LD-90-005, Design Certification Licensing Review Basis Document
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9 Enclosure I to LD-90-005 COMBUSTION ENGINEERING. INC.
SYSTEM 80+" STANDARD DESIGN DESIGN CERTIFICATION LICENSING REVIEW BASIS DOCUMENT January 1990 Q ~^l3 Wgg
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2 ABSTRACT Title 10 of the Code of Federal Regulations, Part 52, provides the regulatory framework for issuance of early site permits, standard design certifications, and combined licenses for nuclear power plants.
Consistent with this rule, combustion Engineering, Inc. (C-E) has applied for the design certification of the System 80+TM Standard Design, which is described in the Combustion Engineering Standard Safety Analysis Report-Design Certification (CESSAR-DC).
This Licensing Review Basis document has been developed to (1) identify key technical issues and C-l's proposed approach for their resolution and (2) summarize the schedule and process established by C-E and NRC staff for the final design approval and design certification review. This document is consistent with Staff Requirements memoranda " Resolution Process for Severe Accident Issues on Evolutionary LWRs" and " Recommended Priorities for Review of Standard Plant Designs", dated December 15, 1989.
TABLE OF CONTENTS Section ELqa ABSTRACT
1.0 INTRODUCTION
1 1.1 Scope of the System 80+ Standard Design 1
1.2 Regulatory Compliance and Policy Issues 2
1.2.1 Design Features and Methods Departing 2
from Current Regulations 1.2.2 Design Features and Methods which 3
may Involve Questions of Commission Policy 2.0 SCHEDULE 8
3.0 CONTENT OF APPLICATION 12 3.1 CESSAR-DC Format 12 3.2 CESSAR-DC Amendment Identification 12 3.3 Incorporation of Key Requirements 12 4.0 NRC STAFF REVIEW 'iSCEDURES 15 5.0 ACRS PARTICIPATION 17 6.0 SEVERE ACCIDENT ISSVES 18 6.1 Introduction 18 6.2 TMI Requirements for New Plants 18 6.3 Resolution of USIs and GSIs 19 6.4 Probabilistic Risk Assessment 20 6.5 Severe Accident Performance Goals 21 6.5.1 Prevention of Core Daniage 21 6.5.2 Offsite Consequences of Severe Accidents 22 6.5.3 Expected Containment Performance 22 7.0 OTHER SPECIFIC ISSUES 25 7.1 Comparison of the System 80+ Design 25 with the EPRI ALWR Requirements Document 7.2 Physical Security and Sabotage 26 7.3 Site Envelope Parameters 27 7.4 Completeness of Design Documentation 27 7.5 Program for the Assurance of Quality in Design 29 7.6 Maintenance, Surveillance and Reliability 29 7.7 Safety Goal Policy Statement 30 7.8 Sixty-Year Life 30 7.9 Fire Protection 31 7.10 Station Blackout 31 7.11 Leak-Before-Break 32 i
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Section EAR 7.12 Source Term 32 7.13 Operating Basis Earthquake 33 7.14 Type C Containment Leak Rate 33 l
I 7.15 Hydrogen Generation 33 7.16 Severe Accident Containment Vents 33 7.17 Mid-Loop Operations 34 7.18 Interfacing System LOCA 34 7.19 Anticipated Transients Without Scram (ATWS) 35 7.20 Electrical System Design 35 7.21 Degraded Core Behavior 36 Appendix A System 80+ Design Differences from the A-1 EPRI ALWR Requirements Document LIST OF TABLES Tables ERM 1
Nuclear Power Plant Structures, Systems, and Components 4
Currently Included in the System 80+ Standard Design l
2 Nuclear Power Plant Structures, Systems, and Components 7
for which a Conceptual Design will be Provided 3
CESSAR-DC Major Submittals and NRC Review Schedule 9
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1.0 INTRODUCTION
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Combustioh Engineering, Inc. (C-E)_has applied for design certification of the System 80+ # Standard Design in accordance with the Commission's regulations (10 CFR, Part 52).*
R C-E is enhancing the System 80 -standard design [ described in the Combustion Engineering Standard Safety Analysis Report - FSAR (CESSAR-F), Docket No. STN 50-470F].to_ meet _the requirements of 10 CFR, Part 52, and the guidance of the NRC's Severe Accident and-Safety Goal Policy Statements.
The scope of_this improved design, the System 80+
Standard Design, covers an essentially complete nuclear power plant.
It includes all structures, systems, and components that can affect safe operation, except for site-specific elements. This design is described in the Combustion Engineering j
Standard Safety Analysis Report-Design Certification (CESSAR-DC).
Sufficient information will be provided to enable the NRC taff to
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issue the Final Design Approval (FDA) as a prerequisite to Commission certification of the System 80+ Standard Design.
Both C-E and the NRC staff believe that the NRC staff safety review of CESSAR-DC will proceed more smoothly if certain licensing review bases are established. This' Licensing Review Basis (LRB) document, therefore, has been developed to (1)-identify key technical issues and C-E's proposed approach for their resolution and (2) summarize the schedule and process established by C-E and NRC staff for the FDA and design certification review.
1.1 Stone of the System 80+ Standard Desian The scope of the System 80+ Standard Design includes an essentially j
l complete nuclear power plant.
Table 1 lists the structures, systems, and components included in the scope of the design.
The application for design certification was submitted to the NRC via'C-E letter LD-89-035, dated March 30, 1989.
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1 Table 2 lists the structures, systems and components for which a conceptual design will be provided, consistent with the requirements of 10 CFRi Part 52.
Interface requirements for structures, systems,.and components not j
included in the design scope will be provided in CESSAR-DC.
The NRC staff review of the information presented in CESSAR-DC and preparationofaSafetyEvaluationReport(SER)wifk'ensurethatall safety issues are fully addressed and that all regulatory requirements are met.
Since C-E wishes to obtain an FDA and a design certification for the System 80+ Standard Design before any applicant, site, or equipment suppliers are identified, C-E will provide the necessary level of detailed information to enable the NRC staff to complete its review without preempting competitive bidding on any future project chat references the certified design. The format and content of CESSAR-DC are described in Sections 3, 6, and 7.
1.2 Reaulatory Comoliance and Policy Issues This section lists (1) System 80+ design features and/or methods that depart from current NRC regulations and (2) other features and/or methods which go beyond current regulations and which C-E l
believes may involve questions of Commission policy.
These design features and methods are described in more detail in the LRB sections indicated.
l 1.2.1 Desian Features and Methods Deoartino from Current Reoulations o
The magnitude of the Operating Basis Earthquake is not specified to be one-half of that for the Safe Shutdown Earthquake as required by 10 CFR, Part 100, Appendix A (LRB Section 7.13). _
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o For hydrogen generation and control inside containment, the assumed extent of metal-water reaction (75%) and the maximum 4
i allodable hydrogen concentration (13%) are different than those specified by 10 CFR 50.34(f) (LRB Section 7.15).
4 The requirements for requesting exemptions from current regulations j
(10 CFR 50.12).4111 be applied where appropriate.
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1.2.2 Desion Features and Methods which may involve Ouestions of Commission
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Policy
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o The need for, and formulation of, a goal for expected i
containment performance (Section 6.5.3) j The degree to which, if any, new-design features are required o
i to improve protection against sabotage (Section 7.2)
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The degree to which, if any, new design features are required i
for improved fire protection (Section 7.9) l o
The degree to which realistic source terms can be used for PRA and severe accident evaluations (Section 7.12) o Whether there is a need for containment venting capability for severe accident mitigation beyond that required by current regulations (Section 7.16) o Whether proposed design features for prevention and mitigation of degraded core conditions are adequate (Section 7.21) )
I TABLE 1 NUCLEAR POWER PLANT STRUCTURES. SYST[MS. AND COMPONENTS CURRENTLY INCLUDED IN THE SYSTEN 80+ STANDARD DESIGN REACTOR SYSTEMS o
o FLel System j
o Furl Storage and Handling Systems i
o Chemical and Volume Control System j
o Process Sampling System i
SAFEGUARDS SYSTEMS l
o Shutdown Cooling System o
Safety Injection System i
o Safety Depressurization System o
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STEAM AND POWER CONVERSION SYSTEM 1
o Main Steam Supply o
Condensate and feedwater System o
Steam Generator Blowdown System s
o Main Condenser System o
Condensate Storage System o
Condensate Cleanup System o
Main Vacuum System i
o Demineralized Water Makeup System TURBINE GENERATOR SYSTEMS
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Turbine Generator o
Turbine Bypass System o
Turbine Gland Sealing System o
Turbine Lube Oil System i
o Turbine Control System o
Turbine Cooling System WASTE MANAGEMENT SYSTEMS o
Liquid Waste Management System o
Gaseous Waste Management System o
Solid Waste Management System o
Process and Effluent Radiation fionitoring System INSTRUMENTATION AND CONTROL SYSTEMS o
Reactor Protective System o
Engineered Safety Features Actuation System o
Control Systems Not Required For Safety o
Discrete Indication and Alarm System o
Integrity Monitoring System o
Data Processing System 4
TABLE 1 (Cont'd)
NUCLEAR POWER PLANT STRUCTURES. SYSTEMS. AND COMPONENTS CURRENTLY INCLUDED IN THE SYSTEM 80+ STANDARD DESIGN ONSITE POWER SYSTEM o
Non-lE AC Power Systems o
Diesel Generator o
Non-lE Alternate AC Power Source o
Protect've Relaying System o
Non-1E DC Power Systems CONTAINMENT STRUCTURE AND SUPPORT SYSTEMS o
Containment Pressure Boundary i
o Reactor Building o
Containment Heat Removal System o
Containment Isolation System o
Containment Spray System o
Containment Combustible Gas Control System COOLING WATER SYSTEMS o
Component Cooling Water System o
Station Service Water System o
Turbine Building Service Water System o
Turbine Building Cooling Water System o
Chilled Water Systems (Essential and Normal) o Condenser Circulating Water System SUPPORT SYSTEMS o
Compressed Air Systems (Instrument, Station, and Breathing) o Compressed Gas System o
Fire Protection System o
Communication Systems Lighting Systems (Normal, Security, and Emergency) o o
Diesel Generator Support Systems o
Equipment and Floor Drainage System o
Control Building Ventilation System o
Fuel Building Ventilation System o
Auxiliary and Radwaste Building Ventilation System o
Diesel Building Ventilation System o
Containment Purge Ventilation System o
Containment Cooling and Ventilation System o
Turbine Building Ventilation System o
Station Service Water Pump Structure Ventilation System l
i TABLE 1 (Cont'd) i NUCLEAR POWER PLANT STRUCTURES. SYSTEMS. AND COMPONENTS CURRENTLY TNCLUDED IN THE SYSTEM 8D+ STANDARD DESIGN CONTROL BUILDING o
Advanced Control Complex Master Control Console Auxiliary Console Safety Console Supervisor Console Integrated Process Status Overview Display o
Remote Shutdown Panel o
Safety Channel Equipment Rooms o
Computer Room OTHER BUILDINGS AND STRUCTURES o
Radwaste Building o
Turbine Building o
Auxiliary Building o
Fuel Building o
Emergency Diesel Generator Enclosures f
TABLE 2 1
NUCLEAR PO@ER PLANT STRUCTURES. SYSTEMS. AND COM*0NENTS FOR WHICH A CONCEPTUAL DESIGN WILL BE PROVIDED l
o Offsite Power System (Including Switchyard) o Emergency Operations Facility o
Operational Support Center o
Training Facilities (Including Simulator) o Office Space Outside the Control Complex o
Laboratory Facilities l
o Decontamination Facilities o
Ultimate Heat Sink and Intake Structures o
Warehouses o
Sewage Treatment Facilities i
o Potable and Sanitary Water Systems o
Normal Heat Sink and Intake Structures o
Security System l
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- Conceptual design descriptions and interface requirements will be provided in CESSAR-DC, consistent with the requirements of 10 CFR Part 52.
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' l 2.0 SCHEDULE The schedtile for submitting sections of CESSAR-DC chapters is shown i
in Table 3 along with the schedule for NRC review of those submittals. The major milestones are design completion by December 1990, an FDA by December 1991, and design certification by December 1992.
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i TABLE 3 l
i CESSAR-DC MAJOR
- SUBMITTALS AND NRC REVIEW SCHEDULE CESSAR-DC EXPECTED CESSAR-DC DESCRIPTION SUBMITTAL DRAFT SER CHAPTER AND SUBSECTION DATE DATE 1
Introduction and i
General Description of Plant o Sections 1.1 - 1.9 SEP 1987 o Sections 1.1 - 1.9(RI)
APR 1988 o Sections 1.1 - 1.9(R2)
MAR 1989 o Section 1.2(R3)
AUG 1990 JUN 1991 2
Site Envelope Characteristics o Sections 2.1 - 2.4 SEP 1988 o Section 2.5 AUG 1990 JUN 1991 3
Design of Structures, Components, Equipment, and Systems o Sections 3.1 - 3.5 SEP 1988
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o Sections 3.6 and 3.9 MAR 1989 o Section 3.10 APR 1990 l
o Sections 3.7, 3.8 and 3.11 AUG 1990 o Section 3.7(RI)
DEC 1990 JUN 1991 l
4 Reactor o Sections 4.1 - 4.6, 4A, 4B APR 1988 o Section 4.5(RI)
SEP 1988 o Sections 4.1 - 4.4(RI),
4.5(R2),4.6(RI)
DEC 1989 o Section 4.4(R2)
AUG 1990 MAR 1991 l
and Connected Systems o Sections 5.1 - 5.4 APR 1988 l
0 Sections 5.1 - 5.4(RI),
t 58, SC SEP 1988 l
o Sections 5.1 - 5.4(R2)
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o Section 5A AUG 1990 MAR 1991 i
6 Engineered Safety Features o Section 6.3 JUN 1988 o Sections 6.1, 6.3(RI),
6.6, 6.7 SEP 1988 l
c Sections 6.2(RI), 6.3(R2),
6.4 - 6.6(RI)
MAR 1989 o Sectirsns 6.2(R2), 6.3(R3)
AUG 1990 JUN 1991 i
- Minor subit'tals are made as necessary to reflect design changes and l
responses to NRC questions..
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TABLE 3 (Cont'd)
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l CESSAR-DC MAJOR SUBNITTALS AND NRC REVIEW SCHEDULE CESSAR-DC EXPECTED CESSAR-DC DESCRIPTION SUBMITTAL DRAFT SER CHAPTER AND SUBSECTION DATE DATE 7
Instrumentation and Controls o Sections 7.1, 7.4 - 7.7 SEP 1988 o Sections 7.2, 7.3 MAR 1989 DEC 1990 8
Electric Power o Sections 8.1 - 8.3 MAR 1989 o Section 8.3(RI)
APR 1990 MAR 1991
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9 Auxiliary Systems o Sections 9.2, 9.3 APR 1988 o Sections 9.1,'9.2(RI),
9.3(RI),9.4,9.5 MAR 1989 DEC 1990 10 Steam and Power Conversion System o Sections 10.1, 10.3, 10.4 NOV 1987 o Section 10.4(RI)
JUN 1988 o Sections 10.2(RI),
10.3(RI),10.4(R2),10A MAR 1989 DEC 1990 11 Radioactive Waste l
Management o Sections 11.1 - 11.4 MAR 1989 o Sections 11.l(RI), 11.5, 11A AUG 1990 JUN 1991 l
12 Radiation Protection o Sections 12.1 - 12.3 MAR 1989 JUN 1991 o Sections 12.1 - 12.3(RI),
12.5 AUG 1990 i
13 Conduct of Operations o Sections 13.1 - 13.6, 13A MAR 1989 DEC 1990 14 Initial Test Program o Sections 14.1, 14.2 MAR 1989 o Section 14.2(RI)
AUG 1990 JUN 1991 15 Accident Analyses o Sections 15.0- 15.7, AUG 1990 JUN 1991 15A - ISD 16 Technical Specifications DEC 1990 JUN 1991. -
TABLE 3 (Cont'd)
CESSAR-DC MAJOR SUBMITTALS AND NRC REVIEW SCHEDULE CESSAR-DC EXPECTED CESSAR-DC DESCRIPTION SUBMITTAL DRAFT SER CHAPTER AND SUBSECTION DATE DATE 17 Quality Assurance I
o Section 17.0 NOV 1987 o Section 17.0(RI)
HAR 1989 DEC 1990 18 Human Factors Engineering o Sections 18.1 - 18.4 SEP 1988 o Sections 18.1 - 18.4(RI),
MAR 1989 DEC 1990 18.5 - 18.9 i
APP. A Closure of Unresolved and Generic Safety Issues o Part I DEC 1989 o Part II APR 1990 APR 1991 APP. B Probabilistic Risk Assessment o Methodology DEC 1989 o Results AUG 1990 JUN 1991 l
Inspections, Tests Analyses, and Acceptance Criteria DEC 1990 JUN 1991 l
Maintenance and l
Reliability Guidelines DEC 1990 JUN 1991 i
NRC Q/R Revisions o Sections as Necessary JUN 1991 o Sections as Necessary DEC 1991 SEP 1991 l
Draft of Final (Integrated)
SER OCT 1991 i
Design Certification DEC 1992 l
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3.0 CONTENT OF APPLICATION 7
3.1 CESSAR-DC Format The format of CESSAR-DC will be consistent with the guidance of the Standard Review Plan (NUREG-0800) and the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (Regulatory Guide 1.70, Revision 3).
The numbering of CESSAR-DC sections related-to the Nuclear Steam Supply System will be consistent with CESSAR-F, since the System 80+ design is an evolutionary improvement to the System 80 design described in CESSAR-F.*
3.2 CESSAR-DC Amendment Identification The CESSAR-DC submittals outlined in Table 3 consist of changes or additions to existing CESSAR-F material in chapter-by-chapter l
packages. Bars with amendment identifiers will be provided in the margins to indicate all areas of change relative to CESSAR-F. The CESSAR-DC amendment identifier and date will be provided at the bottom of each amended page.
All CESSAR-F material in CESSAR-DC will ba reviewed specifically for applicability to the System 80+ Standard Design, will be modified if appropriate, and will then be identified as CESSAR-DC material.
3.3 Incorocration of Key Reauirements Applications for design certification must contain, pursuant to 10 CFR 52.47, the following items:
CESSAR-F, Docket No. STN 50-470F, was granted a Final Design Approval on December 21, 1983.
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(1) The technical information which is required of applicants for construction permits and operating licenses by 10 CFR, Part 20, 2
Part 50 and its appendices, and Parts 73 and 100, and which is technically relevant to the design and not site-specific; (2) Demonstration of compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f);
(3) The site parameters postulated for the design, and an analysis and evaluation of the design in terms of such parameters; (4)
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 six months prior to application and which are technically relevant to the design (see Section 6.3);
(5) A design-specific probabilistic risk assessment; (6)
Proposed tests, inspections, analyses, and acceptance criteria which are necessary and sufficient to provide reasonable assurance that, if the tests, inspections, ar.d analyses are performed and the acceptance criteria met, a plant which references the design is built and will operate in accordance with the design certification; (7) The interface requirements to be met by those portions of the plant for which the application does not seek certification.
l These requirements must be sufficiently detailed to allow completion of the final safety analysis and design-specific l
probabilistic risk assessment; l
(8) Justification that compliance with the interface requirements is verifiable through inspection, testing (either in the plant or elsewhere), or analysis; and t
l (9) A representative conceptual design for those portions of the plant for which the application does not seek certification to aid the NRC staff in its review of the final safety analysis, probabilistic risk assessment, and interface requirements.
CESSAR-DC will contain the above information.
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4.0 NRC STAFF REVIEW PROCEDURES The staff'will follow the review procedures of the Standard Review Plan, supplemented and modified as follows:
(1) CESSAR-DC will be submitted as shown in Table 3.
Accordingly, i
the NRC staff Safety Evaluation Report (SER) will be issued in sections in draft form and in accordance with the schedule shown in Table 3.
Each draft SER will contain a target j
schedule for closing outstanding SER issues that is compatible with the target date for the FDA. C-E will provide the information required by 10 CFR 52.47, and will respond to NRC staff requests for additional information in a timely manner to support the draft SER schedules.
l (2) The NRC staff will advise the Commission of any new policy issues which arise during the review of CESSAR-DC.
Prior to documentation in the draft SER, the NRC staff will provide the Commission an analysis and rationale for any proposed policy and cite how it would be applied in the SER. The ACRS will be requested to review those issues and NRC staff positions, and provide comments for the Commission.
(3) The draft SERs will be made publicly available.
However, well in advance of their issuance, the NRC staff will submit these draft SERs to the Commission for information. All significant policy issues will be highlighted and discussed in the staff submittal.
(4) At the completion of the review of individual CESSAR-DC chapters, including the appendices for USI/GSI and PRA, the NRC staff will issue a draft of the final (integrated) SER. Then, upon completion of review by the ACRS, the staff will issue a composite final SER in accordance with the schedule shown in Table 3.
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It will be important to document the open or unresc1ved issues that may be identified early in the review process, but wh'ich cannot be resolved until the completion of later chapters.
Each draft SER section will contain a description of such-issues.
In addition, C-E will maintain an updated checklist (proprietary) which will identify outstanding issues and the future chapter (s) in which resolution is anticipated. This checklist will be available to the NRC Project Manager.
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5.0 ACRS PARTICIPATION TheNRCsfaffwillprovidetheACRSwithacopyoftheapplication (CESSAR-DC), will keep the ACRS informed on the progress of the review, and will request a report on those portions of the application which concern safety, in accordance with 10 CFR 52.53.
In addition, the NRC staff will request the ACRS to review policy issues and NRC staff positions, arising from the review, that will be submitted to the Commission.
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6.0 SEVERE ACCIDENT ISSUES 6.1 Introduct1on Severe accident issues are addressed in 10 CFR 52.47, and in the Severe Accident Policy Statement.
On August 8, 1985, the Commission issued a policy statement on j
l severe accidents (50 FR 32138, " Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants"). The policy statement provides general criteria and procedures for the licensing of new plants and for the systematic examination of existing plants.
The Commission encouraged the development of new designs that might realize safety improvements and stated that the i
Commission intended to take all reasonable steps to reduce the chances of occurrence of a severe accident and to mitigate the consequences of such an accident, should one occur.
The following sections describe the approach to meeting the requirements of 10 CFR, Part 52, and the guidance of the policy statement on severe accidents.
6.2 TMI Recuirements for New Plants C-E will comply with all regulations applicable to the System 80+
Standard Design including those listed in 10 CFR 50.34(f), except l
for those identified in Section 1.2.1 (for which C-E will provide l
information to support an alternate approach, consistent with 10 CFR 50.12).
The THI requirements of 10 CFR 50.34(f) which are not applicable to the System 80+ design but which must be met by the l
operator of a System 80+ nuclear power plant are listed below:
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o Simulator capability [50.34(f)(2)(i)]
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o Opera' ting procedures and training [50.34(f)(2)(ii)]
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[50.34(f)(2)(xxv)]
o Leakage control program during plant operation
[50.34(f)(2)(xxvi)]
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Administrative procedures (50.34(f)(3)(i)]
o Plant QA program [50.34(f)(3)(iii)]
o Construction management [50.34(f)(3)(vii)]
6.3 Resolution of USIs and GSIs A total of 734 issues are identified in a "A Prioritization of Generic Safety Issues" (NUREG-0933),* along with a summary of the status of each issue. Of the 734 issues, 386 are considered to be not applicable to the design of evolutionary Light Water Reactors, based on the EPRI Regulatory Stabilization Program (see NUREG-1197).
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l The version of NUREG-0933 required by 10 CFR 52.47 is that version current six months prior to application.
For the System 80+ Standard Design, the application is dated March 30, 1989.
Six months prior to this date, NUREG-0933 through Supplement 8, June 1988, was in effect. However, for the System 80+ Design Certification Program, C-E has elected to use NUREG-0933 through Supplement 9, April 1989.
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i The remaining 348 issues are considered to be applicable to the design of evolutionary Light Water Reactors.
Further review will be performed by C-E to determine the subset of issues applicable to the 2
System 80+ design. The basis for determining the list of applicable issues (numbering 114 initially) will be described in Appendix A of CESSAR-DC.
As implementation of resolutions for these issues progresses, including NRC review, that list of issues (applicable to the System 80+ design) may be revised and CESSAR-DC, Appendix A, will be revised accordingly.
f Proposed acceptance criteria and design features for resolution of applicable USIs and GSIs will be documented by C-E in Appendix A to CESSAR-DC. The NRC will review this appendix and C-E will provide additional information necessary for NRC approval. Acceptance criteria established by the NRC staff, either based on C-E's proposed input or on its own, will be used to review C-E's resolution of the applicable USIs and GSIs.
i 6.4 Probabilistic Risk Assessment The System 80+ Probabilistic Risk Assessment (PRA) will be a Level III PRA which addresses both internal and external initiators of accident sequences which lead to core damage.
Bounding plant site characteristics will be used for the evaluation of external events (seismic events and tornado strikes only) and for evaluating public risk.
[ Sabotage is addressed in the System 80+ design as described in Section 7.1 of this document.
Sabotage will not be addressed quantitatively in the System 80+ PRA.) The PRA methodology will be described in CESSAR-DC, Appendix B.
The System 80+ PPA has three primary purposes. The first purpose is to identify the dominant contributors to severe accident risk.
The second purpose is to provide an analytical tool for evaluating the impact of design modifications on core damage probability and the overall risk to the health and safety of the public. The final.-
purpose is to calculate the core damage mean frequency, large release mean frequency, and expected containment performance for the System 8Di design (as described in Section 6.5).
Results of this f
l PRA will also be presented in CESSAR-DC, Appendix B.
6.5 avere Accident Performance Goals This subsection describes the goals established by C-E for severe accident performance.
6.5.1 Prevention of Core Da:nace It is C-E's goal that the estimated mean annual core damage frequency (including both internal and external events) will be less than 1.0E-5 events per reactor-year. This goal is consistent with that recommended by EPRI.
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For the System 80+ PRA, C-E has adopted the following criteria for the onset of potential severe core damage. A potential for severe core damage shall be assumed to exist if and only if both of the
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following have occurred:
l (A) The collapsed level in the RCS has decreased such that active fuel in the core has been uncovered; and, i
(B) A temperature of 2200'F or higher is reached in any node of the core as defined in a realistic thermal-hydraulic calculation.
The above criteria are consistent with the EPRI definition provided in the EPRI ALWR Requirements Document.
If the above criteria for potential severe core damage are exceeded, predictions of actual core damage and resulting radioactive releases will be calculated using the MAAP code.
The initial analyses will be done with MAAP-38 and final analyses will be performed using a version of MAAP-DOE.
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6.5.2 Offsite Conseauences of Severe Accidents
-l C-E has aBopted the following large-offsite-release goal for the System 80+ Standard Design:
i In the event of a severe accident, the dose beyond a one-half mile radius from the reactor shall not exceed 25 rem. The mean frequency of occurrenc.e for higher offsite doses shall be less than once per l
million reactor-years, considering both internal and external events.
The C-E goal stated above is consistent with the EPRI goal.
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6.5.3 Excetted Containment Performance The containment is one of the principal barriers to the release of i
radioactivity.
Consistent with this defense-in-depth principle, the System 80+ design will provide protection against containment failure in the event of a release of radioactivity to the containment atmosphere.
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The containment design features include:
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a large dry steel containment (the System 80+ containment has l
an ultimate strength which is approximately four times the design strength -- best-estimate calculations show actual failure at a pressure of approximately 220 psig vs. a design i
f pressure of 49 psig);
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measures to reduce the probability of early containment failure, including the safety-grade Containment Spray System and the safety-grade Safety Depressurization System on the reactor coolant system (primary side);
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a conservative design basis accident (guillotine pipe break);
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severe accident hydrogen control; e.
an ih-containment refueling water storage tank for scrubbing radioactivity out of reactor-coolant-system releases and for providing a reliable source of water for flooding the reactor cavity; f.
reliable containment heat removal systems, (e.g., the non-safety-grade Containment Cooling and Ventilation and Normal Chilled Water Systems and the safety-grade Containment Spray System);
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consideration of severe accidents in the design of the reactor vessel cavity configuration, including entrainment of a hypothetical molten core; and h.
design features (e.g., use of relief valves, increased design pressure) to reduce the probability of an interfacing system LOCA (and containment bypass) relative to that for currently I
operating plants.
During NRC staff review of the System 80+ design, the approach for demonstrating containment integrity under severe accident conditions may be revised or supplemented.
C-E's initial expectations, l
however, are that containment integrity will be demonstrated based j
on the probabilistic reliability approach summarized below.
l Any quantitative reliability prediction of the containment function must be stated together with the corresponding definition of the methodology used in that prediction.
The reliability of containment performance, in the context of the EPRI ALWR Requirements Document, is embodied in the goals stated in Sections 6.5.1 and 6.5.2:
(1) The estimated mean annual core damage frequency (including both internal and external events) will be less that 1.0E-5 events per reactor-year, and I
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In the event of a severe accident, the dose beyond a one-half mile radius from the reactor shall not exceed 25 rem.
The mean freg'uency of occurrence for higher offsite doses shall be less than once per million reactor-years, considering both internal and external events.
The robust containment design selected for the System 80+ design permits C-E to state its expectations for containment performance,*
based on the following definitions:
(1)
" Credible core damage sequences" is defined as all core damage event sequences with a frequency greater than 1.0E-6 per i
reactor-year.
External events which would cause both core j
damage and concurrently fail the containment and which have a frequency of less than 1.0E-5 per reactor-year will not be considered in this evaluation.
(2)
" Containment failure" is defined as a post-core-damage release resulting in a dose greater than 25 rem beyond one-half mile from the reactor.
Based on the above, the System 80+ containment design is expected to be such that the containment conditional failure probability, when weighted ever credible core damage sequences, will be less than one in ten (1.0E-1).
Based on methodology consistent with the EPRI PRA Key Assumptions and Groundrules Document (Appendix A to Volume II, Chapter 1 of the EPRI ALWR Requirements Document)..
l-l 7.0 OTHER SPECIFIC ISSUES l
\\
The follow'ing subsections identify dher specific issues (and the
(
general approach to their resolution).hich are identified in NRC regulations, guidance, or policy statements or which are of special j
interest to NRC staff.
l l
7.1 Comparison of the System 80+ Desian With the EPRI ALWR Recuirements j
Document i
C-E is responsible for the development of the System 80+ Standard Design, even though assistance may be obtained from other organizations during the design process and NRC staff review.
The design bases for the System 80+ Standard Design include performance and safety criteria established by C-E, industry codes and standards, System 80 design information, operating plant experience,
(
NRC regulations and guidance, and input from the EPRI ALWR
(
Requirements Document.
1 Consistent with NRC Staff Requirements memoranda,* C-E will identify to the NRC staff any regulatory requirements that System 80+
{
satisfies in a manner different than that identified in the EPRI ALWR Requirements Document.
C-E believes that the System 80+ design currently complies with all EPRI design requirements necessary to meet NRC regulatory l
requirements. Should this conclusion change eithe-sult of l
completion or modification of the EPRI ALWR Requirement
- cument, completion of NRC staff review of the EPRI ALWR Requirements Document, or completion of the System 30+ Standard Design, C-E will inform the NRC staff.
Staff Requirements - Resolution Process for Severe Accident Issues on Evolutionary LWRs, dated December 15, 1989 and Staff Requirements -
l Recommended Priorities for Review of Standard Plant Designs, dated December l
15, 1989.
Regarding EPRI requirements directed at improved plant performance and operation (and not required for regulatory compliance), C-E has chosen to address selected design attributes in a manner that l
differs from specific EPRI requirements.
Such differences are based on C-E's own evaluation of the desirability, effectiveness, and commercial viability of the feature as it pertains specifically to the System 80+ design. Appendix A to this LRB provides a current list of System 80+ design features which differ from EPRI requirements.
l 7.2 Physical Security and Sabotaae The System 80+ Standard Design is being developed in accordance with current NRC regulations and guidance regarding the physical security of nuclear power plants and the prevention of sabotage. A program i
to identify both existing and potential new design features for sabotage protection was completed and results will be summarized in CESSAR-DC (Chapter 13, Appendix A).
l CESSAR-DC will identify physical barriers sufficient to protect j
l vital equipment in accordance with 10 CFR 73.55(c), " Physical Barriers," and to identify access control points t, all vital areas l
in accordance with 10 CFR 73.55(d), " Access Requirements."
i CESSAR-CC will also include a summary of insider / outsider sabotage scenarios and design features to provide sabotage protection l
(Appendix A to Chapter 13).
The System 80+ design will permit the development and implementation of a security plan which will comply with 10 CFR 73.55,
" Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage," and other applicable portions of 10 CFR, Part 73.
Due to site-specific and operation-specific features, CESSAR-DC will not include all information required by 10 CFR 73.55. Any design i
l I
interface requirements or inputs to site development or plant l
operation will, however, be identified in CESSAR-DC.
Examples of such site-ipecific or operation-specific items in 10 CFR 73.55 are:
1 (1)
Physical Security (2) Access Requirements (3) Detection Aids (4) Communication Requirements (5) Testing and Maintenance (6)
Response Requirements 7.3 Site Enveloce Parameters The System 80+ Standard Design is based on assumed site-related parameters that are selected so as to be applicable to the majority of potential nuclear power plant sites in the United States. The rite envelope parameters for the System 80+ design will be presented i
i Chapter 2 of CESSAR-DC.
7.4 Comoleteness of Desion Documentation i
Consistent with the requirements of 10 CFR 52.47, the level of design information in CESSAR-DC will be sufficient to enable the i
Commission to judge C-E's proposed means of assuring that i
construction conforms to the design and to reach a final conclusion l
on all safety questions associated with the design before the certification is granted. The information will include performance requirements and design information sufficiently detailed to permit the preparation of acceptance and inspection requirements by the NRC, and procurement specifications and construction and installation specifications by C-E.
The information normally contained in certain procurement specifications and construction e.-
,rp--
3-,,r-
4 i
and installation specifications will be completed and available for audit if such information is necessary for the Commission to make its safety determination. Consistent with the above, CESSAR-DC will include, as a minimum, the following:
a.
design. basis criteria b.
analysis and design methods functional desppn and physical arrangement of systems c.
d.
plant physical arrangements sufficient to accommodate systems and components e.
functional and/or_ performance specifications for components and materials sufficiently detailed to become a part of associated procurement specifications f.
acceptance / test requirements g.
accident analyses h.
proposed technical specifications 1.
summary of Probabilistic Risk' Assessment methodology and results Consistent with 10 CFR 52.47(a)(2), design documentation supporting CESSAR-DC and available for NRC audit should include, as l
l appropriate:
l a.
design basis criteria b.
plant general arrangements of structures and components, including piping system layouts c.
process and instrumentation diagrams, electrical system layouts, and major conduit and cable tray layouts d.
control logic diagrams e.
system functional descriptions and supporting studies and analyses f.
sufficient detail to permit preparation of component and procurement specifications, including acceptance criteria and test requirements l _
g.
sufficient detail to permit preparation of construction /
installation specifications, including acceptance criteria and testhequirements h.
program for the assurance of quality i.
design-related aspects for the emergency plans j.
supporting design documentation such as site envelope data and calculations sufficient to support the level of design detail noted above k.
design-related aspects of the physical security program 1.
ALARA/ radiation protection plan m.
detailed Probabilistic Risk Assessment i
7.5 Proaram for the Assurance of Ouality in Desian The C-E Quality Assurance Program is described in topical report CENPD-210, Revision 5, " Quality Assurance Program". Supplemental information will be provided in Chapter 17 of CESSAR-DC.
7.6 Maintenance. Surveillance. and Reliability 1
Certification of a design will be based in part upon a Probabilistic Risk Assessment (PRA). The validity of a PRA is dependent, in part, on a reliability of systems, structures, and components.
The NRC staff requires assurance that programs can be implemented by the plant owner which will ensure that the reliability of those systems, structures, and components (assumed in analyses) will be maintained throughout plant life. Therefore, program input to assure design and operational reliability will be provided during the FDA and design certification review. These program inputs will provide assurance that the reliability of a System 80+ plant can be maintained consistent with the plant design and PRA assumptions and.
~
will include items such as (1) the proposed Technical Specifications, (2) In-Service Inspections and Tests, (3) mainte' nance guidelines, (4) procedure guidelines, and (5) security guidelines.
The proposed Technical Specifications will be submitted for review and approval by the NRC staff as part of the CESSAR-DC submittal.
The proposed Technical Specifications will be developed based upon risk and reliability considerations. C-E will identify (in CESSAR-DC) design features for testinti and maintenance during operation without challenging safety systems.
7.7 Safety Goal Policy Statement On August 4 and 21, 1986, the Commission published a Policy State-ment on " Safety Goals for the Operation of Nuclear Power Plants" (51 FR 28044 and 51 FR 30028). This policy statement focuses on the risks to the public from nuclear power plant operations.
Its objective is to establish goals that broadly define an acceptable level of radiological risk.
l l
l C-E will address the Safety Goal Policy Statement by demonstrating that the core damage and large release goals of Section 6.5 have l
been met.
7.8 Sixty-Year life The staff will review the System 80+ design for a 60-year life notwithstanding the fact that a 40-year license term limitation is presently in the regulations.
C-E will identify the components and systems which are affected. CESSAR-DC will contain information to support the review for a 60-year design life including information on fatigue, corrosion, thermal aging, and reactor vessel embrittlement.
As a resuit of its review, the NRC staff may identify additional information necessary to support a 60-year design life.
7.9 Fire Protection TheNRCsiaffrequiresthatimprovedfireprotectioncriteriabe implemented for the System 80+ Standard Design. The current Branch Technical Position 9.5-1 guidance (e.g., 20 ft. separation) will be supplemented by a criterion for safe shutdown capability in the event of a complete loss of any fire area, assuming that re-entry into the fire area is not possible (except for the containment, where physical separation will be maximized to the extent practical).
Fire protection for control room shutdown v.apability is provided by independent alternate shutdown capability that is physically and electrically independent of the control room.
Fire protection for redundant shutdown systems in the Reactor Containment Building will ensure, to as great an extent as possible, that one shutdown division will be free of fire damage.
In addition, it will be demonstrated that smoke, hot gases, or the fire suppressant will not migrate into other fire areas to the extent that safe shutdown capabilities, including operator actions, could be adversely affected.
i l
7.10 Station Blackout l
The System 80+ Standard Design includes improved design features and l
electrical systems to ensure a safe shutdown of the reactor in the event of a station blackoat. These improvements are summarized below:
(1) One turbine-driven emergency feedwater pump is included for each steam generator.
(These are in addition to the two motor-driven emergency feedwater pumps.)
In previous designs one turbine-driven pump was shared by both steam generators.
f ;
(2)
Each of the four safety-related instrument channels has a battery backup.
In addition, Class IE Electrical Divisions I and 11, which include the two emergency diesel generators, have their own batteries.
(3) The design has full load rejection capability and the capability to subsequently provide electrical power from the turbine generator.
(4) An alternate source of AC power _which is diverse from the safety-grade emergency diesels is included (this alternate AC source is a control-grade gas turbine). This AC source has its own battery.
7.11 Leak-Before-Break Leak-Before-Break (LBB) may be employed where justified. Any use of LBB on the System 80+ design will comply with revised General Design Criterion 4 (effective November 27,1987) and the intent of draft SRP Section 3.6.3.
7.12 Source Term The NRC staff will use the licensing basis source term " TID 14844" for the review of the System 80+ safety analysis.
C-E expects to adopt realistic source terms, based on industry programs, for the System 80+ PRA and severe accident evaluations.
If NRC staff agrees that realistic source terms can also be applied to safety analyses, C-E will revise the safety analysis accordingly. -
7.13 Ooeratino Basis Earthauake
~
The NRC staff agrees that the OBE should not necessarily control the design of safety systems, which now occurs when the criteria of 10 CFR, Part 100, Appendix A, are applied.
The System 80+ design will be consistent with the EPRI ALWR Requirements Document with respect to definition of OBE, SSE, and analysis methodology.
It is expected that the OBE will be less than one-half of the SSE, which is a departure from 10 CFR, Part 100, Appendix A.
The NRC staff will review the OBE design basis and will consider C-E's request to decouple the OBE from the SSE, subject to Commission approval.
7.14 TYDe C Containment Leak Rate Containment leakage is acknowledged by the NRC staff as being a function of containment pressure.
This pressure dependence will be reflected in predictions of leak rate for the System 80+
containment.
7.15 Hydroaen Generati7n C-E will provide information to justify a System 80+ containment design consistent with the EPRI ALWR Requirements Document and NRC staff review thereof. That information will include justification for the assumed extent of metal-water reaction (75%) and the maximum allowable hydrogen concentration (13%).
7.16 Severe Accident Containment Vents C-E will ensure that the System 80+ containment design includes the capability to add containment vents at c future time. This approach is in compliance with 50.34(f)(3)(lv).
NRC staff will review System 80+ severe accident issues including containment overpressure analysis and, subject to Commission approval, will determine if there is a need for special containment vents..
)
a' I
I 7.17 Mid-looo Ooerations i
This issuie addresses the potential loss of decay heat removal capability when the reactor is shut down for refueling or i
maintenance, the reactor coolant system is drained to the "mid-loop"
)
level, and the reactor vessel head has not yet been removed. The phenomenon of concern is a buildup of pressure in the reactor vessel and hot leg which could result after a loss of decay heat removal capability. This pressure buildup could cause a rapid loss of
]!
coolant inventory if there is an opening in the cold leg (e.g.,
during reactor coolant pump repair).
C-E will specifically address this issue through analysis and consideration of specific design features and/or operational restrictions (such as removal of the f
pressurizer manway during mid-loop operation) which would resolve j
the root cause of concern (i.e., precluding pressure buildup above j
the core during mid-loop conditions).
i 1
i 7.18 Interfacino System LOCA i
4 An Interfacing System LOCA is a loss of primary coolant outside
]
i containment via a system which interfaces with the RCS and for which j
l the pressure boundary is outside containment. The interfacing system LOCA is presumed to result from exposing low pressure piping of the interfacing system to full primary system pressure due to l
failure of multiple pressure barrier valves.
i The most significant interfacing system LOCAs would occur in the Safety Injection and Shutdown Cooling Systems since these systems s
i have the largest pipe sizes for interfacing systems.
In the development of the System 80+ Standard Design the probability of an j
interfacing system LOCA was decreased significantly by eliminating the low-pressure Safety Injection System and by increasing the 1
design pressure of the Shutdown Cooling System from 650 psi to 900 j
i j !
1 l
psi. With this higher design pressure, the shutdown cooling system l
is expected to maintain its integrity even when exposed to full reactor cholant system pressure.
l i
In addition to the above design improvements, evaluation of interfacing system LOCAs are included in the Probabilistic Risk Assessment for the System 80+ design. As expected, results to date indicate that interfacing system LOCAs provide only a minor contribution to the core damage frequency (i.e., a contribution of i
approximately 3.0E-9 relative to the core damage frequency goal of 1.0E-5).
7.19 Anticioated Transients Without Scrar,. 'TWS) 4 The System 80+ Standard Design includes a control grade Alternate Protection System (APS) to address the requirements of the ATWS Rule (10 CFR 50.62). The APS includes an Alternate Reactor Trip Signal and an Alternate Feedwater Actuation Signal which are separate and diverse from the safety-grade reactor trip system. The APS, therefore, addresses both the prevention and mitigation requirements of the ATWS Rule.
4 i
7.20 Electrical System Desian The System 80+ Standard Design will be connected to a switchyard and to the transmission system via two separate and independent 1
transmission lines. The generator circuit breaker, along with the unit main transformers, allows one of these lines not only to supply power to the transmission system during normal operation, but also to serve as an immediately available source of preferred onsite power.
The other separate transmission line is connected, via the switchyard and a standby auxiliary transformer, to provide a second independent immediately available source of offsite power to the onsite power distribution system for safety and permanent non-safety loads.
1 The onsite power system for the System 80+ Standard Design consists j
of the main generator, the generator circuit breaker, unit main j
transformers, two unit auxiliary transformers, one standby auxiliary j
transformer, two safety-grade diesel generators, a control grade alternate AC power source, the batteries, and the auxiliary power system.
The Class IE safety loads are divided into two redundant and independent locd group Divisions I and II. Each Load Division is capable of being supplied power from the following sources (listed in decreasing order of priority):
A.
Unit Main Turbine Generator B.
Unit Main Transformers (Offsite Preferred Bus-1)
C.
Standby Auxiliary Transformer (Offsite Preferred Bus-2)
D.
Emergency Diesel Generators E.
Alternate AC Power Source (diverse from the diesel generators)
If the unit main generator, both the offsite power sources, and the diesel generators are all unavailable, either one of the Safety Divisions may be powered from the alternate AC power source.
These design improvements enhance the capability of coping with a station blackout event, as described in Section 7.10.
A description of the electrical power system is presented in Chapter 8 of CESSAR-DC.
7.21 Deoraded Core Behavior The System 80+ Standard Design includes design features for prevention and mitigation of a degraded core. These features include the Safety Depressurization System for reactor coolant system depressurization. This system, when used in conjunction with the Safety Injection System, provides a backup to the Shutdown.
I-Cooling System to decrease the probability of core damage.
The l
Safety Depressurization System also minimizes the possibility of l
ejecting ifrom the vessel) molten core material under high-pressure conditions.
The System 80+ reactor vessel cavity design includes features to l
mitigate the effects of a degraded core which penetrates the reactor i
2 vessel.
First, a large floor area (0.02 m /MWt) enhances debris dispersal and coolability. The second feature is an indirect (labyrinthine) cavity vent path, including a debris collection chamber, which is configured to trap solid core debris and minimize direct containment heating.
The third feature is an In-Containment Refueling Water Storage Tank which provides a source of water for flooding the reactor cavity and cooling core debris.
1 i
l l l
I
)
l~
l l
l l
i APPENDIX A i
1 System 80+ Design Differences 1
from the EPRI ALWR Requirements Document l
l l
1 l
I l
l t
l i
i i
l l
l A-1 l
l l
Regarding EPRI requirements directed at improved plant performance and operation (and not required for regulatory compliance), C-E has 2
chosen to address selected design attributes in a manner that differs from specific EPRI requirements.
Such differences are based on C-E's own evaluation of the desirability, effectiveness, and 1
commercial viability of the feature as it pertains specifically to the System 80+ design. A current list of System 80+ design features which deviate from EPRI requirements is provided below.
System 80+ reactor coolant hot leg temperature will be reduced o
to 615'F (versus 600*F).
Skirt-type supports will be used for the System 80+ pressurizer o
and steam generators (versus pedestal or open-frame supports).
System 80+ steam generators will have handholes at the bottom, o
tubesheet elevation of the secondary side (versus every U-tube support elevation).
o System 80+ Contrei Element Drive Mechanisms will not have anti-ejection latches.
o The System 80+ Emergency Feedwater System will have a cross-connect between the two independent trains.
o The System 80+ design will retain redundant feedwater isolation valves (versus a using the feedwater control valve for isolation).
o The System 80+ main feedwater pumps will be turbine-driven (versus motor-driven).
The System 80+ design will not include the main steam isolation o
signal on pressure rate-of-change.
l A-2
o o
The System 80+ sensor and cable failures may not meet the
~
time-to-detect-and-repair criterion of eight hours in all cases.
o The System 80+ design will includt some polyvinyl choloride and neoprene insulation.
o The System 80+ alternate AC power source (a combustion turbine) will not meet IEEE Standard 387(diesel generators).
o The Syste.o 80+ design includes the Reactor Vessel Level Measurement System. Compliance with the EPRI requirement to not have such a syster would lead to a non-compliance with 10 CFR 50.34(f)(2)(xxviii).
I o
The System 80+ emergency diesel generators'will automatically start but not load'on' loss of offsite power without turbine-aenerator trio (versus automatic start and load).
o The System 80+ design has a spherical (versus cylindrical) steel dual containment.
l o
The System 80+ nuclear steam supply system will not be significantly offset from the center of the containment due to l
the enlarged workspace provided by the spherical geometry of the containment (versus a 15-20 foot offset in a cylindrical containment).
o The System 80+ containment equipment hatch will be at the operating floor level. (versus grade' level).
l I
A-3 l
l l
.. ~.
1
4 E'
FiEDED ASEA BROWN BOVERI August 28, 1990 LD-90-060 Project No. 675 l
Mr. Thomas V. Wambach Project Manager Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Attn:
Document Control Desk l
Washington, DC 20555 I
Subject:
Licensing Review Basis for the System 80+= Standard Design l
Reference:
Letter LD-90-005, A.'E. Scherer (C-E) to l
R. Singh (NRC), dated January 22, 1990 i
Dear Mr. Wambach:
l l
Enclosed are proposed changes to the System 80+* Liscensing Review Basis document (sabnitted by the reference letter) which we expect to include in the next revision.
We are providing this information at this time to facilitate your review.
If you have any questions on the enclosed revisions, please call me or Mr. S. E. Ritterbusch at (203) 285-5206.
Very truly yours, COMBUSTION EN EERING, INC.
E. H. Kennedy Manager Nuclear Systems Licensing EHK:lw
Enclosure:
As Stated cc:
C. Miller (NRC)
W. Miller (NRC)
F. Ross (DOE-Germantown)
ABB Combustion Engineering Nuclear Power Bom Fa 20 PS 9S ?
Wresor. ConnectcJ 06095 0500 Teen 99297 COMBEN WSOR
l l.
.~
l l
o For hydr n generation a ontrol inside co inment, t assum extent of meta' ter reaction (75 and the mum al' able hydrogen entratira (13%)
differe than those ecified by 10 C 50.34(f) (LRB Se on 7s,15)
The requirements for requesting exemptions from current regulations (10 CFR 50.12) will be applied where appropriate.
1.2.2 Desion Features and Methods which may involve Ouestions of Commission Poliev l
o The need for, and formulation of, a goal for expected containment performance (Section 6.5.3) o The degree to which, if any, new design features are required to improve protection against sabotage (Section 7.2) o The degree to which, if any, new design features are required for improved fire protection (Section 7.9) o The degree to which realistic source terms can be used for PRA and severe accident evaluations (Section 7.12)
Whether there is a need for containment venting capability for o
~
severe accident mitigation beyond that required by current regulations (Section 7.16)
Whether proposed design features for prevention and mitigation o
of degraded core conditions are adequate (Section 7.21)
I l
l.
7.13 Operatino Basis Earthouake The NRC sfaff agrees that the OBE should not necessarily control the design of safety systems, which now occurs when t.he criteria of 10 CFR, Part 100, Appendix A, are applied. The System 80+ design will be consistent with the EPRI ALWR Requirements Document with respect to definition of OBE, SSE, and analysis methodology.
It is expected that the OBE will be less than one-half of the SSE, which is a departure from 10 CFR, Part 100, Appendix A.
The NRC staff will review the OBE design basis and will consider C-E's request to decouple the OBE from the SSE, subject to Commission approval.
7.14 Tvoe C Containment leak Rate Containment leakage is acknowledged by the NRC staff as being a function of containment pressure. This pressure dependence will be reflected in predictions of leak ratg for the System 80+
containment.
"fd g, JPm f0+ der.jn iac. des 4 h h.f. ;fnikr sfA (e,~%i 7.15 Hydrooen Generation 14 t 14
- e. a u n a.
em
..r C-E will provide info ation to justify a System
+ con a t
design consistent with he EPRI ALWR Requir'ements Document and NRC staff review thereof.
T_.
- t':-
.J.J. j..;:':.....n f:
t' : :r:f " ' :' nt;
. t;
- i':: p ;;; _. f ".: -
2 J..:.,..
r.,
_ __.t._;.... A,.
7.16 Severe Accident Containment Vents
!E
'" r
"-t Ye System 80+ containment design C.l.1x 'he n
-ru
,n capability to add containnfent,ventz at a ruture time.
This approach g
j is in compliance with 50.34(f)(3)(iv).
NRC staff will review System 80+ severe accident issues including containment overpressure analysis and, subject to Commission approval, will determine if
- ...!containmentvehnt d ahgsp gggggg there is a need 'n
- eerde,& res w.
4 a)J.
l l.
~
l INSERT A has a dual containment system. The steel containment pressure boundary is l
surrounded by a concrete shield building, with an annului in between. Such i
a design presents no significant problems to the addition of a large (three-foot diameter) containment penetration in the future. Therefore, the System 80+ design has l
1 l
l i
I l
l I
Regarding EGI requirements dirteted at improved plant performance and operation (and not required for regulatory compliance), C-E has 2
chosen to address selected design attributes in a manner that
.diffeis from specific EPRI requirements. Such differences are based on C-E's own evaluation of the desirability, effectiveness, and commercial viability of the feature as it pertains specifically to the System 80+ design. A current list of System 80+ design features which deviate from EPRI requirements is provided below, o
System 80+ reactor coolant hot leg temperature will be reduced to 61S*F (versus 600*F).
o Skirt-type supports will be used for the System 80+ pressurizer and steam generators (versus pedestal or open-frame supports).
o System 80+ steam generators will have handholes at the bottom,
~
tubesheet elevation of the secondary side -(versus every U-tube support elevation).
o System 80+ Control Element Drive Mechanisms will not have anti-ejection latches.
o The System 80+ Emergency Feedwater System will havb a cross-connect between the two independent trains.
l o
The System 80+ design will retain redundant feedwater isolation valves (versus a using the feedwater control valve for isolation).
Th: Sy:ter 80: ::!- fe: h:ter p;;;; vill 5: turbin: d r i ;;r, ry m. e., e mm+m a n,m q o
The System 80+ design will not include the main steam isolation signal on pressure rate-of-change.
4
'5: Ty:ter 20; :::::r :nd ::ble f:ilure: ::; ::t :::t th:
ti : t: d:t::t :nd re;:'r cr't:r':n :f ci;ht h: r: i.. :11 o
The System 80+ design will include some poly' vinyl choloride and neoprene insulation.
o
't; C, tx 20: cit:r.:t: "C ;:::: ::r :: ': ::-5 ::- turt' ::
.:t
- t !EEE ri:nd:cf ??? 'fi:::1 ;;n:::t:::;.
o The System 80+ design includes the Reactor Vessel Level
~
Measurement System. Compliance with the EPRI requirement to not have such a system would lead to a non-compliance with 10 QFR50.34(f)(2)(xxviii).
: !; r * :- ?? ' --- ;: :; d':::1 ;;:.:. ;t:..
..t
_ '... 11, m-a t.
,_tr.,._
.r.<_...
7 it-t' : ::-- :t:r tr': ';;r::: ;;t:::ti; :t:rt ;.,d 1::f;.
o The System 80+ design has a spherical (versus cylindrical) steel dual containment.
o The System 80+ nuclear steam supply system will not be significantly offset from the center of the containment due to, the enlarged workspace provided by the spherical geometry of the containment (versus a 15-20 foot offset in a cylindrical containment).
o The System 80+ containment equipment hatch will be at the operating floor level (versus grade level).
4 A-3
~
~
O Svatem 80+ =crety anniveia vill da complettd using the conservativo source term of TID 14844 (versus an EP."I-proposed
" realistic" source term).
M.
System 80+ safety analysis will demon-strate that 10 CFR 100 limits can be met with a containment design leak rate of 0.3 wtt per day (versus 0.5.
1 Lt&%
g System 80+ will employ alloy 690 in the pressurizer heater sleeves and instrument wells (versus restricting use to steam generator tubes).
For a loss of all feedwater core un-covery would occur in approx,imately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes without operator action (versus 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).
1 g.
System 80+ containment purge valves require closure in 15 seconds (versus 30 seconds).
p.
System 80+ minimum diesel start and i
load sequence time is 20 seconds.(ver-sus 40 seconds).
p The System 80+ design places only the first stage of low pressure feedwater i
heaters in the condenser neck (versus all low pressure heaters).
g The System 80+
turbine exhaust to condenser inlet connection uses either a rubber flexible seal or aM I8 C
seal with spring mounted condenser (versus stainless steel flexible seal).
^
g System 80+ uses 6 stages of feedwater i
re-heating in the reference design with the actual number to be deter-mined by the site specific heat bal-j ance (versus 7 stages).
The System 80+ atmospheric dump valves ess e
actuated with variable setpoint).
~
g The third main feedwater pump and booster pump are normally in standby (versus normally operating).
J
V g.
Th2 rcpleesmant of Systsm 80+ prascur-izar haator closvoa rcquiros cutting k
and walding (vsrsua no cutting er welding).
% lb 5 W am&m&v
/ Feed j@
and bleed on System 80+ can he delayed for up o 30 minutes followin'g Y.sf~yout(versus.60 minutes).
dr
. g,_
The System 80+ control room pressure boundary will include fans and filters (versus excluding HVAC equipment).
g The System 80+ Advanced Control (tom-plex (called Nuplex 80+)
integrates spatially dedicated display and con-l trol with compact work stations (ver-sus exclusive use of redundant, com-pact work stations).
l System 80+ point-to-point communica-tion is by sound powered phones (ver-sus a wireless system).
System 80+ arrangements locate kitchen and rest com facilities convenient toj but outhidejthe Main Control Room area j
(versus inside Main Control Room ar-ea).
System 80+
uses separate standby-source transformers ft.,r inverters (versus common transformer).
i System 80+ interfaces do not require separate switchyard for main and re-serve offsite circuits (versus sepa-i rate switchyards).
i
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E UNITED STATES h
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E NUCLEAR REGULATORY COMMISSION
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I W ASHINGTON, D.C. 20 lie 6 I,
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JAN 151991 3
3 MEMORANDUM FOR: Chairman Carr w*
FROM:
James M. Taylor Executive Director for Operations
SUBJECT:
ADDITIONAL RESPONSE TO CHAIRMAN'S QUESTIONS REGARDING SECY-90-353,"LICENSINGREVIEWBASIS(LP,B) DOCUMENT FOR THE COMBUSTION ENGINEERING, INC. (CE), SYSTEM 80+
EVOLUTIONARY LIGHT WATER REACTOR" By memorandum of November 9,1990, the staff responded to two of the three questions in your memorandum of October 31, 1990, regarding the sub.iect document.
In response to the third question regarding the items listed in Appendix A to SECY-90-353, the areas in System 80+ design that differ from the l
advanced light water reactor (ALWR) Requirements Document, the staff indicated j
that the information requested would be obtained from CE.
1 CE has responded by letter of December 21, 1990, which is enclosed with l
this memorandum. Enclosure I to that letter lists the deviations and the i
basis for each. Enclosure II lists those items identified in Appendix A to SECY-90-353 as deviations that CE now believes are not deviations and gives 1
the basis for each.
In regard to item 17 in Enclosure I to its letter, CE i
does not clearly identify the deviations of System 80+ from the ALWR Require-ments Document. The requirements of the ALWR Requirements Document with regard to the main control room would provide the following:
- 1. N Redundancy between workstations in the control room. This ensures that, in the event of a console failure, access is still available to controls and instrumentation important to safety.
2.
Workstation designs that can support the division of tasks between two operators.
In the event of a console failure, this design approach ensures that both operators can execute their procedures from one console.
3.
A dedicated workstation in the control room for the supervisor.
4.
A capability to generate hardcopies of the information on the workstation I
displays.
The staff understands that the CE advanced control complex design does not provide for redundancy of controls and instrumentation between the System 80+
workstations, The programmable displays, which are Class IE, are dedicated to displaying specific parameters.
In the event of an instrument or console failure, the parameters shown on the programmable displays cannot be re-routed to other consoles.
A CRT is provided at each workstation for displaying information. However, the CRT, the supporting computer hardware, and the software are not Class IE.
CONTACT:
T. Wambach, NRR/PDST 492-1103 i
4 r
l l
~
l Chairman Carr l CE provides a console in the control room for the control room supervisor.
CE does not currently provide a capability at the consoles for making hardcopies of the CRT displays.
The staff continues to believe that none of these issues involve a policy issue at this time, with the exception of the possible need for prototype testing of the System 80+ Advanced Control Complex. During the design review, the staff will review all of these issues in both Enclosures to the CE letter against the appropriate positions in the Standard Review Plan and Regulatory Guides. As stated in SECY-90-353, the staff will address the differences in the draft safety evaluation report for System 80+ and will promptly notify the Commission if any possible policy issues arise.
J es M xecutive Director l
for Operations i
Enclosure:
Ltr. dtd. 12/21/90 fm l
E.H. Kennedy to C.L. Miller i
cc w/ enclosure:
Commissioner Rogers h issioner Curtiss Commissioner Remick SECY i
MED D ASEA BROWN BOVERI December 21, 1990 l
LD-9 0-097 Charles L. Miller, Director Standardization Project Directorate U. S. Nuclear Regulatory Commission i
Attn:
Document Control Desk Washington, D.C.
20555
Subject:
Differences Between the EPRI Utility Requirements Document and the System 80+= Standard Design
References:
(1) Letter, C.
L. Miller (NRC) to E. H. Kennedy (C-E), dated November 6, 1990 (2) Letter LD-90-060, E. H.
Kennedy (C-E) to T. V. Wambach (NRC), dated August 28, 1990 1
Dear Mr. Miller:
This letter responds to your request [ Reference (1)) to discuss l
in more detail the differences between certain provisions of the EPRI ALWR Utility Requirements Document and the System 80+
Standard Design.
Enclosure I responds to your request.
In addition, a number of differences identified in Appendix A of our draft LRB document [ Reference (2)) have been removed, either due to a change in the design / analysis or a change in the application of the EPRI criteria.
Enclosure II provides, for your information, a listing of the previously identified differences which have been removed.
Notwithstanding these differences, the i
System 80+ design continues to have a very high degree of compliance with EPRI criteria.
If you have any questions on the attached material, please call-
~
me or Mr. S. Ritterbusch of my staff at (203) 285-5206.
Sincerely, COMBUSTION ENGINEERING, INC.
hWUi E. H. 'K edy Manager Nuclear Systems Licensing EHK:lw Enclosures As Stated ABB Combustion Engineering Nuclear Power E14 CLOSURE Combusison Engneerng inc 1000 Prospect Helt Road Teephone (203) 6881911 Post Ottce Boa 500 Faw (20h 285 9512
/_.,
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- wnasor. Connectcut 06095-0500 Tem 99297 cot /BCN WSOR
Enclosure I DIFFERENCES BETWEEN THE SYSTEM 80" DESIGN AND THE EPRI EVOLUTIONARY PLANT REOUIREMENTS DOCUMENT
- AS OF DECEMBER, 1990 1.
Hot Leo Temperature: The EPRI URD, Chapter 3, paragraph 3.2.3, states "The reactor coolant system shall be desigr.e;d so the average het leg temperature is no higher than 600F."
The EPRI criterion (600F) was selected based on the following considerations:
1.
Increased thermal margin in the reactor core
- 2. Reduced likelihood of steam generator tube failure due to stress corrosion cracking.
The System 80+ design has a maximum hot leg temperature of 615F (compared to 620F for the System 80 design).
The EPRI l
considerations are satisfied in the following manner:
- 1. The System 80+ design meets the EPRI criterion for 15%
thermal margin in the reactor core
- 2. Corrosion resistance of the steam generator tubes has been l
addressed by specifying thermally-treated Alloy 690 (in I
lieu of Alloy 600). Laboratory testing of Alloy 690 has i
shown it to be highly resistant to stress corrosion l
cracking at operating conditions.
j i
To maintain equivalent plant efficiency and performance when the hot leg temperature is decreased, the steam generator size must be increased. Primarily as a result of this consideration, the System 80+ steam generator heat transfer area has been increased by about 10%.
The steam generator size cannot be increased significantly more than this amount and still remain within proven manufacturing capabilities.
- Also, increases in steam generator size and accompanying increases in secondary side inventory make the l
consequences of steam line break accidents more adverse.
Based on the above considerations, C-E believes that a hot leg temperature of 615F represents an optimal balance.
Based on the EPRI URD as of mid-summer, 1990.
The EPRI URD Roll-up Document submitted to the NRC has been held proprietary by EPRI and has not been released to Combustion Engineering.
I
2.
Skirt-tvoe Component Suenorts:
The EPRI URD, Chapter 3,
paragraph 2.2.6, states " Major reactor coolant system equipment supports shall use pedestal type or open frame supports in preference to skirt mountings, which more severely limit access."
As in previous C-E designs, the System 80+ design uses skirt type supports for the two steam generators and the pressurizer. Reasons for this design decision include use of a proven design, inherent strength of the skirt design, and absence of experience indicating problems with access for inspection.
The steam generator skirt is small compared to the diameter of the steam generator and does not c
restrict access to any nozzles or manways. The pressurizer support I
skirt is a full-diameter skirt and permits access to the heater nozzles (from below).
l 3.
Redundant Feedwater Isolation Valves: The EPRI URD, Chapter 2, l
paragraph 4.2.2.4, states " Double valve feedwater isolation is l
required. The feedwater control valve (Chapter 3) serves as one of I
these valves."
t The System 80+ design retains redundant feedwater isolation valves i
in addition to the control valve and two check valves in each feedwater line for the following reasons.
First, use of double feedwater isolation valves requires no reliance on check valves or control valves to meet the single failure criterion in events such as a feedwater line break.
Second, double isolation valves ensure a safety-related means to terminate a steam generator overfill i
transient, without making the control valves or check valves safety-grade.
Finally, two isolation valves provide a tighter shutoff since control valves are not typically designed for isolation.
4.
Main Steam Isolation on Pressure Rate-of-chance: The EPRI URD, Chapter 2, paragraph 3.5.3, states "High containment pressure and low steamline pressure... to signal the MSIVs to close.
During periods of low steamline pressure, high steamline pressure rate shall be used...."
The System 80+ design includes a main steam isolation signal on low steam generator pressure.
The actuation setpoint is manually variable when steamline pressure is being decreased and automatically variable when the pressure is being increased.
This design provides equivalent protection to that for actuation on pressure rate-of-change.
This System 80+ design feature has been used in operating plants and provides protection over the full range of conditions.
5.
Containment Geometrv:
The EPRI URD, Chapter 6,
paragraph 4.3.4.1, states "The primary containment structure shall be a large
i dry type containment with cylindrical steel pressure vessel (150-foot internal diameter)
The corresponding rationale indicates that other designs, including steel spheres, have been i
employed successfully and could meet the ALWR top-tier criteria.
The System 80+ design includes a spherical steel pressure vessel with a 200-foot diameter.
C-E believes that the spherical containment offers operational, constructability, and cost i
advantages and meets all EPRI URD functional criteria.
For example, the operating floor is located at the elevation of maximum diameter (200 feet) which provides significantly more open floor space for maintenance and other operation activities than the 150-foot cylindrical design.
6.
Hatch at Operatina Floor Level: The EPRI URD, Chapter 6,
paragraph 4.3.4.7, states "A maintenance hatch shall be provided at j
grade...." The corresponding rationale indicates that the concern is to provide space at the operating deck level for other activities.
The System 80+ maintenance hatch is located at the level of the operating floor, 54 feet above grade.
Location at this elevation results from use of the spherical geometry (which provides large maintenance and laydown areas at the operating deck level) and general arrangement considerations such as selection of embedment depth and convenient access to the maintenance bay for truck loading and off-loading.
The large floor area at the operating i
deck level ensures that maintenance and other activities will not be inhibited by activity in the vicinity of the equipment hatch.
While there is a difference from the EPRI criterion, the intent of the criterion indicated in the rationale is met by providing maintenance-staging floor area in the adjacent maintenance / outage building at the elevation of the co.stainment operating deck.
The adjacent maintenance building includes equipment for handling heavy components and truck access at grade level.
- 7. Source Term for Radioactivity Release Predictions: The EPRI URD, Chapter 1, paragraph 2.4.1.2, states "... source term... shall be somewhat more realistic than analyses to date on current LWRs."
The System 80+ design currently uses, for the design basis safety analysis, the same source term methodolgy approved by NRC staff and used to date on current LWRs. Realistic source term methodology is used, however, for severe accident calculations.
There are significant benefits to plant operation and siting when a
realistic source term is used (e.g.,
a larger allowable containment leak rate, a smaller Exclusion Area Boundary) and C-E supports and encourages EPRI and the NRC staff to continue their efforts to agree upon more realistic source term assumptions for design basis safety analysis calculations.
Such agreement has not yet been obtained, as indicated in the draft SER for Chapter 5 of
the EPRI URD. Therefore, the current System 80+ safety analysis uses NRC-approved methodology.
C-E strongly encourages continued efforts to reach concurrence on realistic source term assumptions, especially for the evoluticnary LWRs.
It is anticipated that the System 80+
safety analysis would be revised to reflect more realistic methodology if concurrence between EPRI and NRC staff is obtained prior to the certification of the System 80+ Standard Design.
4
- 8. Containment Desian Leak Rate: The EPRI URD, Chapter 1, paragraph 2.4.1.1, states "It shall be demonstrated that 10CFR100 exposure limits can be met with a containment design leak rate of not less than.5 percent The System 80+ Chapter 15 offsite and control room dose analysis is based on a design leak rate of 0.34%.
This represents a relaxation from current values of typically 0.1%.
A larger leak rate would be justified if more realistic source term assumptions were implemented, as indicated in item (7) above on the source term.
9 Allov 690 for Pressurizer Heaters: The EPRI URD, Chapter 1.
use of Alloy 690 shall be paragraph 5.3.1.3.1.3, states "...
restricted to steam generator tube applications."
The System 80+ design also uses Alloy 690 for pressurizer heater sleeves.
Use of Alloy 690 decreases the potential for stress corrosion cracking, relative to previously used materials, and it is considered to be highly resistant to stress corrosion cracking, ba. sed on laboratory tests.
- 10. Pressurizer Heater Sleeve Replacement Method: The EPRI URD, Chapter 3, paragraph 3.4.3.4.5, states "The design and arrangement of pressurizer heater sleeves shall allow replacement of bundles of heater sleeves without cutting and welding on the pressurizer shell.
The System 80+ design uses Alloy 690 for pressurizer heater sleeves to minimize the possibility of having to replace them due to corrosion.
C-E believes that use of Alloy 690 decreases the need for heater sleeve replacement and, therefore, current removal and replacement techniques are considered adequate.
- 11. Feedwater Heater Location: The EPRI URD, Chapter 2, paragraph 4.3.1.5, states "All low pressure feedwater heaters shall be located within the condenser neck."
In the System 80+ reference design, only the first stage is located inside the condenser neck.
This is a proven design and does not result in an overly congested condenser design.
Sufficient space is provided such that location of the heaters outside the condenser
i i
t will not hamper plant maintenance operations.
- 12. Turbine Exhaust Connection: The EPRI URD, Chapter 2, paragraph 4.4.3.13, states "A stainless steel expansion joint shall be provided.
A solid connection is permitted if the condenser is spring-mounted."
l The System 80+ design allows for either a flexible rubber seal or l
a rigid seal with a spring-mounted condenser.
The experience database for the rubber seals does not merit, in C-E's opinion, preclusion of their use.
l
- 13. Number of Feedwater Heatina Staces: The EPRI URD, Chapter 2, paragraph 4.3.1.4, states "Seven heating stages... shall be used for the PWR."
The System 80+ reference design uses six feedwater heaters.
For the most efficient plant operation, the actual number of feedwater heaters will be determined by a site-specific heat balance.
i 14.
Atmospheric Dumo Valve Control: The EPRI URD, Chapter 2,
l paragraph 3.4.3.3.1, states "Each main steam line shall be provided with... two PORVs for a two-loop plant."
l The System 80+ design includes safety grade, manually controlled l
(from the control room) dump valves which are used only for plant l
cooldown and decay heat removal under post-accident conditions.
A l
control grade, automatic steam dump and bypass system (with a total I
capacity of 55% of full power steam flow) is provided for control of over-pressure conditions. In conjunction with the Reactor Power Cutback System, the System 80+ design can accomodate a 100% load rejection and, therefore, automatic operation of the dump valves is not necessary.
4.3.5.1, states "The feedwater system shall include three main feed l
- pumps, All three pumps shall be normally operating."
The corresponding rationale indicatos that such an arrangement requires l
the pumps to be run somewhat below their design point, but this arrangement is preferrable because it provides a smoother transient I
following a pump trip and lessens the risk of a plant trip.
The rationale also indicates that an arrangement of two operating pumps with an installed spare has the advantage of running the pumps at l
their design point.
The System 80+ design includes three installed main feed pumps, with two operating and the third in standby status.
Keeping the third pump in standby status reduces wear and tear and maintenance.
Also, running the two operating pumps at their design point is more l
efficient. The System 80+ den i.gn includes the proven Reactor Power l
Cutback System to responf to plant transients and decrease the i
l
i likelihood of a plant trip during events such as the loss of a main feedwater pump.
If one of the main feedwater pumps is lost the Reactor Power Cutback System would decrease reactor power to about i
75% until the spare feedwater pump is brought on line (each of the feedwater pumps has adequate capacity at runout conditions to support 75% reactor power).
16.
Incation of Control Room Suonort Facilities: The EPRI URD, Chapter 10, paragraph 4.9.1, states "The main control room shall include within its security boundary...
An operator's
- area, i
including a restroom and kitchen...."
j The System 80+ design has the non-critical control room operator support facilities close to and easily accessible from the control
- room, but outside its security boundary.
This reduces the potential impact of failures in the non-safety plumbing and electrical systems from affeccing the control room and minimizes the need for special venting of drain lines.
It also minimizes personnel activity within the control room security boundary, including food and janitorial services.
- 17. Advanced Control Complex Desian: The EPRI URD, Chapter 10, paragraph 2.2.10, states "... Each work station shall have the full capability to perform main control room functions...."
It is C-E's understanding that in order to comply with the above j
criterion, each work station would have to have the capability to perform an control and monitoring functions for the plant.
Because computer failures or seismic events could impact the entire man-machine interface for all work stations, the work station and supporting computers must be qualified (safety-related).
Otherwise, the safe shutdown and other control functions would likely have to be performed at a separate, qualified station (which the operator would not use in day-to-day operations).
C-E has taken a more conservative approach.
The System 80+
Advanced Control Complex (called Nuplex 80+") integrates spatially distributed (and dedicated), seismically qualified monitoring and control panels with non-qualified compact work stations. This approach meets all regulatory criteria for separation and independence of redundant safety system channels.
The Nuplex 80+
design ensures that, under accident conditions, the operator will be familiar with the instrumentation used for accident monitoring and mitigation since that same instrumentation is used for day-to-day operations.
The Nuplex 80 + design also uses proven, off-the-shelf digital computer components which are configured for improved online testing, mode-dependent alarm prioritization, validated signal display, and core thermal margin monitoring.
Use of existing technology ensures that Nuplex 80+ is an evolutionary design (vs.
revolutionary) which is compatible with existing training and maintenance programs, while at the same
- time, components are configured using state-of-the-art human factors methodology.
J 16 Use of Sound Powered Phones:
The EPRI URD, Chapter 10, paragraph 4.6.2, states " Portable, Wireless Communication... shall be designed as the primary, dedicated means of communication...."
and "... fixed telephone stations shall be provided... to support may be general communication needs."
Also, "These systems supplemented with additional communication systems such as sound-powered phones...."
The System 80+ design includes sound-powered phones as the primary means of communication.
Separate, dedicated circuits are provided for maintenance, refueling, and emergency activities.
Reliance on sound-powered phones avoids problems which have been experienced with wireless communication, e.g.,
interference with control systems and continuous coverage.
Wireless phones can be used, howe.ver, where needed to ensure communication access to all locations.
19.
Seoarate Switchvards: The EPRI URD, Chapter 11, paragraph 3.3.4, states "The main and reserve off-site power circuits shall be connected to switching stations which are independent and separate."
The off-site power system is site-specific.
Therefore, the System 80+ design includes only a conceptual design in the safety analysis report. The interface requirements provided for the off-site power system do not preclude separate switchyards, but separation is not recuired.
Separate and independent power lines are required, however.
While separate switchyards may be desirable for some sites, there may be site-specific considerations for using a single switchyard.
- 20. Control Room Pressure Soundarv for the HVAC: The EPRI URD, Chapter 6, paragraph 4.2.5.1, states that "All air conditioning...
equipment required for the control room shall be located within...
the control room pressure envelope." The rationale states that the intent is to eliminate in-leakage of unfiltered air (presumably into the control room) and to minimize out-leakage.
The System 80+ control room HVAC equipment is located outside the control room pressure boundary to provide shielding for the filters which may become radioactive subsequent to an accident.
It should be noted that locating the HVAC equipment inside the control room pressure boundary would not eliminate the potential for in-leakage since there would still be penetrations for intake air and refrigeration equipment.
21.
No Fuel Damace for Loss of All Feedwater and No Operator Action:
The EPRI URD, Chapter 1, paragraph 2.3.3.2, states "There I
shall be no fuel damage for at least two hours after sustained loss
{
of all feedwater with no operator action (PWRs only)."
j
The analyr;is for assessing compliance with this criterion shows that - core. uncovery would not occur for at least 90 minutes.
Exteasion beyond this time would require.a larger reactor coolant system. volume.
This would result in more severe mass and energy releases to the containment during a LOCA and a corresponding increase in the containment volume.- It is believed that the System -
80+ design reper sents an optimal balance between containment size, s
reactor coolant system volume, and margin to fuel camage.for loss of feedwater events.
- 22. Safety Deeressurization System canacity:'The EPRI URD, Chapter 5,
paragraph 5.5.2.3.2, states "The SDVS bleed paths shall have sufficient total flow capacity...
to prevent core uncovery following a TLOFW [ total loss of.feedwater) if. feed and bleed is delayed up to 60 minutes from the time the primary safety valves lift.
Analyses.shall'show a margin to core uncovery of at least two feet, using best estimate methods."
The analysis for assessing ~ compliance with this criterion shows that, for the SDS valve size selected, feed and bleed would have to be' initiated at about 30 minutes to maintain a two-foot cover of water above the core.
Based on engineering. judgment, it was
-decided to not have an SDS valve size larger than-the size of the ASME Code pressurizer safety valves. The primary considerations in l
this judgment were 1) restricting reactor coolant system thermal-l hydraulic perturbations when the SDS is used nr.12) minimizing the loss of coolant.should.the SDS remain open longer than intended.
It is believed that 30 minutes is sufficient time.for trained l
operators to initiate feed and bleed.
- 23. Reactor vessel Level Measurement: The EPRI URD,. Chapter 1,
l Appendix B,
Section 2.4.1.1, states "
it is unnecessary to l
specify a reactor pressure vessel (RPV level' instrumentation system for the AIRR. " The EPRI URD, Chapter 4, parargaph 6.3.3.2, states i
"The instrumentation... shall provide for detection of voiding in the upper head of the RPV."
The System 80+ design includes the Reactor Vessel Level Measurement System for detecting voids in the reactor vessel upper head and determining an equivalent liquid level.
i
Enclosure II DIFFERENCES WTCH NO LONGER EXIST BETWEEN THE EPRI URD AND THE SYSTEM 80+ STANDARD DESIGN
- 1. Anti-eiection Latch for Control Element Assemblies (CEAs):
The I
System 80+ design does not have anti-ejection latches.
An early l
I version of the EPRI Utility Requirements Document (URD) considered f
the use of anti-ejection latches to lower the probabability of a CEA ejection and remove the need to analyze the event in the design i
basis. safety analysis.
The current EPRI URD does not include a recommendation for such latches.
There is, therefore, no longer a deviation.
2.
Location of Steam Generator Handholes: The EPRI URD, Chapter 3, paragraph 4.4.1.4.2, states " Access openings and/or inspection ports shall be provided in the secondary shall in the vicinity of the tube sheet surface."
The corresponding rationale states "...
openings may be neaded at each tube support elevation to inspect for sludge accumu3ation and OD corrosion...."
The System 80+ steam generators have handholas at the tube sheet elevation and thus the design is in compliance with the EPRI URD.
With respect to the supporting rationale on the potential need for inspection ports at tube support elevations above the tubesheet, it should be noted that C-E's "aggerate" tube supports have more favorable thermal-hydraulic characteristics than those of other vendors in operating plants.. Moreover, the eggerate supports for the System 80+ design are of Type 409 stainless steel, which is resistant to chemical attack that produces denting..
Inspection l
ports at elevations above the tubasheet have, therefore, no practical application in the current C-E design.
Recent communication with EPRI has confirmed that the C-E design meets the l.
EPRI URD and is consistent with the supporting rationale.
3.
Cross-connection Between Trains of the Emercency Feedwater System:
The EPRI URD, Chapter 5,
paragraph 5.3.3.1.3, states
" Arrangement of the four pumps in two divisions shall minimize cross-connections between individual trains...."
The System 80+ design includes ' a cross-connection between the discharge lines of each pump. This provides a more reliable system for scenarios that go beyond a single failure.
In addition, a cross-connection between divisions allows either of the motor-driven pumps to feed either steam generator when the steam generators are depressurized (the turbine driven pumps are not available when the steam gene::ctors are depressurized).
1 l
i l
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i
)
Since the motor-driven emergency feedwater pumps are credited for keeping the steam generator tubes covered with water during long-1 term post-LOCA conditions, the cross-connection between the motor-driven pumps is required to meet the single failure criterion.
This minimizes the potential for escape of radioactive leakage from the primary system.
In summary, these cross-connections make the system more reliable and are required for safety reasons.
Therefore, this design complies with the EPRI URD.
- 4. Senarate Power Supolv Transformers for Safety Channels: The EPRI URD, Chapter 11, paragraph 7.3.1.5, states that "... vital AC power supply system shall minimize the number of system components...."
l The rationale for this criterion references Figure 11.7-1 as a way of meeting the criterion. Figure 11.7-1 shows a single transformer that powers both safety channels within a division.
The System 80+ design includes an additional transformer within each division such that each safety channel in that division can be powered from a different transformer.
This is required to meet channel separation criteria and ensures that a single transformer failure cannot affect more than one safety channel.
Therefore, this design meets the EPRI criterion to minimize the number of components in this system.
The rationale for the EPRI criterion indicates that use of a common transformer is acceptable for an ALWR, but use of more than one transformer is not precluded.
5.
Offset of the Reactor Coolant System from the Containment Center: The EPRI URD, Chapter 6, paragraph 4.3.4.2, states that
...RCS loop offset dimension from containment centerline shall be optimized for cylindrical containments to provide a large laydown space...."
The corresponding rationale indicates that a 15-to 20-foot offset is feasible.
The System 80+ containment is a 200-foot spherical steel pressure vessel surrounded by a shield building.
Because the operating floor is at the elevation where the diametar is 200 feet, there is ample laydown area without offsetting the reactor coolant system from the containment conterline. The EPRI criterion is applicable to cylindrical (not spherical) containments.
The System 80+
design, however, meets the intent of the EPRI URD.
- 6. Emeroency Diesel Generator Start Time: The EPRI URD, Chapter 11, paragraph 5.3.2.1, states "... combined starting and relevant load sequencing time of each EDG shall be less than 40 seconds...."
In the System 80+ design the diesels start and begin accepting load within 20 seconds and critical emergency equipment is loaded within an additional 20 seconds.
Therefore, this design meets the EPRI URD criterion.
1 7.
Neoorene and Polv-vinyl Chloride Insulation Inside Buildinas:
The EPRI URD, Chapter 11, paragraph 2.6.3.2, states "... polyvinyl chloride (PVC) and neoprene shall not be used..."
The System 80+ procurement specifications for com.ponents inside j
buildings will state that PVC and neoprene insulation should not be used.
8.
Containinent Purce Valve Closure Tifne: The EPRI URD, Chapter 1, Table 1.2-5, states " Automatic actuation or isolation of fluid systems.... sha31 not be required in less than 30 seconds."
Recent calculations have confirmed that the purge valve closure time for System 80+ has been increased to 30 seconds from the previous value of 5 seconds.
The previous estimate of the System 80+ closure time was based on a conservative estimate rather than l
on a specific calculation. The System 80+ design, therefore, meets the EPRI criterion.
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