Standard Review Plan for the Review of

NUREG-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
NUREG-0800
(formerly issued as
NUFREG75/087)
Standard Review Plan
for the Review of
Safety Analysis Reports
for Nuclear, Power Plants
LWR Edition
(This June 1987 update includes all revisions
issued between July 1981 and June 1987.)
U.S. Nuclear Regulatory•
Commission
Office of Nuclear Reactor Regulation
June 1987
*"
",dC
.,wrlwIoe
)
INTRODUCTION
The
Standard
Plan (SRP)
prepared for
guidancesafety
of staff
reviewers
in the
Office Review
of Nuclear
Reactoris Regulation
in the
performing
reviews
of
applications to construct or operate nuclear power plants. The principal
purpose of the SRP is to assure .the quality and. uniformity of staff reviews
and to present a well-defined base from which to evaluate proposed changes in
the scope and requirements of reviews. It is also a purpose of the. SRP to
make information about regulatory matters widely available and to improve
communication and understanding of the staff review process by interested
members of the public and the nuclear power industry.
The safety review is primarily based on the information provided by an applicant
in a Safety Analysis Report (SAR).
Section 50.34 of 10 CFR 50 of the Commission's
regulations requires that each application for a construction permit for a
nuclear facility shall include a Preliminary Safety Analysis.Report (PSAR) and
that each application for a license to operate such a facility shall include a
Final Safety Analysis Report (FSAR).
The SAR must be sufficiently detailed to
permit the staff to determine whether the plant can be built and operated
without undue risk to the health and safety of the public. Prior to submission
of an SAR, an applicant should have designed and analyzed the plant in sufficient
detail to conclude that it can be built and operated safely. The SAR is the
principal document in which the applicant provides the information needed to
understand the basis upon which this conclusion'has been reached.
Section 50.34 specifies, in general terms, the information to be supplied in a
*SAR.
The specific information required by the staff for an evaluation of an
application is identified in Regulatory Guide 1.70, "Standard Format and
Content of Safety Analysis Reports for Nuclear Power Plants
-
LWR Edition."
The SRP sections are keyed to the Standard Format, and the SRP sections are
numbered according to the section numbers in the Standard Format. Review
plans have not been prepared for SAR sections that consist of background or
design data which are included for information or for use in the review of
other SAR sections.
The Standard Review Plan is written so as to cover a variety of site conditions
and plant designs. Each section is written to provide the complete procedure
and all*acceptance criteria for all of the areas of review pertinent to that
section. However, for any given application, the staff reviewers may select
and emphasize particular aspects of each SRP section as is appropriate for the
application. In some cases,.the major portion of the review of a plant feature
may be done on a generic basis with the designer of that feature rather than
in the context of reviews of particular applications from utilities. In other
cases a plant feature may be sufficiently similar to that of a previous plant
so that a de novo review of .the feature is not needed. For these and other
similar reasons, the staff may not carry out in detail all of the review steps
listed in each SRP section in the review of every application.
the
review,
the is.
address,
sections the
individual
The
accomplished,
the review
howperforms
review, who
for detail,
basis in
that areSRPreviewed,
matters
25
by
is
performed
review
safety
The
sought.
and the conclusions that are
primary branches. One of the objectives of the SRP'is to assign the reviewI
responsibilities to the various branches and to define the sometimes complex
interfaces between them. Each SRP section identifies the branch that has the
primary review responsibility for that section• In some review areas the
primary branch nay require support, and the branches that are assigned these
secondary 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 by
the branch having primary review responsibility. This subsection contains a
description of the systems, components, analyses, data, or other information
that is reviewed as part of the particular Safety Analysis Report .section in
question. It also contains a discussion of the information needed or the
review expected from other branches to permit the primary review branch to
complete its review.
I I.
Acceptance Criteria
This subsection contains a statement of the purpose of the review, an identifica -I
tion of which NRC requirements are applicable, and the technical basis forI
determining the acceptability of the design or the programs within the scope
of the area of review of the SRP section. The technical bases consist of
specific criteria such as NRC Regulatory Guides, GeneralIDesign Criteria,
Codes and Standards, Branch Technical Positions, and other criteria.
The technical bases for some sections of the SRP are provided in Branch Technical
Positions or Appendices which are included in the SRP. These documents typically
set forth the solutions and approaches determined to be acceptable in the past
by the staff in dealing with a specific safety problem or safety-related
design area. These solutions and approaches are codified in this form so that
staff reviewers can take uniform and well-understood positions as the same
safety problems arise in future cases. Some Branch Technical Positions and
Appendices may be converted into Regulatory Guides if it appears that this
step would aid the review process. Like Regulatory Guides, the Branch Technical Positions and Appendices represent solutions and approaches that are
acceptable to the staff, but they are not required as the only possible solutions and approaches. However, applicants should recognize that, as in the
case of Regulatory Guides, substantial time and effort on the part of the
staff has gone into the development of the Branch Technical Positions and
Appendices and that a corresponding amount of time and effort will probably be
required to review and accept new or different solutions and approaches.
Thus, applicants proposing solutions and approaches to safety problems or
safety-related design areas other than those described in the Branch Technical
Positions and Appendices must expect longer review times and more extensive
questioning in these areas. The staff is willing to consider proposals for
other solutions and approaches on a generic basis, apart from a specific
license application, to avoid the impact of the additional review time on
individual cases.
2
III. Review Procedures
This
subsection
discusses procedure
how the review
is accomplished.
The section
is
generally
a step-by-step
that the
reviewer goes through
to provide
reasonable verification that the applicable safety criteria have been met.
IV.
Evaluation Findings
This subsection presents the type of conclusion that is sought for the particular
review area. For each sectjon, a conclusion of this type is included in the
staff's Safety Evaluation Report in which the staff publishes the results of
their review. - The SER also contains a description of the review including
such 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 design
or the applicant's programs deviate from the criteria stated in the SRP; and
the 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 di~rected toward water-cooled reactor power
plants. Staff reviewers will adapt the SRP for use in the reviews of other
reactor types where .applicable.
The Standard Review Plans result from many years of experience by the staff in
establishing and using regulatory requirements in evaluating the safety of
nuclear power plants and in reviewing Safety Analysis Reports. A great deal
of progress has been made in the methods of review and in the development of
regulations, guides, and standards since the early years of review. This
Standard Review Plan may be considered a part of a continuing regulatory
standards development activity that not only documents current methods of
review but also provides, the base of orderly modifications of the review
process in the future.
In 1981, the Standard Review Plan was revised in entirety and published as
NUREG-0800. The revision program had three major objectives, i.e., to more
completely identify the NRC requirements that are germane to each review
topic, to more fully describe how the review effort determines satisfaction of
the requirement, and to incorporate the large number of new and revised
regulatory positions (primarily 111I-related) that had already been established.
To accomplish this and to conform to the revised NRR organization, some SRP
sections were added, deleted, split," and/or combined.
The SRP will be revised and updated periodically as the need arises to clarify
the content or correct errors and to incorporate modifications approved by the
Director of the Office of Nuclear Reactor Regulation. A revision number and
publication 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 as
needed, the revision numbers and dates will not be the same for all sections.
The Table of Contents indicates the revision numbers of the currently effective
sections. As necessary, corresponding changes to the Standard Format will
3
be considered
improvement
and suggestions
also
U.S. and
Regulation,
Reactorwill
Office offorNuclear
to the Director,
sent Commnents
shouldbe bemade.
Nuclear Regulatory Commission, Washington, DC 20555. Notices of errors or
omissions should also be sent to the same address.
4
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 Conimission (NRC)
has published a revision to
the "Table of Contents" of NUREG-0800, "Standard Review Plan for the Review
of SafetyAnalysis 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.
I
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, flUREG-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 TI NUCLEAR REGUL TRY COMMISSION
t' Edson G. Case, Acting Director
Office of Nuclear Reactor Regulation
2
STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS
TABLE OF CONTENTS
Issued
Year/Month
Applicable
Revision
SRP No.
-INTRODUCTION..............................................
1
75/11
81/?
-1
2
3
4
5
75/11
79/1
79/3
80/5
81/7
84/32
0
81/7
Table of Contents.........................................
Compilation of Branch Technical Positions ..............
CHAPTER 1
1.8
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
Interfaces for Standard Design ....
CHAPTER 2
Site Location and Description ....
2.1.2
Exclusion Area Authority and
Control...................................
2.1.3
Population Distribution .......
2.2.1-2.2.2
Identification of Potential Hazards
in Site Vicinity ..........
2.3.1
2.3.2
78/12
81/7
-1
2
75/11
78/7
81/7
-1
2
75/11
78/12
81/7
-1
2
75/11
78/12
81/7
-1
2
75/11
78/7
81/7
---
1
2
75/11
78/12
81/7
1
2
75/li
78/4
81/7
SITE CHARACTERISTICS
2.1.1
2.2.3
0
1
Evaluation of Potential Accidents
...
Regional Climatology .................---
75/11
Local Meteorology .....................--1
2
2.3.3
Onsi te Meteorological Measurements
programs............................
78/4
81/7
75/11
78/5
81/7
---
1
2
t
IRev. 5
-
December 1984
TABLE OF CONTENTS (Continued)
Issued
.Year/Month
Applicable
Revision
SRP No.
Appendix A......................
75/11
---
2.3.4
Short-Tern Diffusion Estimates For
Accidental Atmospheric Releases ....
2.3.5
Long-Tern Diffusion Estimates ....
1
2
78/5
8117
-1
75/11
81/7
---
75/11
78/5
81/7
1
2
2.4.1
75/11
Hydrologic Description ................---
78/6
.8117
1
2
Appendix A......................
75/111
---
78/6
81/7
1
2
2.4o.2
Floods..............................
2.4.3
Probable Maximt= Flood (PHF) on
Streams and Rivers .........
2.4.4
Potential Darn Failures................
75/11
---
1
2
7816
81/7
-1
2
75/11
78/6
81/7
---
75/11
78/6
8117
1
2
2.4.5
Probable Naximuma Surge and Seiche
Flooding ..........................
75/11
---
78/6
81/7
1
2
2.4.6
Probable Maximuw Tsunami Flooding
2.4.7
Ice Effects .........................
...
75/11
---
78/6
81/7
1
2
75/11
---
78/5
1
2
2.4.8
Cooling Mater Canals and
Reservoirs ...................
.....
.
81/7
75/111
---
1
2
78/6
81/7
2.4.9
Channel Diversions ..........
-1
2
75/11
78/6
81/7
2.4.10
Flood Protection Requirements ....
---
75/11.
78/5
81/7
1
2
2.4.11
Cooling Water Supply..................
2.4.12
Groundwater .........................
75/11
---
78/5
81/7
1
2
tt
75/fl
---
78/7
8117
1
2
iiRev. 5
-
December 1984
TABLE OF CONTENTS (Continued)
SRP 14o.
BTP HNB/GSB 1 ................
BTP HGEB 1 ...................
2.4.13
2.5.1
2.5.2(
Technical Specifications and
Emergency Operation
Requirements....................
Bastc Geologic and Seismic
Information.....................
1*
2
75/11
78/7
81/7
--1
2
75/117
78/5
81/7
-1
2
75/11
78/6
81/7
-1
2
75/ 11
78/11
81/7
1
75/11
81/7
-1
2
75/11
78/111
81/7
-1
2
-1
2
75/11
78/11
81/7
VibratoryGround Motion.......
2.5.3
Surface Faulting ..................
2.5.4
Stability of'SubSurface Materials
and Foundations .................
2.5.5
Issued
ear/Month
YE
Accidental Releases of Liquid
Effluents in Ground and Surface
Waters .........................
2.4.14
Applicable
Revision
Stability of Slopes..........
CHAPTER 3 DESIGN OF STRUCTURES
*COMPONENTS.
3.2.1
Seismic Classification .............
3.2.2
Syste. Quality Group
Classification ..................
75/11
78/11
81/7
EQUIPMENT, AND SYSTE MS
Appendix A (Formerly
8Th RSB 3-1) ...............
Appendix B (Formerly
BTP RSB 3-2) ...............
1
75/11
81/7
-1
75/11
81/7
-"
75/11
81/7
-1
75/11
81/7
...........
Appendix C ..........
None
81/7
1
Appendix P.....................
None
81/7
0
3.3.1
Wind Loadings.....................
-1
2
75/11
78/8
81/7
3.3.2
Tornado Loadings ..................
-I
2
75/11
78/8
81/7
I1t
iii
Rev. 5
-
December 1984
I
I
TABLE OF CONTENTS (Continued)
SRP No.
Applicable
Revision
Issued
Year/Month
3.4.1
Flood Protection ...........
-1
2
75/11
78/3
81/7
3.4.2
Analys~s Procedures .........
-1
2
75/il
None
-1
2
75/11
78/4
81/7
-1
2
7/11
78/8
81/7
-2.
2
75/111
78/7
-1
2
75/11
78/7
81/7
BTP NMB 3-2 ...........
--
75/n1
BTP ASB 3-2......................
2
81/7
-1
75/11
81/7
1
2
75/il
None
8117
3.5.1.1
Internally Generated Missiles
(Outside Contai nment) .......
3.5.1. 2
Internally Generated Mis siles
(Inside Containment) ........
3.5.1.3
Turbine Missiles ...........
3.5.1. 4
M!ssiles Generated by Natural
Phenomena...............................
Site Proximity Missiles (Except
Ai rcraft)...............................
3.5.1. 6
Aircraft Hazards.
3.5.2
Structures, Systems, and Components
to be Protected from Externally
.........
Generated Missiles..................
3.6.1
75/n1
---
78/3
81/7
1
2
75/11
a
81/
0
81/7
Plant Design for Protection Against
Po'stulated Piping Failures in
Fluid Systems Outside
Containment .......................
8TP ASB-3-1.....................
3.6.2
None
Barrier Design Procedures .............--Appendix A........................
Determinati on of Rupture Locations
and Dynamic Effects Associated
with the Postulated Rupture of
......
Piping ........
BTP MEB-3-1.....................
iv
.
81/7
1
3.5.1.5
3.5.3
81/7
75/11
---
81/7
1
75/11
---
1
81/7
-1
75/11
81/7
---
75/111
81/7
1
iv
Rev. 5
-
December 1.984
TABLE OF CONTENTS (Continued)
Applicable
Revision
SRP No.
3.7.1
Seismic Design Parameters ..........
3.7.2
Seismic System Analysis............
3.7.3
Seismic Subsystem Analysis .........
75/ 11
1
75/11
81/7
1
3.7.4
Seismic Instr~umentation............
3.8.1
Concrete Containment ..............
1
75/11
81/7
1
75/11
81/7
1
0
Appendix..............
3.8.2
Stee~l Containment .................
3.8.3
Concrete and Steel Internal
Structures of Steel or Concrete
Containments.....................
3.8.4
75/11
81/7
1
75/11
81/7
1
Other Seismic Category I
Structures......................
Appendix
Appendix
Appendix
Appendix
75/111
81./7
1
*0
0
0
0
A ......................
...........
B...........
C......................
0 .....................
3.8.5
Foundations.......................
3.9.1
Special Topics for M4echanical
Components •.....................
81/7
8117
8117
81/7
75/11
81/7
75/11
78/4
81/7
1
2
3.9.2
Dynamic Testing and Analysis of
Systems, Components, and
,...............
Equipment .......
75/11
78/8
81/7
1
2
3.9.3
Issued
Year/Month
ASME Code Class 1, 2, and 3
Components, Component Supports,
and Core Support Structures ....
75/11
Appendix A......................
3.9.4
Control Rod Drive Systems ..........
3.9.5
Reactor Pressure Vessel Internals
3.9.6
Inservtce Testing of P~ups and
Valves .........................
V
1
81/7
0
1
8117
84/4
1
2
75/11
8117
84/4
75/u1
...
78/4
' 8117
1
2
751/u
7814
81/7
1
2
VRev.
5
-
December 1984
TABLE OF CONTENTS (Continued)
SRP No.
3.10
3.11
Issued
Year/Month
1
2
75/11
78/4
8147
1
2
75/11
78/7
81/7
of Category I
Seismic Qualification
Instrunentation and Electrical
Equipment.......................
Environm~ental Design of Mechanical
and Electrical Equipment .........
CHAPTER 4
4.2
Applicable
Revision
REACTOR
Fuel System Design...................
Appendix A.........................
4.3
75/11
m.
1
2
0O
78/9
81/7
81/7
1
2
75/f1
78/4
81/7
--
75/u
1
2
78/4
81/7
Nuclear Design ................................
BIP CPB 4.3-1 ..........
4.4
--
Thermal and Hydraulic Design .....
75/111
81/7
-1"
75/11
81/7
Appendix .................................
1
4.5.1
Contro1 Rod Drive Structural
Materials.•..........................---
75/11
78/1
81/7
1 2
4.5.2
Reactor Internal and Core
Support Materials...................
75/11
78/1
81/7
---
1
2
4.8
Functional Design of Control Rod
Drive System ........
..............
75/111
---
81/7
1
CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS
5.2.1.1
Compliance with the Codes and Standard
Rule, 10 CFR § 50.55a...............
75/11
78/1
--1
81/7
2
5.2.1.2
Applicable Code Cases.................
75/111
---
78/1
81/7
1
2
5.2.2
Overpressure Protection .......
BTPRSB 5-2 ..........................
5.2.3
75/n1
-1
81/7
0
81/7
Reactor Cool ant Pressure Boundary
Materials .........................
vi
751/f
78/4
81/7
---
1
2
Rev. 5
-
December 1984
TABLE OF CONTENTS (Continued)
Applicable
Revision
SRP No.
BTP KTE8 5-7 ...........
-:1
2
75/11
78/4
81/7
Reactor Coolant Pressure Boundary
5.2.4
Inservice Inspection and Testing
...
75/111
---
1
81/7
Reactor Coolant Pressure Boundary
5.2.5
Leakage Detection ..................
75/11
---
1
81/7
Reactor Vessel Materials ...............---
5.3.1
75111
1
81/7
75/U1
Pressure-Temperature Limits ............---
5.3.2
1
BTP MTE8 5-2 ......................---
81/7
75/11
1
81/1
75/f1
Reactor Vessel Integrity ...............---
5.3.3
81/7
1
5.4
Preface .............................
75/11
---
1
81/7
Pump Flyw~heel Integrity (PWR) ..........-.--
5.4.1.1
75/11
1
81/7
75/11
Steam Generator Materials ..............---
5.4.2.1
1
2
MTEB 5-3 ....................
-BTP
78/11
81/7
75/11
---
1
2
5.4.2.2
78/111
81/7
Steam Generator Tube Inservice
Inspection.........................
75/11
---
81/7
1
5.4.6
Reactor Core Isolation Cooling
System (BWR) .......................
75/11
---
1
2
3
5.4.7
Residual Heat Reoval "(RHR) System
...
.....
BTP RSB 5-1 ......
78/3
81/7
84/4
75/11
---
1
2
3
78/8
81/7
84/4
-1
75/fl
78/8
81/7
---
75/11
'2
5.4.8
Reactor Water Cleanup System
(BWR) ............
.................
1
2
5.4.11
5.4.12
Issued
Year/Month
Pressurizer Relief Tank...............
Reactor Coolant System
High Point Vents....................
Vii
78/7
81/7
75/11
---
1
2
78/8
81/7
0
81/7
VII
Rev. 5
-
Deceadber 1984
TABLE OF CONTENTS (Continued)
Applicable
Revision
SRP No.
CHAPTER 6
6.1.1
ENGINEERED SAFETY FEATURES
Engineered Safety Features
Materials.......................
-1
*2
75/11
78/12
81/7
*
-1
2
75/11
78/12
81/7
*
1
2
75/11
78/12
81/7
1
2
75/111
78/4
81/7
~
1
2
75/11
718/
8117
-1
2
75/11
78/8
81/7
-1
2
6
75/11
78/5
78/8
79/2
81/7
83/1
84/8
0
1
2
0
79/2
81/7
83/1
83/1
1
2
75/11
78/8
81/7
1
75/11
81/7
1
75/11
81/7
BTP MTEB 6-1 .................
6.1,2
Protective Coating Systems
(Paints)
-
Organic Materials ....
6.2,1
Containment Functional Design .......
6. 2,.1.lA
PWR Dry Containments, Including
Subatmospheric Contai nuents .......
6. 2.1. 1.9
Ice Condenser Containments .........
6.2.1.1. C
Pressure-Suppression TypeBU
Contai nments ....................
4
Appendix I ...................
Appendix A ...................
Appendix B ...................
6.2.1.2
6.2.1.3
6.2.1.4
6.2.1.5
Issued
Year/Month
Subcoeipartment Analysis ............
Mass and Energy Release Analysis for
Postulated Loss-of-Coolant
Accidents.......................
Mass and Energy Release Analysis for
Postulated Secondary System Pipe
Ruptures........................
Hinimwn Containment Pressure
Analysis for Emergency Core
Cooling System Performnce
Capabilitty Studies ...............
BTP CSB 6-1 ...................
viii
75/n1
1
2
78/8
81/7
1
2
75/11
78/8
81/7
Rev. 5
-
D)ecenber 1984
I
I
TABLE OF CONTENTS (Continued)
SR
N.Applicable
Issued
Revision
_RPN___
6.2.2
6.2.3
Containment Heat Removal Systems
....
Year/Month
-1
2
3
75/11
---
75/11
78/4
78/8
81/7
Secondary Containment Functional
Design............................
3.
2
•BTP CSB 6-3 ...........
6.2.4
Containment Isolation System .....
BTP CSB 6-4 ...........
6.2.5
78/8
81/7
-1
2
75/fl.
7818
81/7
-1
2
75/11
78/5
81/7
-1
2
75/11
78/5
81/7
---
75/11
Combustible Gas Control in
Containment .......................
1
2
Appendix A. ....................
78/5
81/7
75/11
---
BTP CSB 6-2.....................
1
78/5
2
81/7
---
1
- ..
2
6.2.6
6.2.7
6.3
75/11
Fracture Prevention of Containment
Pressure Boundary...
............
Emergency Core Cooling System ....
Control Room Habitabli~ty Systems
1
2
78/9
81/7
0
81/7
75/11
---
*
6.4
81/7
Containment Leakage Testing ...........---
BTP RSB 6-1................ .....
...
1
2
--
1
81/7
84/4
"
75/U1
81/7
75/111
---
1
78/12
2
Appendix A......................
82/7
75/11
---
1
2
6.5.1
ESF Atmosphere Cleanup Systems ....
78/12
81/7
75/11
---
1
2
6.5.2
75/11
78/5
78/'7
81/7
Containment Sprey as a Fission
Product Cleanup System ..............---
75/11
1
tx
ixPRev. 5
83/7
-
D~ecember 1984
...
TABLE OF CONTENTS (Continued)
Applicable
Revision
SRP No.
6.5.3
6.5.4
6.6
6.7
Fission Product Control Systems
and Structures ...........
-1
2
75/11
78/7
81/7
---
75/11
Ice Condenser as a Fission Product,
Cleanup System.....................
Inservice Inspection of Class 2
and 3 Components ..........
M4ain Steam Isolation Valve Leakage
Control System (BUR)
...
....
Issued
Year/Month
1
2
78/4
81/7
-1
75/11
81/7
-1
•2
75/11
78/3
81/7
-1
2
3
75/li
78/7
81/7
84/2
CHAPTER 7 INSTRUMIENTATION AND CONTROLS
7.1
Instrumentation and Controls
-
Introduction ............
Table 7-1 Acceptance Criteria
and Guidelines for Instrumentation and Controls Systems
75/11
Important to Safety ............---
1
2
3
78/7
0
81/7
Appendix A.....................
0
1
81/7
84/2
Appendix B.....................
0
81/7
Table 7-2 THI Action Plan
Requirements for Instrumentation and Controls Systems
Important to Safety .............
7.2
Reactor Tri~p System..................
81/7
84/2
75/11
---
78/7
81/7
1
2
Appendix A ...........
7.3
Engineered Safety Features Systems
..
75/11
78/7
---
75/U1
81/7
78/7
81/7
1
2
Appendix A ...........
7.4
-1
2
Safe Shutdown Systems.................
-"75/11
1
2
78,/7
83/7
---
75/11
78/7
81/7
1
2
X
xRev. 5
-
December 1984
TABLE OF CONTENTS (Continued)
SRP No.
7.5
7.6
7.7
Applicable
Revision
Issued
Year/Honth
I
2
3
75/11
78/7
81/7
84/2
1
2
75/11
78/7
81/7
Information Systems Important to
Safety...........................
Interlock Systems Important to
Safety...........................
Control Systems ....................
75/11
78/7
1
2
3
Appendix 7-A
81/7
84/2
Branch Technical Positions (ICSB)..
BTP ICSB
BTP ICSB
BTP ICSB
BTP ICSB
BTP ICSB
1
2
75/11
78/7
81/7
1
2
75/11
78/7
81/7
1
2
75/11
78/7
81/7 -•
1
2
75/11
78/7
81/7
1
2
75/'11
78/7
81/7
1
2
75/11
78/7
81/7
1
2
75/11
78/7
81/7
I
2
75/11
78/7
81/7
1
2
75/11
78/7
81./7
1
2
75/.11
78/7
81/7
1
2
75/11
78/7
8117
1
2
75/11
78/7
81/7
1
2
75/11
78/7
81/7/
1 (DOR) .............
3 ...................
4 (PSB) ..............
5 ...................
9 ...................
BTP ICSB 12 ...................
BTP ICSB 13 ....................
BTP ICSB 14 ...................
BTP ICSB 16 ...................
BTP ICSB 19 ....................
BTP ICSB 20 ...................
BTP ICSB 21............
.......
x!
xl
Rev. 5
-
December 1984
TABLE OF CONTENTS (Continued)
Applicable
Revision
SRP No.
BTP I[CSB 22 :..........
-"75/11
1
2
BTP ICSB 25 ...........
-1
2
BTP ICSB 26 .............
Appendix 7-B
CHAPTER 8
8.1
Electric Power-Introduction .....
1
2
75/11
78/7
81/7
-1
75/11
81/7
-1
2
75/11
78/4
81/7
-1
2
75/11
78/4
81/7
75/11
Offsite Power System ...................---
78/4
81./7
1
2
3
0
Appendix A.........................
8.3.1
78/7
81/7
ELECTRIC POWER
Table 8-1 Acceptance Cr1trtra
and Guidlelines for Electcric
Power Systems .........
8.2
A-C Power Systems (Onsite) ............--Appendix...........................
8.3.2
D-C Poweer Systems (Onsit~e) ............-
Appendix BA
Branch Technical Positions (PSB)
83/7
83/7
75/11
78/5
81/7
81/7
1
2
2
75/fl.
78/4
81/7
1
2
BTP ICSB
....
2 (PSB)..............
75/11
---
78/6
81/7
1
2
75/12
---
78/6
81/7
1
2
BTP ICSB
78/7
81/7
75/n1
,.........---
General Agenda, Station Site
Visits....................................
Issued
Year/Mont~h
4 (PSB) ..................--
BTP ICSB 8 (P5B)................
BTP ICSB 11 (PSB)................
75/n1
1
2
78/6
81/7
-"'-
75/n1
78/6
81/7
1
2
75/111
---
78/6
1
2
BTP ICSB 15 (PSB)................
81/7
-"-"
75/11
1
2
78/5
8117
xiiRev. 5
-
Deceeber 1984
•
TABLE OF CONTENTS (Continued)
e ":"
]k •:"•
" :..."
"
"
l"
SSRP No.
BTP ICSB 17 (PSB) .................
DTP ICSB 18 (PSB) ........
BTP ICSB 21 (PSB) ........
BTP PSB 1.......................
BTP PSB 2...........
Appendix 8B
............
General Agenda, Station Site Vistts
Applicable
Revision
Issued
Year/Month
1
75i/6
2
81/7
-1
2
75/11
78/6
-1
2
75/11
0
81/7
0
81/7
0
81/7
81/7
78/6
81/7
CHAPTER 9 AUXILIARY SYSTEMS
9.1.1
9.1.2
9.1.3
9.1.4
New Fuel Storage ...........
Spent Fuel Storage
......
1
2
78/2
81/7
,-75/11
1.
S 2
3
78/3
79/3
81/7
1
81/7
Spent Fuel Pool Cooling and Cleanup
System....................................--75/11
Light Load Handling System (Related
to Refueli ng) ...........
-1
2
-,
BTP ASS 9"1
9.1.5
75/u1
--
.,.
...
75/11
78/4
81/7
75/11
.--
Overhead Heavy Load Handling
Systems..................
........
9.2.1
Statton Service Water System .....
9.2.2
Reac:tor Auxiliary Cooling Water
1.
2
78/4
81/7
0
81/7
-1
2
75/11
78/3
81/7
75/n1
Systems.................................
--
9.2.3
9.2.4
9.2.5
Demineral ized Water Makeup System
Potable and Sanitary W/ater Systems
Ultimate Heat Sink ..........
xttt
...
..
1
2
81/7
84/4
-1
2
75/fl
-1
2
75/11
-1
2
75/fl
78/3
81/7
xiii
Rev. 5
78/3
81/7
78/3
81/7
-
December 1984
TABLE OF CONTENTS (Continued)
SRP No.
Applicable
Revision
Issued
Year/Month
-1
2
75/11
78/3
81/'7
8TP ASB 9-2 ...........
9.2.6
CondensateStorage Facilities ....
-1
2
75/11
78/3
861/7
9.3.1
Compressed Air System ........
--
75/11
9.3.2
Process and Post-Accident Sampling
Systems ..............................
---
75/11
9.3.3
9.3.4
9.4.1
9.4.2
9.4.3
System ...........................
75/11
78/3
---
1
2
81/7
Chemical and Volume Control System
(PWR) (Including Bloron Recovery
751/U
---
78/3
81/7
1
2
Standby Liquid Control System
(BR) .............................
---
Control Room Area Ventilation
•....................
System .......
---
Spent Fuel. Pool Area Venti lati on
System ...............
---
75/11
78/3
81/7
1
2
Auxiliaryj and Radwastae Area
Ventilation System..................
Turbine Area Ventilation System ...
9.4.5
Engineered Safety Feature
Ventilation System ... ..............
Fire Protection Program .......
...
BTP CuEB 9.5.1 ..........
75/1l
78/3
81/7
1
2
9.4.4
9.5.1
81/7
Equipment and Floor Drainage
System) ...........................
9.3.5
78/7
1
2
. ....
75/11
78/3
81/7
1I
2
75/n1
---
78/3
81/7
S1
2
75/11
---
78/3
81/7
1
2
75/11
---
1
2
78/3
81/7
-1
2
3
75/11
76/5
78/3
81/7
--
7/
1
2
Appendix A......................
--
---
81/7
1
xfv
xiv
Rev. 5
78/3
81/7
76/11
-
December 1984
TABLE OF CONTE•S (Continued)
Applicable
Revision
SRP No.
9.5.2
Coamuni cartions Systes ........
9.5.3
Li g~titng Systems ...........
9.5.4
Emergency Diesel Engine Fuel Oil1
Storage and Transfer System ....
9.5.5
Emergency Diesel Engine Cooling
Water System ............
9.5.6
9.5.7
9.5.8
Emergency Diesel Engine Starting
System.................................
Emaergency Diesel Engine Lubrication
System.................................
Emergency Diesel Engine Combustion
Air Intake and Exhaust System ...
CHAPTER 10
Issued
Year/Month
-°75/fl
1
2
7814
81/7
-1
2
75/11
78/4
81/7
-1
2
75111
78/4
81/7
-1
75/11
78/4
2
81/7
-1
2
75/11
78/4
81/7
-1
2
75/11
78/4
81/7
-1
2
75/11
78/4
81/7
STEAM AND POW/ER CONVERSION SYSTEM
10.2
Turbine Generator ..........
-1
2
75/11
78/4
81/7
10.2.3
Turbine Disk Integrity ........
-1
75/11
81/7
10.3
Main Steam Supply System .......
-1
2
75/11
78/4
81/7
84/'4
3
10.3.6
Steam and Feedwater System
-°75/11
Materials...............................
1
2
10.4.1
Main Condensers ...........
10.4.2
Main Condenser Evacuation System
10.4.3
Turb~ne Gland Sealing System .....
10.4.4
Turbine Bypass System........
....
78/4
81/7
-1
2
75/11
78/4
81/7
-1.
2
75/11
78/7
81/7
-"75/11
1
2
78/7
81/7
-
75/11
78/4
1
281/7
XV
xvRev. 5
-
December 1.984
TABLE OF CONTENTS (Continued)
10.4.5
Circulating Water' System....... .....
10.4.6
Condensate Cleanup System ..........
10.4.7
Condensate and Feedvater System ...
DIP ASB 10-2 .......
10.4.8
10.4.9
1
2
75/11
78/3
81/7
1
2
75/ 11
78/3
81/7
75/n
1
2
78/3
81/7
84/4
1
2
3-
75/11
78/~3
81/7
84/4
1
2
75/111
78/7
•81/7
1
2
75/11
78/4
81/7
1
2
75/11
78/4
81/7
..........
Steam Generator Blowdown System
(pwR)
..........................
Auxiliary Feedwater System (PWR).
BTP ASB 10-1 .................
11.1
CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT
-..............
Source Terms ........
1
2
11.2
Liquid Waste Management Systems
11.3
Gaseous Waste 1Management Systems
....
Solid Waste Management Systems ....-
BTP ETSB U3-3 .................
75/lI
78/7
81/7
1..
-
BTP ETSB 11-5.......................
11.4
*Issued
.Year/Month
Applicable
Revision
SRP No.
75/11
78/7
81/7
75/111
-1
2
78/7
81/7
81/7
0
751/u1
1
78/7
81/7
-
75/11
78/7
81/7
8,1/7
Appendix 11.4-A..................C
11.5
Process and Effluent Radiological
Monitoring Instrumentation and
Sampling Systems ................
Appendix 11.5-A..................
xvi
-1
2
3
75/11
78/7
79/4
81/7
0
1
79/4
81/7
Rev. 5
-
December 1984
I
TABLE OF CONTENTS (Continued)
SRP No.
CHAPTER 12
12.1
12.2
12.3-12.4
1
2
75/11
78/5
81/7
1
2
75/11
78/5
81/7
2
75/11
78/5
81/7
ThatareOccupational
Radiation
Ecposures
As Low As Is
Reasonably Achievable ............
lAssuring
Radiation Sources .................
Radiation Protection Design
Features ........................
Dose Assessment ...................
12.5
Operational Radiation Protection
Program...........................
CHAPTER 13
13.1.2-13.1.3
Issued
Year/Month
RADIATION PROTECTION
12,4(1)
13.1.1
Applicable
,Revi sion
CONDUCT OF OPERATIONS
Management and Technical Support
Organization....................
Operating Organization .............
75/11
78/5
1
75/11
78/5
8)./7
1
2
75/11
79/4
81/7
1
2
75/11
79/4
81/7
13.1.3(2)
1323
13:.2.1
13.2.2
Qualifications of Nuclear Plant
Personnel.......... ........
.....
Training.
1R
1
2
7511
79/4
0
75/11
78/3
81/7
........................
Reactor Operator Training ..........
Training For Non-Licensed Plant
Staff ..........................
81/7
1
2
:13.3
Emergency Planning ................
13.4
Operational Review ................
1
2
1354
Plant Procedures ..................
1
2
81/7
75/11
78/3
81/7
75/11
79/2
81/7
81/7
(1)SRP Section has been combined with SRP Section 12.3.
(2)$RP 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 1.3.5.2.
jcvii
Rev. 5
-
December 1984
TABLE OF CONTENTS (Continued)
Applicable
Revision
SRP No.
Issued
Year/Month
13.5.1
Administration Procedures ..............
C
81/7
13.5.2
Operating and Maintenance
Procedures.........................
0
81/7
2
75/11
81/7
1
2
75/11
79/2
8117
1
2
75/11
7912
81/7
0
1
79/2
81/7
13.6
Physical Security .........
14.1
Initial Plant Test• Programs
14.2
Initial Plant Test Programs
CHAPTER 14
;........
INITIAL TEST PROGRAN
-
-
PSAR .
FSAR .
Standard Plant Designs, Initial Test
Program - Final Design Approval
(FDA) ..........................
14.3
CHAPTER 15
ACCIDENT ANALYSIS
Introduction .......................
15.0
75/n1
78/8
1
2
INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
15.1
15.1.1-15.1.4
81/7
Decrease in Feedwater Temperature,
Increase In Feedvater Flow,
Increase In steam Flow, and
Inadvertent Opening of a Steam
Generator Relief or Safety Valve .
1
75/11
81/7
2
75/11
78/8
81/7
1
2
75/11
78/8
81/7
* Steam System Piping Failures Inside
and Outside of Containment
;......................
(PR) .....
15.1.5
Appendix A ...............
15.2
15.2.1-15.2.5
DECREASE IN HEAT REMlOVAL BY THE SECONDARY SYSTEM
Loss of External Load, Turbine Trip,
Loss of Condenser Vacuum,
Closure of 1Main Steam Isolation
Valve (BWR), and Steam Pressure
Regulatory Failure (Closed) .......
75/11
81/7
1
15.2.6
15.2.7
to the Station Auxiliaries ....
Loss of Normal*Feedwater Flow....
xvfff
75/11
81/7
1
75/n1
81/7
1
xviii
Rev. 5
-
December 19B4
TABLE OF CONTENTS (Cont~inued)
Applicable
Revision
SRP No.
15.2.8
Feedwater System Pipe Breaks
Inside and Outside Contaitwment
(PWR) ..........................
75/11
81/7
1
DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE
15.3
15.3.1-3 .5.3.22 Loss of Forced Reactor Coolant Flow
Including Trip of Pump and FlowControl ler Malfunctions ..........
15.3.3-3 .5.3.44 Reactor Cool ant Pumip Rotor Seizure
and Reactor Coolant Punmp Shaft
Break .....
. ...
...
. ..
15.4
15.4.1
15.4.3
..
1
75/11
81/7
-1
2
75/111
78/8
81/7
REACTIVITY AND POWER DISTRIBUTION ANOMALIES
Uncontrolled Control Rod Assembly
*Withdrawal from a Subcritical
or Low Power Startup Condition
15.4.2
/
..
,.
1
2
75/111
78/4
81/7
1
2
75/11
78/4
81/7
1
2
75/111
78/4
81/7
1
83/7
1
75/11
81/7
Uncontrolled Control Rod Assembly
Withdrawal at Power .............
Control Rod Misoperation (System
Malfunction or Operator Error) ..
15.4,4-1 .5.4.5 Startup of an Inactive Loop or
Recirculation Loop at an Incorrect
Temperature, and Flow Controller
Malfunction Causing an Increase
in BWR Core Flow Rate.
....
15.4.6
15.4.7
15.4.8
Issued
Year/Month
75/11
Chemical and Volume Control System
Malfunction That Results in a
Decrease in the Boron Concentration in the Reactor Coolant
(PWR)...........................
Inadvertent Loading and Operation
of a Fuel Assembly in an
Improper Position ...............
Spectrum cif Rod Ejection
Accidents (PWR) .................
75/11
1
81/7
1
2
75/11
78/4
81/7
1
75/11
81/7
Appendix A ...................
xfx
xix
Rev. 5
-
December 1984
TASLE OF CONTENTS (Continued)
Appl icabl e
Revision
SRP No.
15.4.9
Spectrum of Rod Drop Accidents
(BR) ..........................
1
2
75/11
78/4
81/7
1
2
75/fl
78/4
8317
Appendix A....................
15.5.1-15.5.2
15.5 INCREASE IN REACTOR COOLANT INVENTORY
Inadvertent Operation o1f ECCS and
Chemical and Volume Control System
Malfunctilon That; Increases Reactor
-Coolant Inventory................
1
15.6
15.6.1
15.6.2
75/n1
81/7
DECREASE IN REACTOR COOLANT INVENTORY
Inadvertent Opening of a PWR.
Pressurizer Relief Valve
or a BWR Relief Valve ............
1
75/11
81/7
1
2
75/11
78/7
81/7
1
2
75/11
78/12
81/7
-2
75/11
78/7
81/7
-1
2
75/11
78/8
81/7
-1
751/n
Appendix B ...................
"
1
75/il
81/7
Appendix C ...................
-1
2
75/11
78/7
81/7
-I
75/11
81/7
Radiological Consequences of" t~he
Failure of Small Lines Carryring
Primary Coolant. Out~side
Conteanment. ....................
15.6.3
Radiological Consequences of' Steam
Generator Tube Failure (PR)....
15.6.4
Radiological Consequences of' Main
Steam Line Failure Outsidle
Cont~ainment, (BLWR)................
15.6.5
Issued
Year/Month
Loss-of-Coolant. Accident~s Resultilng
from Spectrum of Postulated
Piping Breaks Within the Reactor
Coolant. Pressure Boundary ........
..........
Appendix A .........
;....
Appendix D...............
XX
xxRev.
81/7
5
-
December 1984
TABLE OF CONTENTS (Continued)
Applitcabl e
SRP No.
15.7
15.7.1
15.7.2
15.7.3
15.7.4
15.7.5
Issued
Year/Month
Revision
RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COM4PONENT
Waste Gas System Fai lure .......
-1
75/11
8117
Radioactive Liquid Waste System
Leak or Failure (Release to
Atmosphere).............................
-1
75/11.
81/7
Postulated Radioactive Release Due
to Liquid-Containing Tank
Fail1ures .............................
Radiological Consequences of Fuel
Handling Accidents .........
Spent Fuel Cask Drop Accidents ....
..........
1
2
/1
7817
81/7
-1
75/11
8117
-1
2
75/fl.
75/12
81/7
1.5.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM.
•15.8
Anticipated Transients Without
Scram....................................
-1
75/1:1
81/7
-1
75/11
81/7
-1
75/1:1
81/7
-1
2
75/11
79/2.
81/7
-1
2.
75/11
79/2
83/7
0
1
81./7
Control Room.........................
0
84/9
Appendix A.........................
0
84/9
0
84112
0
84/12
Appendix..............................
CHAPTER 16 TECHNICAL SPECIFICATIONS
16.0
Technical Specifications .......
CHAPTER 17 QUALITY ASSURANCE
17.:1
17.2.
Q•uality Assurance During the Design
and Construction Phases ......
Quality Assurance During the
Operations Phase ..........
CHAPTER :18 HlU4AN FACTORS ENGINEERING
18.0
28.1
18.2Z
Hunan Factors Engineering/Standard
Review Plan Development ..............
Safety Parameter Display System ...
Appendix A..............
...........
xxt
xxi
Rev. 5
84/9
-
December :1984
POEM:a
NRmC
RGU'LATORY• CO'MISSION,
U.S NuCLEAR•
"'
I
REPORT NUMBER (Aw:.e,,w T,WC, ,dE WPIA!0*,P,,l
BIBLIOGRAPHIC DATA SHEET
tIUREG-OB00
4."RECIPIENT'S ACCEfSSION INUMBE-R"
3. TITLE AND SUITITLE
Standard Review Plan for the Review of Safety Analysis
Reports f~or Nuclear Power Plant, LWR Edition. Revision 5
_______________
D
ATEREPORTCOMPL'TED
,oNTEA
to SRP' Table of Cont~ents.
1984
December.
7. GATE REPORT ,ziSSUED
6. AUTNORISI
Jlanuary
"198s
I. PROWECTrITASKIWRK' UNIT NUMBER
8. PERFORMING ORGANIZATION NAME AND MAILING AGODRESS fl•cAld
Zi, cavdel
Office of Nuclear Reactor Regulations
U. S. Nuclear Regulatory Commission"
W~ashington, DC 20555
,o FN•UBE•R"
12.•I•.TYPEc OP REPORT -
SPONSORINO ORGANIZATION NAME AND MAILING ADDRES•S I/nmEu
II.
Office of Nuclear Reactor Regulations
U. S. Nuclear Regulatory Commission
Washington, DC 20555
13. SUPPLE:MENTARy MOTES
SRP Section (Guide)
,RIoD
PE,
covERDVe~i
a,...
...
SRP Table of Contents, R~evision 5
14 ABSTACTr 1200 -. ,,
, hi~s)
Revision 5 to SRP Table of Contents.
IS.
KEY WORDS AND DOCU•MENT ANALYSIS
16 AVAILABILITY STATEMENT
|160. OESCR:PTORS
I?. SECURITy CLASSIFIC.ATION
18, NUMBER OP' P•AGES
Unc'as~sified
Unl imi1ted
,9.SECURITY CLASSIF ICA.TION
I~e,
20 PRICE'
S
Compilation of Branch Technical Positions
Branch Technicsl
Position CBTP) No.,
Title of
BTP
ASB 3-1
(Formerly APCSB 3-1)
"Protection Against Postulated Piping
Failures in Fluid Systems Outside
Contatnmuent"
ASB 3-2k
(Formerly AAB 3-2)
."Tornado Design Classiftcation"
BTP
Locati on
3.6.1
3.5.1.4
ASB 9-1k
"Overhead Handling Systems For
Nluclear Power Plants"
9.1.4
ASB 9-2
"Residual Decay Energ;y for LightWater Reactors for Long-Term Cooling"
9.2.5
ASB 10-1
"Design Guidelines For Auxiliary
10.4.9
Feedwater System Pumps Drive and
Power Supply Density For PWRs"
ASB 10-2
CMEB 9o5-1
(Formerly ASB 9.5-1)
"Design Guidelines For Water Hanmers
in Steam Generators with Top Feedring
Designs".
"Guidelines For Fire Protection For
Nuclear Power Plants"
10.4.7
9.5.1
CSB 6-1L
"Minimum Containment Pressure Model
For PWR ECCS Performance Evaluation"
CS8 6-2*
"Control of Coelnastible Gas Concentra.tions In Containment Following a Loss
of Coolant Accident"
6.2.5
CSB 6-3
"Dletermination of Bypass Leakage Paths
in Dual Containment Plants"
6.2.3
CSB 6-4
"Containment Purging During Normal
Plant Operations".....
8.2.4
CPB 4.3-1
"Westinghouse. Constant Axial Offset
Control (CADC)"
ETSB 11-3
"Design Guidance For Solid Radioac-tive
Waste Management Systems Installed In
Light-Water-Cool ant Nuclear Reactor
Pl ants"
11.4
ETSB 11-5
"Postulated Radioactive Releases Due
to a Waste Gas Syst~em Leak or Failure"
11.3
HOEB 1
(Formerly HMB/GSB 1)
USafety-.Related Permanent Dewatering
Systems"
6.2.1.5
4.3
2.4.12
ICSB 1
"Backfitting of the Protection and
Emergency Power Systems of Nuclear
Powner Reactors"
Appendi x 7-A
ICSB 3
"Isolation of Low Pressure Systems
From the High Pressure Reactor
Cool ant System"
Appendix 7-A
ICSB 4
"Requirements of Motor-Operaeld Valves
in the ECCS Accumulator Lines"
Appendix 7-A
-1-
-1Rev. 0
-
July 1981
Branch Technical
Position (BTPI No.
Title of
ETP
BTP
Locati on
1CSB 5*
S$cram Blreaker Test RequirementsTechni ca1 Specifi cati ons"
Appendi x 7-A.
ICSB 9*
"Definition of" Use of Channel
Cal ibrati on-Technical Sped'fI cati on"
Appendix 7-A
XCSB 12
"Protection System Trip Point Changes
For Operation with Reactor Coolant
Pumps Out of Seriwce"
Appendix 7-A
ICSB 13
"Design Criteria for Auxiliary
Feedwater Systems"
Appendi x 7-A
!CSB 14
"Spacious Withdrawal of Single Control
Rods in Pressurized Water Reactors"
Appendix 7-A
ICSB 16
"Control Elemnent Assembly (CEA)
Interlocks in Combustion Engineering
Reactors"
Appendi x 7-A
ICSB 19
"Acceptability of Design Criteoria For
Hydrogen Mixing and Drywell Vacuum~
Relief Systems"
Appendix 7-A
zC5u 20
"Design of Instrinentation and
Controls Provided to Accomplish
Changeover From Injection to
Red rculation Made"
Appendix 7-A
ICSB 21
"Guidance For Application of
Regulatory Guide 1.47"
Appendix 7-A
ICSD 22
"Guidance For Application of
Regulatory Guide 1.22"
Appendix 7-A
ICSB 25*
"Guidance For the Interpretation of
General Design Criterion 37 For Testing
the Operability of the Emergency Core
Cooling System as a Whole"
Appendix 7-A
ICSB 26
"Requirements for Reactor Proltectton
System Anticipatory Trips"
Appendix 7-A
ICSB 2
(PSB)
"Diesel-Generator Reliability
Qualification Testing"
Appendix 8-A
ICSB 4
"Requirements on Motor-Operated Valves
in the ECCS Accumulator Lines"
Appendix 8-A
"Use of Diesel-Generator Sets i-or
Peaking"
Appendix 8-A
(PSO)
IC.san
"Stability of Otfsfite Power Systems"
•Appendix 8-A
"Reactor Coolant Pumps Breaker
Clua11 ficati ons"
Appendix 8-A
"Diesel-Generator Protective Trip
Circuit• Bypasses"
Appendix 8-A
uApplication of the Single Failure
Criterion to M4anually Controlled
Electri cally-Operated Valves"
Appendi x 8-A
(PSB)
ICSB 21
"Guidance For Application of
Regulatory Guide 1.47"
Appendix 8-A
(PSB)
ICSB 8
(PSB)
ICSB 15
(PsB)
ICSB 17
(P55)
ICSB 18
-2-
Rev. 0
-2-
-
Jluly 1981
Branch Technical
Position (BTP) No.
-MTEB 5-2
Title of
BTP
"Fracture Toughness Requirements"
BTP
Lo catio n
5.3.2
MTEB 5-3
"Monitoring of Secondary Side Water
Chemistry In PWR Steam Generators"
5.4.2.1
HTEB 5-7.*
"Material Selection and Processing
Guidelines For BWR Coolant Pressure
Boundary Piping"
5.2.3
MTEB 6-1
"PH For Emergency Coolant Water for
PWPs"
6.1.1
MEB 3-1
"Postulated Rupture Locations In Fluid
System Piping Inside and Outside
Contal nments"
3.6.2
PSB 1
"Adequacy of Shutdown Electronic
Distribution System Voltages"'
Appendix 8-A
PSB 2
"Criteria for Alarms and Indicators
Associated with Diesel-Generator Unit
Bypassed and Inoperable Status"
Appendix 8-A
RSB 3-1
"Classification of Main Steam
Components Other than the Reactor
Coolant Pressure Boundary For BWR
P1 ants"
Appendix A
to 3.2.2
RSB 3-2
"Classification of BWR/6 Main Steam
and Feedwater Components Other Than
the Reactor Coolant Pressure Boundary"
Appendix B
to 3.2.2
RSB 5-1
"Design Requirements of the Residual
Heat Removal System"
5.4.7
RSB 5-2
"Overpressurizatlon Protection of
Pressurized Water Reactors While
Operating at Low Temperatures"
5.2.2
RSB 6-1
"Piping From the RWST (or BWST) and
Containment SuWp(s) to the Safety
Injection Pumps"
6.3
MBTP has been superceeded.
-3-
-3Rev. 0
-
July 1981
NUREG-0800
.STANDARD REVIEW PLAN (SRP) FOR THE REVIEW OF
SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS
TABLE OF CONTENTS
SectionlRevision
Rev. 6
Rev. 2
Title "Date
Table of Contents
03/2007
•Introduction
03/2007
-
1.0
CHAPTER 1
Introduction and General Description of Plant
Introduction and Interfaces
03/2007
CHAPTER 2
Site Characteristics
2.0
Site Characteristics and Site Parameters
03/2007
2.1.1, Rev. 3
Site Location and Description
03/2007
2.1.2, Rev. 3
Exclusion Area Authority and Control
03/2007
2.1.3, Rev. 3
-Population Distribution
2.2.1-2.2.2, Rev. 3
Identification of Potential Hazards in Site Vicinity
03/2007
2.2.3, Rev. 3
Evaluation of Potential Accidents
03/2007
2.3.1, Rev. 3
Regional Climatology
03/2007
2.3.2, Rev. 3
Local Meteorology
03/2007
2.3.3, Rev. 3
Onsite Meteorological Measurements Programs
03/2007
2.3.4, Rev. 3
Short Term Atmospheric Dispersion Estimates for Accident Releases
03/2007
2.3.5, Rev. 3
Long-Term Atmospheric Dispersion Estimates for Routine Releases
03/2007
2.4.1, Rev. 3
Hydrologic Description
2.4.2, Rev. 4
Floods
03/2007
2.4.3, Rev. 4
Probable Maximum Flood (PMF) on Streams and Rivers
03/2007
2.4.4, Rev.3
Potential Dam Failures
03/2007
2.4.5, Rev. 3
Probable Maximum Surge and Seiche Flooding
03/2007
2.4.6, Rev. 3
Probable Maximum Tsunami Flooding
03/2007
2.4.7, Rev. 3
Ice Effects
03/2007
2.4.8, Rev. 3
Cooling Water Canals and Reservoirs
03/2007
2.4.9, Rev. 3
Channel Diversions
03/2007
2.4.10, Rev. 3
Flooding Protection Requirements
03/2007
2.4.11, Rev. 3
Low Water Considerations
03/2007
2.4.12, Rev. 3
Groundwater
03/2007
"03/2007
.03/2007
Table of Contents
-
Page 1Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Date
2.4.13, Rev. 3
Accidental Releases of Radioactive Liquid Effluents in Ground and Surface Waters
03/2007
2.4.14, Rev. 3
Technical Specifications and Emergency Operation Requirements
03/2007
2.5.1, Rev. 4
Basic Geologic and Seismic Information
03/2007
2.5.2, Rev. 4
Vibratory Ground Motion
03/2007
2.5.3, Rev. 4
Surface Faulting
03/2007
2.5.4, Rev. 3
Stability of Subsurface Materials and Foundations
03/2007
2.5.5, Rev. 3
Stability of Slopes
03/2007
CHAPTER 3
Design of Structures, Components, Equipment, and Systems
3.2.1, Rev. 2
Seismic Classification
03/2007
3.2.2, Rev. 2
System Quality Group Classification
03/2007
3.3.1, Rev. 3
Wind Loading
03/2007
3.3.2, Rev. 3
Tornado Loads
03/2007
3.4.1, Rev. 3
Internal Flood Protection for Onsite Equipment Failures
03/2007
3.4.2, Rev. 3
Analysis Procedures
03/2007
3.5.1.1, Rev. 3
Internally Generated Missiles (Outside Containment)
03/2007
3.5.1.2, Rev. 3
Internally Generated Missiles (Inside Containment)
03/2007
3.5.1.3, Rev. 3
Turbine Missiles
03/2007
3.5.1.4, Rev. 3
Missiles Generated by Tornadoes and Extreme Winds
03/2007
3.5.1.5, Rev. 4
Site Proximity Missiles (Except Aircraft)
03/2007
3.5.1.6, Rev. 3
Aircraft Hazards
03/2007
3.5.2, Rev. 3
Structures, Systems, and Components To Be Protected From Externally-Generated
Missiles
03/2007
3.5.3, Rev. 3
Barrier Design Procedures
03/2007
3.6.1, Rev. 3
Plant Design for Protection Against Postulated Piping Failures in Fluid Systems
Outside Containment
03/2007
3.6.2, Rev 2
Determination of Rupture Locations and Dynamic Effects Associated with the
Postulated Rupture of Piping
03/2007
3.6.3, Rev. I
Leak-Before-Break Evaluation Procedures
03/2007
3.7.1, Rev. 3
Seismic Design Parameters
03/2007
3.7.2, Rev. 3
Seismic System Analysis
03/2007
3.7.3, Rev. 3
Seismic Subsystem Analysis
03/2007
3.7.4, Rev. 2
Seismic Instrumentation
03/2007
3.8.1, Rev. 2
Concrete Containment
03/2007
Table of Contents
-
Page 2Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Date
3.8.2, Rev. 2
Steel Containment
03/2007
3.8.3, Rev. 2
Conlcrete and Steel Internal Structures of Steel or Concrete Containments
0312007
.3.8.4, Rev. 2
Other Seismic Category I Structures
03/2007
3.8.5, Rev. 2
Foundations
03/2007
3.9.1, Rev. 3
Special Topics for Mechanical Components
03/2007
3.9.2, Rev. 3
Dynamic Testing and Analysis of Systems, Structures, and C•omponents,
03/2007
3.9.3, Rev. 2
ASME Code Class 1, 2, and 3 Components, and Component Supports, and Core
Support Structures
03/2007
3.9.4, Rev. 3
Control Rod Drive Systems
03/2007
3.9.5, Rev. 3
Reactor Pressure Vessel Internals
03/2007
3.9.6, Rev. 3
Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves,
and Dynamic Restraints
03/2007
3.9.7
Risk-Informed Inservice Testing of Pumps and Valves
08/1 998
3.9.8
Risk-Informed Inservice Inspection of Piping
09/2003
3.10, Rev. 3
Seismic and Dynamic Qualification of Mechanical and Electrical Equipment
03/2007
3.11, Rev. 3
Environmental Qualification of Mechanical and Electrical Equipment
03/2007
3.12
ASME Code Class 1, 2, and 3 Piping Systems, Piping Components and Their
Associated Supports
03/2007
3.13
Threaded Fasteners
03/2007
Branch Technical
Position 3-1, Rev. 2
Classification of Main Steam Components Other Than the Reactor Coolant Pressure
Boundary for BWR Plants
03/2007
Branch Technical
Position 3-2, Rev. 2
Classification of Main Steam Components Other Than the Reactor Coolant Pressure
Boundary
03/2007
Branch Technical
Position 3-3, Rev. 3
Protection Against Postulated Piping Failures in Fluid Systems Outside Containment
03/2007
Branch Technical
Position 3-4, Rev. 2
Postulated Rupture Locations in Fluid System Piping Inside and Outside
Containment
03/2007
-
ASME Code Class 1, 2, and 3
CHAPTER 4
Reactor
4.2, Rev. 3
Fuel System Design
03/2007
4.3, Rev. 3
Nuclear Design
03/2007
4.4, Rev. 2
Thermal and Hydraulic Design
03/2007
4.5.1, Rev. 3
Control Rod Drive Structural Materials
03/2007
4.5.2, Rev. 3
Reactor Internal and Core Support Structure Materials
03/2007
4.6, Rev. 2
Functional Design of Control Rod Drive System
03/2007
Branch Technical
Position 4-1, Rev. 3
Westinghouse Constant Axial Offset Control (CAOC)
03/2007
Table of Contents
-
Page 3Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
CHAPTER 5
Reactor Coolant System and Connected Systems
5.2.1.1, Rev. 3
Compliance With the Codes and Standards Rule, 10 CFR 50.55a
03/2007
Applicable Code Cases
03/2007
5.2.2, Rev. 3
Overpressure Protection
03/2007
5.2.3, Rev. 3
Reactor Coolant Pressure Boundary Materials
03/2007
5.2.4, Rev. 2
Reactor Coolant Pressure Boundary Inservice Inspection and Testing
03/2007
5.2.5, Rev. 2
Reactor Coolant Pressure Boundary Leakage Detection
03/2007
5.3.1, Rev. 2
Reactor Vessel Materials
03/2007
5.3.2, Rev. 2
Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock
03/2007
5.3.3, Rev. 2
Reactor Vessel Integrity
03/2007
5.4, Rev. 2
Reactor Coolant System Component and Subsystem Design
03/2007
5.4.1.1, Rev. 2
Pump Flywheel Integrity (PWR)
03/2007
5.4.2.1, Rev. 3
Steam Generator Materials
03/2007
5.4.2.2, Rev. 2
Steam Generator Program
03/2007
5.4.6, Rev. 4
Reactor Core Isolation Cooling System (BWR)
03/2007
5.4.7, Rev. 4
Residual Heat Removal (RHR) System
03/2007
5.4.8, Rev. 3
Reactor Water Cleanup System (BWR)
03/2007
•5.4.11, Rev.•3
Pressurizer Relief Tank
03/2007
5.4.12, Rev. 1
Reactor Coolant System High Point Vents
03/2007
5.4.13
Isolation Condenser System (BWR)
03/2007
Branch Technical
Position 5-1, Rev. 3
Monitoring of Secondary Side Water Chemistry in PWR Steam Generator's
03/2007
Branch Technical
Position 5-2, Rev. 3
Overpressure Protection of Pressurized-Water Reactors While Operating at Low
Temperatures
03/2007
Branch Technical
Position 5-3, Rev. 2
Fracture Toughness Requirements
03/2007
Branch Technical
Position 5-4, Rev. 4
Design Requirements of the Residual.Heat Removal System
5.2.1.2, Rev. 3
-
.03/2007
CHAPTER 6
Engineered Safety Features
6.1.1, Rev. 2
Engineered Safety Features Materials
03/2007
6.1.2, Rev. 3
Protective Coating Systems (Paints) - Organic Materials
03/2007
6.2.1, Rev. 3
Containment Functional Design
03/2007
6.2.1.1 .A, Rev. 3
PWR Dry Containments, Including Subatmospheric Containments
03/2007
6.2.1.1 .B, Rev. 3
Ice Condenser Containments
04/1996
DRAFT
______
Table of Contents
-
Page 4Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Date
6.2.1.1 .C, Rev. 7
Pressure-Suppression Type BWR Containments
03/2007
6.2.1.2, Rev. 3
Subcompartment Analysis
03/2007
6.2.1.3, Rev. 3
Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents
(LOCAs)
03/2007
6.2.1.4, Rev. 2
Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures
6.2.1.5, Rev. 3
Minimum Containment Pressure Analysis for Emergency Core Cooling System
Performance Capability Studies
03/2007
6.2.2, Rev. 5
Containment Heat Removal Systems
03/2007
6.2.3, Rev. 3
Secondary Containment Functional Design
03/2007
6.2.4, Rev. 3
Containment Isolation System
03/2007
6.2.5, Rev. 3
Combustible Gas Control in Containment
03/2007
6.2.6, Rev. 3
Containment Leakage Testing
6.2.7, Rev. 1
Fracture Prevention of Containment Pressure Boundary
03/2007
6.3, Rev. 3
Emergency Core Cooling System
03/2007
6.4, Rev. 3
Control Room Habitability System
03/2007
6.5.1, Rev. 3
ESF Atmosphere Cleanup Systems
03/2007
6.5.2, Rev. 4
Containment Spray as a Fission Product Cleanup System
03/2007
6.5.3, Rev. 3
Fission Product Control SystemS and Structures
03/2007
6.5.4, Rev. 4
DRAFT
Ice Condenser as a Fission Product Cleanup System
04/1 996
6.5.5, Rev. 1
Pressure Suppression Pool as a Fission Product Cleanup System
03/2007
6.6, Rev. 2
Inservice Inspection and Testing of Class 2 and 3 Components
03/2007
*6.7, Rev. 3
DRAFT
Main Steam Isolation Valve Leakage Control System (BWR)
04/1996
Branch Technical
Position 6-1
pH For Emergency Coolant Water for Pressurized Water Reactors
03/2007
Branch Technical
Position 6-2, Rev. 3
Minimum Containment Pressure Model for PWR ECCS Performance Evaluation
03/2007
Branch Technical.
Determination of Bypass Leakage Paths in Dual Containment Plants
03/2007
Containment Purging During Normal Plant Operations
03/2007
.03/2007-
•03/2007
Position 6-3, Rev. 3
Branch Technical
..
Position 6-4, Rev. 3
Branch Technical
Position 6-5, Rev. 3
Currently the Responsibility of Reactor Systems Piping from the RWST (or BWST)
and Containment Surnp(s)to the Safety Injection Pumps
03/2007
CHAPTER 7
Instrumentation and Controls
7.0, Rev. 5
Instrumentation and Controls - Overview of Review Process
03/2007
7.0-A, Rev. 5
Review Process for Digital Instrumentation and Control Systems
03/2007
Table of Contents
-
Page 5Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Date
7.1, Rev. 5
Instrumentation and Controls
7.1-T, Second Rev. 5
Table 7-1 Regulatory Requirements, Acceptance Criteria, and Guidelines for
Instrumentation and Control Systems Important to Safety
03/2007
Appendix 7.1-A,
Second Rev. 5
Acceptance Criteria and Guidelines for Instrumentation and Control Systems
Important to Safety
03/2007
Appendix 7.1-B,
Rev. 5
Guidance for Evaluation of Conformance to IEEE Std. 279
03/2007
Appendix 7.1-C,
Rev 5
Guidance for Evaluation of Conformance to IEEE Std. 603
03/2007
Appendix 7.1-D
Second Issuance
Guidance for Evaluation of Conformance to IEEE Std. 7-4.3.2
03/2007
7.2, Rev. 5
Reactor Trip System
03/2007
-
introduction
03/2007
7.3, Rev. 5
.Engineered Safety Features Systems
03/2007
7.4, Rev. 5
Safe Shutdown Systems
03/2007
7.5, Rev. 5
Information Systems Important to Safety
03/2007
7.6, Rev. 5
Interlock Systems Important to Safety
03/2007
7.7, Rev. 5
Control Systems
03/2007
7.8, Rev. 5
Diverse Instrumentation and Control Systems
03/2007
7.9, Rev. 5
Data Communication Systems
03/2007
Appy:-5dxA
D
Apped,'•-x7.•B,
General Agenda, Station Site Visits
0
nICh]•.I
Th';llll
Appendix 7-A, Rev.
5, Branch
Technical
Positions (BTP)
(02/20/2007), has
been separated
into individual
sections.
rG1IJ;Lt•.sI
Re.v.-5
Appendix 7-A,
Rev. 5
Appehall× -C,
Rev:Appendix 7-B,
Rev. 5
Acronyms, Abbreviations, and Glossary
Branch Technical
Position 7-1, Rev. 5
Guidance on Isolation of Low-Pressure Systems From the High-Pressure Reactor
Coolant System
Branch Technical
Position 7-2, Rev. 5
Guidance on Requirements of Motor-Operated Valves in the Emergency Core
Cooling System Accumulator Lines
Branch Technical
Position 7-3, Rev. 5
Guidance on Protection System Trip Point Changes for Operation With Reactor
Coolant Pumps Out of Service
Branch Technical
Position 7-4,
Second Rev. 5
Guidance on Design Criteria for Auxiliary Feedwater Systems
Table of Contents
-
Page 6Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Branch Technical
Position 7-5, Rev. 5
Guidance on Spurious Withdrawals of Single Control Rods in Pressurized Water
Reactors
Branch Technical
Position 7-6, Rev. 5
Guidance on Design.of Instrumentation and Controls Provided to Accomplish
Changeover from Injection to Recirculation Mode
Branch Technical
Position 7-8, Rev. 5
Guidance for Application of Regulatory Guide 1.22
Branch Technical
Position 7-9, Rev. 5
Guidance on Requirements for Reactor Protection System Anticipatory Trips
Branch Technical
Position 7-10, Rev. 5
Guidance on Application of Regulatory Guide 1.97
Branch Technical
Position 7-11, Rev. 5
Guidance on Application and Qualification of Isolation Devices
Branch Technical
Position 7-12, Rev. 5
Guidance on Establishing and Maintaining Instrument Setpoints
"Branch Technical
Position 7-13, Rev. 5
Guidance on Cross-Calibration of Protection System Resistance Temperature
Detectors
Branch Technical
Position 7-14, Rev. 5
Guidance on Software Reviews for Digital Computer-Based Instrumentation and
Controls Systems
Branch Technical
Position 7-16
Withdrawn
Guidance on Level of Detail Required for Design Certification Applications Under 10
CFR Part 52
Branch Technical
Position 7-17, Rev. 5
Guidance on Self-Test and Surveillance Test Provisions
Branch Technical
Position 7-18, Rev. 5
Guidance on the Use of Programmable Logic Controllers in Digital Computer-Based
Instrumentation and Control Systems
Branch Technical
Position 7-19, Rev. 5
Guidance for Evaluation of Diversity and Defense-in-Depth in Digital
ComPuter-Based Instrumentation and Control Systems
Branch Technical
Position 7-21, Rev. 5
Guidance on Digital Computer Real-Time Performance
.
see ML070450253
CHAPTER 8
Electric Power
8.1, Rev. 3
Electric Power - Introduction
8.2, Rev. 4
Offsite Power System
8.3.1, Rev. 3
AC Power Systems (Onsite)
8.3.2, Rev. 3
DC Power Systems (Onsite)
8.4
Station Blackout
8-A, Rev. 1
General Agenda, Station Site Visits
Branch Technical
Position 8-1, Rev. 3
Requirements on Motor-Operated Valves in the ECCS Accumulator Lines
Branch Technical
Position 8-2, Rev. 3
Use of Diesel-Generator Sets for Peaking
Table of Contents
-
Page 7Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Date
Branch Technical
Position 8-3, Rev. 3
Stability of Offsite Power Systems
03/2007
Branch Technical
Position 8-4, Rev. 3
Application of the Single Failure Criterion to Manually Controlled Electrically
Operator Valves
03/2007
Branch Technical
Position 8-5, Rev. 3
Supplemental Guidance for Bypass and Inoperable Status Indication for Engineered
Safety Features Systems
03/2007
Branch Technical
Position 8-6, Rev. 3
Adequacy of Station Electric Distribution System Voltages
03/2007
Branch Technical
Position 8-7, Rev. 3
Criteria for Alarms and Indications Associated with Diesel-Generator Unit Bypassed
and Inoperable Status
03/2007
CHAPTER 9
Auxiliary Systems
9.1.1, Rev. 3
Criticality Safety of Fresh and Spent Fuel Storage and Handling
03/2007
9.1.2, Rev. 4
New and Spent Fuel Storage
03/2007
9.1.3, Rev. 2
Spent Fuel Pool Cooling and Cleanup System
03/2007
9.1.4, Rev. 3
Light Load Handling System (Related to Refueling)
03/2007
9.1.5, Rev. I
Overhead Heavy Load Handling Systems
03/2007
9.2.1, Rev. 5
Station Service Water System
03/2007
9.2.2, Rev. 4
Reactor Auxiliary Cooling Water Systems
03/2007
9.2.3 - Withdrawn
Demineralized Water Makeup System
see ML063320108.
9.2.4, Rev. 3
Potable and Sanitary Water Systems
03/2007
9.2.5, Rev. 3
Ultimate Heat Sink
03/2007
9.2.6, Rev. 3
Condensate Storage Facilities
03/2007
9.3.1, Rev. 2
Compressed Air System
03/2007.
9.3.2, Rev. 3
Process and Post-Accident Sampling Systems
03/2007
9.3.3, Rev. 3
Equipment and Floor Drainage System
03/2007
9.3.4, Rev. 3
Chemical and Volume Control System (PWVR) (Including Boron Recovery System)
03/2007
9.3.5, Rev. 3
Standby Liquid Control System (BWR)
03/2007
9.4.1, Rev. 3
Control Room Area Ventilation System
03/2007
9.4.2, Rev. 3
Spent Fuel Pool Area Ventilation System
03/2007
9.4.3, Rev. 3
Auxiliary and Radwaste Area Ventilation System
03/2007
9.4.4, Rev. 3
Turbine Area Ventilation System
03/2007
9.4.5, Rev. 3
Engineered Safety Feature Ventilation System
03/2007
9.5.1, Rev. 5
Fire Protection Program
03/2007
9.5.2, Rev. 3
Communications Systems
03/2007
9.5.3,. Rev. 3
Lighting Systems
032007
Table of Contents
-
Page 8Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Title
Date
9.5.4, Rev. 3
Emergency Diesel Engine Fuel Oil Storage and Transfer System
03/2007
9.5.5, Rev. 3
Emergency Diesel Engine Cooling Water System
03/2007
9.5.6, Rev. 3
Emergency Diesel Engine Starting System
03/2007
9.5.7, Rev. 3
Emergency Diesel Engine Lubrication System
03/2007
9.5.8, Rev. 3
Emergency Diesel Engine Combustion Air Intake and Exhaust System
03/2007
CHAPTER 10
Steam and Power Conversion System
.10.2, Rev. 3
Turbine Generator
03/2007
10.2.3, Rev. 2
Turbine Rotor Integrity
03/2007
10.3, Rev. 4
Main Steam Supply System
03/2007
10.3.6, Rev. 3
Steam and Feedwater System Materials
03/2007
10.4.1, Rev. 3
Main Condensers
03/2007
10.4.2, Rev. 3
Main Condenser Evacuation System
03/2007
10.4.3, Rev. 3
Turbine Gland Sealing System
03/2007
10.4.4, Rev. 3
Turbine Bypass System
03/2007
10.4.5, Rev. 3
Circulating Water System
03/2007
10.4.6, Rev. 3
Condensate Cleanup System
03/2007
10.4.7, Rev. 4
Condensate and Feedwater System
03/2007
10.4.8, Rev. 3
Steam Generator Blowdown System
03/2007
10.4.9, Rev. 3
Auxiliary Feedwater System (PWR)
03/2007
Branch Technical
Position, 10-1, Rev. 3
Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply
Diversity for Pressurized Water Reactor Plants
03/2007
Branch Technical
Position, 10-2, Rev. 4
Design Guidelines for Avoiding Water Hammers in Steam Generators
03/2007
CHAPTER 11
Radioactive Waste Management
11.1, Rev. 3
Source Terms
03/2007
11.2, Rev. 3
Liquid Waste Management System
03/2007
11.3, Rev. 3
Gaseous Waste Management System
03/2007
11.4, Rev. 3
Solid Waste Management System
03/2007
11.5, Rev. 4
Process and Effluent Radiological Monitoring Instrumentation and Sampling
Systems
03/2007
Branch Technical
Position 11-3, Rev. 3
Design Guidance for Solid Radioactive Waste Management Systems Installed in
Light-Water-Cooled Nuclear Power Reactor Plants
03/2007
Branch Technical
Position 11-5, Rev. 3
Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure
03/2007
Table of Contents
-
Page 9Reion6-Mrh20
Revision 6 - March 2007
Section/Revision
Branch Technical
Position 11-6
Title
Date
Postulated Radioactive Releases Due to Liquid-Containing Tank Failures
03/2007
CHAPTER 12
Radiation Protection
12.1, Rev. 3
Assuring that Occupational Radiation Exposures Are As Low as is Reasonably
Achievable
03/2007
12.2, Rev. 3
RadiationSources
0312007
12.3-12.4, Rev. 3
Radiation Protection Design Features
03/2007
12.5, Rev. 3
Operational Radiation Protection Program
03/2007
CHAPTER 13
Conduct of Operations
13.1.1, Rev. 5
Management and Technical Support Organization
03/2007
13.1.2-13.1.3, Rev: 6
Operating Organization
03/2007
13.2.1, Rev. 3
Reactor Operator Requalification Program; Reactor Operator Training
03/2007
13.2.2, Rev. 3
Non-Licensed Plant Staff Training
03/2007
13.3, Rev. 3
Emergency Planning
03/2007
13.4, Rev. 3
Operational Programs
03/2007
13.5.1.1
Administrative Procedures
-
General
03/2007
Administrative Procedures - Initial Test Program
113.5.1.2
DRAFT
(Content subsumed into SRP Section 14.2)
13.5.2.1, Rev. 1
Operating and Emergency Operating Procedures
*13.5.2.2
DRAFT
03/2007
Maintenance and Other Operating Procedures
(Content subsumed into SRP Section 17.5)
13.6
Physical Security
03/2007
13.6.1
Physical Security - Combined License
03/2007
*13.6.2
Physical Security - Design Certification
03/2007
.13.6.3
Security - Early Site Permit
*Physical
03/2007
CHAPTER 14
Initial Test Program and iTAAC-Design Certification
14.2, Rev. 3
Initial Plant Test Program - Design Certification and New License Applicants
14.2.1
Generic Guidelines for Extended Power Uprate Testing Programs
08/2006
14.3
Inspections, Tests, Analyses, and Acceptance Criteria
03/20_0_77
[Reserved]
03/2007
03/2007
14.3:1
..
14.3.2
Structural and Systems Engineering
-
Inspections, Tests, Analyses, and
.03/2007
Acceptance Criteria
14.3.3
.
Piping Systems and Components - Inspections, Tests, AnalyseS, and Acceptance
03/20_07
Criteria
Table of Contents
-
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Revision 6 - March 2007
Section/Revision
Title
.14.3.4
14.3.5
Date
Reactor Systems - inspections, Tests, Analyses, and Acceptance Criteria
03/2007
Instrumentation and Controls
03/2007
-
Inspections, Tests, Analyses, and Acceptance
Criteria
14.3.6
Electrical Systems - Inspections, Tests, Analyses, and Acceptance Criteria
03/2007
14.3.7
Plant Systems - Inspections, Tests, Analyses, and• Acceptance Criteria
03/2007
14.3.8
Radiation Protection Inspections, Tests, Analyses, and Acceptance Criteria
03/2007
14.3.9
Factors Engineering - Inspections, Tests, Analyses, and Acceptance Criteria
.Human
14.3.10
Initial Test Program and D-RAP - Inspections, Tests, Analyses, and Acceptance
03/2007
0312007
Criteria
14.3.11
Containment Systems and Severe Accidents
-
Inspections, Tests, Analyses, and
03/2007
Acceptance Criteria
14.3.12
Physical Security Hardware
-
Inspections, Tests, Analyses, and Acceptance Criteria
03/2007
CHAPTER 15
Accident Analysis
15.0, Rev. 3
Introduction
15.0.1
Radiological Consequence Analyses Using Alternate Source Terms
07/2000
15.0.2
Review of Transient and Accident Analysis Methods
01/2006
15.0.3
Design Basis Accidents Radiological Consequence Analyses for Advanced Light
Water Reactors
03/2007
15.1.1
-
15.!.4, Rev. 2
S
-
Transient and Accident Analyses
03/2007
Decr~ease in Feedwater Temperature, Increase in Feedwater" Flow, Increase in
Steam Flow, and Inadvertent OPening of a Steam Generator Relief or Safety Valve
03/2007
15.1.5, Rev. 3
Steam System Piping Failures Inside and Outside of Containment (PWR)
03/2007
15.1.5.A, Rev. 2
Radiological Consequences of Main Steam Line Failures O~utside Containment of a
PWR
07/1981
15.2.1-15.2.5, Rev. 2
Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; .Closure of Main
Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)
03/2007
________
15.2.6, Rev. 2
Loss of Non-Emergency AC Power to the Station Auxiliaries
15.2.7, Rev. 2
Loss of Normal Feedwater Flow
15.2.8, Rev. 2
Feedwater System Pipe Breaks Inside and Outside Containment (PWR)
03/2007
15.3.1-15.3.2, R~ev. 2
Lossof Forced Reactor Coolant Flow Including Trip of Pump Moto)r and Flow
03/2007
03/2007
.03/2007
Controller Malfunctions
15.3.3-15.3.4, Rev. 3
15.4.1, Rev. 3
Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break
03/2007
Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power
03/2007
Startup Condition
15.4.2, Rev. 3
Uncontrolled Control Rod Assembly Withdrawal at Power
03/2007
15.4.3, Rev. 3
Control Rod Misoperation (System Malfunction or Operator Error)
03/2007
15.4.4-15.4.5, Rev. 2
Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and
Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate
03/2007
Table of Contents
-
Page 11Reion6-Mrh20
Revision.6 - March 2007
Section/Revision
Title
Date
15.4.6, Rev. 2
Inadvertent Decrease in Boron Concentration inl the Reactor Coolant (PWR)
03/2007
15.4.7, Rev. 2
Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position
03/2007
15.4.8, Rev. 3
Spectrum of Rod Ejection Accidents (PWR)
03/2007
15.4.8.A, Rev. 2
Radiological Consequences of a Control Rod Ejection Accident (PWR)
07/1 981
15.4.9, Rev. 3
Spectrum of Rod Drop Accidents (BWR)
03/12007
15.4.9.A, Rev. 2
Radiological Consequences of Control Rod Drop Accident (BWR)
07/1 981
15.5.1-15.5.2, Rev. 2
Inadvertent Operation of ECCS and Chemical and Volume Control System
Malfunction that Increases Reactor Coolant Inventory
03/2007
15.6.1, Rev. 2
Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve or a BWR
Pressure Relief Valve
03/2007
15.6.2, Rev. 2
Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant
Outside Containment
07/1981
15.6.3, Rev. 2
Radiological Consequences of Steam Generator Tube Failure (PwR)
07/1 981
15.6.4, Rev. 2
Radiological Consequences of Main Steam Line Failure Outside Containment
07/1 981
______________(BWR)______
15.6.5, Rev. 3
Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks
Within the Reactor Coolant Pressure Boundary.
03/2007
...
15.6.5.A, Rev. 2
Radiological Consequences of a Design Basis Loss-of-Coolant Accident Including
Containment Leakage Contribution
07/1981
1 5.6.5.B, Rev. 2
Radiological Consequences of a Design Basis Loss-of-Coolant Accident Leakage
Engineered Safety Feature Components Outside Containment
07/1981
15.6.5.D, Rev. 2
Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Leakage
From Main Steam Isolation Valve Leakage Control System (BWR)
07/1981
15.7.3, Rev. 2
Postulated Radioactive Releases Due to Liquid-Containing Tank Failures (content of
this section has been relocated to BTP 11-6)
07/1 981
15.7.4, Rev. 2
Radiological Consequences of Fuel Handling Accidents
07/11981
15.7.5, Rev. 2
Spent Fuel Cask Drop Accidents
07/1981
15.8, Rev. 2
Anticipated Transients Without Scram
03/2007
15.9
Boiling Water Reactor Stability
03/2007
~From
CHAPTER 16
Technical Specifications
16.0, Rev. 2
16.1, Rev. I1
17.1, Rev. 2
17.2, Rev. 2
[Technical Specifications
Risk-Informed Decision Making: Technical Specifications
[Quality
CHAPTER 17
Quality Assurance
Assurance During the Design and Construction Phases
Quality Assurance During the Operations Phase
Table of Contents
-
Page 12Reion6-Mrh20
If
03/2007
j
03/2007
If
07/1981
j
07/1981
Revision 6 - March 2007
Section/Revision
17.3
17.4
..
Title
Date
Quality Assurance Program Description
07/1981
Reliability Assurance Program (RAP)
03/2007
17.5
Quality Assurance Program Description
New License Applicants
17.6
Maintenance Rule
-
Design Certification, Early Site Permit and
03/2007
03/2007
CHAPTER 18
Human Factors Engineering
18.0, Rev. 2
Human Factors Engineering
03/2007
CHAPTER 19
Severe Accidents
19.0, Rev. 2
Probabilistic Risk Assessment and Severe Accident Evaluation
19.1, Rev. 2
Determining the Technical Adequacy of Probabilistic Risk Assessment Results for
Risk-Informed
19.2
Review of Risk Information Used to Support Permanent Plant-Specific Changes to
the Licensing Basis: General Guidance
Table of Contents
-
Page 13Reion6-Mrh20
Revision 6 - March 2007
NUREG-0800
U.S.
INT
DUCLERRGLTOR
OMSION
Purpose of the Standard Review Plan
The Standard Review Plan (SRP) provides guidance to US Nuclear Regulatory Commission
(NRC) staff in performing safety reviews of construction permit (CP) or operating license (OL)
applications (including requests for amendments) under 10 CFR Part 50 and early site permit
(ESP), design certification (DC), combined license (COL), standard design approval (SDA), or
manufacturing license (ML) applications under 10 CER Part 52 (including requests for
amendments).
The principal purpose of the SRP is to assure the quality and uniformity of staff safety reviews.
It is also the intent of this plan to make information about regulatory matters widely available and
to improve communication between the NRC, interested members of the public, and the nuclear
power industry, thereby increasing understanding of the NRC's review process.
Background
The NRC first issued the SRP in 1975 as NUREG-75/087. It was developed from many years
of NRC experience in establishing safety requirements and staff experience in applying those
requirements in evaluating the safety of various designs for nuclear facilities. NRR Office Letter
No. 2, dated August 12, 1975, established the SRP as a routine tool for the NRC staff to use in
evaluating the safety of nuclear power, plant designs. Specifically, that office letter described the
SRP as representing "the integrated result of the hundreds of conscious choices made by the
staff and by the nuclear industry in developing design criteria and design requirements for nuclear
power plants" and "the most definitive basis available for specifying the NRC's interpretation of an
acceptable level of safety for lig ht-water reactor facilities."
Following an extensive revision program, the NRC reissued the SRP as NUREG-0800 in
July 1981. This revision identified all NRC requirements that were relevant to each review
topic; described how a reviewer would determine that safety requirements had been met; and
incorporated a number of newly established regulatory positions, including those related to the
Three Mile Island (TMI) Action Plan.
In 1991, the NRC established the Standard Review Plan Update and Development Program
(SRP-UDP) to update NUREG-0800 for use in reviewing future reactor design applications. The
staff subsequently issued an "Implementing Procedures Document (IPD)," NUREG-1447, in
May 1992 to describe the SRP-UDP and establish procedures for updating the SRP. This update
reflected the experience of the safety reviews conducted on design certification applications for
evolutionary nuclear power plant designs. The SRP-UDP resulted in a draft revision to the SRP
-1-
-1Revision 2 - March 2007
in 1996. NRC staff used acceptance criteria and procedures introduced in the 1996 draft in
reviewing license amendment applications and new applications submitted under 10 CFR Part 52,
provided that the changes embodied in it were based on new regulations or regulatory guidance
approved• through other means. In addition, new SRP sections issued as part of the 1996 draft
were used as the primary means to evaluate new applications submitted under 10 CFR Part 52
(e.g., Section 14.3, "Inspections, Tests, Analyses, and Acceptance. Criteria - Design Certification")
since these sections represented the only guidance available for the given review area. Applicants
under 10 CFR Part 52, however, were not required to address these new SRP sections in their
applications.
In 2005, the Commission directed the staff to revise applicable sections of the NUREG-0800,
other guidance documents and office procedures to ensure up-to-date guidance would be
available for the next generations of staff that would be responsible for reviewing and licensing
.new sites and new reactors. The staff was to develop an integrated and continuing plan for
updating licensing review guidance and provide the plan, along with a schedule for completion,
to the Commission. "Briefing of Status of New Site and Reactor Licensing," (M050406) Staff
Requirements Memorandum dated May 10, 2005 (ML051 300673). The staff response to this
SRM is contained in SECY-06-0019, "Semiannual Update of the Status of New Reactor Licensing
Activities and Future Planning for New Reactors," dated January 31, 2006. In the next semiannualupdate, SECY-06-0187 dated August 25, 2006, the staff informed the Commission that they had
accelerated the SRP schedule to March 2007.
Some of the changes incorporated into this revision include:
*
*
*
•
*
*
*
*
*
*
extended applicability to 10 CFR 52 licensing processes;
technical rationale was developed and added to each SRP section to provide a basis for
the acceptance criteria;
assigning the review responsibilities by function,, with-the responsible organizations
maintained separately from the SRP itself to minimize impacts of office reorganizations,
consistent format applied to each section
NRC metrication policy was implemented;
resolution of unresolved safety issues (USIs) and generic safety issues (GSls) were
incorporated within the applicable sections;
consideration of operating experience insights from Generic Letters and Bulletins was
incorporated within the applicable sections;
TM! Action Plan requirements 1 were reconciled;
where applicable,, staff affirmed the technical accuracy of the draft SRP issued in 1996; and
staff renumbered branch technical positions to remove dated branch acronyms.
Changes to specific sections are detailed at the end of this Introduction.
NRR Office Instruction LIC-200, "Standard Review Plan Process," was used as the guiding
document in performing the March 2007 revision to the SRP.
'For 10 CF~R Part 50 applicants not listed in 10 CFR 50.34(f), "Additional TMI-related
requirements," the applicable provisions of 10 CFR 50.34(f) will be made a requirement during
the licensing process.
-2-
-2Revision 2 - March 2007
Objectives of the SRP
The SRP is intended to be a comprehensive and integrated document that provides the reviewer
with guidance that describes methods or approaches that the staff has found acceptable for
meeting NRC requirements. Implementation of the criteria and guidelines contained in the SRP by
staff members in their review of applications provides assurance that a given design will comply
with NRC regulations and provide adequate protection of the public health and safety. The SRP
also makes the staff's review guidance for licensing nuclear power plants publicly available and is
intended to improve industry and public stakeholder understanding of-the staff review process. It
should be noted that the SRP is not a substitute for NRC regulations, and compliance with the
SRP is not required.
In addition to documenting current methods of review, the SRP provides a basis for orderly
modification of the review process. The NRC disseminates information regarding current safety
issues and proposed solutions through various means, such as generic communications and the
process for treating generic safety issues. When current issues are resolved, it is necessary to
determine the need, extent and nature of revision that should be made to the SRP to reflect new
NRC guidance.
The staff should use the SRP as superseded or supplemented by new or revised regulations,
regulatory guidance, staff analyses of previous applications, and other published staff positions
to perform its review of a power reactor operating license application and a proposed change to
an existing operating license under 10 CFR Part 50, or a new reactor license application under
10 CFR Part 52.
Scope of Review of License Applications (Initial Applications and Amendments)
Because the staff's review constitutes an independent audit of the applicant's analysis, the staff
may emphasize or de-emphasize particular aspects of an SRP section, as appropriate, for the
application being reviewed. Prior to the initiation of a review, the technical branch chief and
assigned reviewer establish the scope and depth of the review to be performed, including the use
of acceptahnce criteria and review guidelines to be used. In some cases, the staff may propose
justification for not performing certain reviews called for by the SRP. These areas of increased
or decreased emphasis are acceptable, if the reviewer has management approval and documents
the scope and depth of the review in the SER. Examples of acceptablevariations in the scope of
a review include reduced emphasis on design reviews that the design and its underlying conditions
of acceptance are identical to that of another unit that was recently reviewed and approved or
increased emphasis on certain aspects of the design review as a result of recent operating
experience or consideration of unique design features that are not addressed in the SRP. Riskinsights can also be used in determining the depth of review. The staff should generally limit its
review of a proposed amendment to an existing operating license to those parts of the SRP that
are directly affected by the proposed change.
The SRP will provide pertinent review guidance to the staff for review of new license applications
submitted under 10 CFR Part 52. This will include ESP, DC, COL, SDA, and ML applications.
The SRP sections applicable to a COL application for a new light-water reactor (LWR) are based
on Regulatory Guide (RG) 1.206, "Combined License Applications for Nuclear Power Plants
(LWR Edition)." The SRP sections applicable to an ESP and a DC application are based on the
site-related sections and design-related sections of RG 1.206. Furthermore, RG 1.206 delineates
different content based on whether the COL application references an ESP, a DC, both or neither.
-3-
-3Revision 2 - March 2007
In general, review of a SDA or a ML application will be similar to that of a DC.
The SRP was originally written for 10 CFR Part 50 license applications. For DC and COL
applications submitted under 10 CFR Part 52, the level of design information reviewed should be
consistent with that of a final safety analysis report (FSAR) submitted in an OL application.
However, verification that the as-built facility conforms to the approved design is performed
through the inspections, tests, analyses, and acceptance criteria (ITACC) verification process.
For the review of COL applications, specific sections of the SRP will be used to-review operational
programs. The review will be performed consistent with guidance contained in SECY-05-01 97,
"Review of Operational Programs in a Combined License Application and Generic Emergency
Planning Inspections, Tests, Analyses, and Acceptance Criteria," and the related SRM dated
February 22, 2006. Consistent with this guidance, the staff will review and obtain a reasonable
assurance~finding on the program and its implementation schedule. In addition, .the staff will
include a license condition on subsequent implementation milestones for each program for which
specific implementation requirements are not specified in the regulations. In lieu of the
implementation schedule the applicant may propose inspections, tests, analyses, and acceptance
criteria for the program.
Deviation from the SRP by Applicants
Because the SRP generally describes an acceptable means of meeting the regulations, but not
necessarily the only means, applications may deviate from the acceptance criteria in the SRP.
On March 10, 1982, the Commission approved 10 CFR 50.34(g), "Conformance with the Standard
Review Plan (SRP)." 10 CFR 50.34(g) was subsequently renumbered as 10 CFR 50.34(h).
Specifically, § 50.34(h) requires applications for light water cooled nuclear power plant operating
licenses docketed after May 17, 1982, to include an evaluation of the facility against the SRP in
effect on May 17, 1982 or .the SRP revision in effect six months prior to the docket date of the
application, whichever is later. The evaluation must include an identification and description of all
differences in design features, analytical techniques, and procedural measures proposed for a
facility and those corresponding features, techniques, and measures given in the SRP acceptance
criteria. Where such a difference exists, the evaluation shall discuss how the alternative proposed
provides an acceptable method of complying with those rules or regulations of the Commission, or
portions thereof, that underlie the corresponding SRP acceptance criteria. Similar provisions are
in 10 CFR Part 52 contents of application sections of the different license processes contained in
the Subparts to 10 CFR Part 52. Staff guidance for reviewing the applicant's evaluation is
'contained in SRP Chapter 1.0, "Introduction and Interfaces."
The General Design Criteria (GDC) do not apply to the plants, that received construction .permits
(CPs) before 1971. For these plants, the Principal Design Criteria (PDC) in the CP, which are
discussed in the FSAR, apply. For amendment requests for plants to which the GDC do not apply,
the review should follow the SRP in light of applicable plant-specific PDC. In addition, certain
identified SRP acceptance Criteria are not readily applicable to new light-water reactor designs that
use simplified, passive, or other innovative means to accomplish their safety functions.
-4-
-4Revision
2 - March 2007
Organization of SRP.
Each SRP section is organized as follows:
Review Responsibilities: This subsection identifies the primary and, as applicable,
secondary review functions. The organizational review responsibilities are maintained
separate from the SRP.
I. Areas of Review
The areas of review subsection describes the scope of review by the branch having primary
responsibility for the identified functional area. Specifically, this subsection contains a
description of the systems, components, analyses, data, or other information that is reviewed
as part of the particular Safety Analysis Report (SAR) section. It also contains a ;discussion
of the information needed or the review expected from other branches to permit the primary
review branch to complete its review, as well as a list of applicable interfacing sections.
I1.
Acceptance Criteria
The acceptance criteria subsection identifies the applicable NRC requirements including
specific regulations, NRC orders, and industry codes and-standards referenced by
regulations. Note, NRC orders are temporal in nature and are not applied to applicants.
NRC orders are imposed when an applicant is issued a license.
For new reactor license applications submitted under 10 CFR Part 52, the applicant is also
required to address the proposed technical resolution of unresolved safety issues (USIs) and
medium- and high-priority generic safety issues (GSls) that are identified in the version of
NUREG-0933 current on the date 6 months before application and that are technically
relevant to the design, TMI requirements, and relevant operating experience 2. These
requirements are not identified within specific SRP sections, rather, these requirements are
identified within SRP Chapter 1., "Introduction and Interfaces." An applicant will tabulate
information within Chapter 1, but will address the technical issues to satisfy the requirements
within the specific sections, themselves.
This subsection also identifies the regulatory guidance which th~e staff has determined to
provide an acceptable approach for satisfying the applicable requirements (i.e., SRP
acceptance criteria). The types of guidance documents include but are not limited to:
Regulatory Guides, Commission policy as described in SECY papers and corresponding"
Staff Requirement Memoranda, NRC-approved or endorsed industry codes and standards,
certain technical reports (e.g., NUREGs and topical reports and corresponding safety
evaluations), and Branch Technical Positions (BTPs), which are provided as appendices to
the SRP. BTPs typically set forth Solutions and approaches previously determined to be
acceptable by the staff in dealing with a similar safety or design matter. These solutions and
approaches are documented in this form so that staff reviewers can take uniform and
2 Consideration
of operating experience for design certification applications only is
currently addressed in a SRM, dated February 15, 1991, on SECY-90-377, "Requirements for
Design Certification under 10 CFR Part 52."
-5-
-5Revision 2 - March 2007
well-understood positions as similar matters arise in the review of other applications. Each
SRP section explicitly states that the SRP is not a substitute for the NRC's regulations, and
compliance with them is not required. However, applicants are required to identify
differences from the SRP acceptance criteria and evaluate how the proposed alternatives to
the SRP acceptance criterialprovide an acceptable method of complying with the NRC's
regulations.
Lastly, this subsection also contains, as necessary, the technical bases for applicability of the
requirements to the subject areas of review or relationship of regulatory guidance to the
associated requirement.
Ill.
Review Procedures
This subsection discusses how the review is accomplished. The subsection is a step-by-step
procedure to be implemented by the reviewer to obtain reasonable assurance that the
applicable regulatory requirements have been met. These review procedures are based
on the identified SRP acceptance criteria. For deviations from these speciflc acceptance
criteria, the staff should review the applicant's evaluation of how the proposed alternatives
to the SRP criteria provide an acceptable method of complying with the relevant NRC
requirements identified in specific review areas.
For new reactor license applications submitted under 10 CFR Part 52, this subsection should
address staff review procedures for how operating experience insights identified in generic
letters and bulletins or equivalent international operating experience has been incorporated
into the plant design.
IV. Evaluation Findings
This subsection presents the type of conclusion that is sought for the particular review area.
•For each SRP section, the staff's conclusion is incorporated into a published Safety
Evaluation Report (SER). The SER describes the review and the aspects of the review the
staff emphasized, and identifies (1).the changes the applicant made to the application,
(2) the .matters addressed by additional information, (3) the matters for which additional
information is expected to be forthcoming, (4) the matters remaining unresolved, and
(5) deviations from the SRP in design and operational programs, and the bases for the
acceptability of such deviations. The SER also clearly identifies any requested exemptions
from the regulations and the staff's basis for its determinations on these requests.
V.
Implementation
This subsection provides guidance to applicants and licensees regarding the NRC's plans
for using the SRP section. 10 CFR 50.34(h) and similar provisions in 10 CFR Part 52
require each application to include an evaluation of the facility against the SRP of record6 months prior to docketing, including all differences between the design features,_analytical
techniques, and procedural measures proposed for a facility and those in the SRP
acceptance criteria.
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-6Revision 2 - March 2007
While the applicant's evaluation is performed against the SRP in effect 6 months prior to the
docket date of the application, the NRC staff will use the SRP in effect at the time of the
application review.
VI.
References
This subsection lists the references used in the review process.
Maintenance of the SRP
The SRP will be revised and updated periodically as the need arises to clarify the content or
correct errors and to incorporate modifications approved by the Director of the Office of Nuclear
Reactor Regulation or the Director of the Office of New Reactors.
A revision number and publication 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 as needed, the
revision'numbers and dates will not be the same for all sections. As necessary, corresponding
changes to the RG 1.206 will also be made. Comments may be submitted electronically by email
to NRR SRP(~nrc.,qov. Notices of errors or omissions should also be sent to the same address.
Comment resolution will be addressed in subsequent SRP revisions. Prior to revision to individual
sections, comment resolution may establish a basis for how alternatives to the N UREG-0800
acceptance criteria provide an acceptable method of complying with the NRC's regulations.
Specific Changes to SRP Sections
New Sections
-
in March 2007
* SRP Chapter 1, "Introduction and Interfaces"
-this
incorporates guidance previously
contained in SRP Section 1.8;
*
SRP Section 2.0, "Site Characteristics and Site Parameters;"
_• SRP Section 3.12, "ASME Code Class 1, 2, and 3 Piping Systems and Associated Supports
Design;"
*
SRP Section 3.13, "Threaded Fasteners -ASME Code Class 1, 2, and 3;"
*
SRP Section 5.4.13, "Isolation Condenser System (BWR);"
*
Appendix 7.1-D, "Guidance for Evaluation of Conformance to IEEE Std. 7-4.3.2;"
*
SRP Section 8.4, "Station Blackout;"
*
Branch Technical Position (BTP) 11-6, "Postulated Radioactive Releases Due to
.Liquid-Containing Tank Failures;"
*
SRP Section 13.4, "Operational Programs;"
*
*
*
SRP Section 13.6.1, "Physical Security - Combined License;"
SRP Section 13.6.2, "Physical Security - Design Certification;"
SRP Section 13.6.3, "Physical Security - Early Site Permit;"
*
SRP Section 14.3 and associated subsections on inspections, Tests, Analyses, and
Acceptance Criteria;
* SRP Section 15.0.3, "Design Basis Accident Radiological Consequence Analyses for
Advanced Light Water Reactors;"
*
SRP Section 15.9, ~"BWR Core Stability;"
• SRP Section 17.4, "Reliability Assurance Program;"
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-7Revision
2 - March 2007
* SRP
Section 17.5, "Quality Assurance Program Description
Permit and New License Applicants;"
-
Design Certification, Early Site
* SRP Section 17.6, "Maintenance Rule"
Reorganization of Content
Several sections have reorganized review content to better align with functional review
responsibilities:
* SRP Section 9.1.'1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling" and
SRP Section 9.1.2, "New and Spent Fuel-Storage" - the content was reorganized from new
.fuel in SRP Section 9.1.1 and spent fuel in SRP Section 9.1.2 to have criticality issues
addressed in one SRP section and the other review topics in the other section.
Sections not revised in March 2007:
SRP sections that are considered current for intended application:
*
•
*
SRP Section 3.9.7, "Risk-Informed Inservice Testing of Pumps.and Valves," August 1998;
SRP Section 3.9.8, "Risk-Informed Inservice Inspection of Piping," September 2003;
SRP Section 14.2.1, "Generic Guidelines for Extended Power Uprate Testing Programs,"
August 2006;
* SRP Section 15.0.1, "Radiological Consequence Analyses Using Alternate Source
Terms," July 2000;
* SRP Section 15.0.2, "Review of Transient and Accident Analysis Methods," January 2006
Guidance relocated:
* Branch Technical Position 7-16, "Guidance on Level of Detail Required for Design
Certification Applications Under 10 CFR Part 52," see Regulatory Guide 1.206;
* SRP Section 13.5.1.2, "Administrative Procedures - Initial ,Test Program," see SRP
Section 14.2;
* SRP Section 13.5.2.2, "Maintenance and Other Operating Procedures," see SRP
Section 17.5;
* SRP Section 15.1 .5.A, "Radiological Consequences of Main Steam Line Failures Outside
Containment of a PWR," see SRP Section 15.0.3;
* SRP Section 15.4.9.A, "Radiological Consequences of Control Rod Drop Accident
(BWR)," see SRP Section 15.0.3;
* SRP Section 15.6.2, "Radiological Consequences of the Failure of Small Lines Carrying
Primary Coolant Outside Containment," see SRP Section 15.0.3;
* SRP Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure
(PWR)," see SRP Section 15.0.3;
* SRP Section 15.6.4, "Radiological Consequences of Main Steam Line Failure Outside
Containment (BWR)," see SRP Section 15.0.3;
* SRP Section 15.6.5.A, "Radiological Consequences of a Design Basis Loss-of-Coolant
Accident Including Containment Leakage Contribution," see SRP Section 15.0.3;
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-8Revision 2 - March 2007
•. SRP Section 15.6.5.8, "Radiological Consequences of a Design Basis Loss-of-Coolant
Accident Leakage From Engineered Safety Feature Components Outside Containment,".
see SRP Section 15.0.3;
* SRP Section 15.6.5.0, "Radiological Consequences Of a Design Basis Loss-of-Coolant
Accident: Leakage From Main Steam Isolation Valve Leakage Control System (BWR),"
see SRP Section 15.0.3;
* SRP Section 15.7.3, "Postulated Radioactive Releases Due to Liquid-Containing Tank
Failures," see Branch Technical Position 11-6.
* SRP Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents," see
SRP Section 15.0.3;
*
SRP Section 15.7.5, "Spent Fuel Cask Drop Accidents," see SRP Section 15.0.3;
*
SRP Section 17.1, "Quality Assurance During the Design and Construction Phases,"
see SRP Section 17.5;
* SRP Section 17.2, "Quality Assurance During the Operations Phase," see SRP
Section 17.5;
* SRP Section 17.3, "Quality Assurance Program Description," see SRP Section 17.5.
SRP Sections not necessary for intended applications:
*
*
*
SRP Section 6.2.1.1.B, "Ice Condenser Containments;"
SRP Section 6.5.4, "Ice Condenser as a Fission Product Cleanup System;"
SRP Section 6.7, "Main Steam Isolation Valve Leakage Control System (BWR)."
SRP Sections withdrawn
*
SRP Section 9.2.3, "Demineralized Water Makeup System"
The March 2007 SRP revision is located in ADAMS. The package accession number is
ML070660 036.
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-9Revision 2 - March 2007
NUREG-0800
"m't,
•
_
~STANDARD
U.S. NUCLEAR REGULATORY COMMISSION
REVIEW PLAN
INTRODUCTION
-
PART 2
Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants: Light-Water Small Modular Reactor Edition
PURPOSE OF THE STANDARD REVIEW PLAN
The Standard Review Plan (SRP) provides guidance to U.S. Nuclear Regulatory Commission
(NRC) staff in performing safety reviews of light-water nuclear reactor power plants. The SRP
scope includes construction permit (CP) or operating license (OL) applications (including
requests for amendments) submitted under Title 10 of the Code of FederalRegulations
(10 CFR) Part 50. The scope also includes applications for early site permits (ESP), design
certifications (DC), 'combined licenses (COL), standard design approvals (SDA), or
manufacturing licenses (ML) under 10 CFR Part 52 (including requests for amendments).
The principal purpose of the SRP is to assure the quality and uniformity of staff safety reviews.
It is also the intent of this plan to make information about regulatory matters transparent, widely
available, and to improve communication between the NRC, interested members of the Public,
and the~nuclear power industry, thereby increasing understanding of the NRC review process.
Revision 0 - January 20.14
USNRC STANDARD REVIEW PLAN
This Standard Review Plan (SRP), NUREG-0800, has been prepared to establish criteria that the U.S. Nuclear Regulatory
Commission (NRC) staff responsible for the review of applications to construct and operate nuclear power plants intends to use in
evaluating whether an applicant/licensee meets the NRC regulations. The SRP is not a substitute for the NRC regulations, and
compliance with it is not required. However, an applicant is required to identify differences between the design features, analytical
techniques, and procedural measures proposed for its facility and the SRP acceptance criteria and evaluate how the proposed
alternatives to the SRP acceptance criteria provide an acceptable method of complying with the NRC regulations.
The SRP sections are numbered in accordance with corresponding sections in Regulatory Guide (RG) 1.70, "Standard Format and
Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition)." Not all sections of RG 1.70 have a corresponding
review plan section. The SRP sections applicable to a combined license application for a new light-water reactor (LWR) are based
on RG 1.206, "Combined License Applications for Nuclear Power Plants (LWREdition)."
These documents are made available to the public as part of the NRC policy to inform the nuclear industry and the general public of
regulatory procedures and policies. Individual sections of NUREG-0800 will be revised periodically, as appropriate, to
accommodate comments and to reflect new information and experience. Comments may be submitted electronically by e-mail to
NRO SRP.Resource(i.nrc.qov
Requests for single copies of SRP sections (which may be reproduced) should be made to the U.S. Nuclear Regulatory"
Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by
email to DlSTRlBUTlON•,nrc..qov. Electronic copies of this section are available through the NRC public Web site at
http:llwww.nrc..qovlreadinq-rm/doc-collectionsfnureqs/staff/sr0800/, or in the NRC Agencywide DocumentsAccess and
Management System (ADAMS) at http:llwww.nrc.,qov/readinq-rmladams.html, under Accession # ML13207A315.
This part of the SRP Introduction describes and incorporates the review philosophy and
framework to be applied by the staff for new light-water Small Modular Reactor (SMR)
applications made under 10 CFR Part 52, and incorporates staff commitments made in
SECY-1 1-0024 (see Background).
In current terminology, SMRs may be either light-water or nonlight-water designs, with an
electrical generation capacity of 300 MWe or less per module. The 300 MWe classification is
consistent with the International Atomic Energy Agency (IAEA) definition used for small and
medium sized reactors ("SMR" in IAEA terminology) found in IAEA-TECDOC-999 and other
IAEA publications. For the purposes of this NUREG, an SMR is a light-water power reactor
design, with the same electrical generating capacity limitation per module described above.
Nonlight-water designs are not included in this revision of the SRP introduction.
This SMR review framework (the "framework") is distinct from the approach used for non-SMR
applications and license amendments; but it satisfies the same SRP purposes described above.
Incorporation of this framework in the SRP does not change NRC requirements for applications
or applicants.
Applicants 1 are no~t required to engage~with the NRC in the pre-application activities described
herein. Submittals by applicants that choose not to engage the NRC in pre-application activities
associated with development of a Design-Specific Review standard-(DSRS) will be reviewed by
the staff using current SRP guidance and methods rather than using a DSRS in the riskinformed and integrated review framework discussed in this part of the SRP Introduction.•
However, it is the staff's belief that early-engagement with the NRC as described in this review
framework will positively benefit all review process stakeholders. The extent of benefits realized
will depend directly on the depth and timing of pre-application engagement by applicants. All
applicants are encouraged to engage the NRC in pre-application coordination, regardless-of the
application review methodology chosen.
A summary Of the changes in Revision 0 appears on the last page of the document.
Backaqround
The NRC first issued the SRP in 1975 as NUREG-75/087. Itwas developed from many years
of Atomic Energy Commission experience in establishing safety requirements and staff
experience in applying those requirements in evaluating the safety of various designs for
nuclear facilities. The Office of Nuclear Reactor Regulation (NRR), in Office Letter No. 2 dated
August 12, 1975, established the SRP as a routine tool for the NRC staff to use in evaluating
the safety of nuclear power plant designs. Specifically, that Office Letter described the SRP as
representing "the integrated result of the hundreds of conscious choices made by the staff and
by the nuclear industry in developing design criteria and design requirements for nuclear power
plants" and "the most definitive basis available for specifying the NRC's interpretation of an
acceptable level of safety for light-water reactor facilities. "
Following an extensive revision program, the NRC reissued the SRP as NUREG-0800 in
July 1981. This revision identified all NRC requirements that were relevant to each review topic,
described how a reviewer Would determine that safety requirements had been met, and
]For convenience throughout the introduction, the term "applicant" also includes entities interested in engaging the NRC in
pre-application activities which may lead to application under 10 CFR Part 52.
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-2Revision 0 - January 2014
incorporated a number of newly established regulatory positions, including those related to the
Three Mile Island (TMI) Action Plan.
In 1991, the NRC established the Standard Review Plan Update and Development Program
(SRP-UDP) to update NUREG-0800 for use in reviewing future reactor design applications. The
staff subsequently issued an "Implementing Procedures Document (IPD)," NUREG-1447, in
May 1992 to describe the SRP-UDP and establish procedures for updating the SRP. This
update reflected the experience of the safety reviews conducted on design certification
applications for evolutionary nuclear power plant designs. The SRP-UDP resulted in a draft
revision to the SRP in 1996. The NRC staff used acceptance criteria and procedures
introduced in the 1996 draft in reviewing license amendment applications and new applications
submitted under 10 CFR Part 52, provided that the changes embodied in it were based on new
regulations or regulatory guidance approved through other means. In addition, new SRP
sections issued as part of the 1996 draft were used as the primary means to evaluate new
applications submitted under 10 CFR Part 52 (e.g., Section 14.3, "Inspections, Tests, Analyses,
and Acceptance Criteria,) since these sections represented the only guidance available for the
given review area.
In 2005, the Commission directed the staff to revise applicable sections of NUREG-0800, other
guidance documents, and office procedures to ensure up-to-date guidance would be available
for staff responsible for reviewing and licensing new sites and new reactors. The staff was to
develop an integrated and continuing plan for updating licensing review guidance and provide
the plan, along with a schedule for completion, to the Commission 2. The staff response to this
SRM is contained in SECY-06-001 9, "Semiannual Update of the Status of New Reactor
Licensing Activities and Future Planning for New Reactors," dated January 31, 2006. In the
next semiannual update, SECY-06-0187 dated August 25, 2006, the~staff informed the
Commission that they had accelerated the SRP revision schedule to March 2007. The staff
completed the revision of all SRP sections per the SECY-06-0187 schedule.
In 2010, the Commission provided direction to the staff on the preparation for, and review of,
SMR applications, with a near-term focus on integral pressurized water reactor (iPWR) designs.
As used in this document, iPWRs are a subset of SMRs. The Commission directed the staff to
more fully integrate the use of risk insights into pre-application activities and the review of
applications and, consistent with regulatory requirements and Commission policy statements, to
align the review focus and resources to risk-significant structures, systems, and components
(SSCs) and other aspects of the design that contribute most to safety in order to enhance the
effectiveness and efficiency of the review process. The Commission directed the staff to
develop a design-specific, risk-informed review plan for each SMR to address pre-application
and application review activities 3. In 2011i, the staff responded to this SRM in SECY-1 1-0024,
"Use of Risk Insights to Enhance the Safety Focus of Small Modular Reactor Reviews," dated
February 18, 2011 (Agencywide Documents Access and Management System (ADAMS)
Accession No. MLI110110688). On May 11, 2011, the Commission issued an SRM approving
the use of the risk-informed and integrated review framework for staff pre-application and
application review activities pertaining to SMR/iPWR design applications (ADAMS Accession
2 Refer to Staff Requirements Memorandum (SRM) M050406, "Briefing of Status of New Site and Reactor Licensing," dated
May 10, 2005 (ADAMS Accession No. ML05I1300673).
SRefer to SRM
-
COMGBJ-1 0-0004/COMGEA-1 0-0001, "Use of Risk Insights to Enhance Safety Focus of Small Modular Reactor
Reviews," dated August 31, 2010 (ADAMS Accession No: ML1 02510405).
-3-
-3Revision 0 - January 2014
No. ML111320551). This SRP Introduction, Part 2, incorporates the Commission-approved
SMR risk-informed and integrated review framework described in SECY-1 1-0024.
Obiectives of the SRP
The SRP is intended to be a comprehensive and integrated document that provides the
reviewer with guidance describing methods or approaches that the staff has found acceptable
for meeting NRC requirements. Implementation of the criteria and guidelines contained in the
SRP by staff members in their review of applications provides assurance that a given design will
comply with NRC regulations, and provide adequate protection of the public health and safety.
The SRP also makes the staff's review guidance for licensing nuclear power plants publicly
available and is intended to improve industry and public stakeholder understanding of the staff
review process. It should be noted that the SRP is not a substitute for NRC regulations, and
compliance with the SRP is not required. However, when using methods or"approaches other
than described in the SRP, applicants are expected to provide sufficient information for the staff
to conduct independent evaluations to confirm the results and conclusions are in compliance
with the regulations.
In addition to documenting current methods of review, the SRP provides a basis for orderly
modification of the review process. The NRC disseminates information regarding current safety
issues and proPOsed solutions through various means, such as generic communications and
the process for treating generic safety issues. When current issues are resolved, it is necessary
to determine the need, extent and nature of revision that should be made to the SRP to reflect
new NRC guidance.
The staff should use the SRP as superseded or supplemented by new or revised regulations,
regulatory guidance, staff analyses of previous applications, and other published staff positions
to perform its review of a power reactor application or a proposed change to an existing license
under 10 CFR Part 50, or a new reactor license application or amendment under 10 CFR
Part 52.
For SMR applications submitted by applicants that agree to participate in risk-informed and
integrated review framework pre-application activities, DSRSs are developed by the staff
specifically for the SMR design. The DSRS serves the same purpose and has the same
objectives that the SRP has for non-SMR application reviews. Each DSRS includes a "Safety
Review Matrix" as a cross-reference indicating which SRP sections are "use-as-is" (no
corresponding DSRS section required), which SRP sections are usable with minor
modifications, which SRP sections will be replaced by new DSRS sections, and which SRP
Sections do not apply to the particular SMR design being reviewed.
Scope of Review of License Applications (Initial Applications and Amendments)
Because the staff's review constitutes an independent audit of the applicant's analysis, the staff
may emphasize or de-emphasize particular aspects of an SRP section, as appropriate, for the
application being reviewed. Prior to the initiation of a review, the technical branch chief and
assigned reviewer establish the scope and depth of the review to be performed, including the
use of acceptance criteria and review guidelines to be used. In some cases, the staff may
propose justification for not performing certain reviews called for by the SRP. These areas of
increased or decreased emphasis are acceptable, if the reviewer has management approval
and documents the scope and depth of the review in the Safety Evaluation Report (SER).
-4-
-4Revision 0
-
January 2014
Examples of acceptable variations in the scope of a review include reduced emphasis on design
reviews ifthe design and its underlying conditions of acceptance are identical to that of another
unit that was recently reviewed and approved or increased emphasis on certain aspects of the
design review as a result of recent operating experience or consideration of unique design
features that are not addressed in the SRP. Risk-insights can also be used in determining the
depth of review. The staff should generally limit its review of a proposed amendment to an
existing license to those parts of the SRP that are directly affected by the proposed change.
The staff review scope and flexibilities described above are further detailed below under "SMR
Design Pre-Application Activities and Application Reviews."
In addition to the guidance provided for applications and amendments submitted under 10 CFR
Part 50, the SRP provides pertinent review guidance to the staff for review of new license
applications submitted under 10 CFR Part 52. This includes ESP, DC, COL, SDA, and ML
applications. The SRP sections a'pplicable to a COL application are consistent with the
organization of guidance contained in Regulatory Guide (RG) 1.206, "Combined License
Applications for Nuclear Power Plants (LWR Edition)." The SRP sections applicable to an ESP
and a DC application are consistent with the site-related sections and design-related sections of
RG 1.206. Furthermore, RG 1.206 delineates different content based on whether the COL
application references an ESP, a DC, both or neither. In general, review of a SDA or a ML
application will be similar to that of a DC.
For DC, SDA, and COL applications submitted under 10 CFR Part 52, the level of design
information reviewed should be consistent with the level of review performed for a Final Safety
Analysis Report (FSAR) submitted in a 10 CFR Part 50 OL application. For COL applicants,
verification that the as-built facility conforms to the approved design is performed through the
inspections, tests, analyses, and acceptance criteria (ITAAC) verification process.
For the review of COL applications, applicable sections of the SRP or DSRS will be used to
review the operational program descriptions submitted by the applicant. The review will be
performed consistent with guidance contained in SECY-05-01 97, "Review of Operational
Programs in a Combined License Application and Generic Emergency Planning Inspections,
Tests, Analyses, and Acceptance Criteria," and the related SRM dated February 22, 2006.
Consistent with this guidance, the staff will review and obtain a reasonable assurance finding on
the operational program and its implementation schedule. In addition, the staff will include a
license condition on subsequent implementation milestones for each operational program
description for which specific implementation requirements are not specified in the regulations.
Deviation from the SRP/DSRS by Applicants
Because the SRP and the DSRS generally describe an accePtable means of meeting the
regulations, but not necessarily the only means, applications may deviate from the
acceptance criteria in the SRP or the DSRS. On March 10, 1982, the Commission approved
10 CFR 50.34(g), "Conformance with the Standard Review Plan (SRP)." 10 CFR 50.34(g) was
subsequently renumbered as 10 CFR 50.34(h). Specifically, paragraph 10 CFR 50.34(h)
requires applications for light-water cooled nuclear power plant operating licenses docketed
after May 17, 1982, to include an evaluation of the facility against the SRP in effect on May 17,
1982, or the SRP revision in effect six months prior to the docket date of the application,
whichever is later. The evaluation must include an identification and description of all
differences in design features, analytical techniques, and procedural measures proposed for a
5
-5Revision 0 - January 2014
facility and those corresponding features, techniques, and measures given in the SRP
acceptance criteria. Where such a difference exists, the evaluation shall discuss how the
alternative proposed provides an acceptable method of complying with those rules or
regulations of the Commission, or portions thereof that underlie the corresponding SRP
acceptance criteria.
Similar provisions regarding contents of applications for the different license processes are
contained in the Subparts to 10 CFR Part 52. Staff guidance for reviewing the applicant's
evaluation is contained in SRP Chapter 1.0, "Introduction and Interfaces."
Alternatively, SMR applicants may evaluate the facility against the OSRS revision in effect six
months before the docketed date of the application. If a final version of the DSRS is not
available, the applicant may refer to the latest public draft version of the document. This is
sufficient to meet the intent of the regulations cited above.
As stated in the Introduction to 10 CFR Part 50, Appendix A, the General Design Criteria
(GDCs) establish minimum requirements for the principal design criteria for nuclear power
plants similar in design and location to plants for which construction permits and operating
licenses have been issued by the Commission. The GDCs are also considered to be generally
applicable to other types of nuclear reactor designs and are intended to provide guidance in
establishing the principal design criteria for such other units.
The modification of existing GDCs or development of new ones may be necessary for some
new SMR designs for which the existing GDCs are not sufficient or for which additional criteria
must be identified and satisfied in the interest of public safety. Given this recognition, their
omission or lack of specificity for some aspects of new reactor designs does not relieve
applicants from considering these matters in the design of a specific facility and in satisfying the
necessary safety requirements. It is expected that the GDCs may need to be augmented or
changed as important new requirements for these design features are identified by the technical
staff.
Finally, there may be instances for which compliance with some GDCs may not be necessary or
appropriate. In such cases, departures must be identified and justified by the applicant. The
DSRS described below is intended to identify specific acceptance criteria that are applicable for
the review of individual SMR designs.
Organization of SRP/DSRS
Each SRP/DSRS section is organized as follows:
Review Responsibilities: This subsection identifies the primary and, as applicable, secondary
review functions.
I.
Areas of Review
The Areas of Review subsection describes the scope of review by the branch having primary
responsibility for the identified functional area. Specifically, this subsection contains a
description of the systems, components, analyses, data, or other information that is reviewed as
part of the particular Safety Analysis Report (SAR) section. It also contains a discussion of the
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information needed or the review expected from other branches to permit the primary review
branch to compiete its review, as well as a list of applicable interlacing sections.
I1.
Acceptance Criteria
The Acceptance Criteria subsection identifies the applicable NRC requirements including
specific regulations, NRC orders, and industry codes and standards referenced by regulations.
For new reactor license applications submitted under. 10 CFR Part 52, the applicant is also
required to address:..
*
the proposed technical resolution of unresolved safety issues and medium and high
priority, generic safety issues that are identified in the version of NUREG-0933 current on
the date six months before application, and that are technically relevant to the design
*
Three Mile Island (TMI) requirements
*
relevant operating experience
These requirements are not identified within specific'SRP or DSRS sections; rather, these
requirements are identified within SRP Chapter 1, "Introduction and Interfaces." An applicant
will tabulate information within Chapter 1, but will address the technical issues to satisfy the
requirements within the specific sections, themselves.
This subsection also identifies the regulatory guidance which the staff has determined provides
an acceptable approach (i.e., SRP acceptance criteria) for satisfying the applicable
requirements. For the purposes of this NUREG, these criteria can be generally classified as
design-based acceptance criteria or as performance-based acceptance criteria.
Examples of design-based acceptance criteria include those acceptance criteria related to SSC
basic design, materials, and suitability for service conditions.
Examples of performance-based acceptance criteria include those acceptanc~e criteria related to
SSC capabilities, reliability, and availability.
The Guidance documents include but are not limited to: RGs, Commission policy as described
in SECY papers and corresponding SRMs, NRC approved or endorsed industry codes and
standards, certain technical reports (e.g.,' NUREGs and topical reports and corresponding safety
evaluations), and Branch Technical Positions (BTPs), which are provided as appendices to the
SRP. BTPs typically set forth solutions and approaches previously determined to be acceptable
by the staff in dealing with a similar safety or design matter. These solutions and approaches
are documented in this form so that staff reviewer's can take uniform and well understood
positions as similar matters arise in the review of various applications.
....
.
Each SRP and DSRS section explicitly states that the SRPIDSRS is not a substitute for the
NRC regulations, and compliance with it is not required. However, applicants are required to
identify differences from the SRP or DSRS acceptance criteria and evaluate how the proposed
alternatives to the acceptance criteria provide an acceptable method of complying with the. NRC
regulations.
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Lastly,
this subsection
also contains,
as necessary,
the technical
bases for
applicability
requirements
to the subject
areas of review
or relationship
of regulatory
guidance
to the of the
associated requirement.
Ill.
Review Procedures
This subsection discusses how the review is accomplished. The subsection is a step-by-step
procedure to be implemented by the reviewer to obtain reasonable assurance that the
applicable regulatory requirements have been met. These review procedures are based on the
identified SRP/DSRS acceptance criteria. For deviations from these specific acceptance
criteria, the staff should review the applicant's evaluation of how the proposed alternatives to the
acceptance criteria provide an acceptable method of complying with the relevant NRC
requirements identified in specific review areas of Subsection I1.
For new reactor license applications submitted under 10 CFR Part 52, this subsection
addresses staff review procedures for how pertinent operating experience insights identified in
generic letters and bulletins or equivalent international operating experience have been
incorporated into the plant design.
IV.
Evaluation Findingqs
This subsection presents the type of conclusion that is sought for the particular review area. For
each SRP/DSRS section, the staff's conclusion is incorporated into a published SER. The SER
describes the review and the aspects of the review the staff emphasized, and identifies (1) the
changes the applicant made to the application (if any),. (2) the matters addressed by additional
information (if applicable), (3) the matters for which additional information is expected to be
forthcoming, (4) the matters remaining unresolved as open items, and (5) deviations from the
SRP/DSRS acceptance criteria in design and operational programs, and the bases for the
acceptability of such deviations. The SER also clearly identifies any requested exemptions from
the regulations and the staff's bases for its determinations on these requests.
V.'
Implementation
..
. .
.:.
This subsection provides guidance to applicants and licensees regarding the NRC plans for
using the SRP/DSRS section. The NRC regulations in 10 CFR 50.34(h) and similar provisions
in 10 CFR Part 52 require each application to include an evaluation of the facility against the
SRP of record six months prior to docketing, including"all differences between the design
features, analytical techniques and procedural measures proposed for a facility and those in the
SRP acceptance criteria.
The NRC staff will use the SRP/DSRS version in effect at the time of the application review.
VI.
References
This subsection lists the references used in the review process.
Maintenance of the SRP
The SRP will be revised and updated periodically as the need arises to clarify the content or
correct errors and to incorporate modifications approved by the Director of the Office of Nuclear
Reactor Regulation or the Director of the Office of New Reactors.
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A revision number and a publication 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 as needed,
the revision numbers and dates will not be the same for all sections. As necessary,
corresponding changes to RG 1.206, "Combined License Applications for Nuclear Power
Plants," will also be made. Comments may be submitted electronically by email to
NRO SRP.Resource(~nrc.,qov. Notices of errors or omissions should also be sent to the same
address.
Comment resolution will be addressed in subsequent SRP revisions. Prior"to revision to
individual sections, comment resolution may establish a basis for how alternatives to the
NUREG-0800 acceptance criteria provide an acceptable method of complying with the NRC
regulations.
SMR DESIGN PRE-APPLICATION ACTIVITIES AND APPLICATION REVIEWS
This portion of the SRP Introduction, Part 2, describes the licensing review philosophy and the
risk-informed, integrated review framework to be applied by the staff for new SMR applications
under 10 CFR Part 52; and incorporates staff commitments made to the Commission in
SECY-1 1-0024 (see Background).
This framework is distinct from the review approach used for non-SMR applications and license
amendments. The review framework described below is not intended for use with current
non-SMR licensees or applicants.
In~corporation of this framework in the SRP does not change NRC requirements for SMR
applications or applicants. As previously noted, applicants are not required to engage with the
NRC in the pre-application activities described herein. The submittals by applicants that choose
not to engage the NRC in pre-application activities will be reviewed by the staff using current
SRP guidance and methods rather than a DSRS.
Use of this framework does not relieve the requirement for SSCs that are important to safety to
meet NRC regulations to perform their safety functions, unless granted an exemption or subject
to an application-specific order.
Potential benefits of implementing the framework for SMR applications include:
*
gains in early awareness of unique or non-traditional SMR generic issues and design or
operational features through pre-application exchanges with applicants, stakeholders,
and the NRC staff
*
enhanced safety focus for SMR application reviews through the use of risk insights
*
improved cross-disciplinary staff reviews and interactions
This framework also advances, where appropriate, the use of a performance-based regulatory
approach, which is consistent with longstanding goals of the agency.
As used throughout this discussion of the framework, the term "reviewer" means all NRC staff in
all disciplines involved with the pre-application and post-application reviews of specific
application sections and creation of the associated SERs.
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Overview
As described in the "Background" section of this introduction, the Commission directed the staff
to more fully integrate the use of risk insights into pre-application activities and the review of
applications and, consistent with regulatory requirements and Commission policy statements, to
align the review focus and resources to risk-significant SSCs and other aspects of the design
that contribute most to safety in order to enhance the efficiency of the review process.
In response, the staff developed a framework to enhance the efficiency of the review process,
and to align the staff review focus and resources with risk-significant SSCs and other aspects of
an SMR design that contribute most to safety. The framework builds upon the review process
used for non-SMR applications, resulting in a risk-informed and integrated process for the
review of SMR applications. The staff implements the framework as they conduct SMR
pre-application and post-application activities. The success of this approach depends on the
'
ag~ilability of detailed SSC safety and risk information from the prospective applicant(s).
Three major elements comprise the framework (see Figure 1). First,, it incorporates a
risk-informed review approach by considering both the safety classification and the risk
significance of SSCs in order to determine the appropriate level of review (i.e., the framework
uses a "graded review" approach).
Second, the framework incorporates an integrated review approach by using the satisfaction of
selected requirements to provide reasonable assurance of some aspects of SSC performance
(for example, performance-based acceptance criteria related to SSC capability, reliability, and
availability). Examples of requirements that could be applied for this purpose include 10 CFR
Part 50, Appendix A (general design criteria, overall requirements, criteria 1 through 5), 10 CFR
Part 50, Appendix B (quality assurance program), 10 CFR 50.49 (electric equipment
environmental qualification program), 10 CFR 50.55a (code design, inservice testing and
inservice inspection programs), 10 CFR 50.65 (maintenance rule), Technical Specifications
(TSs), Availability Controls for SSCs subject to Regulatory Treatment of Non-Safety Systems
(RTNSS), the Initial Test Program (ITP), and ITAAC. In preparing the safety evaluation for the
application, the staff may use the satisfaction of these selected requirements to augment or
replace, as appropriate, technical analysis and other evaluation techniques to obtain reasonable
assurance that the performance-based acceptance criteria are satisfied. Under the framework,
the staff also has the flexibility to use these selected requirements to demonstrate satisfaction of
design-based acceptance criteria for the SSCs with low risk significance. The staff will verify the
demonstration of the design-basis capabilities of SSCs that are important to safety as part of the
ITAAC completion review prior to plant operation.
Third, the results of the safety/risk categorization and the integrated review approach described
above are documented in the DSRS created by the staff for each SMR design. The DSRS
serves the same purpose and has the same objectives that the SRP has for non-SMR
application reviews.
The framework is applicable to the review of all SSCs, but it is not applicable to the review of
programmatic, procedural, organizational, or other non-SSC topics. This is because under the
current risk analysis state-of-the-art, it is not yet possible to assign risk metrics to non-SSCs.
However, the application of the selected requirements to SSCs is considered when reviewing
the risk significance of individual SSCs that fall within the scope of the requirements. Non-SSC
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topics screened out of the framework are reviewed by the staff using traditional evaluation
methods to reach a finding of reasonable assurance. Examples of these non-SSC topics
include quality assurance programs, training, human factors engineering, health physics
programs, and operating procedures.
While it may not yet be 'possible to assign quantitative risk metrics to non-SSCs, the technical
branches responsible for these topics are encouraged to identify and consider alternate
methods of risk-informing the reviews of these sections.
Additional guidance on implementing the framework is provided in subsequent sections of this
introduction.
Implementation of the SMR Framework
The major activities required to implement the SMR framework are described in this section.
The intent of describing these activities here is to provide staff with guidance that, when used in
conjunction with applicable, detailed internal procedures, will result in a review framework that
can be consistently and objectively applied across SMR designs and application reviews.
These activities may be broadly categorized as pre-application activities and post-application
activities.
Pre-Application Activities
The framework is implemented as soon as the staff determines that an applicant has sufficient
commercial intent, organizational capacity, and design maturity to support commencement of
meaningful regulatory interactions and that there is reasonable expectation of an application
submittal. The major factors that will govern the level and timing of the staff's pre-application
activities include the maturity of the SMR design and associated Probabilistic Risk Assessment
(PRA), and the willingness of the applicant to coordinate with the NRC prior to submitting an
application.
A number of critical applicant inputs will determine the ability of the staff to formulate its review
strategy and create useful draft DSRS documents during the pre-application period. The quality
and timeliness of these inputs are key to the effectiveness of the staff's pre-application activities.
Early submittal of finalized or near-final design information for reference use ,by the staff will
minimize revisions of the DSRS sections. Preliminary PRA results and Reliability Assurance
Program (RAP) list categorizations will assist the staff in gaining an understanding of the
applicant's safety/risk categorization strategy for the SMR SSCs. If the SMR applicant intends
to use innovative design features such as passive systems, simplified control features, or other
similar approaches, early identification of these features to the NRC will facilitate timely
identification of unique regulatory issues that may arise as a result.
Additionally, "white papers," Topical Reports, Technical Reports, or other types of information
documents may be submitted by the applicant to the NRC for review during the pre-application
period. Documents such as these will assist the NRC in understanding the SMR design as
early in the design cycle as possible. Requirements for stakeholder engagement, and the
receipt and processing of documents, are not changed by implementation of the framework.
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DSRS Preparation
The principal risk-informed and integrated review framework pre-application activity for the stiff
is the preparation of the overall design-specific review plans, including the DSRS, for use as
guidance for performing SMR application reviews. The overall plans include identification of the
specific pre-application and post-application review activities, development of the schedule for
those activities, and creation of the DSRS itself. Each DSRS provides guidance to support the
staff's application review activities by tailoring the SRP to the specific SMR design.
To the extent afforded by the cooperation of the SMR applicant, the staff's DSRS development
occurs in parallel with SMR design development. Since the design is expected to evolve from
conceptual design through final design, preparation of the DSRS is expected to be iterative.
The staff's goals are to complete and publish the public draft DSRS one year prior to submittal
of the application, and to issue the final DSRS for use not later than the time of docketing of the
application.
During development of the individual DSRS sections by the staff, each corresponding section of
the SRP is reviewed to determine whether it can be referenced for use-as-is, needs
modifications for use, whether an entirely new DSRS section needs to be created, or whether
the corresponding SRP section should be deleted from the DSRS (i.e., the SRP section is not
applicable to the SMR design). This assessment is documented by the staff in a "Safety Review
Matrix" that is developed for and included in the DSRS prepared for each SMR design.
Development of the DSRS provides a mechanism for ongoing communications and interactions
,
among the staff, applicant, and other stakeholders to support the early identification and
resolution of both technical and regulatory issues.
Each DSRS is prepared by the staff in a format that corresponds with the format/content of the
SRP previously described under "Organization of SRP/DSRS." Similar to the SRP, each SSC or
topic section/subsection includes a description of the scope of review, identification Of the
acceptance criteria to be satisfied, and relevant references for the reviewer to use determine
whether there is reasonable assurance that the applicant has adequately addressed the NRC
regulations and requirements listed in the DSRS section/subsection.
The DSRS will incorporate lessons learned from past NRC large light-water reactor application
reviews and applicable published Interim Staff Guidance documents. This information will be
incorporated into the SRP sections as applicable, during future regular upd~ate cycles.
Six Month Pre-Application SRP/DSRS Reviews by SMR Applicants
NRC regulations in 10 CFR 52.17(a)(1)(xii), 10 CFR 52.47(a)(9), 10 CFR 52.79(a)(4,1),
10 CFR 52.137(a)(9), and 10 CFR 52.157(c)(30) state that the applicant for an early site permit,
design certification, combined license, standard design approval, or manufacturing license
respectively, must include in its application an evaluation of the facility against the SRP revision
in effect six months before the docketed date of the application. ••
Alternatively, SMR applicants may evaluate the facility against the DSRS revision in effect six
months before the docketed date of the application. If a final version of the DSRS is not
available, the applicant may refer to the latest public draft version of the document. This is
sufficient to meet the intent of the regulations cited above.
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Additional guidance on the timing of the SRP/DSRS evaluation submission is given in SRP
Chapter 1.0, Item Number 9. Additional information on the disposition of differences between
the applicant's design and the SRP/DSRS is given in the Deviation from the SRP/DSRS by
Applicants section of this document.
Following submittal of the application, the NRC staff will determine if design and operational
details in the application require adjustments to the DSRS guidance and NRC review approach.
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FIGURE 1
RISK-INFORMED AND
INTEGRATED REVIEW
FRAMEWORK
DSRS Section/Subsection
.. is for Programmatic,
Procedural, Organization, or
Other Non-SSC Topic.
Safety/Risk Categorization
is Outside of Framework
Scope. (Note 1)-.
Yes
Yes.•L• Significant?
Risk-Informed
Activities
Integrated Review
£'•l=•,LI
V l LI•
Al
-
SR,
Risk significant
Apply current review
process (i.e., analysis/
evaluation techniques)
to determine if designbased and performancebased acceptance
criteria have been
satisfied.-
A2 .- SR,
Non-Risk Significant
Commence graded
approach. Consider use
of Selected requirements
to determine if designbased and performancebased acceptance
criteria have been
satisfied.
BI
NSR,
Risk Significant
B2 -NSR,
Non-Risk Significant
Further extend graded
approach from A2. Apply
Al review processes for
design-based acceptance
criteria. Wherever possible,
use selected requirements
to determine if performancebased acceptance criteria
have been satisfied.
Further extend graded
approach from 'B1.
Wherever possible, use
selected requirements to
determine ifdesignbased and performancebased acceptance
criteria have been
satisfied.
-
Note 1: Programmatic, procedural, organization, or other non-SSC topics (e.g., quality assurance,
training,
human factors engineering, health physics programs, operating procedures) are outside of the risk-informed
and integrated review framework scope and are not subject to the safety/risk categorization process shown in
Figure i. These non-SSC topics will be evaluated using traditional methods as appropriate.
The risk significance associated with these non-SSC topics may be difficult to quantify and evaluate. In these
cases, the responsible technical organizations will determine the most appropriate method for demonstrating
satisfaction of the acceptance criteria on a case-by-case basis. In doing so, the organizations are encouraged
to identify and consider alternate methods of risk-informing reviews of these sections.
-
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Risk-Informed Categorization of SSCs
Performance of the risk-informed categorization of SMR SSCs is a key framework activity in the
development of the DSRS, which is risk-informed through identification of the safety and risk
attributes of SSCs. In order for the staff to implement the categorization process
depicted in Figure 1, the applicant must first categorize SSCs as (1) either safety-related or
nonsafety-related using the criteria in 10 CFR 50.2, and (2) either risk significant or not risk
significant using the process developed for the RAP; normally documented in Section 17.4 of
the DC or COL FSAR. The staff expects to receive preliminary results of the categorization
activities as they become available from the applicant in the pre-application phase of the staffs
review. The staff will conduct pre-application meetings or audits as necessary to obtain and
review the information.
•: •:.,.'••:i
The staff will assign each SSC to one of the four categories shown on Figure 1 based on its
review of the information developed by the applicant., It is important that the staff receive the
information a~nd complete the initial verification of SSC safety and risk significance as early in
the pre-application review, process as possible to enable assignment of each SSC to one of the
fourcategories. Complete results of the applicant's categorization activities will be .provided in
the DC or COL ESAR when the application is submitted to the NRC for review. The staff will
review these categorization results as a part of its review of the application. Should the results
change as a result of the staff's review, or for other reasons, the staff will adjust its previous
category selections accordingly and conduct any additional review dictated by the changes as
necessary.
Initial staff activities for this portion of the framework are similar to the existing review practices.
Both require a general understanding of the functions of a specific SSC, an overview of design,
modes of operation, relationships to other systems, and contributions to risk significance in
terms of event initiation or mitigation in order to evaluate information developed by the
pre-applicant effectively.
As discussed above, the final safety/risk categorization of SSCs will not be known until the
applicant's detailed design and PRA results have been finalized and communicated to the NRC
•in its application. Therefore, the staff will make bestefforts to use the applicant's preliminary
categorization assessment to pre-classify SSC safety and risk categorization in order to begin
writing the draft DSRS. As the design evolves and the applicant communicates additional
information to the staff, the draft DSRS will be reviewed and modified as appropriate.
With regard to risk significance, applicants are responsible for determining which SSCs are
candidates for RTNSS, and which are included in the RAP list. The staff assesses and verifies
the applicant's categorization once sufficient design detail, PRA information, and RAP list
information are available. The verification of whether an SSC is safety-related (i.e., satisfies any
of the criteria in 10 CFR 50.2), risk-significant, or both is accomplished through current
evaluation and decision, processes. Risk significance is measured relative to the likelihood and
consequences of severe accidents which involve core damage and can lead to containment
failure with a large release of radioactivity. Consequently, risk significance may be determined
with the use of insights from the list of risk-significant SSCs included in the applicant's RAP list.
The staff reviews the methods and results used by the applicant to establish the list of SSCs
included in RAP using guidance in SRP Section 17.4. Guidance for reviewing the selection of
•SSCs for RTNSS is provided in SRP Section 19.3 and on an SSC-specific basis in the
applicable DSRS for a given SSC.
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This determination/verification will be documented in the final DSRS prepared for the SMR
design. When final, the applicable DSRS sections provide reviewers with SSC design
information to guide the determination of whether an SSC meets the definition of safety-related
in 10 CFR 50.2 or not. An SSC is considered risk-significant if it has been included in the
applicant's RAP or RTNSS program. SSCs that are not included in RAP and RTNSS, but that
are still within the scope of the risk analyses (whether modeled or screened out), are considered
to have low risk-significance.
Once the safety and risk categorization of the SSC0 has been provided by the applicant, the
NRC requirements specific to the SSC are identified by the technical staff and listed in .the
SSC-specific section of the DSRS. This list includes the selected requirements that are
-assigned based on the safety and risk categorization previously developed. The SSC-specific
section of the DSRS also lists the acceptance criteria to be met in order to demonstrate
satisfaction of the requirements.
Post-Application Activities
.
Post-application activities for SMR applicants participating in the risk-informed and integrated
review framework are similar to those performed for all applicants. Technical reviewers tasked
with performing reviews of application sections confirm the applicable SSC safety/risk
categorization shown in the DSRS and make adjustments if required based on changes in
design information received after initial re~ceipt of the application or resulting from Request for
Additional Information responses. Figure 1 will be used as a guide to verify the appropriate
framework categorization and associated review approach for the SSC based on the SSC
safety classification and risk significance evaluation.
Application of the Integqrated Review Approach
Four review levels (labeled as A, A2, 81, and B2 in Figure 1) correlate to the safety
classification and risk significance of the SSC under review. Using a graded approach, the staff
applies the most rigorous review techniques to SSCs with the highest safety and risk
significance (analogous to the typical review process using the SRP), and a progressively
less-detailed review to other SSCs as the assigned safety/risk significance declines.
In the SMR review framework, satisfaction of design-based acceptance criteria for categories
Al and 81 continues to be demonstrated using current methods, including independent analysis
and evaluations, confirmatory calculations, computer modeling, and other similar techniques.
Satisfaction of design-based acceptance criteria for categories A2 or B2 may also be
demonstrated using these current methods, or by the use of selected requirements as
discussed below.
Satisfaction of performance-based acceptance criteria in the framework may be demonstrated
by use of traditional methods as described above, through the use of test or performance data
from selected requirements, or through a combination of these techniques. The blend of
techniques selected by the DSRS technical writers and the reviewers are guided by the SSC
safety/risk categorization determined by the applicant and verified by the staff.
The NRC requirements that must be met by an SSC do not change under the SMR framework.
Under the graded approach, the NRC staff may rely on the applicant's submittal with selected
requirements to demonstrate satisfaction of performance-based acceptance criteria in lieu of
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detailed independent analyses. They may also be used to demonstrate satisfaction of designbased acceptance criteria for category A2 and B2 SSCs. For example, satisfaction of
acceptance criteria related to the capability, availability or reliability of an SSC may be
addressed through the satisfaction of these selected requirements, to an extent consistent with
the safety/risk categorization of the SSC. The staff will verify the demonstration of the designbasis capabilities of SSCS that are important to safety as part of the ITAAC completion review
prior to plant operation.
The staff preparing the DSRS, using safety/risk categorization inputs from the applicant as
verified by the staff, makes an initial determination of which selected requirements could be
used as an alternate method for demonstrating the satisfaction of the design-based or
performance-based acceptance criteria. The review, including decisions on the use of selected
requirements and analysis/evaluation techniques, should focus on the functions and
characteristics of the SSC that pertain to its safety/risk significance.
.Examples of requirements that may apply to an. SSC and that could be used to demonstr~ate the
satisfaction of design-based or performance-based acceptance criteria include:
*
10 CFR Part 50, Appendix A, General Design Criteria, Overall Requirements, Criteria 1
through 5
•
10 CFR Part 50, Appendix B, Quality Assurance (QA) Program
*
10 CFR 50.49, Environmental Qualification of Electric EqUipment (EQ) Program
*
10 cFR 50.55a, Code Design, Inservice Inspection and Inservice Testing (ISI/IST)
Programs
*
10 CFR 50.65, Maintenance Rule requirements (MR)
*
Reliability Assurance Program (RAP)
*
Technical Specifications (TSs)
*
Availability Controls for SSCs subject to Regulatory Treatment of Non-Safety Systems
(RTNSS)
*
Initial Test Program (ITP)
*
10 CFR 52.47, Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
This list of examples above is not intended to be all-inclusive. During preparation of the DSRS
by the staff, the list of selected requirements for specific SSCs is determined. This list is
included in the "Review Procedures" subsection of each DSRS section. Following receipt of the
application, it is the responsibility of the technical reviewers to determine how best to apply the
list of selected requirements in order to determine whether design-based and performancebased acceptance criteria have been met.
Once an application has been received, reviewers retain flexibility and discretion in selecting the
appropriate review methods to be applied to an SSC based on unique characteristics or
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circumstances. For example, the level of review methods to be applied to an SSC initially
categorized by the applicant and confirmed by the staff preparing the DSRS as "BI" includes
evaluation and analysis techniques for design-based acceptance criteria and the use~of selected
requirements "wherever possible" to determine the satisfaction of performance-based
acceptance criteria. The reviewer may determine that additional analyses are needed to
augment the use of selected requirements for a particular SSC to reach a conclusion of
reasonable assurance.
When reliance on a selected requirement is used to demonstrate satisfaction Of acceptance
criteria and SSC performance, the specific requirement sub-element and implementation
milestone are to be identified by the reviewer.
The four safety/risk categories in Figure 1 are described below with examples of the integrated
review approach.
-,•
Ft•!FiS.S~s etermine.edto be both safety-related and risk-significant, the_ level of
r....•i•
ieviewis
denote"d as Al. •:!For such •SSCs," th'ereview is consistent with ithe typical review
process :using the SRP in that the' review typically involves detailed analysis and
evaluation techniques to demonstrate Satisfaction of the DSRS design-based and
performance-based acceptance criteria. In addition, the DSRS identifies those selected
requirements applicable to the SSC.
Al,-
For example, the staff will verify as part of the application review that Al SSCs satisfy all
special treatment requirements applicable to those SSCs including QA, EQ,
10 CFR 50.55a, MR, RAP, ITP, and ITAAC.
*A2
-
For SSCs determined to be both safety-related and not risk-significant, the level of
review is denoted as A_22. Similar to Al, the NRC staff continues to be required to reach
a reasonable assurance finding for the capability of Safety-related SSCs categorized as
A2 to perform their safety-related functions prior to issuing a license or design approval.
However, the graded review approach commences at the A2 level for design-based and
performance-based acceptance criteria. The reviewer identifies selected requirements
that may be considered for use in lieu of some analysis and evaluation techniques to
demonstrate satisfaction of specific acceptance criteria.
*Under the framework for category A2 SSCs, the staff has flexibility in determining how
best to apply the selected requirements listed above to demonstrate satisfaction of
acceptance criteria. For example, .the applicant may provide a certification in its
submittal that NRC requirements for design-basis capability will be are satisfied with
because of the applicant's reliance on selected requirements[ such as QA, and others as
applicable. The reviewer may determine that for a particular SSC, the applicant's
certification commitment to these requirements is sufficient to reach a finding of
reasonable assurance for the SSC being reviewed and the reviewer may include an
ITAAC to verify that the A2 SS0 is built as designed. System performance of the A2
SSC will be verified during pre-operational testing to satisfy the ITAAAC combined with
demonstration verification of the design-basis capability of the A2 SSC during the review
of pre-operational testing to verify ITAAC completion prior to plant operation, is sufficient
to reach a finding of reasonable assurance for the SSC being reviewed.
-18-
-18Revision 0 - January 2014
*
81
- For SSCs determined to be both nonsafety-related and risk-significant, the level of
review
is denoted as 81. For design-based acceptance criteria, the review is similar to
the review for Al SSCs.
For performance-based acceptance criteria, the graded review approach is further
extended from the A2 level. The review emphasis shifts from applying analysis and
evaluation techniques to identifying those selected requirements that satisfy DSRS
acceptance criteria wherever possible. If any of the proposed selected requirements
satisfies the acceptance criteria, it can be used to augment or replace some of the
review procedures. For those acceptance criteria that cannot be satisfied, either in
whole or in part, by performance-based activities (e.g., tests or inspections) within
selected requirements, the appropriate analysis and evaluation techniques are applied
(i.e., relying on existing review methods described in the DSRS). Note that for SSCs
determined to be highly risk-significant, it may be appropriate to perform more detailed
reviews using methods associated with reviews performed at the Al level.
*
82 - For SSCs determined to be both nonsafety-related and not risk-significant, the level
of review is denoted as 82. The graded review approach is further extended from the
81 level. At the B2 level, both the design-based and the performance-based acceptance
criteria are anticipated to be minimal. The review is focused on identifying those ...
performance-based activities (e.g., tests or inspections) within the selected requirements
that can be used to satisfy the design-based or performance-based acceptance Criteria.
If any of the proposed requirements satisfies the acceptance criteria, it can be used to
replace some of the review procedures.
However, there may be SSC design-based acceptance criteria that cannot be satisfied
solely through the use of selected requirements. For such sscs, the reviewer considers
application of appropriate analysis and evaluation techniques to be the alternative review
.method.
Review levels Al through 82. reflect a graded approach to reviews in that performance-based
activities within selected requirements are increasingly applied to satisfy DSRS acceptance
criteria in lieu of applying traditional analysis and evaluation techniques. This approach involves
the professional judgment of the reviewer and, therefore, the extent to which selected
requirements are applied to-satisfy the acceptance criteria during A2, 81, and 82 reviews will
vary, as do the traditional review approaches given the flexibilities with the SRP.
In addition, in cases where SMR designs include features that differ significantly from large
LWR designs, the staff considers the risk significance of the subject SSCs in the implementation
of the additional analysis and testing requirements required by 10 CFR 50.43(e).
"
When a technical reviewer has determined that a particular requirement will be used to satisfy a
specific acceptance criterion, the reviewer ensures that the documentation submitted by the
applicant includes the specific method to be used to satisfy the criterion. The use of the
selected requirement to satisfy the criterion is also documented in the final SER. Ifthe
application does not include the specific requirement used as a basis for satisfaction of the
design criterion, the NRC requests that the application be revised to include the commitment in
the design basis of the plant. An example could be a request for a particular test or inspection
to be included in the plant initial test program if it was not already included.
-19-
-
19-Revision 0 - January 2014
For
example, a technical reviewer may determine that an "A2" system flow rate needs to be at
least 40 gallons per minute to support a finding of reasonable assurance. The reviewer may
determine, based on the safety and risk significance classification of the SSC, that a detailed
analysis or independent calculation is not necessary for this parameter and the information
provided in the applicant's submittal is sufficient to support the safety finding. System
performance will be verified during pre-operational testing to satisfy the ITAAC associated with
the minimum system flow rate.
The reviewer has options regarding the best way to incorporate the requirement for the
performance test. These options are informed by the safety and risk categorization of the
particular SSC. The test requirement could, be included in the applicant's ITAAC submittal as a
Tier I or Tier 2 item in the Initial Test Program (ITP), the test requirement could be added as a
COL action item, or the reviewer could request the applicant to add the test requirement to the
application submittal.
-
20
-
Revision 0 - January 2014
Paperwork Reduction Act Statement
The infomiation collections contained in the Standard Review Plan are covered by the requireme~nts of 10 CFR Part 50 and 10 CFR
Part 52, and were approved by the Office of Management and Budget, approval numbers 3150-0011 and 3150-0151.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information
collection requirement unless the requesting document displays a currently valid 0MB control number.
-
21
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SRP Introduction - Part 2
Summary of Changes
"STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR
NUCLEAR POWER PLANTS: SMALL MODULAR REACTOR (SMR) EDITION"
Standard Review Plan Introduction - Part 2 is a new SRP section not previously included in
NUREG-0800. It has been developed to provide an overview of the "Risk-informed and
Integrated Review Framework' review methodology to be used for SMR applications under
10 CFR Part 52, when applicants choose to participate in pre-application coordination with the
NRC.
-
22
-
Revision 0
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NUREG-0800
•"••
• ,,"°•
1. 0
U.S. NUCLEAR REGULATORY COMMISSION
SSTANDARD REVIEW PLAN
INTRODUCTION AND INTERFACES
REVIEW RESPONSIBILITIES
Primary :i":•-:Secondary
Licensing project manager
-
.All review organizations
-.
•..
.•
-
.•i
.,.,:ii
.. .,
I. AREAS OF REVIEW
This section provides guidance to the licensing project manager and: all review organizations
performing the review of the introductory material contained in Chapter 1 of the applicant's
safety analysis report. This is ageneral chapter for an application for a construction permit (CP)
or an operating license (OL) submitted in accordance with Title 10 of the Code of Federal
Regulations (10 CFR), Part 50 or an early site permit (ESP), a design certification (DC), or
combined license (COL) submitted in accordance with 10 CFR Part 52. This chapter is also
'
applicable to a standard design approval (SDA) or a manufacturing license (ML) application
submitted in accordance with 10 CFR Part 52. The scope of information to be reviewed in this
Standard Review Plan (SRP) chapter is that for a COL application unless otherwise noted.
Revision 2 - December 2011
USNRC STANDARD REVIEW PLAN
This Standard ,Review Plan, NUREG-,0.80.0., has been prep~aredl to establish criteria that t~he.U.S. Nuclear Regulato~ry C.ommission
staf responsible for the review ofapplications toconstruc[ and operate nuclear power pla~nts intends tO, use in evaluating whether an
applicantllicensee, meets the NRCL'S regulations. The Standard Review Plan is not a substitute fOr the NRCL'S regtulatons, and]
compliance with. it is not required. However, an applicant is required to identify differences between the d]esign leatures, analytical
techniq~ues, and proced.ural m~easures proposed f~or its facility and the SRP acceptance criteria and] evaluatee..ow the proposed]
altematives to the SR acceptance criteria provid]e an acceptable method of complying with the NRC regulations.
The standard review p~lan sections, are numbered in acc~orda..nce with corresloondingq sections in Regulatory Guide 1.70, "Stan~dard
,Format and Content of Safety Analysis Reports for Nuclear Power.Plants (LWR Ecdfition)." Not all sections of" Regulatory. Guide 1.70
have a correspond]ing review p1tan section. The SR sections ap~plicable to a combined license appltication for a new Ii ht-water
reactor (LWR) are based on Regulatory Guide 1.206, "Combinedl License Applications for Nuclear Power Plants(LRdion.
These ,documentsare made .ava~ilable to t~he .publicas part of the NRC's p~olicy to inform t~he nuclear industry and the general public
Of regulato~ry procedures and p~olicies., Individ]ual sections o~fNUREG-0800.will be revised period~ically, as. ap~propriat•e to
accommrodate comments and to reflect new information and experience. C;omments may be submi~tted electronically by email to
NtRRStRP•nrc.gov."•
Requests for single copies of SRP sections ,(which may .be reproduced) s.ho.uld .bemade to. th.e U.S..Nuclear Re~qulatory
Com~mission Wvas.hing on DC 2055 Ateon: Reprodu~ction a~nd Distribut[ion, Services S•ection o.r by fax.o (3!)4.-229 rb
email.to DuS'FRIBUTI nuNcnrc.gov. E-iectronic copies of this section are available through the N•b's public vve site at
,orb
.http://www.nrc.nov/read•inq-rm/doc-co.lections/nuregs/staff/sr08Ou,h or in t~he N.RC'S Agencywide Duoc~uments Access andi
Management S•ystem (ADAM) at http://www. nrc.gov/reading-rm/adams.html, und]er Accession # M~L'I 127303931
There are two types of information presented:
*
General information that enables the reviewer or reader to obtain a basic understanding
of the overall facility without having to refer to the subsequent chapters. Review of the
remainder of the application can be accomplished with a better perspective and
recognition of the relative safety-significance of each individu~al item in the overall plant
description.
*
Specific information that relates to regulatory considerations that applies throughout the
balance of the application (e.g., conformance with the SRP acceptance criteria).
The specific areas of review are as follows:
1.
Introduction
The principal aspects of the 0verall-application are reviewed. These principal aspects
include: the type of license requested, the number• of plant units, a brief description of
the proposed plant location, the type of containment structure and its designer, the type
of nuclear steam supply system and its designer, the core thermal power levels (both
rated and design), the corresponding net electrical output for each thermal power level,
and the scheduled completion date and anticipated commercial operation date of each
unit.
2.
General Plant Description
A summary description of the principal characteristics of the site and a concise
description of the facility is reviewed. The facility description should include a brief
discussion of the principal design criteria, operating characteristics, and safety
considerations for the facility; the engineered safety features and emergency systems;
the instrumentation, control, and electrical systems; the power conversion system; the
fuel handling and storage systems; the cooling water and other auxiliary systems; and
the radioactive waste management system. The general arrangement of major
structures and equipment should also be indicated by the use of plan and elevation
drawings in sufficient number and detail to provide a reasonable understanding of the
general layout of the plant. Those features of the plant that are likely to be of special
interest because of their relationship to safety should also be identified. In addition, such
items as unusual site characteristics, solutions to particularly difficult engineering and/or
construction considerations (e.g., modular construction techniques or plans), and
significant changes in technology represented by the design should be highlighted.
3.
Comparison with Other Facilities
A comparison with other facilities of similar design and comparable power level is
reviewed.
4.
Identification of Agents and Contractors
1.0-2
1.0-2Revision 2 - December 2011
The primary agents or contractors for the design, construction, and operation of the
nuclear power plant are reviewed. The principal consultants and outside service
organizations (such as those providing audits of the quality assurance program) are also
reviewed. The division of responsibility between the reactor/facility designer(s),
architect-engineer(s), constr'uctor(s), and plant Operator should also be delineated by the
applicant.
5.
Performance of New Safety Features
For a DC application or COL application that does not reference a certified design, this
review addresses information or references to the location of information that
demonstrates the performance of new safety features for nuclear power plants that differ
significantly from light-water reactor (LWR) designs licensed before 1997, or use
simplified, inherent, passive, or other innovative means to accomplish their safety
functions.
6.
-
Material Referenced
A table of all topical reports and technical reports that are incorporated by reference as
-part of the application is reviewed. In this context, "topical reports" are defined as
reports that have been prepared by reactor designers and manufacturers,
architect-engineers, or other organizations, and filed separately with the U.S. Nuclear
Regulatory Commission (NRC) in support of this application or other applications or
product lines. For each topical report, this table should include the report number and
title, the date on which the report was submitted to the NRC, and the sections of the.
COL application in which the report is referenced. For any topical reports that have
been withheld from public~disclosure as proprietary documents pursuant to 10 CFR
2.390(b), this table should also reference nonproprietary summary descriptions of the
general content of each such report.
A table of any documents submitted to the Commission in other applications that are
incorporated in whole or in part by reference in the application is reviewed. If any
information submitted in connection with other applications is incorporated by reference
in this application, summaries of such information should be included in appropriate
sections of this application, as necessary, to provide clarity and context.
Results of test and analyses may be submitted as separate reports. In such cases,
these reports should be referenced in this section and summarized in the appropriate
section(s) of the final safety analysis report (FSAR).
7.
Drawings and Other Detailed Information
A table of all instrumentation and control (I&C) functional diagrams, as well as electrical
one-line diagrams cross-referenced to the related application section(s), including
legends for electrical power, I&C, lighting, and communication drawings is reviewed.
1.0-3
1.0-3Revision 2
-
December 2011
diagrams) and system
(e.g., pipingto and
drawings
table of system
A
the instrumentation
related section(s) of the application is
cross-referenced
that are
designators
reviewed. This information should include the applicable drawing legends and notes.
1.0-4
1.0-4Revision 2 - December 2011
8.
Interfaces with Standard Designs
For a DC application or a COL application referencing a DC, this SRP section addresses
interface requiremerits contained in a DC application and a COL application that
references a certified design. For a DC, this review will address interface requirements
for those design features that are outside the scope of the certified design as identified
by the applicant and a representative conceptual design for those portions of the plant
for which the application does not seek certification; 10 CFR 52.47(a)(24) requires a
conceptual design and 10 CFR 52.47(a)(25) sets forth interface requirements for
out-of-scope portions of the design. Inspection, test, analysis, and acceptance criteria
(ITAAC), required by 10 CFR 52.47(b)(1), apply only to in-scope portions Of the design
and are not related to 10 CFR 52.47(a)(24) and (25). For a COL, this review will
address how a COL application addresses the interface requirements established for the
design. The COL.review will be based on complete design information, as any
.
•conceptual design information included in a DC FSAR will be replaced by site-specific
infor'mation.
COL Action Items
A table of information demonstrating how COL action items were addressed, or
appropriate FSAR section references as to where this information is provided, is
reviewed. COL applicants may also include a consolidation, in an appropriate section of
the COL application, of those COL action items that cannot be completely resolved
before the CCL is issued, as well as any post-licensing information commitments made
to the NRC as part of the license application review. The CCL applicant may propose
such post-licensing commitments as ITAAC, license conditions or FSAR commitments to
ensure completion of these items.
Departures
A table listing the departures and applicable FSAR section(s) is reviewed along with the
departure report submitted in accordance with the applicable appendix to 10 CFR
Part 52.
9.
Conformance with Regulatory Guidance
Regulatory Guides (RGs)
A table of conformance with the NRC's RGs that are applicable to the application is
reviewed. The table should also include an identification and description of deviations
from the guidance contained in the NRC's RGs, as well as suitable justifications for any.
alternative approaches proposed by the COL applicant with appropriate references to
the FSAR sections where they are addressed.
Conformance with the Review Guidance
An evaluation of the facility against the SRP in effect 6 months before the docket date of
the application is reviewed. The evaluation should include an identification and
1.0-5
1.0-5Revision 2 - December 2011
description of all differences in design features, analytical techniques, and procedural
measures proposed for the facility and those corresponding features, techniques, and
measures given in the acceptance criteria in the review guidance. Where differences
exist, the evaluation should discuss or provide references to the FSAR section that
describes how the proposed alternative provides an acceptable method of complying
with the Commission's regulations that underlie the corresponding acceptance criteria.
The regulations specify that this evaluation is made against the SRP in effect 6 months
before the docket date of the application; however, as a practical matter the evaluation
should be performed against the guidance in effect 6 months before the submittal date of
the application ...
Generic Issues and Three Mile Island (TM I) Requirements
A table that identifies proposed technical resolutions for those unresolved safety issues
and medium- and high-priority generic safety issues that are identified in the version of
NUREG-0933 current on the date up to 6 months before the submittal date of the
application and that are technically relevant to the design and identifies FSAR section
references where the resolutions are addressed is reviewed. The table also identifies
TMI requirements set forth in 10 CFR 50.34(f).
Part 21 Notification of Failure to Comply or Existence of a Defect and its Evaluation
An evaluation by the applicant of all defects and noncompliance reports submitted under
10 CFR Part 21 to determine their applicability and potential impacts on applications for
design certification (DC), DC renewal, and combined licenses (COLs) that reference a
DC is reviewed. For DC renewals and COLs that reference a DC, the evaluation should
address those notifications issued between the original certification and the DC renewal
or COL application as stipulated in 10 CFR 21.21.
Operational Experience (Generic Communications)
Information from the applicant that demonstrates how operating experience insights from
generic letters and bulletins issued after the most recent revision of the applicable
standard review plan and 6 months before the docket date of the application, or
comparable international operating experience, have been incorporated into the plant
design is reviewed.
Advanced and Evolutionary Light-Water Reactor Design Issues
A table that identifies information addressing applicable issues developed by the NRC
and documented in SECY-93-087 and the associated staff requirements memorandum
for advanced and evolutionary LWR designs is reviewed.
10.
Nuclear Power Plants to be operated on Multi-Unit Sites
This section addresses the review of an evaluation of potential hazards to the structures,
systems, and components (SSCs) important to safety of the operating units resulting
from construction activities, as well as a description of the managerial and administrative
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1.0-6Revision 2 - December 2011
controls to be used to provide assurance that the limiting conditions for operation are not
•exceeded as a result of construction activities at multi-unit sites.
Review Interfaces
Other SRP sections interface with this section as follows:
1.
The general information provided in each area of review enables the reviewer or reader
to obtain a basic understanding of the overall facility without •having to refer to the
subsequent chapters. Review of the detailed chapters that follow can then be
accomplished with a better understanding of the relative safety-significance of each
individual item in the overall plant design.
2.
The specific information provided in each area of review provides references to where
the regulatorY considerations are addressed throughout the balance of the application.
The specific acceptance criteria and review prociedures are contained in the applicable SRP
sections.
I1.
ACCEPTANCE CRITERIA
Requirements
Acceptance criteria are based on meeting the relevant requirements of the following
Commission regulations:
1.
10 CFR 50.33, 10 CFR 50.34, 10 CFR 52.16, 10 CFR.52.17, 10 CFR 52.46,
10 CFR 52.47, 10 CFR 52.77, and 10 CFR~ 52.79, as they relate to general introductory
matters.
2.
Regulations governing Interfaces with standard designs, including:
A.
10 CFR 52.47(a)(24) requires the DC application to contain a representative
conceptual design for those portions of the plant for which the application does
not seek certification, to aid the NRC in its review of the design control document
(DCD) and to permit assessment of the adequacy of the interface requirements
in paragraph (a)(25) of 10 CFR 52.47.
B.
10 CFR 52.47(a)(25) requires the DC FSAR to contain the interface requirements
to be met by those portions of the plant for which the application does not seek
certification. These requirements must be sufficiently detailed to allow
completion of the FSAR.
C.
10 CFR 52.47(a)(26) requires the DC FSAR to contain justification that
compliance with the interface requirements of paragraph (a)(25) of 10 CFR 52.47
is verifiable through inspections, tests, or analyses. The method to be used for
verification of interface requirements must be included as part of the proposed
ITAAC required by paragraph (b)(2) of 10 CFR 52.47.
1.0-7
1.0-7
Revision 2 - December 201,1
0.
10 CFR 52.79(d)(2) requires that for a COL referencing a standard DC, the
FSAR demonstrate that the interface requirements established for the design
under 10 CFR 52.47 have been met.
3.
10 CFR 50.34(h), 10 CFR 52.17(a)(1)(xii), 10 CFR 52.47(a)(9), and 10 CFR 52.79(a)(41)
as they relate to an evaluation of the application against the applicable NRC review
guidance in effect 6 months before the docket date of the application.
4.
10 CER 52.47(a)(21) and 10 CFR 52.79(a)(20) as they relate to proposed technical
resolutions of those unresolved safety issues and medium and high priority generic
safety issues, which are identified in the version of NUREG-0933 current on the date up
to 6 months before the docket date of the application and which are technically relevant
to the design.
5.
10 CFR 50.34(f)1 , 10 CFR 52.47(a)(8) and 10 CFR 52.79(a)(1 7) as they relat~to
S compliance with technically relevant positions of the TMI requirements.
6.
10 CFR 52.47(a)(22) and 10 CFR 52.79(a)(3-7) as they relate to the information
necessary to demonstrate how operating experience insights have been incorporated
into the plant design.
7.
10 CFR 50.43(e) as it relates to requirements for approval of applications for a DC, COL,
ML, or OL that propose nuclear reactor designs, which differ significantly from LWR
designs that were licensed before 1997, or use simplified, inherent, passive, or other
innovative means to accomplish their safety functions.
8.
10 CFR 52.79(a)(31) regarding nuclear power plants to be operated on multi-unit sites,
as it relates to-an evaluation of the potential hazards to the SSCs important to safety of
operating units resulting from construction activities, as well as a description of the
managerial and administrative controls to be used to provide assurance that the limiting
conditions for operation are not exceeded as a result of construction activities at the
multi-unit sites.
9.
10 CFR 21 .21 as it relates to reviews of failure notifications and evaluations of the
impacts from operational experience and implementation of lessons learned on
engineering design for the review of DC and COL applications. The applicability,
relevancy and significance of these failure notifications in DC and COL reviews shall be
determined by the individual applicant and specific to each design center with emphasis
on significant notifications. The applicant's evaluation shall include all defects and
noncompliance reports submitted under 10 CFR 21.21 to determine 'their applicability
and potential impact on the application under review by the staff. For DC reviews, the
scope of the applicant's review should include notifications issued prior to submittal of
the DC application. For DC renewals, and COL applications that reference a DC, the
scope of the applicant's review should include those notifications issued between the
1For Part 50 applicants not listed in 10 CFR 50.34(f), the applicable provisions of 10 CFR 50.34(f) will be
made a requirement during the licensing process.
1.0-8
1.0-8Revision 2
-
December 2011
original design certification rule (DCR) and submittal of the DC renewal, or COL
application that references the OCR, as applicable.
1.0-9
1.0-9Revision 2
-
December 2011
SRP Acceptance Criteria
Specific SRP acceptance criteria acceptable to meet the relevant requirements of the NRC's
regulations identified above are as follows for the review described in this SRP section. The
SRP is not a substitute for the NRC's regulations, and compliance with it is not required.
However, an applicant is required to identify differences between the design features, analytical
techniques, and procedural measures proposed for its facility and the SRP acceptance criteria
and evaluate how the proposed alternatives to the SRP acceptance criteria provide acceptable
methods of compliance with the NRC regulations.
1.
There are no specific SRP acceptance criteria associated with these general
requirements.
2.
For regulatory considerations, acceptance is based on addressing the regulatory
requirements as discussed within this FSAR section or within the referenced FSAR
section. The SRP acceptance criteria associated with the referenced section will be
reviewed within the context of that review.
"3.
For performance of new safety features, the information is sufficient to provide
reasonable assurance that (1) these new safety features will perform as predicted in the
applicant's FSAR, (2) the effects of system interactions are acceptable, and (3) the
applicant, provides sufficient data to validate analytical codes. The design qualification
testing requirements may be met with either separate effects or integral system tests;
prototype tests; or a combination of tests, analyses, and operating experience.
Ill.
REVIEW PROCEDURES
The reviewer will select material from the procedures described below, as may be appropriate
for a particular case.
:.
These review procedures are based on the identified SRP acceptance criteria. For deviations
from these acceptance criteria, the staff should review the applicant's evaluation of how the
proposed alternatives provide an acceptable method of complying with the relevant NRC
requirements identified in Subsection II.
1.
General information
The licensing project manager will review the information for sufficiency to enable the
reviewer or reader to obtain a basic understanding of the overall facility without having to
refer to subsequent chapters.
2.
Regulatory considerations
The licensing project manager will review the information for sufficiency. The licensing
project manager will coordinate the reviews of the Specific technical issues as
referenced.
3.
Potential hazards from construction to SSCs important to safety on an operating unit.
1.0-10
1.0-10Revision 2 - December 2011
The licensing project manager will review the evaluation and-consult with the
organization responsible for the review of site hazards and the operating reactor project
manager.
4.
Post-licensing commitments.
The licensing project manager will review any post-licensing commitments proposed by
the COL applicant and will consult with the organization responsible for the review of the
technical areas associated with these commitments. The project manager should
ensure that no post-licensing information commitment involves information that is
necessary for the staff's determination regarding COL issuance.
IV.
EVALUATION FINDINGS
The licensing project manager, with support from the identified technical reviewers, verifies that
the applicant has provided sufficient information and that the review and evaluations
(if applicable) support conclusions of the following type to be included in the staffs safety
evaluation report (SER). The reviewer also states the bases for those conclusions.
As applicable to the type of license application, the applicant has provided sufficient information
to enable the reviewer or reader to obtain a basic understanding of the overall facility without
having to refer to subsequent chapters.
The applicant provided sufficient information to address the regulatory considerations, including
potential hazards to SSCs of the operating reactor as a result of construction (if applicable).
The staff concludes the requirements identified above have been met.
The licensing project manager, with support from the identified technical reviewers, determines
the most appropriate post-licensing commitment option for any COL action items that cannot be
completely resolved before license issuance, as well as any post-licensing information
commitments made to the NRC as part of the license application review. The project manager
should ensure that no post-licensing information, commitment involves information that is
necessary for the staff's determination regarding COL issuance. Guidance for making this
determination is provided in Appendix A. This evaluation of post-licensing commitments should
be included in the SER associated with the NRC staff's review of the COL application.
For DC applications, the licensing project manager, with support from the identified technical
reviewers, verifies that COL action items are identified correctly and that the scope of
responsibility is appropriately defined for the COL applicant. See definition of "COL action item"
in Section VI for further discussion.
V.
IMPLEMENTATION
The staff Will use this SRP section in performing safety evaluations of DC applications and
license applications submitted by applicants pursuant to 10 CFR Part 50 or 10 CFR Part 52.
Except when the applicant proposes an acceptable alternative method for complying with
1.0-11
1.0-11Revision 2 - December 2011
specified portions of the Commission's regulations, the staff will use the method described
herein to evaluate conformance with Commission regulations.
The provisions of this SRP section apply to reviews of applications submitted 6 months or more
after the date of issuance of this SRP section, unless superseded by a later revision.
VI.
DEFINITIONS
The following definitions are used in the context of ESPs:
Site Characteristics:
Based on site investigation, exploration, analysis and testing, the
applicant initially proposes a set of site characteristics. These site
characteristics are the actual physical, environmental and
demographic features of a site. Site characteristics, if reviewed
and approved by the staff, are specified in the ESP. In general,
site characteristics may fall into one of four categories, namely:
(1) severe natural phenomenaL(e.g., tornado wind speed, probable
maximum flood, maximum groundwater levels); (2) physical
features of the site (e.g., soil strength, topography); (3) boundaries
or locations controlled by the applicant (e.g., exclusion area
boundary, low population zone); and (4) characteristics relating to
nearby human activities (e.g., x/Q for the nearest resident, meat
animal, or vegetable garden; distances to nearby man-made
hazards to the new plant).
Plant Parameter Envelope:
A plant parameter envelope (PPE) sets forth postulated values of
design parameters that provide design details to support the NRC
staff's review of an ESP application. A controlling PPE value, or
bounding parameter value, is one that necessarily controls the
value of a site characteristic. As the PPE is intended to bound
multiple reactor designs, the actual design selected in a COL or
CP application referencing an ESP would be reviewed to ensure
that the design fits within the bounding parameter values.
Otherwise, the COL or CP applicant would need to demonstrate
that the design, given the site characteristics in the ESP, complies
with the Commission's regulations. Should an applicant reference
an ESP for a design that is not certified, the applicant would need
to demonstrate that the design's characteristics fall within the
bounding parameter values.
Permit Condition:
The Commission's regulation in 10 CFR 52.24 authorizes the
inclusion of limitations and conditions in an ESP. The staff should
recommend a permit condition in three typical circumstances: (1)
the staff's evaluation in the SER rests on an assumption that is not
currently supported, and which is practicable to support only after
ESP issuance (e.g., subsurface conditions discovered upon
excavation for foundation construction); (2) a site physical attribute
is not acceptable for the design of SSCs important to safety (such
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a condition may call for action to remedy the deficiency, e.g.,
cracked or weathered rock that is not acceptable for bearing
-foundation
loads is replaced or filled with lean concrete or
otherwise treated so as to be acceptable) (the attribute may be
deficient only with respect to a particular type of reactor); or (3) the
staffs evaluation depends on a future act (e.g., a state regulatory
approval may be called for). A permit condition is not needed
when an existing NRC regulation requires a future regulatory
review and approval process .to ensure adequate safety during
design, construction, or inspection activities for a new plant.
The following definitions are used in the context of ESPs and DC reviews:
COL action item:
COL action items identify certain matters that shall be addressed
in the FSAR by an applicant who submits a CDL application that
references a DCand/or an ESP. The term "COL holder item" is
not defined and shall not be used. CDL action items constitute
information requirements, but do not form the only acceptable set
of information in the ESAR. An applicant may depart from or omit
these items, provided that the departure or omission is identified
and justified in the FSAR. In addition, these items do not relieve
an applicant from any requirement in 10 CFR.Parts 50 and 52 that
govern the application. That is, DCs were not intended to identify,
as CDL action items, all the requirements that a CDL applicant
needs to meet to demonstrate compliance with 10 CFR'Part 52,
"Subpart C - Combined Licenses." Therefore, for a CDL
application that references a DC or an ESP, it may not be
sufficient for the CDL applicant to address only those CDL action
items contained in the referenced DC or ESP. The CDL applicant
must demonstrate compliance with all the regulatory requirements
in 10 CFR 52.79 and 10 CFR 52.80 whether they are addressed
by. a CDL action item or not.
CDL action items may contain requirements for information that
are necessary for the NRC to review to make its license
determination. This information must be provided as part of the
CDL application and cannot be deferred until after license
issuance. CDL action items may also include requirements for
providing updated FSAR information or updates to other licensing
basis documents. Completion of these types of CDL action, items
may be deferred as post-licensing commitments. After issuance
of a construction permit or CDL, these items are not requirements
for the licensee unless such items are restated as conditions of
the license.
Further, the staff may identify CDL action items with respect to
individual site characteristics in order to ensure that particular
significant issues are tracked and considered during the CDL
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application phase. For example, since control room air intake
design and location are not yet specified, a COL action item is
warranted with respect to the evaluation of the dispersion of
airborne radioactive materials to the control room.
The COL action items need not and should not be exhaustive.
Rather, COL action items should focus on matters that may be a
significant issue in any COL application referencing the particular
ESP. COL action items should not normally be needed for
matters controlled by permit conditions, or explicitly covered by
the postulated design parameters (i.e., within a PPE or design
described in the ESP application).
VII.
REFERENCES
..
10 CFR Part 50, as noted.
10 CFR Part 52, as noted.
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Appendix
-
A
Guidance for NRC Review of Post-Combined License
Commitments on Completion of COL Items
,Background:
COL action or information items may be included in ESPs, D~s or applications for ESPs and
D~s. Although the terms COL action item and COL information item are used interchangeably
by the NRC staff, historically, DC applicants have included the term "COL information item" in
their DCDs, while the NRC staff has used the term "COL action item" in its SERs and
regulations. This is also discussed in RG 1.206, Section C.II1.4, "COL Action or Information
Items." Applicants for COLs that reference ESPs or DCDs are required to address these COL
action or information items in their applications. The scope of information typically requested in
these COL action or information items is beyond the scope of information requirements
necessary to obtain an ESP or DC. This information typically includes site specific facility
design information and operational information for the facility such as programs and procedures.
Information Required for License Determination:
COL action or information items contained in an ESP or DOD may include information
requirements that are necessary for the NRC staff to make findings that are necessary to issue
a COL and information requirements that are not necessary for license issuance. Information
necessary for the NRC staff to issue a license cannot be deferred by a COL action or
information item and must be provided during the COL application review. COL action or
information items may also include information requirements that are not necessary for license
issuance, and therefore, may be deferred. Deferred actions may include such items as
providing as-built design information or to provide notifications to the NRC regarding schedules
for implementation of programs or for commencement of certain activities. During reviews of
applications, the NRC staff may request that applicants not combine the two types of
information requirements (i.e., licensing and post-licensing) into one COL action or information
item, but rather include them in separate COL items.
No "COL Holder Items":
Although the timing for providing the t•wo categories of information (i.e., licensing and
post-licensing) may be reasonably determined by an applicant for an ESP or DCD, it is not
the purview of these applicants to determine the appropriate timing for the COL applicant to
complete these items. Recently, attempts have been made by applicants to distinguish COL
action or information items by the timing of their completion. Those COL action or information
items that could not be completed until after the license was issued were sometimes identified
as "COL holder items." The term "COL holder item" is not defined in NRC regulations or
guidance; therefore, during the development of ESP and DC applications, the applicants should
refrain from using the term "COL holder item." Although some designs that were previously
certified may still include this term, the NRC staff should ensure that during its review of DC
applications the term "COL holder item" is not used.
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Regulatory Requirements and Guidance:
The regulations in 10 CFR Part 52 and the guidance provided in RG 1.206, "Combined License
Applications for Nuclear Power Plants (LWR Edition)," provide several options for a CCL
applicant to provide the information necessary for a license application. A COL application may
incorporate by reference an ESP, a DOD, neither, or both. As such, during its reviews of license
applications, the NRC staff may encounter COL applications that include different combinations
of the permitted options. In addition, the regulation in 10 CFR 52.55(c)-permits a CCL applicant
to reference, at its own risk, a design for which a DC application has been docketed, but not
granted. Only a few of the designs that have been certified by the NRC are currently being
referenced by COL applicants. Certification of those designs took place several years ago and,
as a result, the scope and nature of CCL action or information items included in those certified
designs may vary from those currently being proposed in DC applications that have been
submitted to the NRC more recently. The NRC staff should expect that the nature and scope of
CCL action or information items included in more recent applications are more clearly defined.
This expectation is a reasonable outcome of the implementation and on-going maturation of the
new licensing process specified in 10 CFR Part 52. The NRC staff should be cognizant of the
differing nature and scope of CCL action or information items that were included in previously
certified designs that are now being referenced in CCL applications and the ramifications on
completion of these items. For example, in more recent applications, ITAAC have been used to
verify the as-built reconciliation of piping designs, whereas CCL action or information items may
have been used for this purpose in previously certified designs. For COL action or information
items that cannot be completed until after license issuance, appropriate post-lcensing...
commitments should be identified for these items.
The following review guidance is provided for the NRC staff in determining post-licensing
commitment options and includes examples that illustrate differences in CCL items as
discussed above.
Guidance on Post-combined License Commitment Options:
A CCL applicant that references a certified design is required to provide information that
addresses the CCL action items (see Section IV.A.2.e of the DCRs). Likewise, an ESP may
contain terms and conditions that must be satisfied by a CCL applicant referencing an ESP to
allow NRC staff issuance of the CCL. In addition, a CCL applicant may include a commitment
to perform an action following issuance of the license (e.g., update information, provide
schedules, etc.) that is related to site-specific design features or programs for the facility that
were not identified in an ESP or DOD that it references. CCL items associated with information
that is not necessary to issue the license are identified as post-licensing commitments. The
following options are provided for identifying these post-licensing commitments:
*
*
*
ITAAC
License conditions
FSAR (or other licensing basis document) information commitments
The above options are not limited to CCL action items that cannot be completed prior to license
issuance, but may also be used for post-licensing information commitments that were identified
during COL application reviews that were not associated with COL action items. COL
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appiicants may propose one or more of the options above for completing COL items as
post-licensing commitments, but are not required to do so. In the case where a COL applicant
proposes post-licensing commitments, the NRC Staff will review the COL applicant's proposal,
confirm the acceptability of the applicant's proposal or modify it, as appropriate, land document
the final determination in their safety evaluation. If the CDL applicant does not provide a
proposal on post-licensing commitments, the NRC staff, based on its review of the COL
application and other docketed correspondence including request for additional information
responses, may include appropriate post-licensing commitments in the SER. In either case, the
NRC staff should provide the final determination of the most appropriate post-licensing
commitment from the options provided above and the applicant's FSAR should be revised to
conform to the staffs final SER determination, as necessary. To assist with this determination,
the NRC staff should consider the following review guidance:
ITAAC:
The requirement for inclusion of ITAAC in an application for a CDL is set forth in
10 CFR 52.80(a), which states that the application must contain:
The proposed inspections, tests, and analyses, including those applicable to
emergency planning, that the licensee shall perform, and the acceptance
criteria that are necessary and Sufficient to provide reasonable assurance that,
if the inspections, tests, and analyses are performed and the acceptance
criteria met, the facility has been constructed and will be operated in conformity
with the combined license, the provisions of the Act, and the Commission's
•rules and regulations. (Emphasis added)
The licensee is required by regulation to provide notification along with sufficient documentation
to demonstrate successful completion of ITAAC in accordance with 10 CFR 52.99(c). The NRC
is required to ensure that the prescribed ITAAC are performed and to publish notices in the
Federal Register of the NRC staff's determination of the licensee's successful completion of
inspections, tests, and analyses per 10 CFR 52.99(e). Following that, the NRC must find that
the acceptance criteria of the ITAAC are met in order to authorize operation of the facility per
10 CFR 52.103(g)....
Guidance for development of ITAAC, as well as additional considerations for ITAAC, is provided
in RG 1.206, Sections C.I1.1, C.lI1.1, and C.III.7. NRC staff review guidance on ITAAC is
provided in SRP Section 14.3. When determining whether a post-licensing information
commitment or a CDL action item that cannot be completed until after license issuance should
be treated in an ITAAC or not, the NRC staff should use the same guidance and criteria
provided in SRP Section 14.3. ITAAC are considered a post-licensing verification program,
whose focus is on ensuring that the as-built condition of the plant complies with the license for
the facility and the Commission's regulations. Another consideration for ITAAC is that
completion of lTAAC, by definition, must take place prior to fuel load. The licensee must
successfully complete all ITAAC in order for the Commission to make the findings prerequisite
to fuel load as required by 10 CFR 52.103(g).
New ITAAC proposed by a CDL applicant referencing a certified design to address completion
of designs, reconciliation of portions of the as-built facility with the design of the facility, etc.,
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within the scope of the referenced certified design may only be included in a COL application in
accordance with the change process described in Section VIII, Processes for Changes and
Departures, of the associated DCR. The NRC staff should review any new ITAAC proposed by
a COL applicant in accordance with the guidance provided in SRP Section 14.3. In addition,
NRC staff review should include the applicant's use of and compliance with the change
processes described in Section VIII of the associated DCR. COL applicants have typically
included their ITAAC and any necessary departures and exemptions in Part 10 of their
applications. The NRC staff should use caution in attempting to create new ITAAC to address a
COL action item that cannot be completed until after issuance of the license. Section VI.D of
the DC rules contained in the Part 52 Appendices explicitly states:
0. Except in accordance with the change processes in Section VIII of this attachment, the
Commission may not require an applicant or licensee who references this attachment
to:
1. Modify structures, systems, components, or design features as described in the
generic DCD;
2. Provide additional or alternative structures, systems, components, or design
features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing, analyses, acceptance
criteria, or justification for structures, systems, components, or design features
discussed in the generic DCD.
Note that the above requirements do not apply when a COL application references an
application for DC. In this case, the NRC staff has more latitude within the context of the
design-centered working: group to discuss with the DC and COL applicants the potential for
adding new ITAAC in the DCD. For site-specific elements or custom COL applicants, which do
not reference certified designs, the staff should review the application, as appropriate, to
determine if the proposed ITAAC are necessary and sufficient for the Commission to make the
findings required by the Atomic Energy Act.
License Conditions:
The liCense for a nuclear facility contains terms and conditions for operation. Section 50.54 of
the Commission's regulations identifies the standard conditions, with some exceptions, that are
applicable to every COL issued. In addition to those standard conditions, additional license
conditions may be proposed by the COL applicant to address completion of post-licensing
information commitments or COL action items that cannot be completed until after license.
issuance. A license condition, however, is not necessary for. those matters already covered by
the license, including Technical Specifications, or regulations. License conditions may be
proposed by COL applicants; however, there is no requirement to do so. Any license conditions
proposed by the COL applicant shall be reviewed by the NRC staff. The NRC staff will make
the final determination as to the appropriateness of the proposed license conditions, may modify
proposed license conditions or include new conditions. In addition, in cases where COL
applicants have not proposed any license conditions, appropriate license conditions may be
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*imposed
NRC staff. The NRC staff should document any such license conditions
necessarybytothe
complete
COL items in the SER.
The following discussion should be considered by the NRC staff for review of license conditions
proposed by a COL applicant and for those license conditions that the NRC staff determines are
necessary to impose on the licensee:
*License conditions remain in effect for the licensee until satisfactorily completed and
their removal is approved via the license amendment process per 10 CFR 52.98(f)..
*License conditions are enforceable the same way a regulation or order is enforceable.
*In contrast to completion of an ITAAC, where a licensee is required to make a
submission to the NRC staff documenting satisfactory completion of the ITAAC, there
need not be submission requirements asso5ciated~with completion of a license condition
that necessitate further NRC reviews. However, there may be some conditions
specifically included in the license that require the licensee to notify the NRC of the
schedule of availability of information for inspection or implementation schedules of
programs or activities to be inspected. For example, license conditions may be used to
identify notification commitments to the NRC on when activities associated with
completion of SSC design governed by design acceptance criteria (DAC) have been
completed following issuance of the license and are available for inspection by the NRC.
The NRC staff should use caution when including requirements in license conditions
such as "submission" and "staff review" since these typically describe actions taking.
during the license review. The NRC staff should instead consider use of terms like
"reporting requirements" or "make available for inspection," as more appropriate.
*License conditions may be used to include operational restrictions for the facility, impose
restrictions on operating power levels, require the performance of special tests, impose
operational constraints associated with implementation of specific design features (e.g.,
containment sump screen sweepers, etc.).
*
License conditions may be used to include implementation schedules for operational
programs as discussed in RG 1.206, Sections C.I and C.II1.1, Table 13.4.-
Exampies:
(1)
In a section of a previously certified design describing spent fuel racks, the DCD
identifies that the COL holder will implement a spent fuel rack Metamic coupon
monitoring program when the plant is placed into commercial operation. This
program will include tests to monitorbubbling, blistering, cracking, or flaking;, and a
test to monitor for corrosion, such as weight loss measurements and or visual
examination. In this example, the commitment was previously characterized as a
"COL holder item" since it cannot be completed until after license issuance. Based
on the guidance above, either the COL applicant could propose or the NRC staff
could impose a license condition to address this item. The licensee should develop
a program for performing spent fuel rack coupon monitoring and evaluation.
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Although
program
not C.II1.1,
considered
operational
programofasthis
discussed
RG 1.206,this
Sections
C.lisand
Tablean13.4,
implementation
program in
following issuance of a license can be imposed using a schedule milestone
contained in a license condition.
(2)
In a section of a previously certified design describing turbine design and the
requisite maintenance and inspection that form part of the basis for turbine missile
generation assumptions, the DCD identifies that the COL holder will submit to the
NRC staff for review prior to fuel load, and then implement a turbine maintenance
and inspection program. The program will be consistent with the maintenance and
inspection program plan activities and inspection intervals identified in another
section of the DCD. The COL holder will have available plant-specific turbine rotor
test data and calculated toughness curves that support the material property
assumptions in the turbine rotoranalyses after the fabrication of the turbine and prior
to fuel load. in this example, the commitment was previously characterized as a
"COL holder item" Since it cannot be completed until after license issuance. Based'
on the guidance above, either the COL applicant could pr~opose or the NRC staff
could impose a license condition to address this item. In this example, it is important
to point out the sensitivity and appropriateness of using the phrase "submit to the
NRC staff for review prior to fuel load" in a license condition. A licensing decision
must be based on turbine missile generation information already provided to the
NRC in the DCD or the COL application. The license condition allows for
confirmation by the NRC via inspection that the as-built information is bounded by
the original assumptions regarding turbine missile generation. In this example, the
NRC staff should use more appropriate language such as "available for NRC
inspection" in the final language for the license condition, although a more detailed
reporting requirement may be appropriate. It should be noted that more recent DC
applications have included this as-built confirmation in an ITAAC rather than a COL
action item. In addition, the licensee must implement a maintenance and inspection
program that is not an operational program as discussed in RG 1.206, Sections 0.I
and C.III. 1, Table 13.4, but implementation of the program validates assumptions
related to turbine missile probability. Scheduling the availability-of the confirmatory
evaluation and implementation of the program for NRC inspection following issuance
of a license can be determined using a schedule milestone contained in a license
condition.
FSAR Commitments:
Another way for CCL applicants to address completion of post-licensing information
commitments or CCL actions items that cannot be completed until after license issuance is
-through an FSAR commitment. In this context, an FSAR commitment is a commitment to
provide updated information in the FSAR, which contains the design-basis portion of the
licensing basis, or other licensing basis documents that has been considered appropriate by the
NRC staff to ensure that the licensing basis for the facility is up-to-date. This approach may
also be used for other licensee controlled documents such as Quality Assurance plans,
emergency plans, etc. Based on past experience with currently operating reactors, it is
important for licensees to maintain their licensing bases documents up-to-date. The NRC and
its licensees have dealt with several issues resulting in significant efforts over the years that
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emphasize the importance of maintaining a current licensing basis (CLB) and a discussion on
CLB is provided for information following this section. These efforts have involved issues
related to loss of configuration control, design-basis reconstitution, commitment management
and commitment change control.
The staff has identified two approaches for providing the information necessary to maintain the
design-basis for the facility: 1) include specific design-basis information items in a license
condition, and 2) include design-basis information in FSAR updates required by
10 CFR 50.71(e). In the first approach, the focus is on ensuring that FSAR information that
is identified during the COL review process and is necessary to include in the design-basis is
included in an FSAR update. In the second approach, the focus is on ensuring that routine
FSAR updates that have traditionally occurred following issuance of an OL are performed.
These routine FSAR updates are typically associated with:
*
*
*
*
Changes to the facility in accordance with the requirements of 10 CFR 50.59
Changes to the facility resulting from approved exemptions and departures from a
referenced certified design
Changes to the facility resulting from approved variances from a referenced ESP
Amendments to the license in accordance with the requirements of 10 CFR 50.90
The two approaches for FSAR information commitments are discussed below:
ESAR information commitment included in a license condition
•
The regulations in 10 CER 50.71(e) and the appendices to 10 CFR Part 52 that contain the
DCRs include requirements for holders of COLs to update their FSARs. Specifically,
10 CFR 50.71 (e)(3)(iii) requires that an update of the FSAR be Submitted annually to the NRC
during the period from the docketing of a COL application until the Commission makes the
finding under 10 CFR 52.103(g). In addition, 10 CFR 50.71 (e)(4) requires that subsequent
FSAR revisions be filed annually or 6 months after each refueling outage provided the interval
between successive updates does not exceed 24 months. These revisions must reflect all
changes up to a maximum of 6 months prior to the date of the filing. Although these
requirements for FSAR. updates currently exist, the focus of FSAR information commitment
items included in a license condition is to ensure the inclusion of specific information identified
during the initial licensing review that should be included in the design-basis for the facility. This
includes the information that should be reviewed as part of the design-basis for the facility when
reviews and evaluations such as those performed in accordance with 10 CFR 50.54(f), 10 CFR
50.59, 10 CFR 50.65, etc., are required. The staff believes that use of a license condition for
inclusion of specific FSAR information commitments provides an appropriate enforcement
mechanism for ensuring an up-to-date licensing basis. The license condition should also
include a milestone schedule for ensuring that the specific FSAR information identified is
included in an FSAR update required by 10 CFR 50.71(e).
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Examples of the types of information that may be included in this license condition are:
*
*
*
*
•
ESAR level dlesign information from completed digital I&C DAC
ESAR level design information from as-built reconciliations of piping
Design features installed as a result of the completed pipe break hazards analyses
Update to turbine missile generation analyses, as necessary, based on as-procured
material data
Update to reactor vessel materials data, as necessary, based on as-procured vessel
material data
The NRC staff considers the above types of information to be appropriate to include in a timely
FSAR update on a schedule that is more suitable to ensuring an updated design-basis for initial
operation of new plants than that required by 10 CFR 50.71(e). For example, the updated
information would ensure that the licensing basis for the facility is up-to-date prior to loading
fuel, prior to initial criticality, prior to exceeding 5% .of the authorized power level, etc. The COL
applicant should specifically identify these FSAR information requirements and consolidate
them under a license condition that includes a proposed milestone for implementation.
The NRC staff considers this information to have sufficient relevance and distinction from the
types of information typically included in routine FSAR updates to warrant its inclusion in a
license condition. Together with the requirements of 10 CFR 50.71(e) and Part 52, this type of
license condition furthers the NRC's goal of ensuring that the design-basis for the facility
(i.e.,, the FSAR) is up-to-date when operation of the facility begins. A license condition proposed
by COL applicants that includes such FSAR commitments should be included in an appropriate
section of the COL application to facilitate identification and tracking.
Examples:
(1)
.
In a section of a previously certified design describing pipe rupture hazard
evaluations, the DCD identifies that after the COL is issued and prior to fuel load, the
COL holder will complete the as-built reconciliation of the pipe break hazards
analysis in accordance with the criteria outlined in another section of the DCD. In
this example, the commitment was previously characterized as a "COL holder item"
since it cannot be completed until after license issuance. Based on the guidance
above, either the COL applicant or the NRC staff could propose a license condition
that includes a specific FSAR information commitment to address this item. Note
that in this example, completion of the piping design was part of DAC included in the
ITAAC and other more recent DC applicants have included the as-built reconciliation
of the piping design as ITAAC. In this example, a pipe rupture hazard analysis is to
be completed following completion of the piping DAC. The completed design,
including the as-built reconciliation, is used to identify postulated break locations and
necessary layout changes, support designs and locations, whip restraint designs and
locations, and jet shield designs and locations, as necessary. The piping DAC,
approved and certified in the DCD, was sufficient for the NRC staff to make its
licensing determination. The basis for including the as-built reconciliation in a license
condition with a specific FSAR information commitment is that it provided updated
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information for the licensing basis document on the final as-installed piping, including
any necessary pipe whip restraints and/or jet shields that were installed.
(2)
In a section of a previously certified design describingthe seismic analysis of nuclear
island structures, the DCD identifies that the COL holder will reconcile, priorto fuel
load, the seismic analysis described in another section of the DCD for detail design
changes, such as those due to as-procuredor as-built changes in component mass,
center of gravity, and support configuration based on as-procuredequipment
in formation. In this example, the commitment was previously characterized as a
"COL holder item" since it cannot be completed Until after license issuance. Based
on the guidance above, either the COL applicant or the NRC staff could propose a
license condition that includes a specific FSAR information commitment to address
this item. Please note that other more recent DC applicants have included the
as-built seismic reconciliation in the ITAAC. The basis for including the as-built
seismic reconciliation in a license condition with a specific FSAR information
Commitment in this example is that an analysis was provided either in the COL or
in the referenced DOD that was sufficient for the NRC staff to make its licensing
determination. The FSAR information commitment is for the as-built reconciliation
of this analysis to be included as an update to the licensing basis document.
FSAR information commitments included in routine.FSAR update:
Updated information that does not warrant inclusion in the above categories or that occurs after
the milestone associated with the license condition should be included in the periodic FSAR
updates required by 10 CER 50.71(e). Guidance on FSAR updates is provided in RG 1.181,
"Content of the Update Final Safety Analysis Report in Accordance with 10 CFR 50.71(e),"
which endorses Revision 1 of Nuclear Energy Institute (NEI) 98-03, "Guidelines for Updating
Final Safety Analysis Reports." The guidance for these routine FSAR updates is contained in
RG 1.181 and NEI 98-03 and is typically associated with:
*
*
*
*
Changes to the facility in accordance with the requirements of 10 CFR 50.59
Changes to the facility resulting from approved exemptions and departures fromn a
referenced certified design
Changes to the facility resulting from approved variances from a referenced ESP
Amendments to the license in accordance with the requirements of 10 CFR 50.90
The following additional guidance should be considered by COL applicants when proposing
FSAR information commitments in their application:
*Completion of COL action items via an FSAR commitment cannot be used to provide
information to the NRC that is necessary to make a finding required for license issuance.
However, completion of post-licensing information commitments or a COL action item
that does not include information necessary for licensing via an FSAR commitment could
be used to ensure that the licensing basis for the facility is updated and maintained in a
current state.
1.0-23
1.0-23Revision 2 - December 2011