.1 Steam Generator Materials ......- ...... - 75/111 78/112 81/7
BTP MTEB 5-3 ......... ........... --- 75/111 78/112 81/7
5.4.2.2 Steam Generator Tube InserviceInspection .--- 75/11
1 81/7
5.4.6 Reactor Core Isolation CoolingSystem (BWR).--- 75/11
1 78/32 81/73 84/4
5.4.7 Residual Heat Removal'(RHR) System ... --- 75/111 78/82 81/73 84/4
BTP RSB 5-1 ...................... --- 75/111 78/82 81/7
5.4.8 Reactor Water Cleanup System(BWR) ............................ . 75/11
1 78/72 81/7
5.4.11 Pressurizer Relief Tank 75/1............. - 75/1 78/82 81/7
5.4.12 Reactor Coolant SystemHigh Point Vents .0 8V17
v71 Rev. 5 - December 1984
SRP No.
6.1.1
6.1.2
6.2.1
6. 2. 1. 1.A
6. 2. 1. 1.B
6. 2. 1.1.C
6.2.1.2
6.2.1.3
6.2.1.4
6.2.1.5
TABLE OF CONTENTS (Continued)
ApplicableRevision
CHAPTER 6 ENGINEERED SAFETY FEATURES
Engineered Safety FeaturesMaterials .........................
2
BTP MTEB 6-1 .--12
Protective Coating Systems(Paints) - Organic Materials. ---
12
Containment Functional Design .--12
PWR Dry Containments, IncludingSubatmospheric Containments .-;
12
Ice Condenser Containments. ---12
Pressure-Suppression Type BWRContainments. ---
IssuedYear/Month
75/1178/1281/7
75/1178/1281/7
75/1178/1281/7
75/1178/481/7
75/1178/881/7
75/1178/881/7
75/1178/578/879/281/783/184/8
79/281/783/183/1
75/1178/8817
75/1181/7
75/1181/7
Appendix I .....................Appendix A .....................
Appendix B .....................
Subcompartment Analysis .............
Mass and Energy Release Analysis forPostulated Loss-of-CoolantAccidents .........................
Mass and Energy Release Analysis forPostulated Secondary System PipeRuptures ..........................
Minimum Containment PressureAnalysis for Emergency CoreCooling System PerformanceCapability Studies ................
BTP CSB 6-1 ....................
123456
0120
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75/111 78/82 81/7
--- 75/111 78/82 81/7
Rev. 5 - December 1984viii
TABLE OF CONTENTS (Continued)
SRP No.
6.2.2
6.2.3
6.2.4
6.2.5
Containment Heat Removal Systems ....
Secondary Containment FunctionalDesign ............................
BTP CSB 6-3 ....................
Containment Isolation System ........
BTP CSB 6-4 ....................
Combustible Gas Control inContainment .......................
Appendix A ....................
BTP CSB 6-2 ....................
Containment Leakage Testing .........
Fracture Prevention of ContainmentPressure Boundary .................
Emergency Core Cooling System .......
BTP RSB 6-1 ....................
Control Room Habitability Systems
Appendix A .....................
ESF Atmosphere Cleanup Systems ......
Containment Spray as a FissionProduct Cleanup System ............
ApplicableRevision
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75/1178/881/7
75/1178/881/7
75/1178/5s/n
75/1178/581/7
75/1178/581/7
75/1178/581/7
75/1178/5s8n
75/1178/981/7
81/7
75/1181/784/4
75/1181/7
75/1178/1281/7
75/1178/12817
75/1178n817
75/1181/7
6.2.6
6.2.7
6.3
6.4
6.5.1
6.5.2
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SRP No.
6.5.3
6.5.4
6.6
6.7
7.1
TABLE OF CONTENTS (Continued)
Applicable IssuedRevision Year/Month
Fission Product Control Systemsand Structures .................... --- 75/11
1 78/72 81/7
Ice Condenser as a Fission ProductCleanup System .................... --- 75/11
1 78/42 81/7
Inservice Inspection of Class 2and 3 Components .................. --- 75/11
1 81/7
Main Steam Isolation Valve LeakageControl System (BWR) .............. --- 75/11
1 78/32 81/7
CHAPTER 7 INSTRUMENTATION AND CONTROLS
Instrumentation and Controls -Introduction ...................... --- 75/11
1 78/72 81/73 84/2
Table 7-1 Acceptance Criteriaand Guidelines for Instrumen-tation and Controls SystemsImportant to Safety --- --- 75/11
1 78/72 81/73 84/2
Table 7-2 THI Action PlanRequirements for Instrumenta-tion and Controls SystemsImportant to Safety . 0 0 81/7
Appendix A ................... 0 81/71 84/2
Appendix B. .................. 0 81/7
Reactor Trip System ........ ......... --- 75/111 78/72 81/7
Appendix A 75/1..................... - 75/1 78/72 81/7
Engineered Safety Features Systems .. --- 75/111 78/72 81/7
Appendix A ..................... 75/111 78n2 81/7
Safe Shutdown Systems ....... ........ --- 75/1
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7.3
7.412
78/781/7
x Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
SRP No.
7.5 Information Systems Important toSafety ............................
Interlock Systems Important toSafety ............................
Control Systems .....................
7.6
7.7
Appendix 7-A Branch Technical
BTP ICSB 1
BTP ICSB 3
BTP ICSB 4
BTP ICSB 5
BTP ICSB 9
BTP ICSB 12
BTP ICSB 13
BTP ICSB 14
BTP ICSB 16
BTP ICSB 19
BTP ICSB 20
BTP ICSB 21
Positions (ICSB) .-
(DOR) ..............
(PSB) ..............
i....................
....................
ApplicableRevision
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IssuedYear/Month
75/1178/781/784/2
75/1178/781/7
75/1178/781/784/2
75/1178/781/7
75/1178/781/7
75/1178n81/n
75/1178/781/7
75/1178/781/7
75/1178/781/7
75/117sn817
75/117sn8V7
75/1178/781/7
75/1178n81/7
75/1178/781/7
75/13.78/781/
75/117sn81/
December 1984
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xi
TABLE OF CONTENTS (Continued)
SRP No.
BTP ICSB 22 ...................
BTP ICSB 25 ....................
BTP ICSB 26 ....................
Appendix 7-B General Agenda, Station SiteVisits ............................
CHAPTER 8 ELECTRIC POWER
8.1 Electric Power-Introduction.
Table 8-1 Acceptance Criteriaand Guidlelines for ElectricPower Systems ................
8.2 Offsite Power System.
Appendix A ........................
8.3.1 A-C Power Systems (Onsite).
Appendix ..........................
8.3.2 D-C Power Systems (Onsite).
Appendix BA Branch Technical Positions (PSB)
BTP ICSB 2 (PSB) ..............
BTP ICSB 4 (PSB) ..............
BTP ICSB 8 (PSB) ..............
BTP ICSB 11 (PSB) ..............
BTP ICSB 15 (PSB) ..............
ApplicableRevision
12
12
___
121
2
___
1
2
12
___
3
12
2
___
12
2
12
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12
1
2
12
2
1
2
2
Rv5---
IssuedYear/Month
75/1178/781/7
75/178/781/7
75/1178/781/7
75/1181/7
75/1178/481/7
75/1178/481/7
75/1178/481/783/783/7
75/1178/581s781/7
75/1178/481/7
75/1178/681/7
75/1178/681/7
75/1178/681/7
75/1178/681/7
75/1178/681/7
75/1178/681s7
)ecember 1984
I
xii
TABLE OF CONTENTS (Continued)
SRP No.
Appendix 8B
9.1.1
9.1.2
9.1.3
9.1.4
9.1.5
9.2.1
9.2.2
9.2.3
9.2.4
9.2.5
BTP ICSB 17 (PSB) ..............
BTP ICSB 18 (PSB) ..............
BTP ICSB 21 (PSB) ..............
BTP PSB I ......................
BTP PSB 2 ......................
General Agenda, Station Site Visits
CHAPTER 9 AUXILIARY SYSTEMS
New Fuel Storage ....................
Spent Fuel Storage ..................
Spent Fuel Pool Cooling and CleanupSystem ............................
Light Load Handling System (Relatedto Refueling) .....................
BTP ASB 9-1 ....................
Overhead Heavy Load HandlingSystems ...........................
Station Service Water System ........
Reactor Auxiliary Cooling WaterSystems ...........................
Demineralized Water Makeup System
Potable and Sanitary Water Systems
Ultimate Heat Sink ..................
xiii
Applicable IssuedRevision Year/Month
___ 75/111 78/62 81/7
--- 75/111 78/62 81/7
--- 75/111 78/62 81/7
0 81/7
0 s1n
0 81/7
--- 75/111 78/22 81/7
--- 75/111 78/32 79/33 81/7
75/111 81/7
--- 75/111 78/42 81/7
75/111 78/42 81/7
0 81/7
--- 75/111 78/32 81n3 84/4
--- 75/111 81/72 84/4
75/111 78/32 81/7
--- 75/111 78/32 81/7
75/111 78/32 81/7
Rev. 5 - December 1984
I
I
TABLE OF CONTENTS (Continued)
Applicable IssuedSRP No. Revision Year/Month
BTP AS8 9-2 ......... ........... --- 75/111 78/32 81/7
9.2.6 Condensate Storage Facilities --....... - 75/111 78/32 81/7
9.3.1 Compressed Air System ...... ......... - 75/111 81/7
9.3.2 Process and Post-Accident SamplingSystems . . 75/11
1 78/72 81/7
9.3.3 Equipment and Floor DrainageSystem................ -- 75/11
1 78/32 81/7
9.3.4 Chemical and Volume Control System(PWR) (Including Boron RecoverySystem)............... --- 75/11
1 78/32 81n
9.3.5 Standby Liquid Control System(BWR)................ --- 75/11
1 78/32 81/7
9.4.1 Control Room Area VentilationSystem................75/11
1 78/32 81/7
9.4.2 Spent Fuel Pool Area VentilationSystem ................. 75/11
1 78/32 81/7
9.4.3 Auxiliary and Radwaste AreaVentilation System 75/U................ - 75/
1 78/32 81/7
9.4.4 Turbine Area Ventilation System --..... - 75/111 78/32 81/7
9.4.5 Engineered Safety FeatureVentilation System ................ 75/11
1 78/32 81/7
9.5.1 Fire Protection Program ...... ....... --- 75/111 76/52 78/33 81/7
BTP CMEB 9.5.1 ....... .......... --- 76/51 78/32 81/7
Appendix A 76/11..................... -- 71 81/7
Xiv Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
SRP No.
9.5.2
9.5.3
9.5.4
9.5.5
9.5.6
9.5.7
9.5.8
10.2
10.2.3
10.3
10.3.6
10.4.1
10.4.2
10.4.3
10.4.4
Communications Systems ..............
Lighting Systems ....................
Emergency Diesel Engine Fuel OilStorage and Transfer System .......
Emergency Diesel Engine CoolingWater System ......................
Emergency Diesel Engine StartingSystem ............................
Emergency Diesel Engine LubricationSystem ............................
Emergency Diesel Engine CombustionAir Intake and Exhaust System .....
ApplicableRevision
___
12
12
12
12
___
1
2
1
2
IssuedYear/Month
75/1178/481/7
75/1178/481/7
75/1178/481/7
75/1178/481/7
75/1178/481/7
75/1178/481/7
75/1178/481/7
75/1178/481/7
75/1181/7
75/1178/481/784/4
75/1178/481/7
75/1178/481/7
75/1178n81/7
75/1178/781/7
75/1178/481/7
December 1984
CHAPTER 10 STEAM AND POWER CONVERSION SY2
Turbine Generator ...................
Turbine Disk Integrity ..............
Main Steam Supply System ............
Steam and Feedwater SystemMaterials .........................
Main Condensers .....................
Main Condenser Evacuation System ....
Turbine Gland Sealing System ........
Turbine Bypass System ...............
xv R
STEM
12
1
123 I
___
12
___
12
12
12
__ -
12
Rev. 5 -
TABLE OF CONTENTS (Continued)
SRP No.
10.4.5
10.4.6
10.4.7
Circulating Water System ............
Condensate Cleanup System ...........
Condensate and Feedwater System .....
BTP ASB 10-2 ...................
Steam Generator Blowdown System(PWR) .............................
Auxiliary Feedwater System (PWR) ....
BTP ASB 10-1 ...................
ApplicableRevision
12
12
123
10.4.8
10.4.9
11.1
11.2
11.3
11.4
11.5
123
___
12
12
12
CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT
Source Terms . ....................... ---
2
Liquid Waste Management Systems ..... ---
2
Gaseous Waste Management Systems .... ---
2
BTP ETSB 11-5 ...... ............... 0
Solid Waste Management Systems ...... ---12
BTP ETSB 11-3 . .................
2
Appendix 11.4-A .... ............ 0
Process and Effluent RadiologicalMonitoring Instrumentation andSampling Systems . ..................
123
Appendix 11.5-A .... ............ 01
xvi Rev. 5-
IssuedYear/Month
75/1178/381/7
75/1178/381/7
75/1178/381/784/4
75/1178/381/784/4
75/1178/781/7
75/1178/481/7
75/1178/481n
75/1178n81/7
75/1178n81/7
75/1178/781/7
81/7
75/1178/781/7
75/1178n81/7
81/7
75/1178n79/481/7
79/481/7
December 1984
I
TABLE OF CONTENTS (Continued)
Applicable IssuedRevision Year/MonthSRP No.
12.1
12.2
12.3-12.4
12.4(1)
12.5
13.1.1
13.1.2-13.1.3
13.1.3(2)
13.2(3)
13.2.1
13.2.2
13.3
13.4
23.5(4)
CHAPTER 12 RADIATION PROTECTION
Assuring That Occupational RadiationExposures are As Low As IsReasonably Achievable .............
Radiation Sources ...................
Radiation Protection DesignFeatures ..........................
Dose Assessment .....................
Operational Radiation ProtectionProgram ...........................
CHAPTER 13 CONDUCT OF OPERATIONS
Management and Technical SupportOrganization ......................
Operating Organization ..............
Qualifications of Nuclear PlantPersonnel .........................
Training ............................
Reactor Operator Training ...........
Training For Non-Licensed PlantStaff .............................
Emergency Planning ..................
Operational Review ..................
Plant Procedures ....................
12_;_1
2
12
_;_
1
2
12
___
12
1
1212
1
2
2
75/1178/581/7
75/1178/581/7
75/1178/581/7
75/1178/5
75/1178/581/7
75/1179/481/7
75/1179/481/7
75/1179/4
75/1178/381/7
81/7
81/7
75/1178/3817
75/1179/281/7
75/1178/381/7
I
I
(1)SRP Section has been combined with SRP Section 12.3.(2)SRP Section has been combined with SRP Section 13.1.2.(3)SRP Section has been replaced by SRP Sections 13.2.1 and 13.2.2.(4)SRP Section has been replaced by SRP Sections 13.5.1 and 13.5.2.
xvii Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
SRP No.
13.5.1
13.5.2
13.6
Applicable IssuedRevision Year/Month
0 81/7Administration Procedures ...........
Operating and MaintenanceProcedures ........................
Physical Security ...................
0
2
81/7
75/1181/7
14.1
14.2
CHAPTER 14 INITIAL TEST PROGRAM
Initial Plant Test Programs - PSAR ..
Initial Plant Test Programs - FSAR ..
Standard Plant Designs, Initial TestProgram - Final Design Approval(FDA).............................
CHAPTER 15 ACCIDENT ANALYSIS
Introduction ........................
1__
212
12
01
75/1179/281/7
75/1179/281/7
79/281/7
14.3
15.012
75/1178/881/7
15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
15.1.1-15.1.4
15.1.5
Decrease in Feedwater Temperature,Increase in Feedwater Flow,Increase in Steam Flow, andInadvertent Opening of a SteamGenerator Relief or Safety Valve ..
Steam System Piping Failures Insideand Outside of Containment(PWR) .............................
Appendix A .....................
1
12
2
75/11s8n
75/1178/8s8/
75/1178/8s8n
15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
15.2.1-15.2.5
15.2.6
15.2.7
Loss of External Load, Turbine Trip,Loss of Condenser Vacuum,Closure of Main Steam IsolationValve (BWR), and Steam PressureRegulatory Failure (Closed) .......
Loss of Nonemergency AC Powerto the Station Auxiliaries ........
Loss of Normal Feedwater Flow .......
1
1
1
75/1181/7
75/1181/7
75/118s7
xviii Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
SRP No.
15.2.8
Applicable IssuedRevision Year/Month
Feedwater System Pipe BreaksInside and Outside Containment(PWR) .............................
175/1181/7
15.3.1-15.3.2
15.3.3-15.3.4
15.4.1
15.4.2
15.4.3
15.4.4-15.4.!
15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE
Loss of Forced Reactor Coolant FlowIncluding Trip of Pump and FlowController Malfunctions ...... ..... ---
1
4 Reactor Coolant Pump Rotor Seizureand Reactor Coolant Pump ShaftBreak .............................
12
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES
i
Uncontrolled Control Rod AssemblyWithdrawal from a Subcriticalor Low Power Startup Condition ....
Uncontrolled Control Rod AssemblyWithdrawal at Power .:.............
Control Rod Hisoperation (SystemMalfunction or Operator Error) ....
Startup of an Inactive Loop orRecirculation Loop at an IncorrectTemperature, and Flow ControllerMalfunction Causing an Increasein BWR Core Flow Rate .............
12
___
12
12
1__1
75/1181/7
75/1178/881/7
75/1178/481/7
75/1178/48)17
75/1178/481/7
75/1181/7
15.4.6
15.4.7
15.4.8
Chemical and Volume Control SystemMalfunction That Results in aDecrease in the Boron Concentra-tion in the Reactor Coolant(PWR) .............................
Inadvertent Loading and Operationof a Fuel Assembly in anImproper Position .................
Spectrum of Rod EjectionAccidents (PWR) ...................
Appendix A .....................
1
11
12
l-
75/1181/7
75/1181/7
75/1178/481/7
75/1181/7
xix Rev. S - December 1984
TABLE OF CONTENTS (Continued)
SRP No.
15.4.9 Spectrum of Rod Drop Accidents(BWR) .............................
Appendix A .....................
Applicable IssuedRevision Year/Month
--- 75/111 78/42 81/7
--- 75/111 78/42 81/7
15.5.1-15.5.2
15.6.1
15.6.2
15.6.3
15.6.4
15.6.5
15.5 INCREASE IN REACTOR COOLANT INVENTORY
Inadvertent Operation of ECCS andChemical and Volume Control SystemMalfunction That Increases ReactorCoolant Inventory ........ ......... ---
1
15.6 DECREASE IN REACTOR COOLANT INVENTORY
Inadvertent Opening of a PWRPressurizer Relief Valveor a BWR Relief Valve ...... ....... ---
1
Radiological Consequences of theFailure of Small Lines CarryingPrimary Coolant OutsideContainment .............. ......... ---
12
Radiological Consequences of SteamGenerator Tube Failure (PWR) ...... ---
12
Radiological Consequences of MainSteam Line Failure OutsideContainment (BWR) ........ ......... ---
12
Loss-of-Coolant Accidents Resultingfrom Spectrum of PostulatedPiping Breaks Within the ReactorCoolant Pressure Boundary ..... .... ---
12
Appendix A ..................... ---1
Appendix B ..................... ---1
Appendix C ..................... ---22
Appendix D ..................... ---1
75/n81/7
75/1181/7
75/n7sn81/7
75/1178/1281/7
75/n78/781/7
75/1178/8s8n
75/1181/7
75/1181/7
75/1178/781/7
75/1181n
XX Rev. 5 - December 1984
TABLE OF CONTENTS (Continued)
Applicable IssuedSRP No. Revision Year/Month
15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT
15.7.1 Waste Gas System Failure ............ --- 75/111 81/7
15.7.2 Radioactive Liquid Waste SystemLeak or Failure (Release toAtmosphere) .. . 75/11
1 81/7
15.7.3 Postulated Radioactive Release Dueto Liquid-Containing TankFailures . . --- 75/11
1 78n2 81/7
15.7.4 Radiological Consequences of FuelHandling Accidents ....... ......... --- 75/11
1 8117
15.7.5 Spent Fuel Cask Drop Accidents ...... --- 75/111 78/122 81/7
15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM
15.8 Anticipated Transients WithoutScram ............................. --- 75/11
1 81/7
Appendix ....................... --- 75/111 81/7
CHAPTER 16 TECHNICAL SPECIFICATIONS
16.0 Technical Specifications ...... ...... --- 75/111 81/7
CHAPTER 17 QUALITY ASSURANCE
17.1 Quality Assurance During the Designand Construction Phases .--- 75/11
1 79/22 81/7
17.2 Quality Assurance During theOperations Phase .--- 75/11
1 79/22 81/7
CHAPTER 18 HUMAN FACTORS ENGINEERING
18.0 Human Factors Engineering/StandardReview Plan Development ..... ...... 0 81/7
1 84/9
18.1 Control Room ............. ........... D 89
Appendix A ............. 0........... 84/9
18.2 Safety Parameter Display System 0..... 84/12
Appendix A ............. 0........... 84/12
xx i Rev. 5 - December 1984
NRC FORM tU U.S NUCLEAR REGULATORY COMMISSION I REPORT NUMBER 1ASpOd bY TtOC. OM Vd1 NO. Of ""O
t6-31
BIBLIOGRAPHIC DATA SHEETNNUREG- 0800
2 Leave bo" k-
2 TITLE AND SUBTITLE 4. RECIPIENT'S ACCESSION NUMBER
Standard Review Plan for the Review of Safety AnalysisReports for Nuclear Power Plant, LWR Edition. Revision 5 DATE REPORT COMPLETED
to SRP Table of Contents. MONTH YEAR
December 1984S. AUTHORtSI 7. DATE REPORT ISSUED
MONTH |YEAR
January 19859. PROJECTITASKIWORIC UNIT NUMBER
tL PERFORMING ORGANIZATION NAME AND UAILING ADDRESS 4A ZC t DJ
Office of Nuclear Reactor RegulationsU. S. Nuclear Regulatory Commission 10FINNUMBER
Washington, DC 20555
I1. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS f1t*p* Zw Gtr 12s. TYPE OF REPORT
Office of Nuclear Reactor RegulationsU. S. Nuclear Regulatory Commission SRP Section (Guide)
Washington, DC 20555 12b PERIOD COVERED fII*wvv ds.
13. SUPPLEMENTARY NOTES
SRP Table of Contents, Revision 5
14 ABSTRACT t2 --
Revision 5 to SRP Table of Contents.
15.. KEY WORDS AND DOCUMENT ANALYSIS 15b. DESCRIPTORS
1B AVAILABILITY STATEMENT 17. SECURITY CLASSIFICATION IB. NUMBER OF PAGESITAJI repwo~
UnclassifiedUnlimited 19 SECURITY CLASSIFICATION 20 RICE
S
Compilation of Branch Technical Positions
Branch TechnicalPosition (BTP) No.
ASB 3-1(Formerly APCSB 3-1)
ASB 3-2*(Formerly MB 3-2)
ASB 9-1*
ASB 9-2
ASB 10-1
ASB 10-2
CMEB 9.5-1(Formerly ASB 9.5-1)
CSB 6-1
Title ofBTP
Protection Against Postulated PipingFailures in Fluid Systems OutsideContalnment"
'Tornado Design Classification"
"Overhead Handling Systems ForNuclear Power Plantsu
"Residual Decay Energy for Light-Water Reactors for Long-Term Cooling"
"Design Guidelines For AuxiliaryFeedwater System Pumps Drive andPower Supply Density For PWRs"
"Design Guidelines For Water Hamersin Steam Generators with Top FeedringDesigns"
"Guidelines For Fire Protection ForNuclear Power Plants"
BTPLocation
3.6.1
3.5.1.4
9.1.4
9.2.5
10.4.9
10.4.7
9.5.1
CSB 6-2*
CSB 6-3
CSB 6-4
CPB 4.3-1
ETSB 11-3
ETSB 11-5
"Minimum Containment Pressure ModelFor PWR ECCS Performance Evaluation"
"Control of Combustible Gas Concentra-tions In Containment Following a Lossof Coolant Accident"
"Determination of Bypass Leakage Pathsin Dual Containment Plants"
"Containment Purging During NormalPlant Operationsu
"Westinghouse Constant Axial OffsetControl (CAOC)"
"Design Guidance For Solid RadioactiveWaste Management Systems Installed InLight-Water-Coolant Nuclear ReactorPlants"
"Postulated Radioactive Releases Dueto a Waste Gas System Leak or Failure"
6.2.1.5
6.2.5
6.2.3
6.2.4
4.3
11.4
11.3
HGEB 1(Formerly HMB/GSB 1)
ICSB I
USafety-Related Permanent DewateringSystems"
"Backfitting of the Protection andEmergency Power Systems of NuclearPower Reactors'
"Isolation of Low Pressure SystemsFrom the High Pressure ReactorCoolant System"
'Requirements of Motor-Operated Valvesin the ECCS Accumulator Lines"
2.4.12
ICSB 3
Appendix 7-A
Appendix 7-A
Appendix 7-AICSB 4
-1.- Rev. 0 - July 1981
Branch TechnicalPosition (BTP) No.
ICSB 5*
ICSB 9*
ICSB 12
ICSB 13
ICSB 14
ICSB 16
ICSB 19
ICSB 20
ICSB 21
ICSB 22
ICSB 25*
ICSB 26
ICSB 2(PSB)
ICSB 4(PSB)
ICSB 8(PSB)
ics8 11(PSB)
ICSB 15(PSB)
ICSB 17(PSB)
ICSB 18(PSB)
ICSB 21
Title ofBTP
Scram Breaker Test Requirements-Technical Specifications"
"Definition of Use of ChannelCallbration-Technical Specification"
'Protection System Trip Point ChangesFor Operation with Reactor CoolantPumps Out of Service"
"Design Criteria for AuxiliaryFeedwater Systems"
'Spacious Withdrawal of Single ControlRods in Pressurized Water Reactors"
"Control Element Assembly (CEA)Interlocks in Combustion EngineeringReactors"
"Acceptability of Design Criteria ForHydrogen Mixing and Drywell VacuumRelief Systems"
"Design of Instrumentation andControls Provided to AccomplishChangeover From Injection toRecirculation Mode"
"Guidance For Application ofRegulatory Guide 1.47"
"Guidance For Application ofRegulatory Guide 1.22"
"Guidance For the Interpretation ofGeneral Design Criterion 37 For Testingthe Operability of the Emergency CoreCooling System as a Whole'
"Requirements for Reactor ProtectionSystem Anticipatory Trips"
"Diesel-Generator ReliabilityQualification Testing"
"Requirements on Motor-Operated Valvesin the ECCS Accumulator Linesu
"Use of Diesel-Generator Sets I-orPeaking"
"Stability of Offsite Power Systems"
"Reactor Coolant Pumps BreakerQualifications'
"Diesel-Generator Protective TripCircuit Bypasses"
"Application of the Single FailureCriterion to Manually ControlledElectrically-Operated Valves"
"Guidance For Application ofRegulatory Guide 1.47"
BTPLocation
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 7-A
Appendix 8-A
Appendix 8-A
Appendix 8-A
Appendix 8-A
Appendix 8-A
Appendix 8-A
Appendix 8-A
Appendix 8-A
-2- Rev. 0 - July 1981
Branch TechnicalPosition (BTP) No.
-MTEB 5-2
MTEB 5-3
MTEB 5.7*
MTEB 6-1
MEB 3-1
Title ofBTP
"Fracture Toughness Requirements"
"Monitoring of Secondary Side WaterChemistry In PWR Steam Generators"
"Material Selection and ProcessingGuidelines For BWR Coolant PressureBoundary Piping"
"PH For Emergency Coolant Water forPWPs",
"Postulated Rupture Locations In FluidSystem Piping Inside and OutsideContainments"
BTPLocation
5.3.2
5.4.2.1
5.2.3
6.1.1
3.6.2
PSB 1 "Adequacy of Shutdown ElectronicDistribution System Voltages"
PSB 2
RSB 3-1
"Criteria for Alarms and IndicatorsAssociated with Diesel-Generator UnitBypassed and Inoperable Status"
"Classification of Main SteamComponents Other than the ReactorCoolant Pressure Boundary For BWRPlants"
Appendix 8-A
Appendix 8-A
Appendix Ato 3.2.2
Appendix Bto 3.2.2
5.4.7
* 5.2.2
6.3
RSB 3-2
RSB 5-1
RSB 5-2
RSB 6-1
"Classification of BWR/6 Main Steamand Feedwater Components Other Thanthe Reactor Coolant Pressure Boundary"
"Design Requirements of the ResidualHeat Removal System"
"Overpressurization Protection ofPressurized Water Reactors WhileOperating at Low Temperatures"
"Piping From the RWST (or BWST) andContainment Sump(s) to the SafetyInjection Pumps'
mBTP has been superceeded.
-3- Rev. 0 - July 1981
