ML20036A066
| ML20036A066 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/30/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9305100002 | |
| Download: ML20036A066 (86) | |
Text
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GE Nuclear Energy benew vea:r cwnww
!?S Curtner kenur. Sin kse. Lt< 5bG
April 30,1993 Docket No. STN 52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Review Schedule - USis and GSIs
Dear Chet:
Enclosed is a draft replacement for Appendix 198, " Assessment of Applicable Unresolved Safety Issues and Generic Safety Issues," which addresses Open items 20.1-1 and 20.2-1.
Please provide a copy of this transmittal to Melinda Malloy.
Sincerel',
bCf ack Fox Advanced Reactor Programs cc: Norman Fletcher (DOE)
Bernie Genetti (GE)
Carl Szybalski(GE) 060C50 0 q 930510)002 930430 PDR ADOCK 05200001 A
PDR L-a
O M A!.20.1 -l General Dectne Company MM 2 0. I-- f.
ABM PROPRIETARY INFORMATION 23A6100AS Standard Plant a= m wa 19B.1 INTRODUCTION quirements Document (Reference 3) identified in Im topic papers. Safety issue resolutions from the 19B.1.1 Purpose ALWR requirements are superceded by the NRC g/ fig 7 resolutions as the latter are developed. The ABWR The NRC generic licensing, TMI and new ge-design has been compared to the NRC or ALWR neric safety issues in NUREG-0933 (Reference 1) resolution requirements with the resulting evaluation and associated correspondence were reviewed and leading to resolution of the safety issues for the evaluated for the ABWR. The unresohrd issues for ABWR.
the ABWR were identified from the issues applicable to BWR design, and at the beginning of 19B.1.3 References GFY 1989, the NRC resolution was not fully com-plete in documentation or distribution. Status of the (1) A Prioritisation of Gencric Safety Issues, issues from the NRC (Reference 2) and EPRI were.
NUREG-0933, including SupplemenN reviewed. Issues superceded, dropped, resolved, Garch 1981be-regulatory impact or PWR design specific were 43 excluded. The unresolved safety issues requiring (2) Generic Issue Mana e ent Control System -
ABWR resolution according to the NRC severe acci-Fourth Quarter FY Opdate, Memorandum dent policy are identified in Table 19B.1-1.
for James M. Taylor from E.S. Beckjord dated
@ovemtier 3, lW!D //2.3r(// 70,/993 Generic safety issues (GSI) and unresolved safety issues (USI) of concern to the NRC for reviewing (3) Adranced Light Water Reactor Utility ALWR designs are listed * - able 19B.1-2.
set of Requirements Document, Electric Power GSIs and USIs was provi eo in Enciosure 2 o be Institute, Advanced LWR Program.
Request for Additional Information (RA Reference 20.4.17 in the following categories:
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(1) High and Medium Priority Unresolved GSIs.
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(2) Nearly Resolved GSIs (3) Resolved GSis and USIs for which guidance has been issued in the form of Generic Letters, NUREGs, Reports, Bulletins, etc.
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(4) GSIs and USIs for which resolution has resulted in the issuance of a rule, or the development of a standard review plan revision, regulatory guide, or revision to a regulatory guide.
A second set of GSIs and USIs identified in NUREG-0933 Appendix B as resolved was also provided by Reference 20.4.17. A third NRC set of 1
GSIs and USIs, Beyond Appendix B to NUREG-0933, is also listed in Table 19B.1 1. The table includes remarks for reference to the ABWR.
19B.1.2 Summary The ALWR issue description summary and I
resolution summary are included herein to maintain BWR consistency in the ABWR. Also repeated are the pertinent requirements in the EPRI-ALWR Re 19B 1-1 Amendment 17
i Insert 19B.l.1 The ABWR has proposed technical resolutions of those Uresolved Safety Issues (USI) and medium and high priority Generic Safety Issues (GSI) which are identified in the version of NUREG-0933 (Reference 1) current on the date six months prior to the ABWR application and which are technically relevant to the ABWR design in accordance with 10 CFR 52.47 (a) (iv).
NUREG-0933 and associated correspondence (References 2 & 3) were reviewed and evaluated for the ABWR.
The TMI issues satisfying Section II of NUREG-0800, Standard Review Plan, are addressed in Appendix 1A and those satisfying 10 CFR 50.34 (f) are addressed in Appendix 19A.
The remaining issues satisfying severe accident requirements are addressed in Subsection 19B.2.
i The following guidelines were use in the review of NUREG-0933 to eliminate potentially non relevant issues to the ABWR design:
(1) Priority rating of low, dropped, or not yet prioritized.
(2) Operational, environmental, licensing, or other NRC impact with no plant design content.
(3) No design content applicable to the ABWR design except for five NRC selected issues.
(4) Resolved with no new requirements except for six NRC selected issues.
In addition, the NRC staff assisted in identifying relevant and current issues and resolutions.
The group of issues remaining are identified in Table 19B.1-l'and are evaluated in the referenced Subsection.
The COL applicant will evaluate those issues referencing the COL applicant in accordance with Subsection 19B.3.
The documentation of the issue evaluation is comprised of four sections:
- ISSUE, ACCEPTANCE CRITERIA, RESOLUTION and REFERENCES.
The ISSUE statement is a brief summary description of the issue.
The ACCEPTANCE CRITERIA are taken i
from NUREG-0933 and GIMCS (Reference 2). resolution references and where there is no formal NRC resolution, accepted industry codes and standards and good engineering practices.
The RESOLUTION contains the technical resolution of the issue for the ABWR standard plant design.
The REFERENCES identifies documentation other than the SSAR.-
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ABWR 2346toors arv i Standard Plant Table 198.1-1 SAFETY ISSUES INDEX NRC SSAR Title Priority Subsection Generic lunes A-1 Water Hammer Resolved 198.2.2 A-7 Mark I Long-Term Program Resolved 19B.2.3 A-8 Mark 1 Containment Pool Dynamic Loads - Long Term Resolved 19B.2.4 Program A-9 ATWS Resolved 19B.2.5 A-10 BWR Feedwater Nozzle Cracking Resolved 19B.2.6 A-13 SnubberOperability Assurance Resolved 19B.2.7 A-24 Qualification of Class IE Safety Related Equipment Resolved 19B.2.8 A-25 Non-Safety Loads on Class 1E Power Sources Resolved 19B.2.9 A-31 RHR Shutdown Requirements Resolved 19B.2.10 A-35 Adequacy of Offsite Power Systems Resolved 19B.2.11 A-36 Control of Heavy Loads Near Spent Fuel Resolved 19B.2.12 A-39 Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits Resolved 198.2.13 A-40 Scismic Design Criteria - Short Term Program Resolved 19B.2.14 A-42 Pipe Cracks in Boiling Water Reactors Resolved 19B.2.15 A-44 Station Blackout Resolved 19B.2.16 A-47 Safety Implications of Control Systems Resolved 19B.2.17 A-18 Hydrogen Control Measures and Effects of Hydrogen Burns Resolved 198.2.18 on Safety Equipment B-10 Behavior of BWR Mark Ill Containments Resolved 198.2.19 B-17 Criteria for Safety Related Operator Actions Resolved COL App.
B-36 Develop Design, Testing and Maintenance Criteria for Atmospheric Cleanup System Air Filtration and Absorption Umts for Enginected Safety Feature Systems and for Normal Ventilation Systems Resolved 19B.2.21 B-55 Improved Reliability of Target Rock Safety Relief Valves Resolved 19B.2.22 B-56 DieselReliability Resolved 19B.2.23 B-61 Allowable ECCS Equipment Outage Periods Resolved 19B.2.24 B-63 Isolation of Low Pressure Systems Connected to the Resolved 19B.2.25 Reactor Coolant pressure Boundary B-66 Control Room Infiltration Measurements Resolved 19B.2.26 C-1 Assurance of Continuous Long Term Capability of Hermetic Scals on Instrumentation and Electrical Equipment Resolved 19B.2.27 C-10 Effective Operadon of Containment Sprays in a LOCA Resolved 19B.2.2s C-17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes Resolved 19B.2.24 New Generic Issues 15 Radiation Effects on Reattor Vessel Supports High 19B.2.30 -
23 Reactor Coolant Pump Scal Failures High 198.2.31 25 Automatic Air Header Dump on BWR Scram System Resolved 19B.2.32 40 Safety Concerns Associated with Pipe Breaks in the BWR Resolved 19B.2.33 Scram System WI Amendment
ABWR mums nrv 4 Standard Plant P
Table 19 B.1-1 SAFETY ISSUES INDEX (Continued)
NRC SSAR I
Title Priority Subsection f
P New Generic Issues (Continued) 45 Inoperability of Instrumentation Due to Extreme Cold Weather Resolved 19B.234
l 51 Proposed Requirements for Improving the Reliability of Open Cycle, service Water Systems Resolved 19B.235 57 Effects of Fire Protection System Actuation on Safety.Related Equipment Resolved 19B.236 o7.3.3 Improved Accident Monitoring Resolved 19B.237 75 Generic Implicadons of ATWS Events at the Salem Nuclear Plant Resolved 19B.238 l
78 Monitoring of Fatigue Transient Limits for Reactor Coolant Resolved 19B.239 i
System 83 Control Room Habitability Near Res.
19B.2.40 86 Long Range Plan for Dealing with Stress Corrosion Cracking
- 19. 2.41 B
Near Res.
in BWR Piping 87 Failure of HPCI Steam Line Without Isoladon Near Res.
19B.2.42 Medium 19B.2.43 89 Suff Pipe Clamps 103 Design for Probable Maximum Precipitation dgh 19B.2.44 105 Interfacing Systems LOCA at BWRs Resolved 19B.2.45 1% Piping and Use of Highly Combustible Gases in Vital Areas Medium 19B.2.46
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118 Tendon Anchorage Failure Resolved 19B.2.48 120 On-Line Testability of Protection Systems Mediun-19B.2.49 121 Hydrogen Control for Large, Dry PWR Containments Resolved 19B.2.50 124 Auxiliary Feedwater System Reliability Resolved 19B.2.51 128 ElectricalPowerReliability Resolved -
19B.2.52 142 Leakage Through Electrical Isolators in Instrumentadon Circuits Medium 19B.2.53 143 Availability of Chilled Water Systems High 19B.2.54 Resolved 19B.2.55
- 145 Improve Surveillance and Startup Testing Programs 151 Reliability of Recirculation Pump Trip During an ATWS Resolved 19B.2.56 153 Loss of Essential Service Water in LWRs High 19B.2.57 i
155.1 More Realistic Source Term Assumptions Near Res.
19B.2.58 Human Factors Tccues
' Resolved COL App.
i HF.1.1 Shift Staffing HF.4.4 Guidelines for Upgradmg Other Procedures High COL App.
HF.5.1 1.acal Control Stations High COL App.
HF.5.2 Review Crisena for Human Factors Aspects of Advanced Control and tresu.. station High COL App.
5 Issues Resolved With No New Reonirements Resolved 19B.2.59 A-17 SystemsInteraction A-29 Plant Design for Reduction of Vulneralbility to Sabotage Resolved 19B.2.60 B-5 Ductility of Two.Way Slabs and Shells and Buckling Resolved 19B.2.61 Behavior of SteelContainments 9
M AwallO q N*C I f" E' y
_19B.1/
i Amendment i
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i ABWR uisioo s l
Rrv i Standard Plant -
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f Table 198.1 1
't SAFETY ISSUES INDEX (Continued)
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I.t NRC SSAR h
Title Priority Subsection' f
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New Generic issues
( Contin u ed) -
Iunes Resched With No New Reauirements t
(Continued) f i
- 29 Bolting Degradation or Failures in Nuclear Plants Resolved 19B.2.62 -
82 Beyond Design Bases Accidents in Spent Fuel Pools Resolved 19B.2.63
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113 Dynamic Qualification Testing of Large Bore Hydraulic Snubbers Resolved 19B.2.64
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198.2.2 A-h WATER IIAMMER ISSUE Unresolved Safety issue (USI) A-01 in NUREG-0933 (Reference 1) addresses identifying the probable causes of water hammer and minimizing the susceptibility of fluid systems and components to water hammer by correcting design and operational deficiencies.
Water hammer is defined as a rapid deviation in pressure caused by a change in the velocity of a fluid in a closed volume. There are various types of water harruner, including steam condensation-induced water hammer, which occurs in the secondary side of a PWR steam generator at the connection to the feedwater line. This type of water hammer involves steam generator feedrings and piping. Water hammer has been observed in many fluid systems including residual heat removal, contamment spray, service water, feedwater systems, and main steam lines.
In addition to condensation-induced water hammer, other fonns of initiating events which cause water hammer can occur, such as steam driven slugs of water, pump startup with partially empty lines, and rapid valve cycling.
Regardless of the initiating event, water hammer and the resulting fluid accelerations can cause damage to the affected fluid system. De level of severity of damage depends upon the event, and can range from minor damage such as overstressed pipe hangers to major damage to restraints, piping and components.
According to NUREG-0927 (Reference 2), water hammer can be induced by operator / maintenance actions and by design inadequacies. Experience has shown mat water hammer events reported on LERs are about equally divided between operator or maintenance actions and design deficiencies. He NRC implemented SRP changes relative to the design, operation, and maintenance of new plants to minimize the probabdity and effects of water hammer, and issued a Branch Technical Position (BTP) for pre-operational tests.
ACCEfrTANCE CRITERIA Re acceptance criterion for the resolution of USI A-01 is that safety-related process fluid systems shall be designed to conform to the requirements of 10CFR50 ape).2.1,9.2.2,10.3, and 10.4.7 (includndix A GDC 4 (Refe BTP ASB 10 f)uide lementing the lines identified in SRP (Reference 4) Sections 5.4.7,6.3, De following systems shall be designed to include features minimizing the probability of" water hammer" occurrences as well as features to withstand the adverse dynamic loads imposed by " System (ECCS), Resid water hammer" events:
Condensate and Feedwater Systern, Main Steam System. Emergency Core Cooling Removal (RiiR) Containment Spray Subsystem RIIR Shutdow n Cooling Subsystem, Reactor Building Cooling Water (RCW) System, llVAC Normal Cooling Water (liNCW) System, IIVAC Emergency Cooling Water (IIECW) System, and the Turbine Building Cooling Water (TCW) System.
Operating and Maintenance procedures shall include adequate precautions to minimize the potential occurrence of " water hammer".
RESOI UTION The ABWR SS AR Standard Design adequately addresses system dynamic loads such as may result from " water hammer". The entire Condensate and Feedwater System piping is analyzed for water hammer loads that could potentially result from anticipated flow transients tsee Secuan 10.4.7.3).
He main steam, feedwater, and associated drain lines are protected from potential damage due to fluid jets, missiles, reaction forces, pressures, and temperatures resulting from pi = breaks as well as analyzed for dynamic loadings due to fast closure of the turbine stop valves (see Section 5.4.F)).
Condensate-induced Water llammer (CIWil) phenomenon can potentially occur in BWR ECCS piping during transient and/or accident conditions involving reactor depressurization. A General Electric study and analysis was made to review the ABWR RiiR System, the liigh Pressure Core Flooder (llPCF) System, and the Reactor Core Isolation Cooling (RCIC) System piping configuration from the point of view of CIWii. Based upon this analysis, it was determined that since CIWil can only occur in pip' g used to inject cold water into the reactor, the IIPCF, R1IR-m LPFL (Low Pressure Flooder), RCIC, and Feedwater (FW) injection piping was reviewed. It was concluded that the llPCF and R11R-LPFL mode in the LOCA event may have a potenti.d for water hammer. His was concluded after analyzing the fluid conditions inside the reactor pressure vessel (RPV) at the corresponding nozzle height. Ilowever, detailed study of this piping demonstrated that CIWil would not occur. The effect of depressurization transients during a LOCA (flashing oI saturated water in completely filled upstream LPFL pipe) was analyzed in order to calculate the quantity of water left in the pipe when the pressure drops to 250 psia. It was found that about 80% of water still remains in the pipe. Therefore, slow injection of cold water by the LPFL injection valve into the horizontal LPFL pipe parually filled with saturated water will not cause CIWii. In the llPCF System, the presence 1
of two-phase mixture or saturated water at the nozzle height and in the piping inside the RPV avoids the occurrence of high water hammer loads. In the RCIC and FW Systems. the water level in the reactor is above the nozzle level at the time of system initiation, and therefore, there is no CIWii. The overall conclusion is that the proposed ABWR injection pipmg configuration is not susceptible to CIWii.
Dynamic loads caused by condensation-induced " water hammer" as well as piping arrangements and drainage provision protection against water entrainment are adequately addressed by the ABWR SSAR Main Steam System (see Section 10.33).
Dynamic loads, and the provisions of vents and drains (where appropriate) are also addressed in the design of I
the following systems: RilR Shutdown Cooling System, ECCS, RilR Containment Spra IINCW System, llECW System, and the TCW system (see Sections 5.4.7. 6.3.1,5.4.7.1,y System, RC 9.2.11.1, 9.2.11,9.2.12, 9.2.13, and 9.2.14 respectively).
Electric gupi elines), and require proper precautions to mmimize water hammer" potential. ting and Plant l
Pre-operational Tests are established to be perfonned by the COL applicant which include the guidelines of BTP ASB 10-2. These tests verify that unacceptable " water hammer" does not occur in the following systems: RIIR System RCIC System,ilPCF System RCW System,I!ECW S stem,llNCW System, Condensate and Feedwater System, and the 'FCW System (see Sections 14.2.12.1.8, 14.2. l f. l.9, 14.2.12.1.10, 14.2.12.1.29, 14.2.12.1.3 14.2.12.1.33,14.2.12.1.53, and 14.2.12.1.63 respectively).
s Since the design and testing of the safety systems potentially subject to " water hammer" meet the intent of the acceptance criteria above, this issue is resolved for the ABWR SS AR Standard Iksign.
REFERENCESi 1.
NUREG-0933,"A Status Report on Unresolved Safety Issues". U.S. NRC, April 1989.
2.
NUREG-0927, Revision 1 " Evaluation of Water llammer Occurrences in Nuclear Power Piants", U.S. hTC, April 1984.
3.
10CFR50 Appendix A. " General Design Criteria for Nuclear Power Plants". Code of Federal Regulations, Office of the Federal Register, National Archives and Records Administration.
4.
NUREG-0800," Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants-LWR Edition". U.S. NRC.
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19B.2,3 A-7: MAR K I I ONG-TER51 PROGR AM ISSUE During testing for an advanced BWR containment system design (M ARK llD. suppression pool hydrodynamic loads were identilted which had not been considered in ine original design of the MARK I contamment system. To address this issue, a MARK 1 Owners Group was formed and the assessment was divided into a short-term and long-tenn program. The results of the NRC staff s review of the MARK I Containment Short Term Program are desenbed in NUREG-(M08 (Reference 7). The long-term program (LTP) was conducted to provide a generic basis to define suppression pool hydrodynamic loads and the related structural acceptance criteria, such that a comprehensive reassessment of each M ARK I containment system would be performed. A series of experimental and analytical programs were conducted by the MARK 1 Owners Group to provide the necessary bases for the generic load defimtion and structural assessment techniques. The genenc methods proposed by the MARK I Owners Group, as modified by the NRC staff s requirements, will be used to perform plant-umque analyses, which will identify the plant modifications, if any, that will be needed to restore the originally intended margm of safety in the MARK I containment designs. This item was originally identified in NUREG-0311 (Reference 6) and was later determined to be a Unresolved Safety issue (USI).
ACCEPTANCE CRITERI A The objectives of the LTP were to establish design basis (conservative) loads that are appropriate for the anticipatedlife of each Mark I hoiling water reactor (BWR) facility (40 years) and to restore the originally intended design safety margins for each Mark I containment system. The pn,ncipal thrust of the LTP has been the development of generic methods for the definition o(suppression pool hydrodynamic loadings and the associated structural assessment techniques for the Mark I configuration.
RESOLUTION On the basis of the review of the experimental and analytical programs conducted by the Mark I Owners Group, the NRC staff concluded that, with one exception, the proposed supnression pool hydrodynamic load definition procedures, as modified by the NRC Acceptance Critena m Appendix A of Reference 1. will provide a conservative estimate of these loading conditions. De exception is the lack of an acceptable specification for the downcomer.
condensation oscillation loads. In addition, the staff requested confirmatory programs tojustify the adequacy of the loading specifications in the following three areas: (1) adequacy of the data base for specifying torus wall pressures during condensation oscillations,(2) possibility of asymmetnc torus loadine during condensation oscillations, and (3) effect of fluid compressibility in the vent system on I-swellloads. 'lhese programs were documented in b
Reference 3. This report supJtements the Mark 1 SER. 'UREG-0661) by addressing the outstanding issues relating l
to the Mark I containment L t P, namely the downcomer condensation oscillation load definition and the confirmatory analyses and test programs that are intended tojustify the adequacy of the load specifications.
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t he NRC staff has concluded that the improved load definition submitted by the Mark I Owners Group for downcomer condensation oscillation loads is acceptable. In addition, the staff has concluded that the load specification associated with the confirmatory experimental and analytical programs has been justified.
His USI was RESOLVED (Reference 3) in August 1982 with the issuance of Supplement I to NUREG-0661 (Reference 1) and SRP (Reference 2) Section 6.2.1.lC. De load delmition methodology used for the ABWR
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containment design is similar to that used for prior BWR containment designs. Wherever the ABWR unique design features warranted need for additional information for defining ABWR design loads, ABWR unique analyses and tests were conducted to provide an adequate data base for delming tl~ nertinent hydrodynamic loads (Reference 4).
Herefore this issue is resolved for the ABWR.
j REFERENCES 1.
U.S. NRC, " Safety Evaluation Report, Mark I Long Tenn Program, Resolution of Generic Technical Activity A 7,"NUREGM61 July 1980.
2 NUREG-0800, " Standard Review Plan," U.S. NRC.
3.
NUREG-0661, Supplement 1 " Safety Evaluation Report for the MARK 1 Containment Long-Term Program,"
U.S. NRC, August 1982.
i 4.
ABWR SSAR Section 3B.1 ABWR Containment Design.
5.
NUREG-0933, "A Prioritization of Generic Safety Issues," July 1991 (and Supplements 1-12).
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Task Action Plans for Generic Acuvities Category A.1978.
7.
Mark I Containment Short-Term Program,1978.
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t 198,2.4 A-8: M ARK II CONTAINSfENT POOf DYNAMIC LOADS LONG-TERSI PROGR AM ISSUE As a result of the GE testing program for the M ARK III pressure-suppression containment progmm, new containment loads associated with a postulated LOCA were identified in 1975 which had not been explicitly included in the origin:ll design of M ARK I and M ARK 11 containments. These loads result from the dynamic effects of drywell air and steam being rapidly forced into the suppression pool during a postulated LOCA event. Other pool dynainic loads previously unaccounted for result from the actuation of safety /rehef valves (SRVs) in the M ARK 11 containment. The review and evaluation of the M ARK 1 loads were addressed in USI A-7 and SRV loads for all suppression-type containments were addressed in USI A-39. His item was originally identified in NUREG-0371 (Reference 5) and was later determined to be a USI.
ACCEPTANCE CRITERIA The NRC established an acceptance criteria for Mark 11 LOCA-Related Pool Dynamic Loads addressing pool swell loads, condensation oscillation loads, and chugging loads (Reference 1 Appendix A). He original design of the Mark 11 containment system considered only those loads normally associated with design-basis accidents. These included pressure and temperature loads associated with a LOCA, seismic loads, dead loads, jet impingement loads, hydrostatic loads due to water in the suppression chamber, overload pressure test loads, and construction loads.
Ilowever, since the establishment of the original design criteria, additional loadin that must be considered for the pressure-suppression containment-system design.g conditions have been identi RESOI UTION
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He load definition methodology for defining the LOCA pool swell loads, LOCA steam condensation oscillation loads, and LOCA chugging loads on submerged structures for the ABWR will be consistent with the methodology ued for prior plants (Reference 3). This USI was RESOLVED in August 1981 with the issuance of hlfREG-0808 (Reference 1) and SRP Section 6.2.1.lC (Reference 2).
REFERENCF;S 1.
NUREG-0808, "M ARK 11 Containment Program Evaluation and Acceptance Criteria," U.S. NRC, August 1981.
2.
NUREG-0800, " Standard Review Plan," U.S. NRC.
3.
ABWR SSAR Section 3B.5: Submerged Structure Loads.
4.
NUREG-0933, "A Prioritization of Generic Safety Issues," July 1991 (and Supplements 1-12).
5.
Task Action Plans for Generic Acuvities Category A,1978.
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s 19B.2.5 A-9: ANTICIPATED TR ANSIENTS WITIIOUT SCR AM. ( ATWS)
ISSL'E This issue A-9 (Reference 1) addresses the concern that the reactor can attain safe shutdown after incurring an anticipated transient (such as a loss of feedwater, loss of condenser vacuum, or loss of offsite power) with a failure of the reactor protection system to shutdown the reactor (Reference 1). 'lhe technical report on ATWS (WASil-1270) (Reference 2) discussed the probability of an A1WS event as well as an appropriate safety objective for the event. In 1975 the staff published a status report on each vendor analysis which mcluded guidelmes on analysis models and ATWS safety objectives. This issue was resolved by the NRC with the publication of a final rule, 10CFR50.62 (Reference 3).
ACCEPTANCE CRITERI A I
The acceptance criteria for the resolution of this issue is that the reactor must be capable of reaching a safe shutdown condition as identified in 10CFR50.62 after incurring an anticipated transient and a failure to scnun.
i Specifically,10CFR50.62 requires the BWR to have automatic recirculation pump (s) trip, an alternate rod insertion system and an automatic standby liquid control system.
t RESOI.UTION For ATWS prevention / mitigation for the ABWR. the following are provided:
An ARI system diverse and independent of the reactor protection system, Electric insertion of the fine motion control rod drives which is also diverse and independent of the reactor protection system.
Automatic recirculation pump trip, and Automatic initiation of the standby liquid control system.
These features are desenbed in Section 15.8 and fulfill the requirements of 10CFR50.62 to resolve ting,ue for the ABWR.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety issues"(with Supplements 1-12) U.S. NRC, July 1991.
2.
WASil-1270, " Anticipated Transients Without Scram for Water-Cooled Reactors," U.S. NRC, September 1973.
Events for Light-%q,uirements for Reduction of Risk from Anticipated Transients Without Scra 10CTR50.62, "Re 3.
ater-Cooled Nuclear Power Plants."
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19B.2.6 A-10: BWR FFFDWATER NOZZI E CR ACKING ISSUE Inspections of operating BWRs conducted up to April 1978 revealed cracks in the feedwater nozzles of 20 reactor vessels. Most of these BWRs contained 4 nozzles with diameters ranging from 10 inches to 12 inches.
Although most cracks ranges from 1/2 inch to 3/4 inch in de th (including cladding), one crack penetrated the cladding into the base metal for a total depth of approximate y 1.5 inch It was determined that cracking was due to high-cycle fatigue caused by fluctuations in water temperature within the vesselin the nozzle region. These fluctuations occurred during periods oflow feedwater temperature when flow is unsteady and intermittent Once initiated, the cracks enlarge from hi h ressure and thermal cycling associated with startups and shutdowns. This item was originally identitied in b
-0371 and was later determined to be a unresolved safety issue (USI) (References 1 and 2).
ACCEIFI'ANCE CRITERIA Re acceptance criteria is based on developing a design that provides protection to the feedwater nozzles from the water temperature fluctuations. The feedwater nozzles expenence thermal stress because the incoming feedwater is colder than that in the reactor vessel. It is much colder dunng startups before feedwater heaters are in service and during shutdown after heaters are taken out of service. Turbulent mixing of the hot water retuming from steam separators and dryers and the incoming cold feedwater causes thermal stress cycling of nozzle bore unless it is thoroughly protected by the sparger thermal sleeve.
B rpass leakage past the junction of the thermal sleeve and nozzle safe end is the primary source of cold water impinging upon the nozzle bore. A secondary source is the layer of water that sheds off after being cooled by contact with the outer surface of the sleeve, i
RESOLUTION e
The ABWR utilizes a double feedwater nozzle thermal sleeve. An inner thermal sleeve leading the cooler feedwater to the feedwater sparger is welded to the nozzle safe end. The welded thermal sleeve design was adopted to assure that there is no leaEage of cold feedwater between the thennat sleeve and the safe end. A secondary thermal sleeve is placed concentrically in the annulus between the inner thennal sleeve and the nozzle bore to prevent cooled water that may be shedding from the outside surface of the inner sleeve impinge on the nozzle bore and the inside nozzle corner.
The welded double sleeve design gives a low fatigue usage factor in the nozzle bore and at the inner nozzle corner. The design protects the nozzle from fluctuatigmperatures and, therefore, the issue of high cycle fatigue in the feedwater nozzle has been resolved for the AB REFERENCE 1.
NUREG-0619. "BWR Feedwater Nozzle and Control Rod Drive Retum Line Nozzle Cracking," U.S. NRC, November 1980.
2.
NUREG-0371," Task Action Plans for Generic Activities (Category A), U.S. NRC, November 1978.
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19B.2.7 A-13 SNUBBER OPERABIt ITY ASSURANCE ISSUE Generic Safety Issue (GSI) A-13 in NUREG-0933 (Reference 1), addresses snubber selection and operability for safety related systems and components by identifying the need for:
1.
a consistent means of determining snubber operability through standardized functional testing; I
2.
a set of criteria for selection and specification; and, 3.
preservice and inservice inspection programs.
Snubbers are utilized primarily as seismic and pipe whip restraints at operating plants. Reir safety function is to operate as rigid supports for restraining the motion of systems or components under dynamic load conditions such as earthquakes and severe hydraulic transients, e.g., pipe breaks.
According to NUREG-0933, a substantial number of Licensee Event Reports (LER's), conceming snubber operability, were issued by utilities. A review of these LER's showed that a variety of methods were empl,oyed to determme the operability of the snubbers and that different types of snubbers were used for systems with similar configurations.
ACCEIrrANCE CRITERIA ne acceptance criteria for the resolution of GSI A-13 is that the design, specification, installation, and in-service operability of snubbers must meet the intent of the guidance given m SRP Section 3.9.3 (Reference 2).
consideration should be given as to their umque application,ponents for which snubbers are to be used, su Specifically, during the design of safety systems or com conditions and the effect of these responses on the associated system and,ynse to normal, upset, and faulted i.e., their res or component.
RESOLUTION For the ABWR design, snubbers are minimized by using design optimization procedures. Ilowever, where required, snubber supports aie used as shock arrestors for safety-related systems and components. Snubters are used as structural supports during a dynamic event such as earthquake or pipe break, but during normal operation act as passive devices which accommodate normal expansions and contracuons without resistance.
Assurance of snubber operability for the ABWR design is provided by incorporating analytical, design, installation. in-service, and verification criteria to meet the intent of the draft Regulatory Guide (Reference 3) as described in Section 3.9.3.4.1(3). He elements of snubber operability assurance include:
1.
Consideration of load cycles and travel that each snubber will experience during normal plant operating conditions.
2.
Verification that the thermal growth rates of the system do not exceed the required lock-up velocity of the snubber.
3.
Appropriate characterization of snubber mechanical properties in the structural analysis of the snubber-supported system.
4.
For engineered, large bore snubbers, issuance of a design specification to the snubber supplier, describing the required structural and mechanical performance of the snubber with respect to: activation level, release rate, spring rate, dead band, and drag as specified in the draft Regulatory Guide SC-708-4 (Reference 3).
Subsequent verification that the specified design and fabrication requirements were met.
In summary, during the design of safety-related systems or components for which snubbers are to le used, sufficient consideration is given as to their unique application, (i.e., their response to normal, upset and faulted conditions and the effect 0 Gese responses on the associated system and/or component). Dus the design, specification, installation, and in-service operability of snubbers meets the intent of SRP Section 3.9.3 and this issue is resolved for the ABWR design.
8
E REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety issues"(with Supplernents 1-12), U.S. NRC, 2.
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -
LWR Edition". U.S.NRC.
3.
DRAFT Regulatory Guide (SC-708-4), February 1981.
4 9
19B.2.8 A-24: OUALIFICATION OF CI ASS 1E SAFFTY REI ATED EOUIP51ENT ISSUE Safety issue A-24 in NUREG-0933 (Reference 1) addresses the adequacy of environmental qualification methods and acceptance criteria for Class IE electrical equipment.
The Nuclear Regulatory Commission (NRC) initially required license applicants to qualify all safety-related equipment to IEEE Std 323-1974 (Reference 2). Some of the industry qualilication methods and concepts proposed in accordance with this standard, such as testing margins, atng effects, and the simulation of worst case i
environments, were not resolved to the satisfaction of the hKC. It was therefore decided that a, generic approach should be devekiped under A-24 to expedite the review and assessment of equipment qualificat on methods used by vendors.
All major Nuclear Steam Supply Systems (NSSS) vendon and architect engineers submitted topical reports on their methods of environmental qualification which were reviewed by the NRC and the results documented in NUREG-0588 (Reference 3). In a subsequent rulemaking,10 CFR 50.49 (Reference 4) established the requirement for an environmental qualification program for Class 1E electrical equipment together with rules for its content.
References 2 and 3 comprise the bases for the rules. Reculatory Guide 1.89 was then revised (Reference 5) to described an acceptable method for complying with 10 CFR 50.49.
D 'namic and seismic qualification of Class IE electrical equipment was not included in the scope of 10 - 50.49. Existing dynamic and seismic qualification requirements are identified in Regulatory Guide 1.100 '
(Reference 6).
ACCEI'TANCE CRITERI A The acceptance criteria for the resolution of issue A-24 are that safety-related electrical equipment shall be environmenta ualified in accordance with 10 CFR 50.49, and dynamically and seismically qualified in accordance wi e acceptance criteria of Regulatory Guide 1.100 RESOLUTION Environmental design and qualification are described in Section 3.11. The Class IE electrical equipment (including pump and valve motors and electrical accessories) of the ABWR is environmentally qualified by the methods documented in the NRC-appmved report NEDE-24326-1-P (Reference 7). These methods are in accordance with the guidance of IEEE Std 323-1974 (Reference 8), NUREG-0588. Regulatory Guide 1.89 Revision 1, and the genene requirements of 10 CFR 50.49 as described in Section 3.11.
Equipment required to mitigate the consequences of a design basis accident (DB A) or to attain a safe shutdown of the reactor is identified in Section 6.3 for emergency core cooling function and in Appendix 3L3.2 for typical normal and accident environments in that kication including integrated radiation doses.
i Typical environment conditions (temperature, pressure, humidity, integrated radiation dose, and exposure to -
chemicals) are given in Ap'pendix 31.3.2 and cover the design lifetime. Conditions are tabulated for normal operation in and outside of containment, and for loss-of-coolant accident (LOCA) and Mgh energy line break
[
(IIELB)inside containment.
,t Environmental qualification tests and analyses are addressed in Section 3.11.2. The safety-related equipment in I
the areas of Appendix 31.3.2 is required to remain functional in the environmental conditions expected at the l
equipment location during,and after the limiting DB A. Qualification tests and analyses of electncal equipment for the effects of aging, radiauon, temperature, humidity, chemical spray, submereence, and power supply variation, as l
applicable, are performed and the results documented in accordance with NEDE-24326-1-P.
l l'-
Dynamic and seismic qualification testing and analysis of the electrical equipment in SSAR Appendix 31.3.2 are l
addressed in Section 3.10 except for pump motors and valve motor operators, which are addressed m Section 3.9.
The tests and analyses are performed in accordance with IEEE Std 344-1987 (Reference 9), which is endorsed by Regulatory Guide 1.100.
In summary the Class IE electrical equipment is qualified for (a) the environment in which it is required to operate, includmg a limiting DBA at the enJ of design life, in accordance with 10 CFR 50.49, and (b) the seismic and dynamic conditions which it is required to withstand in accordance with the recommendations of Regulatory Guide 1.100. This issue is therefore resolved for the ABWR.
i 10
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC.
2.
IEEE Std. 323-1974, "IEEE Standard for Qualifving Class 1E Equipment for Nuclear Power Generating Stations " Institute of Electrical and Electronics Engmeers.
3.
NUREG-0588," Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
U.S. NRC, July 1981.
4, 10 CFR 50.49, " Environmental Qualification of Electric Equipment important to Safety for Nuclear Power Plant." Office of the Federal Register, National Archives and Records Administration.
5.
Regulatory Guide 1.89 Revision 1. " Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants " U.S. NRC, June 1984.
6.
Regulatory Guide 1.100 Revision 2. " Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants". U.S. NRC, June 1988.
7.
NEDE-24326-1-P, " General Electric Environmental Qualification Program, January 1983 Proprietary).
8.
IEEE Std. 323-1974. "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers.
9.
IEEE Std. 344-1987, " Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers.
t i1
198.2.9 A-29 NON-SAFETY I OADS ON CLASS 1E POWER SOURCES ISSUE 1
Generic Safety issue (GSI) A-25 in NUREG-0933 (Reference 1), addresses the potential safety degradation of a Class IE Power system caused by its cormection to a non-safety-related power source or load There are two approaches to assuring the reliability of the safety-related system Class IE power supplies for future plants. De first approach is to restrict the connection of primarily safety loads to Class IE power supplies.
[In previous designs, non-safety electrical equipment was connected to Class IE power supplies (i.e., the emergerwy diesel generators) to provide a source of power during loss-of-offsite power (LUUP) events.]
ne second approach is to limit the connection of non-safety-related elecuical equipment to the Class IE power systems and assure that when this equipment is connected to the Class IE power systems that the equipment and the connections conform to the requirements for independence, electrical isolation, and physical separauon. These requirements are identified in IEEE Standard 384-1981 (Reference 2), and guidance is provided in Regulatory Guide 1.75. Revision 2 (Reference 3). [ Supplemental information on Class IE safety systems may be found m IEEE Standard 603-1980, IEEE Standard 279-1971, and IEEE Standard 308-1980, (References 4,5 and 6 respectively).]
Both industry and the NRC, through IEEE Standard 384-1981 and Regulatory Guide 1.75, have determined that these design requirements provide an acceptable means of achieving an adequate level of reliability for the Class IE power supplies. Therefore, a commensurate level of safety for the safety systems is assured.
ACCEPTANCE CRITERI A The acceptance criteria for the resolution of GSI A-25 is that the reliability and level of safety of Class IE power sources and the safety systems which they supply may not be degraded by the sharing of loads between safety-related systems and non-safety-related systems.
Specifically, the second approach, identified in the issue statement, shall be used in establishing an acceptable -
level of reliabihty and safety for Class IE power sources and safety-related systems.
This shall be accomplished by assuring that the interface between safety-related and non-safety-related equipment on Class IE power sources and safety-related systems is adequately controlled by meeung the independence, electricalisolation, and physical separation requirements identified in IEEE Standard'384-1981 and other applicable standards. References 2 and 4 through 6, respectively, taking into consideration the guidance provided in Regulatory Guide 1.75, Revision 2.
RESOLUTION Re ABWR Standard Plant design assures the rrliability and safety of the Class IE power sources and safety-related systems by a highly selective connection (i.e.. only one subsystem) of non-safety-related equipment and strict control of the interface between this subsystem and Class IE power system.
He ABWR design incorporates three independent Class IE diesel generators (DGs) and a non-Class IE combustion turbine generator (CTG). The CrG is designed to automatically and independently assume the plant investment protection (PIP) loads, should a LOOP event occur. This is in much the same manner as the DGs assume the Class IE loads for the same event. Therefore, it is not necessary for the Class IE buses to assume the PIP loads.
De ABWR design excludes non-Class-1E from the Class IE busses, with the exception of the altemate rod insertion (ARI) funcuon which is accomplished by the rod control and information system (RC&lS) and the fine-motion control rod drive (FMCRD) subsystem. The reliability of this subsystem is enhanced for the anticipated transient without scram (ATWS) event by using Class IE power for the dnve motors.
non-Class IE FMCRD loads. In addition to the nonnal overcurrent tripping of these load breakers, powe
' Class IE load breakers in the switchgear are part of the isolation scheme between the Class IE zone selective interlocking (ZAl) is provided between them and the upstream Class IE bus feed breakers..The Class IE load breakers, in conjunction with the ZSI feature, provides the needed isolation between the Class IE bus and the non.
Class IE loads. (See 8.3.1.1.1 for more details on this feature relative to the FMCRD power circuits.)
Sinc both the safety systems and their Class IEpower supplies conform to the requirements of IEEE Standard
+
384-1981 and meet the intent of Regulatory Guide 1.75, Revision 2, an acceptable level of safety exists for both the safety systems and their Class IE power supplies. Therefore, this issue is resolved for the ABWR Standard Plant 12
l REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, April 1989.
2.
IEEE Standard 384-1981, " Criteria for Separation of Class IE Equipment and Circuits," The Institute of Electrical and Electronics Engineers, Inc.
3.
Regulatory Guide 1.75, Rev. 2, " Physical Independence of Electric Systems," U.S. NRC, September 1978.
4.
IEEE Standard 603-1980, " Standard Cnteria for Safety Systems for Nuclear Power Generating Stations," The Institute of Electrical and Electronics Engineers, Inc.
5.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," The Institute of Electrical and Electronic Engineers, Inc.
6.
IEEE Standard 308-1980, " Criteria for Class IE Electric Systems for Nuclear Power Generating Stations,"'Ihe Institute of Electrical and Electronic Engineers, Inc.
13
i l
l J
19B.2.10 A.31; RESIDUAL HEAT REMOVAL fRHR) SHUTDOWN REOUIRafENTS i
I ISSUE i
Unresolved Safety Issue (USI) A-31 in NUREG-0933 (Reference 1), addresses the safe shutdown of the reactor, i
following an accident or abnonnal condition other than a Loss of Coolant Accident (LOCA). from a hot standby condition (i.e., the primary system is at or near normal operating tempe'rature and pressure) to a cold shutdown condinon. Considerable emphasis has been placed on long-tenn cooling which is typically achieved by the residual heat removal system which starts to opemte when the reactor coolant pressure and temperature are substantially lower than the hot standby values.
I Even though it may genemlly be considered safe to maintain a reactor in a hot-standby condition for a long time, experience has shown that there have been abnormal occurrences that reauired long-tenn cooling until the reactor coolant system was cold enough to perfonn inspecnon and repairs. For this reason, the ability to transfer heat from the reactor to the environment, after a shutdown resulting frotn an accident or abnonnal occurrence, is an important safety function. It is essential that a power plant be able to go from hot-standby to cold-shutdown conditions subsequent to any accident or abnormal occurrence condition.
l ACCEPTANCE CRITERIA ne acceptance criterion for the resolution of USl A-31 is that the RHR system shall be designed so that the reactor can be brought from a "llot Standby" to a " Cold Shutdown" condition as described in SRP Section 5.4.7 Revision 3 (Reference 2).
Specifically, the RHR system shall meet the intent of the following functional requirements with respect to cooldown:
1.
The desi n shall be such that the reactor can be taken from normal operating conditions to cold shutdown using o y safety-grade systems. Rese systems shall satisfy 10CFR50 Appendix A (Reference 3) General Design nteria (GDC) I through 5.
2.
The system (s) shall have suitable redundancy in components and features, and suitable interconnections, leak connection, and isolation capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) the system function can be accomplished assuming a single failure.
3.
He system shall be capable of being operated from the control room with either onsite or offsite power available. In demonstrating that the system canperform its function assuming a single failure, limited operator action outside of the control room would be considered acceptable, if suitably justified.
4.
He system (s) shall be capable of bringing the reactor to a cold shutdown condition, with either offsite or onsite power available, within a reasonable period of time following a shutdown, assuming the most limiting single failure.
In addition to the functional requirements listed above, there are certain additional requirements for the RilR system including, pressure relief, pump protection, test and operation.
RESOLUTION Re Residual Heat Removal (RHR) is composed of three electrically and mechanically independent divisions desig'nated as A, B, and C with each division containing the necessary piping, pumps, valves, and heat exchangers (see ABWR SSAR Section 5.4.7).
One of the basic design functions of the RHR System is shutdown. Shutdown cooling to remove decay and sensible heat from the reactor, which also includes tne safety-related requirements that the reactor must be brought to a cold shutdown ecmdition using safety grade equipment.
The design basis for the RHR Shutdown Cooling Subsystem is that it is manually activated by the operator from the control room following insertion of the control rods and nonnal blowdown to the main condenser.
For emergency operations where one of the RHR hops has failed, the RHR system is capable of bringing the reactor to the cold shutdown condition of 100*C within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following reactor shutdown with any two of the three divisions. The subsystem can maintain or reduce this temperature further so that the reactor can be refueled and serviced.
14
t The RIIR system is part of the Emergency Core Cooling (ECCS) System, and therefore is required to be designed with redundancy, piping protection, power separauon. and other safeguards as required of such systems (see ABWR SSAR Section 6.3).
{
Shutdown suction and discharge valves are required to be powered from both offsite and standby emergency power for purposes of isolation and shutdown following a loss of offsite power.
t The RiiR System is designed to meet General Design Criteria (GDC) 1,2. 3,4, and 5 for quality assurance, protection against natural phenomenon, environmental and internally generated missiles, pipe breaks, seismic effects and Tires (see ABWR SS AR Section 5.4.7.1.6).
The RilR Shutdown Cooling System is designed to meet the intent of SRP Section 5.4.7 Rev. 3 with respect to providing a means of bringing the reactor plant from hot standby to cold shutdown under all accident or abnonnal 6
occurrence conditions, as described above. Therefore, this issue is resolved for the ABWR SS AR Standard Design.
REFERENCES 1.
NUREG-0933. "A Status Report on Unresolved Safety Issues", U.S. NRC, April 1989.
l 2.
NUREG-0800. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants--
LWR Edition", U.S. NRC.
3.
10CFR50 Apnendix A. " General Design Criteria for Nuclear Power Plants", Code of Federal Regulations.
Office of the 1 ederal Register, National Archives and Records Administration.
f E
T 1
15
19B.2,11 A-% ADEOUACY OF OFFSITE POWER SYSTEM ISSUE Unresolved Safety issue (USI) A-35 in NUREG-0933 (Reference 1) concerns the protection of safety-related equipment from the effects of a sustained undervoltage condition or a rapid rate of decay of the frequency of the offsite power source as well as interaction effects between offsite and onsite power sources. Associated testing requirements are also addressed.
De plant operator historically has performed transient and steady-state stability analyses of the offsite power system which were documented in the Safety Analysis Report (SAR). Ilowever, abnormal occurrences at several operating phnts indicated that a sustained undervoltage condition of the offsite power source not detected by the existing low of voltage relays could result in failure of redundant safety-related equipment.
De NRC therefore evaluated the power systems of operating plants to determine the susceptibility of safety-related electrical equipment to: (1) a sustained undervoltage condiuon on the offsite power source; (2) a rapid rate of decay of the offsite power source frequency; and (3) interaction for the offsite and onsitepower sources. An addiuonal factor evaluated was (4) the adequacy of testinc req _ irements. New criteria relative to factors (1),(3) and u
f (4) above were issued in Branch Technical Position (BTi ) PSB-1 " Adequacy of Station Electric Distribution System Voltages," which was incorporated in SRP Section 8.3.1 Appendix A (Reference 2). Frequency decay [ factor (2)]
t was found not to be a significant safety issue.
ACCEPTANCE CRITERIA Re acceptance criteria for the resolution of USI A-35 is that the design and capability for test and calibration of the undervoltage protection schemes for the Class IE buses of the onsite power system (while connected to the offsite power source) shall conform to the guidance of BTP PSB-1 in Appendix A of SRP Section 8.3.1.
Specifically, a second level of voltage protection shall be pmvided for Class IE equipment in addition to the existing protection based on detecting the complete loss of offsite power to the Class IE buses. He second level shall have two separate time delays before alerting the control room operator and automatically separating the Class IE buses from the offsitepower source, respectively. Duration of the time delays shall ensure protection from sustained low voltage while avoiding disconnection from the offsite source due to short term transients such as motor starting. The undervoltage protection scheme shall have the capability of being tested and calibrated during power operauon. Voltage levels at the safety related buses shall be optimized for the maximum and minimum load conditions that are expected. thmughout the anticipated range of offsite power source voltage variation. Technical Specifications are to melude limiting conditions of operation, surveillance requirements, and protection equipment setpomts.
RESOLUTION Re conceptual design of an offsite power system and station switchyard (s) for the ABWR Standard Plant design is given m Section 8.2. He interface requirements will ensure that the switchyard (s) provide redundant offsste power feed capability to the nuclear unit, consisting of two prefe Ted power circuits, each capable of supplymg the necessary safety loads and other equipment.
De ABWR onsite power systems are described in Section 8.3, and include three redundant and independent 6.9kV Class IE safety buses. He incoming source breakers trip upon loss of normal power, and emergency power is provided to each Class IE bus by separate and independent diesel generator (DG) units. A combustion turbine generator automatically assumes the plant investment protection kiads, but can be used to manually provide back-up power for any Class IE bus, should a DG fail or be out of service.
The Class IE AC Power Systems are described m 8.3.1.1. Protection against degraded voltage is specifically to the recommendations of addressed in Subsection (8)of 8.3.1.1.7. He protection schemes are designed accordmgB-1, IEEE Standard 741-1986 (Reference 3), which is consistent with the guidance of BTP P Re ABWR Standard Plant Class IE auxiliary power system is des gned in compliance with GDC 18 i
(Reference 4) so that inspection, maintenance, cafibration and testing can be carried out with a minimum of interference with operauon of the nuclear unit, as described in 8.3.1.1.5.3. On-line testing is gready enhanced by the design, which ctilizes three independent Class IE divisions. Indication of the system unavailability is provided m the control room.
A Technical Specification establishes limiting conditions for operations, surveillance requirements. trip setpoints with minimum and maximum limits, and allowable values for the undervoltage protection sensors and '
associated time delay devices.
16
~
Protection of the Class IE power supplies to safety-related equipment from the effects of an undervoltage condition of the offsite power source thus conforms to the guidance of BTP PSB-1, and this issue is therefore resolved for the ABWR Standard Plant design.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues"(with supplements), U.S. NRC. April 1989.
2.
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -
LWR Edition," U.S. NRC, 3.
ANSI /IEEE 741-1986, " Criteria for the Protection of Class IE Power Systems and Equipment in Nuclear Power Gercrating Stations," Institute of Electrical and Electronics Engineers, Inc.
Office of the Ipendix A " General Design Criteria for Nuclear Power Plants," Code of Federal Regulations, 10 CFR 50 Ap 4
ederal Register, National Archives and Records Administration.
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I 19B.2.13 A 39: DETER %flNATION ()F SAFETY REI IEF VAI VE POOL DYNA %flC I OADS AND TEN 1PER ATURE I INilTS ISSUE Operation of BWR primary system pressure relief valves can result in hydrodynamic loads on the suppression pool retaining structures located within the pool. These loads result from initial vent clearing of relief valve piping and steam quenching due to hich local pool temperatures. This issue addresses GE M ARK I,11, and III containments and was originally identified in NUREG-0371 but was later determined to be a USI.
ACCEI'TANCE CRITERI A The acceptance criteria set forth for quencher discharge loads are applicable only to the cross-quencher configuration described in Attachment A to Appendix 3B of GESSAR II, Revision 1. Deviation from this configuration shall be reviewed on plant-unique basis. And acceptability of suppression pool temperature limit (s) shall be based on conformance with the resolution of the issue specined in Section 5 of NUREG-0783 (Reference 3).
t RESOLUTION Safety / Relief Valves (S/RVs) are utilized in a BWR pressure suppression system to provide pressure relief durin pool,g certain reactor transients. S/RV steam flow is routed through discharge lines into the pressu transfer during S/RV actuation between the high temperature compressed air and steam mixture and the cooler water in the suppression pool. This enhances heat transfer while providmg a low amplitude oscillating pressure in the pool and elimmates concem over operation at a high suppression p(Reference 1).ool temperature. For ABWR device is a X-quencher such as has been used in prior plants Following the actuation of a S/RV. water contained initially in the discharge line is rapidly discharged through the X-Quencher discharge device attached at the end of the S/RV discharge line. A highly localized waterjet is formed around the X-Quencher arms. The hydrodynamic load induced outside a sphere c~ircumscribed around the quencher arms by the quencher waterjet is not significant. This is the first phase of loading on the suppression pool boundary due to the S/RV blowdown. There are no submerged structures located within the sphere mentioned atove in the Al3WR arrangement. The induced load for submerged structures located outside the circumscribed sphere by the quencher arm is negligible and is ignored (Reference 5).
After the water discharge, the air initially contained in the discharge line is forced into the suppression pool under high sessure. The air bubbles fonned interact with the surrounding water and pmduce oscillating pressure and velocit fields in the suppression pool. This pool disturbance (air-clearing) gives rise to hydrodynamic loads which are e second phase of S/RV blowdown hiading on submerged structures in the pool and on the pool boundary (Reference 5).
The final stage of S/RV blowdown is the steady steam flow phase. Submer loading is from condensing steam jet oscillations at the quencher (Reference 1).ged structure and pool boundary This USI was resolved with issue of SRP (Reference 6) Section 6.2.1.1.C. NUREG-0763 (Reference 2),
NUREG-0783 (Reference 3), and NUREG-0802 (Reference 4) were also issued for Mark I, II, and til containments, respectively. The load definition methodology for defining the S/RV air bubble loads on submerged structures will be consistent with that used for prior plants. Therefore, this issue is resolved for the ABWR (Reference 5).
REFERENCES 1.
ABWR SSAR Section 3B.2.1: Safety / Relief Valve Actuation.
2.
NUREG-0763, " Guidelines for Confimnatory in-plant Tests of Safety Relief Valve Discharges for BWR Plants," U.S. NRC, May 1981.
3.
NUREG-0783, " Suppression Pool Temperature Limits for BWR Containtnents." U.S. NRC, November 1981.
4.
NUREG-0802. " Safety / Relief Valve Quencher Loads: Evaluation for BWR Mark II and III Containments,"
U.S. NRC, October 1982.
19
.i.
5.
ABWR SSAR Section 3B.5.4: S/RV Submerged Structures l
1 6.
NUREG-0800," Standard Review Plan" U.S. NRC.
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19B.2.14 A-40: SEISMIC DESIGN CRITERIA SHORT-TERM PROGRAM ISSUE Generic, Safety issue (GSI) A-40 in NUREG-0933 (Reference 1) addresses short-term improvements in seismic des,gn entena.
i ne seismic design sequence for recently designed plants included many conservative factors. Although it is believed that the overall sequence was adequately conservative, certain aspects may not have been conservative for all plant sites. The objective of NSI A-40 was to investigate selected areas of the seismic design sequence to determine their conservatism for all types of sites, to investigate attemative approaches where desirable, to quantify the overall conservatism of the design sequence, and to modify the NRC critena in the Standard Review Plan (Reference 2), wherejustified.
Studies were conducted, and the results were documented in NUREG/CR-1161 (Reference 3) with specific recommendations for changes in seismic design requirements. In addition, a NRC/ Industry workshop was held to discuss the complex and controversial subject of soti-structure interaction (SSI) analysis. He adequacy of seismic design of large, above ground, venical, safety-related tanks was also of concem to the NRC, Standard Review Plan (SRP) sections were then revised (Revision 2) with the following principal areas of change: Section 2.5.2, updated to reflect the current NRC staff review practice: Section 3.7.1, design time history critena; Section 3.7.2, development of floor response criteria, damping values. SSI uncertainties, and combination of modal responses; and Section 3.7.3, seismic analysis of above ground tanks, and Category I buried piping.
l De NRC concluded in NUREG-1233 (Reference 4) that these revisions would reflect the current state.of-the-art in seismic design in the licensing process. Implementation of the SRP revisions is expected to contribute to a more uniform and consistent licensmg process and is not expected to have significant impact on recently designed plants.
ACCEIrrANCE CRITERIA Re acceptance criterion for the resolution of NSI A-40 is that future nuclear power plants shall conform to the seismic design acceptance enteria and kuidance of Revision 2 to SRP Sections 2.5.2, Vibratory Ground Motion:
3.7.1, Seismic Design Parameters; 3.7.
Seismic System Analysis; and 3.7.3, Seismic Subsystem Analysis.
Specifically, these SRP Sections respectively cover review of the site characteristics and earthquake potendal, the parameters to be used in seismic design, methods to be used in seismic analysis of the over plant, and methods to be used in seismic analysis of individual systems or components.
RESOLUTION Re design ground motions, site envelope soil parameters, and system and subsystem analyses criteria and methods desenbed in Sections 2.5.2,3.7.1,3.7.2 and 3.7.3 meet the intent of Revision 2 of the corresponding SRP sections, except that the OBE is not a design requirement for the ABWR Elimination of the OBE from the design in advanced reactors is consistent with policy issue SECY-93-087 Reference 5. This issue is therefore resolved for the ABWR standard design.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, December 1989.
2.
NUREG-0800," Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants-LWR Edition," U.S. NRC.
3.
NUREG/CR-1161, " Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria,"
U.S. NRC, May 1980.
4.
NUREG-1233, Regulatory Analysis for USI A-40," Seismic Design Criteria " U.S. NRC, April 1988.
5.
Policy issue SECY-93-087, Policy Technical, and Licensing issues Penaining to Evolutionary and Advanced i
Light-Water Reactor ( ALWR) Designs.
l l
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19B.2.15 A-42 PIPE CR ACKS IN BOILING WATER REACTORS ISSUE Generic Safety Issue (GSD A-42 in NUREG-0933 (Reference 1), addresses the past occurrences of intergranular stress corrosion cracking (ina, that have been highly sensitizedIGSCC) in BWR austenitic steel com inents.
vessel nozzles and the pip furnace heat treatment while attached to vessels during fabrication, were m the late 1960's found to be susceptib to IGSCC.
.ACCElvrANCE CRITERIA The acceptance criteria for the resolution of GSI A-42 are that IGSCC resistant materials and fabrication techniques to minimize sensitization shall be used. In addition, the ABWR water shall be maintained at the lowest '
practically achievable impurity levels. Furthennore, the material and fabrication techniques shall comply with the guidelines of NUREG-0313 (Reference 2).
RESOLUTION For the ABWR, IGSCC resistance is achieved through the use of Type 316 stainless steel and compliance with the guidelines of NUREG-0313. All materials are supplied in the solution heat treated condition. During fabncation, any heating operations (except welding) between 427 and 982 C are avoided, unless followed by solution heat treatment. The ABWR water is maintained at the lowest practically achievable impurity levels to minimize its corrosion potential.
In summary, only stainless steel type 316 material is used and all austenitic steel components are fabricated in accordance witn NUREG-0313. Therefore, this issue is resolved for the ABWR Standard design.
REFERENCES 4
1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC.
2.
NUREG-0313 " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," US. NRC, July 1977, (Revision 1) July 1980, (Revision 2) January 1988.
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.i 198.2.16 A-44: STATION BL ACKOUT ISSUE j
De total loss of ac power (that is, the loss of ac power from both the off-site and on-site sources) is referred to i
~ as a station blackout. In the event of a station blackout, the capability to cool the reactor core is dependent on the availability of systems that do not require ac power and on the ability to restore off-site or on-site ac power before other means of coolin#a the core are lost. De concern is that a prolonged station blackout might result in a core damage accident (Reference 1).
t ACCEFTANCE CRITERI A l
t he acceptance criteria for the resolution of this issue for evolutionary ALWRS is compliance with:
a) SECY-90-016-Evolutionary LWR Certification Requirement (Reference 1)
{
b) NRC Commissioner Policy Statement Certification Requirement (Reference 2)
. c) 10CFR50.63, Loss of all Altemating Current Power (Reference 3) d) Regulatory Guide 1.155, Station Blackout (Reference 4) e) NUMARC-87-00 Guidelines and Technical Basis for Resolution of SBO (Reference 5) f) EPRI-URD-Utility Requirements for Evolutionary LWRS (Reference 6).
g) NUREG-1469-NRC-STAFF DFSER-Section 8.3.9 (Reference 7)
RESOLUTION 1
he ABWR design satisfies the acceptance criteria by demonstrating (in SS AR Appendix IC: Station Blackout l
Performance)in that the ABWR can withstand a station blackout without core damage or loss of containment use of the combustion Turbine Generator or the On-Site DGS or the Off pending on the power recovery through the.
- I integrity for a time period internal from 10 minutes to at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> de Site Power Sources for a wide variety of SBO events. Therefore this issue is resolved for the ABWR.
REFERENCES i
1.
SECY-90-016. " Evolutionary LWR Certification Issues and neir Relationship to Current Regulatory Requirements" January 12,1990.
2.
letter J. Taylor to S. Chilk, " Evolutionary LWR Certification Issues and Their Relationship to Current Regulatory Requirements", June 26,199u.
3.
10CFR50.63, " Loss All Alternating Current Power (Station Blackout-SBO), July 21,1988.
3 t'
4.
RG-1.155, " Station Blackout". July 1988.
5.
NUMARC-87-00," Guidelines and Technical Basis for NUMARC Initiation Addressing Station Blackout at LWR's" Plus Supplemental Q/A, January 4,1990.-
6.
EPRI-URD, "EPRI-Utility Requirements Document for Evolutionary ALRW", July,1990.
7.
NUREG-1469, " Draft Final Safety Evolution Report - Design Certification of GE-ABWR (DFSER), October.
i
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i 19B.2.18 A-48: iWDROGEN CONTROL MEASURES AND EFFECTS OF HYDROGEN BURNS ON SAFFTY EOUIPM ENT ISSUE in the unlikely event of a degraded core accident, or following a LOCA, in a lig postulated that the results is the release of large quantities of combustible gases, pn. ht water reactor pla accumulate inside the primary reactor containment as a result of:
1.
metal-water reaction involving the fuel element cladding. Ilydrogen in significant quantity can be formed as a result of the reaction of zirconium fuel cladding at high temperature with steam.
2.
the radiolytic decomposition of the water in the reactor core and the containment sump; e
3.
the cormsion of certain construction materials by the spray solution; and 4.
any synergistic chemical, thennal, and radiolytic effects of post-accident environmental conditions on contamment prot.ctive coating systems and electric cable msulation.
ACCEPTANCE CRITEM Because of the potential for significant hydrogen generation as a result of an accident,10 CFR 50.44,
" Standards for Combustible Gas Control System m Light Water-Cooled Power Reactors," (Reference 1) and.
General Design Criteria 41," Containment Atmosphere Cleanup,"(Reference 2) in Appendix A to 10 CFR Part 50 (Reference 3), requires that systems be provided to control hydrogen concentrations in the containment atmosphere following a postulated accident to ensure that containment integnty is maintained.
t Paragraph (f)(2Xix) of 10 CFR 50.34 requires that provision be made for a hydrogen control system that can safely accommodate hydrogen generated by the equivalent of a 100 percent fuel-clad metal-water reaction.
An inerted contained and the provision of permanendy installed hydmgen recombiners are acceptable as hydrogen control measures.
RESOLUTION Tne issue of a large amount of hydrocen being generated and burned within containment was resolved as stated in the NRC document SECY 89-122 date3 April 19,1989 (Reference 5). His issue covers hydrogen control measures for recoverable degraded core accidents for all BWRs. Extensive research in this area has led to significant revision of the Commission's hydrogen control regulations, given in 10CFR50.44, published December 2,1981.
He ABWR containment is inerted and per 10CFR50.34 (fX2Xix)(Reference 4) can withstand the pressure and energy addition from a 100% fuel clad metal water reaction. However, in the ABWR, there are no design-basis events that result in core uncovery or core heatup sufficient to cause significant metal-water reaction. GE SS AR Section 6.2.5.3 (Reference 6) states that this is equivalent to the reaction of the active clad to a depth of 0.00023 inches or 0.72% of the active clad. Therefore, this issue is resolved for the ABWR.
REFERENCES 1.
10 CFR 50.44, " Standards for Combustible Gas Control System in Light Water-Cooled Power Reactors" 2.
General Design Criteria 41,'" Containment Atmosphere Cleanup".
3.
4.
10 CFR 50.34 Paragraph (fX2Xix).
5.
SECY 89-122 dated April 19,1989.
6.
1 s
25
i 19B.2.19 B-10: BFli AVIOR OF BWR M ARK III CONTAINMENTS ISSUE Evaluation and approval is required of various aspects of the M ARK 111 containment design which differs from the previously reviewed MARK 1 and MARK 11 designs. De task involves the completion of the staff evaluation of the MARK III containment and documentation of the method used to validate the analytical models and assumptions i
needed to predict the contaitunent pressures in the event of a LOCA (Reference 1).
Following a postulated LOCA, escaping steam forces the suppression pool out of the drywell into the wetwell.
This action results in pool swell and loads from vent clearing, jets, chugging, unpact of water, impact from froth impingement, pool fallback, condensation, and containment pressure.
He concern is that these loadings may damage structures and components located within the wetwell.
Although many of these structures (e.g., walkways) are by themselves not related to safety, the various ECCS sy_ stems take suction from the wetwell and, therefore, damage in the wetwell may affect the performance of the ECCS (Reference 6).
ACCEPTANCE CRITERIA On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III contamment.
He NRC staff and its consultants have reviewed these load definitions and their bases and conclude that, with a few specified changes (see Reference 2), the proposed load definitions provide conservative loading conditions.
Appendix C (Reference 2). ptance criteria for LOCA-related hydrodynamic loads are provided in NUREG-0 ne staff approved acce The staff will review each applicant's use of the NRC acceptance criteria for applicability to their plant design.
Mark Ill applicants for a construction permit (CP) need only fumish a commitment to use the staff s acceptance enteria in the design of their containment. Mark 111 applicants for an operating license (OL) will be required to show how the NRC acceptance criteria were applied and tojustify any devinuons taken. For both CP and OL applicants, the information required shall be submitted in a timely manner to allow for the evaluation to be included m the plant's Safety Evafuation Report, or supplements thereto (Reference 2).
The ABWR horizontal vent confirmatory test program was performed to obtain data which could be used to determine condensation oscillation and chugging loads for design evaluation of containment structures. The test matrix included tests at conditions which produce bounding loads and additional tests to examine the sensitivity of the loads to system parameters. The test specifically documents work perfonned, including general evaluation of the test data and the specification of procedures which can be used to define containment kiads.
RESOLUTION i
some unique design features. These unique features include pressurization of the wetwell airspace,gn, b The ABWR design utilizes a horizontal vent system, which is similar to the prior Mark III desi thepresenceof a lower drywell, the smaller number of honzontal vents (30 in ABWR vs.120 in Mark III), extension of horizontal vents into the pool, vent submergence, and suppress. ion pool width (Reference 3).
He ABWR honzontal vent test (liVT) program produced test data which can be used to confirm condensation oscillation (CO) and chugging (Cli) loads for design application. The test demonstrated that a blowdown test facility can be constructed to be very rigid and thereby eliminate fluid-structure interaction effects. It was also shown that a scaled test facility can be used to obtain condensation data for full-scale design application. Most important, an extensive data base which can be used for confinnation of ABWR CO and Cli loads was obtained (Reference 7).
A spectrum of postulated loss-of-accidents (LOCAs) is considered in assessing the desien adequacy of the ABWR containment system. Each of the accident conditions is described in Reference 4. The load dermition methodology for definmg the LOCA induced loads on submerged stmetures is consistent with the methodology used for prior plants (Reference 5). The ABWR is designed to meet the NRC acceptance for Mark Ill LOCA-related pool dynamic loads. Therefore this issue is resolved for the ABWR plants.
26
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REFTRENCES 1.
NUREGo*71, " Generic Task Problem Descriptions (Categories B, C, and D)," US. NRC, June 1978.
2.
NUREG-0978, " Mark III LOCA-Related Ilydrodynamic Load Definition." U.S. NRC, February 1984 3.
ABWR SSAR Section 3B.1.2: ABWR Contairunent Design Features.
4.
ABWR SSAR Section 3B.2.2: Loss-of-Coolant Accidents.
5.
ABWR SSAR Section 3B.5: Submerged Structure Loads.
6.
NUREG-0933, "A Status Report on Unresolved Safety issues," U.S. NRC, April 1989.
7.
NEDC-31393 " Containment florizontal Vent Confirmatory Test Pan 1 - Final Repon," March 1987, i
)
27
)
i i
19B.2.20 B-17: CRITERIA FOR SAFETY-REI ATED OPERATOR ACTIONS ISSUE
~
This issue involves developing criteria for safety-related operater action (SROA) during the response to or recovery from transients and accidents. The criteria would include a determination of actions that shall be automated in lieu of operator action and the development of a time criterion for SROA. Specifically, to be determined, is whether or not to require an automatic switchover from the injection mode to the recirculation mode following a LOCA (Reference 1).
ACCEPTANCE CRITERI A The acceptance criteria for the resolution of issue B-17 is that the plant transient response time (i.e., time required for safety systems or operator to act) shall be increased over cu Tent plants to improve operability, and that the plant desi n shall permit increased operator response time. Required time before the operator must act shall be not less than 0 minutes with a target of.30 minutes, assuming a single failure. Best estimate methodology shall be used for analysis to show safety limits are not exceeded. Operational inputs should be obtained from experienced operators.
RESOL _UTION De ABWR design satisfies the NRC requirements concerning automation of safetv-related rator actions and operator response times. The ABWR resoluuon is the same as the ALWR resolution. For e
, the ABWR design requires no operator action cadier than thirty minutes for any design bases accident. A BWR design,-by incorporating the RIIR heat exchanger in the ECCS injection loop-has eliminated the need for operator actions for several accidents / transients. In fact, even in the long term, opemtor action is only required for one situation -
initiation of containment cooling transients. His is a relatively simple action and some delay in this action should have no adverse consequences, thus eliminating the need to automate this function. In addition, advance CRTs in the control room shall be utilized for monitoring and alarm functions for safety-related and non-safety-related systems (References 2,3. 4). Therefore, this issue is resolved for the ABWR REFERENCES 1.
NUREG-0933. "A Prioritization of Generic Safety Issues."(and Supplements 1 12), July 1991.
2.
EPRI NP-4361. " Power Plant Alarm Systems: A Survey and Recommended Approach For Evaluating Improvements". December 1985.
3.
EPRI NP-5693P," Evaluation of Alternative Power Plant Alarm Parsentations" 4.
EPRI NP-3448,"A Procedure For Reviewing and Improving Power Plant Alarm Systems" April 19M.
28
19B.2.21 B-% DEVELOP DESIGN. TESTING. AND h1AINTENANCE CRITFRIA FOR ATMOSPHERE _
{
CI F ANUP SYSTE51 AIR FILTRATION AND ADSORI" TION UNITS FOR ENGINFFMFI)
[
S AFETY FEATURES SYSTENIS AND FOR NOR%f AL VENTILATION SYSTE51S i
ISSUE I
This NUREG4171, Reference 1 item involves developing revisions to current guidance and staff technical positions regarding ESF and normal ventilation system air Idtration and adsorption units.
l ACCEI"TANCE CRITERI A L'
Develop revisions to BTP ETSB11,2 Reference 2, and Regulatory Guide 1.52, Reference 3, to address technical advances that have shown that some current positions are unjusufiably conservative some are unnecessary, and in some cases additional positions are necessary.
f RESOLUTION Criteria developed as a result of this issue have been documented in Regulatory Guide 152 Revision 2, Reference 3, issued in March 1978 and in Regulatory Guide 1.140 Revision 1, Reference 4, issued in October 1979, and Reference 5. 'Ihus, this item has been resolved.
REFERENCES j
'f 1.
NUREG-0471. " Generic Task Problem Descripdons (Categories B, C, and D)," U.S. NRC, June 1978.
v 2.
NUREG-0800, " Standard Review Plan," U.S. NRC.
and Maintenance Criteria for Post-Accident Encineered-Safety- -
Regulatory Guide 1.52,"Desi n, Testing, Filtration and Adsorption Units of Light-Water-tooled Nuclear Pow 3.
Feature Atmosphere Cleanup stem Air Plants," U.S. hRC, March 19.
?
4.
Regulatory Guide 1.140, "Desien, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Fihrauon and Adsorption Onits of Light-Water-Cooled Nuclear Power Plants," U.S. NRC.
5.
Memorandum for T. Speis from J. Funches "Prioritization of Generic Issues - Environmental and Licensing Improvements," February 24,1983.
?
I 1
t I
r l
f
~29
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i j
19B.2.22 B-55: IMPROVED RFI I ABil ITY OF TARGET ROCK S AFETY/RFI IEF VALVES
. t 1
1 ISSUE Many of the valves in BWR main steam pressure relief systems are Target Rock safety / relief valves, and a i
significant number of failures of these valves have occurred. Failures include valves (1) failing to open pmperly on i
demand, (2) opening spuriously and then failing to rescat properly, and (3) opening properly and then failing to j
rescat properly. The failure of a pressure relief system valve to open on demand results in a decrease in the total available pressure-relieving capacity of the system. Spurious opemngs of pressure relief system valves,or failures of
)
valves to properly reseat after opemng, can result m inadvenent reactor coolant system blowdown with unnecessary l
thermal transients on the reactor vessel and the vessel internals, unnecessary hydrodynamic loading of the containment systems' pressure-suppression chamber and its intemal components, and potential increases in the release of radioactivity to the environs. In addition, if the valve also serves as part of the ADS, a degradation of the capability of the ADS to perform its emergency core cooling function could result.
ACCElvrANCE CRITERIA l
i la the late 1970s, the NRC staff concluded that the inadvertent blowdown events that had occurred as a result 'of malfunctions of pressure relief system valves had neither significantly affected the structural integrity or capability
{
of the reactor vessel or its internals or the pressure-suppression contamment system, nor resulted in any significant radiation releases to the environment. Even af such events were to occur more frequently, there would not likely be -
any significant effects. The performance of these valves, however, is under continual surveillance and the consequences of their failures are subject to review.
. i RESOLIJTION
. j De B-55 issue is not applicable to the ABWR. De ABWR uses a direct acting safety / relief valve design. This i
design does not have a pilot stage such as that present in the Target Rock 2-stage safety / relief valve. Therefore the.
mechanisms which cause the pilot valve to open spuriously and to fail to open properly are not applicable to the 1
ABWR design. It is these mechanisms which have caused the most serious concems with the Target Rock safetyirelief valve performance. By adopting a direct acting safety / relief valve design, these most serious concerns ~
.i are eliminated in the ABWR.
De B-55 issue is only applicable to the BWRs with Target Rock 2-stage safety / relief valves. GE has identified the princi cause of the most significant concern with these Target Rock 2-stage safety / relief valve and has develo a modification to greauy improve the performance of this valve model.
[
i REFERENCES j
t 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements). U.S. NRC.
2.
Memo from Robert Kirkwood to Robert L Baer, Engineering issues Branch. Division of Safety Issue
- f Resolution. Office of Nuclear Regulatory Research, dated on September 2,1992.
I r
6 f
30
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19B.2.23 B.% DIESEI RFf 1 ABILITY ISSUE Generic Safety Issue (GSI) B-56 in NUREG-0933 (Reference 1), addresses emergency diesel generator reliability. The reliability goal identified in NS AC-108, (Reference 2) for emergency diesel generator stanup, is between 0.95 and 0.975 per demand.
Typical onsite electrical distribution systems for plants use diescl generators as an emergency source of power.
Dese emergency power sources supply safety-related equipment, which is used to prevent or mitigate accidents, in the event of a loss of offsite power.
were developed and placed mnificance of the emergency diesel generators. limiting conditions for ope Because of the safety sig the plant technical specifications. Rese LCOs require periodic testing. Licensee.
Event Reports (LERs) sent to the NRC document problems encountered during periodic testing of the emergency diesel generators (to demonstrate operability). As discussed in NUREG-0933, a review of the LERs conducted by the NRC revealed that a diesel generator's staning reliability is, on the average, about 0.94 per demand. Thus, the NRC determined that there was a need to upgrade the reliability of emergency diesel generators. A new reliability of established in Regulatory Guide 1.9, Revision 3 (DRAFF) generator design, operation and periodic test between 0.95 and 0.975 per demand for emergency diesel (Reference 3).
The specific emergency diesel generator starting reliability identified in Regulatory Guide 1.155 (Reference 4)-
is the same as in Regulatory Guide 1.9, Revision 3 mRAFT) (i.e.,it ranges from 0.95 to 0.975 per demand). De resolution of a related Unresolved Safety issue (US1).A-44, Station Blackout, addresses the plant response to station blackout conditions.
ACCFFTANCE CRITERI A periodic testing shall ensure, as a minimum, a staning reliability of 0.95 per demand, generator de The acceptance criteria for the resolution of GSI B-56. is that emergency diesel as identified m Regulatory Guides 1.9, Revision 3 (DRAFr) and 1.155.
RESOLUTION The ABWR Standard Plant design includes an onsite electrical distribution system which employs three redundant and independent Class IE load group divisions. The Class IE safety loads are capable of being supplied power, in decreasing priority, DGs), and the combustion turbine generator (CTG) [see Figure 8.3-1].from th emergency diesel generators (
Each of the three Class IE divisions can be supplied with emergency standhv power from an independent DG.
He DG is designed and sized with sufficient capacity to operate all the needed Class IE loads powered from its ective Class IE divisional bus. Furthermore, each division can be manually supplied from the non-Class IE res#,G, which is diverse from the DGs. De reliability of the CTG is comparable to that of the DG (see Section Cl 9.5.11).
reliability of 0.986 per demand may be achieved as identified in the EPRI AL%y experience has Each DG is spe'ctTied to start reliably and, with present technology industr R Utility Requirements Document (Reference 5). The time required for the DG to attain rated voltare and frequency, and to begin accepting load, has been cased from 13 to 20 seconds after receipt of a start signal. This reduces their starting stress and contributes to improved reliability over the life of the units. The extended time is still within the limitmg case for opening of the.
RilR valves [sec 8.3.1.1.8.2(4)].
A variety of tests are performed to assure DG reliability and operability, in addition to factory tests, a number of pre-operational and onsite acceptance tests and periodic tests are conducted on each DG system. These tests are identified 8.3.1.1.8.2, and in the technical specifications. Also, conditions for operation are imposed to ensure continual reliability.
In summary, the ABWR Standard Plant design utilizes three independent diesel generators as emergency power sources, which are incorpomted in the onsite electrical distribution system, and which have a diverse backup (i.e., the
. CTU).
The onsite electrical distribution system meets the intent of the guidance given in Regulatory Guides 1.9, Revision 3 (DRAFr), and 1.155. Therefore, this issue is resolved for the ABWR Standard Plant design.
i-31-t a
RFFERENCES l.
NUREG-0933, "A Prioritization of Generic Safety issues" (with supplements), U.S. NRC, January 1989.
2.
NS AC-108, " Reliability of Emergency Diesel Generators at U.S. Nuclear Plants," Electric Power Research
~
Institute, September 1986.
3.
Regulatory Guide 1.9, Revision 3 (DRAFD, " Selection, Design, Qualification. Testing, and Reliability of Diesel Generator Units Used as Onsite Electrical Power Systems at Nuclear Power Plants," U.S. NRC, November 1988.
i 4.
Regulatory Guide 1.155," Station Blackout," U.S. NRC, August 1988.
5.
EPRI, " Advanced Light Water Reactor Utility Requirements Document," Electric Power Research Institute, Chapter 11. April 1989.
9-i L
32
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~. -.
,.~,.
i i
19B.2.24 B-61; ' AI I OWABI E ECCS EOUIP3fENT OUTAGE PERIODS f
I E
-l ISSUE Generic Safety Issue (GSI) B-61 in NUREG-0933 (Reference 1) addresses the potential for an overall reduction in the core damage frequency of a plant by reducing the frequency of surveillance testing and reducing permissible i
outage times for safety-related ECCS eqtnpment.
1 Historica*1y, ECCS equipment outage times and surveillance testing were not established by analysis. Instead, these test requirements were developed using engineenng judgment and equipment operating, performance testing.
and maintenance histories. After development, these test requirements were mcorporated into the plant Technical Specifications as Limiting Conditions for Operation (LCOs)
Studies performed for the NRC on operating reactors indicated that from 30 to 80 percent of the ECCS system i
unavailability was due to testing, maintenance atx! allowed outage periods. The NRC is therefore evaluating whether overall ICCS unavailability, and resulting core damage frequency, can be reduced by extending the m.iervals between testing and maintenance of equipment within a range in which equipment unavailabdity due to testing and e
maintenance is reduced more than the predicted equipment unavailability due to failure is increased. Probabilistic 1
risk assessment (PRA) methods wouldbe used to determine the optimum intervals between ECCS equiprrent tests.
i Surveillance intervals optimized in this manner would then be used in LCOs (Reference 8).
j Asa t
d (References 2 and 3)has been devel d for the time defendent unavailabbof this (IREP and NREP, ystbram a compu er co eis code. using generic data fmm the Intt anal respectively), will be used to verify the capability of the code to determine optimum surveillance.
E intervals and resulting overall risk reduaion. The costs and benefits can then be assessed.
i Because the NRC evaluation of this issue has not yet been com leted the initial LCOs for a future plant design may continue to be based on current industry practice without preiubicing later opt mization when the methods and j
i requirements have been confirmed. He overall plant PRA should take the initial LCOs into account, to establish a r
base against which to measure the effects of later optimization.
f t
ACCEPTANCE CRITERI A l
De acceptance criterion for the resolution of GSI B-61 for future plant designs is that the Technical.
.l Specification LCOs surveillance periods and allowable outage times of ECCS equipment shall be developed in accordarac vith current industry practice.
t I CFR 52 (Reference 4). Any subsequent proposed changes to the LCOs' provisions for ECCS surveillance shall be-He LCOs surveillance periods and outage times shall be accounted for in the overall plant PRA required by a
demonstrated to be within the results of an existing PRA (Reference 8).
j e
RESOLUTION j
a he ABWR Standard Plant Design (Reference 7) incorporates many design enhancements to improve the
~
operation and safety of the plant, and the most significant advances are in the area of engineered safety features. The -
. ECCS conforms to all licensing requirements and good design practices of isolation, separation and common mode failure considerations.
-t In order to meet the above requirements, the ECCS network has built-in redundancy so that adequate can be provided, even in the event of s cific failures. Each s stem of ECCS, including flow rate and sensmg networks, is capable of being tested during t operation,includi logic required to automatically initiate component action.
Provisions for testing the ECC network components ( ectrical, mechanical, hydraulic and pneumauc, as applicable) are installed in such a manner that they are an integral part of the design (Reference 7).
De PRA uses a system fault tree approach to quantify system accident sequences which result in severe core
[
damage. Data related to the engineered safety features used m the quantification includes:
i 1.
Component failure rates 1
2.
Component repair times and maintenance frequencies l
3.
Component inspection and test times and frequencies j
i 4.
Allowable equipment outage times
-l 33 i
The data is used in accordance with the guidance in NUREG/CR 2815 (Reference 5), and basic failure rate data is obtained from the ERPI ALWR Requirements Document (Reference 6) supplemented with other nuclear sources.
Maintenance and repair times are calculated as outlined in NUREG/CR-2815. The inspection and test times and frequencies are as specified in ABWR STS Section 3.5.
The PRA demonstrates that the ABWR Standard Design meets the industry goal of 1.0 x E-5 core dama frequeccy per reactor year for future reactors and indicates that the iaitial LCOs are consistent with this goal.ge The owner-operator may refine the LCOs to achieve further risk reduction or increased operauanal flexibility provided that the resulting overall risk is shown to be withi:i the results of the PRA. This issue is therefore resolved for the ABWR.
REFERENCES 1.
NUREG-0933,"A Prioritization of Generic Safety issues" (with supplements). U.S. NRC, December 1989.
2.
NUREG-0193," FRANTIC - A Computer Code for Time Dependent Unavailability Analysis" U.S. NRC, October 1977.
3.
NUREG/CR-1924, " FRANTIC 11 - A Computer Code for Time Dependent Unavailability Analysis". U.S.
NRC, April 1981.
4.
10CFR52. Early Site Permits; Standard Design Certification; and Combined Licenses for Nuclear Power Reactors". Of fice of the Federal Register, National Archives and Recards Administration.
5.
NUREG/CR-2815. "Probabilistic Safety Analysis Procedures Guide", Brooldiaven National Laboratory, January 1984 6.
" Advanced Light Water Reactor Requirements Document - Chapter 1: Overall Requirements. Appendix A:
PRA Key Assumptions and Groundrules". Electric Power Research Institute. Draft, Apnl 1987 7.
CESS AR Design Certification Amendment 1. December 21.1990.
8.
34 J
l 198.2.26 B-66: CONTROI ROOM INFII.TR ATION MEASUREMEN.'T.S ISSUE Issue B-66 in NUREG-0933 (Reference 1) addresses maintenance of the control room in a safe habitable condition under accident conditions by providing adequate protection for the plant operators against airbome i
radiation and toxic gases.
The rate of hir infiltration into the control room is a significant factor in maintaining habitability, and the NRC measured air exchange rates in selected operating reactor plant control rooms to improve the data base for evaluating P
its effects.
No new design requirements were established by the NRC as a result of this and other work related to control room habitability in an accident. Ilowever, more specific review procedures were incorporated in SRP Sections 6.4.1,9.4.1 and 15.6.5.5 (Reference 2), including the habitability review provisions of TM1 Action Plan item til.D.3.4 (Reference 1) regarding analyses of toxic gas concentrations and operator exposures from airborne radioactive matedal and direct radiauon, to ensure more effective implementation of existing requirements.
ACCEL'f ANCE CRITERI A F
The acceptance criteria for the resolution of issue B-66 is that the control room ventilation and air-conditioning systems be designed to maintain the room's environment within acceptable limits for the operation, testing and maintenance of the unit controls and for uninterrupted safe occupancy during normal and accident conditions.
Specifically, these systems shall be designed to meet the intent of the guidance given in SRP, Sections 6.4.1, Revision 2,9.4.1 and 15.6.5.5 Revision 2.
RESOLUTION I
he ABWR control room is heated, cooled and pressurized by a system mixing recirculated air with filtered outdoor air. It is designed to ensure that the operators can remain in the control room and take actions to operate the plant and maintain it m a safe condition during and fo!!owing an accident. There are two air intakes on the top floor side walls of the control building, one on each end. Radiation monitoring sensors in each air intake warn operators of airborne contamination, and cause the liVAC system to switch automatically to an emergency system employing IIEPA and charcoal filters for cleanup.
His control room heating, ventilating and air-conditioning (llVAC) system is designed:
With redundancy to ensure operation in an emergency with a single, active failure.
For radiation exposure limits not exceedine the guidelines of 10CFR50, Appendix A, General Design Criterion 19 (Reference 3), for any of the Chapter 15 DB As, With provisions to detect and remove smoke and airborne radioactive material, To provide a controlled temperature and pressurized envimnment for continued operation of safety-related equ pment under accident conditions, Protection from toxic chemical and chlorine releases.
His ABWR control room and its design bases are described in Section 6.4, liabitability Systems, and Section 9.4.1. Control Room liabitability Area 11VAC.
Since the control room is monitored, pressurized and filtered by the above described systems, and sitee the r
NRC requirements and the guidance for their design are met, the issue of air infiltration is resolved for the ABWR.
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EEFERENCES 1.
NUREG-0933,"A Prioritization of Generic Safety Issues"(with supplements) U.S. NRC, July 1991.
2.
NUREG-0800. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -
LWR Edition". U.S.NRC.
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3.
10CFR50 Appendix A. " General Design Criteria for Nuclear Power Plants" OfTice of the Federal Register, National Arcti.'es Records Administration.
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19B.2.27 C-f r ASSUR ANCE OF CONTINUOUS I ONG-TERN 1 CAPABILITY OF ITFRNfETIC SEAI S ON INSTRUN1ENTATION AND ELECTRICAL EOUIPS1ENT ISSUE Item C-1 in NUREG-0933 (Reference 12), addresses concems regarding the long-term capability of hermetically-scaled instruments and equipment w hich must function in post accident environments. 'NUREG-(M71 (Reference 2) was developed because of these concerns.
Certain classes of instrumentation incorporate seals. When safety-related components within containment must function during post-LOCA accident conditions, their operability is sensitive to the ingress of steam or water, if the seals should become defective as a result of personnel errors in the maintenance of such equipment, such errors i
could lead to the loss of effective seals and the resultant loss of equipment operability. The establishment of a basis for confidence that sensitive equipment has a seal during the lifetime of the plant is needed.
ACCEPTANCE CRITERIA Re NRC has undertaken a program to reevaluate the qualification of all safety-related electrical equipment at all operating reactors. As part of this program, more definitive criteria for environmental qualification of safety.
related electrical equipment have been developed by the staff. The Division of Operating Reactors' " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors"(DOR Guidelines) were completed in November 1979. The Guidelines are intended as a screening device to catch those equipment which might have qualification problems. In addition, for reactors under licensing review, piec the staff has issued NUREG-0588 (Reference 3). The staffintends to evaluate the qualification of all electrical safety uipment in operating plants pursuant to the Guidelines. If problems arise, the staff shall esolve them using h
-0588 as a gmde for theirjudgment.
On May 27,1980 the NRC issued Commission Memorandum and Order CLI-80-21 (Reference 4) ordering that the above two documents form the requirements which licensees and applicants must meet in order to satisfy those aspects of 10 CFR 50, Appendix A, GDC-4, which relate to the environmental qualification of safety-related electrical equipment. The order established an implementation schedule which set a oal that all safety-related June 30,19 jpment in all operating plants be qualified to the DOR Guidelines or NJae electricale u RESOI UTION Environmental qualification of safety-mlated equipment is described in Section 3.11 of the ABWR SSAR.
abnormal,y related equipment located in a harsh environment must perform its proper safety function during nonnal Safet test, design basis accident and post accident environments as applicable. A list of all safety-related electrical and mechanicalequi ent that is located in a harsh environment area will be included in the Environmental Qualification ument (EQD) to be prepared as indicated in 3.11.6.1.
Environmental conditions for the zones where; safety-related equipment is located are calculated for normal, Environmental Design Critena (EQEDC). Environmental conditions are tabulated by zones,quipment Quali abnormal, test, accident and post-accident conditions and are documented in Appendix 31, E contained m the referenced building arrangements.
Safety-related electrical equipment that is located in a harsh environment is qualified by test or other methods as described in IEEE 323 (Reference 5) and permitted by 10CFR50.49(f) (Reference 6).
He qualification methodology is described in detail in the NRC approved Licensing Topical Report on GE's environmental qualification program (Reference 7). His report also addresses compliance with the applicable portions of the General Design Criteria of 10CFR50, Appendix A, and the Ouality Assurance Criteria of 10CFR50, Appendix B. Additionally, the report describes conformance to NUREG-0588, and Regulatory Guides and IEEE Standards referenced in Section 3.11 of NUREG-0800," Standard Review Plan,"(Reference 81 In summary, the safety-related electrical equipment is qualified in accordance with NRC Guidance, including NUREG-0588, and therefore this item is resolved for the ABWR Standard Plant Design.
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REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements). U.S. NRC, January 1989.
2.
NUREG-0471, " Generic Task Problem Descriptions (Categories B, C, and D)," U.S. NRC, June 1978.
3.
NUREG-0588," Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
U.S. NRC, July 1981.
I 4 NRC Memorandum and Order CLI-80-21, docketed May 27,1980.
o 5.
IEEE Standard 323-1983 " Qualifying Class IE Equipment for Nuclear Power Generating Stations," The Institute of Electrical and Electrome Engineers, Inc.
6.
10 CFR 50, " General Design Criteria for Nuclear Power Plants." Code of Fedemi Regulations Office of the Federal Register, National Archives and Records Administration.
7.
NEDE-24326-1-P, " General Electric Environmental Qualification Program " Proprietary Document. January 1983.
8.
NUREG-0800. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -
LWR Edition", U.S.NRC.
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Spray Subsystem, the Primary Containment System, and the SPCU System are listed in refetence ABWR SSAR 16.7.5,16.9.1, and 16.9.2 (see ABWR SS AR Section 3.6.2 for addition LCOs).
i It should be noted that credit is not taken for any fission product removal provided by the drywell and wetwell I
spray portions of the RHR System. He quantity of fission products released mio the environment following j
postulated accidents is controlled by the standby gas treatment system (SGTS) that has the redundancy and capability to filter the gaseous effluent from the primary and the secondary containment.
3 he ABWR Design fulfills the requirements of General Design Criteria 41,42, and 43 relating to fission product removal, bcriodic inspection, and functional testinb..by conforming to the criteriaI')uidelines of SRP Section 6.5.2 Revision (see ABW R SSAR Sections 3.1.2.4.12.
3.1.2.4.13.2, and 3.1.2.4.14..
In summary, the ABWR design meets the intent of the criteria guidelines of SRP Section 6.5.2 Revision 2, and BTP MTEB 6-1 in order to fulfill the function of reducing the concentration of radioactive iodine and particulates in the containment atmosphere during and after a LOCA, while also minimizing the probability of inittaung stress co:Tosion cracking of stainless steel in the safeguard systems. Design features also minimize the probability of inadvertent actuauon of the RilR Containment Spray Subsystem or the SGTS, thus minimizing possible damage to safety related equipment in the containment. Technical Specifications /LCOs are also provided.
GSI C-10 in NUREG4)933 is therefore resolved for the ABWR Standard Design.
REFTRENCES 1.
NUREG-0933, "A Status Report on Unresolved Safety Issues," U.S. NRC, December 1989.
i 2.
NUREG-0800," Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"
r U.S. NRC.
3.
ANSI /AN $6.5-1979,"PWR and BWR Containment Spray System Design Criteria," American National Standards institute.
4.
10CFR50 Appendix A, Code of Federal Regulations. Office of the Federal Register, National Archives and Records Administration.
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19B.2.29 C-17: INTERIM ACCEPTANCE CRITERIA FOR SOI.IDIFICATION AGENTS FOR l
R ADIOACTIVE SOf ID WASTES 1
ISSUE NUREG-N71 Item C-17 (Reference 1) discusses the Interim Acceptance Criteria for Solidification agents for radioactive solid wastes.
ACCEFTANCE CRITERIA Re acceptance criteria for the resolution of C-17 is that there is no current criteria for acceptability of solidification agents. His NUREG-N71 (Reference 1) task involves the development of critena for acceptability of radwaste soliditication agents to properly implement a process control program for the packaging of diverse plant wastes for shallow land burial.
RESOLUTION 10 CFR Part 61 was published in the Federal Register on December 27,1982 (47 FR 57446) and includes Section 61.56 which addresses waste characteristic (Reference 2). A BTP on waste form has been developed under TMI Action Plan item IV.C.I. He ABWR is committed to meeting the requirements in 10 CFR Part 61 (Subsection 11.4.1.2). Rus this item has been RESOLVED for the ABWR.
REFERENCES 1.
NUREG-(M71, " Generic Task Problem Descriptions (Categories B, C, and D)," U.S. NRC, June 1978.
2.
Memorandum for T. Speis from J. Funches, "Prioritization of Generic Issues - Environmental and Licensing l
Improvements," February 24,1983.
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19B.2.30 m RADIATION EFFECTS ON RE ACTOR VESSEI St_'PPORTS ISSUE Generic Safety Issue (GSI) 015 in NUREG-0933 (Reference 1), addresses the potential for failure of the reactor vessel support structure (RVSS) due to a combination of low temperature and neutron flux irradiation embrittlement.
Neutron irradiation of structural materials used in the RVSS causes embrittlement that may increase the 4
potential for propagation of pre-existing cracks or flaws within these materials. The potential for brittle fracture of these materials is typically raeasured in terms of their nil ductility transition temperature (NDIT). As lonc as the operating environment in which a material is used has a temperature that is sigmficandy higher than the NDIT of the material, no failure by brittle fracture would be expected Many materials, when subjected to neutron irradiation, experience an upward shift in the NDIT, i.e., they become more susceptible to brittle fracture. This effect must be accounted for in the design and fabrication of the RVSS.
During 1988, new data was developed for the RVSS materials at Oak Ridge National Laboratory (ORNL)
(References 2 and 3). This data indicated that neutron flux at low temperatures caused greater embrittlement of the materials used in the RVSS than previously anticipated. This increased material embritdement or " upward shift" in NDIT reduces the fracture toughness of these materials and, under certain specific and conservative transient conditions such as an canhquake or Large-break Imss of Coolant Accident, could conceivably result in the failure of the supports thus permitting the reactor vessel to move.
As a result of the ORNL work, the NRC re-prioritized this issue and is reviewing its regulatory position relative to low temperature and neutron flux radiation embrittlement.
ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI 015 is that the material integrity for the RVSS shall be maintained for the design lifetime of the plant.
Specifically the design of the reactor vessel supports shall address irradiation effects (including low temperature and neutron fluxi and the attendant material embrituement, and be designed to restrain the reactor vessel under the combined Safe Shutdown Eanhquake and branch line pipe break loadiri_gs in accordance with the stress and deflection limits established in Section 111 of the ASME B&PV Code (Reference 4).
RESOLUTION The RVSS for the ABWR is described in Subsection 5.3.3.1.4.1 and 3.9.1.4.2 and shown in Figure 5.3 2. The RVSS consists of a support skin bolted to the support pedestal. 'Ihe skin is located below the corebeltline and slighdy below the co e support plate. As such, the skirt is in a region of low neutron flux which is further reduced since the ABWR water flow region between the vessel shroud and vessel wall is almost 40cm wider than previous BWRs. Therefore neutron embritdement of the skirt is well below any current or potential future limitations. A bounding analysis of neutron flux in these regions is given in Subsecuan 5.3.3.1.4.7. The value in this analysis of 14 6 x 10lTnyt can be compared to the bounding expected value for the skin welds of 3 x 10 nyt for a 60 year exposure.
REFERENCES 1.
NUREG-0933, "A Status Repon on Unresolved Safety Issues " U.S. NRC, December 1989.
2.
ORN1/IM-10444," Evaluation of IIFIR Pressure Vessel Integrity Considering Radiation Embritdement," Oak Ridge National Lalmtory,1988.
3.
ORN1/IM-10966," Impact of Radiation Embrittlement on the Integrity of Pressure Vessel Suppons for Two PWR Plants," Oak Ridge National Laboratory,1988.
4.
American Society of Mechanical Engineers, Boiler & Pressure Vessel Code,Section III (Nuclear), American t
Society of Mechanical Engineers.
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19B.232 m AUTO 51 ATIC AIR IIEADER DU51P ON BWR SCR A51 SYSTE5f ISSUE Ris issue concerns the slow loss of control air pressure in the scram system of BWRs. Air pressure dropping at a certain rate will first allow some of the CRD scram outlet valves to open slighdy, thus filling the scram discharge volume with water but allowing litde or no control rod movement. Eventually, Unit 3 on June 8.1980 the rods will to scram but the scram will be impaired in a manner similar to what happened at Browns Ferry (Reference 1). Meanwhile, the dmpping air pressure can cause a transient (e.g., via feedwater controller lockup) which would normally call for a scrara.
ACCEIFTANCE CRITERIA Re acceptance criteria for this issue is specific to the scram discharge volume design and is not applicable to j
the ABWR See the resolution discussion that follows.
RESOLUTION For the ABWR fine motion control rod drive (FMCRD) design, scram water is discharged through the drive directly into the reactor vessel. There is no scram discharge volume as used in previous BWR designs employing the locking piston control rod drive (LPCRD). Conseguently, the common mode failure or impairment of scram associated with loss of control air pressure and filhng of the scram discharge volume is not applicable to the ABWR.
De analogous concern for the ABWR design is that tie slow loss of control air pressure in the scram air header can allow some of the scram accumulators to leak to the reactor. This action could deplete the accumulators' charge arul impair or prevent their capability to scram the connected control rods, unless specific design features are provided to prevent or mitigate its occurrence. De ABWR design does provide protection agamst this event by mcorporaung the followmg features:
1.
A scram air header low pressure alarm to alert the operator of a low pressure condition in the header. The setpoint value is chosen to be greater than the pressure at which the scram valves could start to open in order to allow the operator the opportunity to take corrective action.
2.
A scram initiated by low pressure in the common header supplying the charging water to the scram accumulators. All the accumulators will have sufficient water volume to scram their associated control rods -
as long as the CRD System pump maintains the pressure in the charging header above the minimum required accumulator chargmg nressure even if multiple scram valves are leaking. The pressure in the t
header will drop only if the total scram v, alve leakage flow is greater than the capability of the charging.
chargm,to provide make-up and rnaintain system pressure. If this should occur, mstrumentation pump g header will sense the loss of pressure and signal the RPS to initiate an immediate scram. He setpomt value is based on the minimum accumulator charging pressure. This automatic feature protects the capability to safely shut down the plant by assuring that scram occurs while adequate accumulator charge is still available.
In summary, the ABWR incorporates design features to prevent the loss or impairment of the scram function due to a slow loss of control air in the scram system. De first is a low pressure alarm to alert the operator to trouble in the scram air header; the second is an accumulator charging header low pressure scram to automatically shut down the plant before the accumulators are depleted. Herefore, this issue is resolved for the ABWR design.
REFERENCES 1.
" Report on the Browns Ferry 3 Parual Failure to Scram Event on June 28,1980." U.S. NRC, July 30,1980.
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19B.2.33 40; SAFETY CONCERNS ASSOCIATED WITH PIPE BREAKS IN TIIE BWR SCR AM SYSTFM ISSUE If a break or leak exists or develops in the scram discharge volume (SDV) piping during a reactor scram, this would result in the release of water and steam at 212'F into the reactor building at a maximum flow rate of $50 gpm and is postulated to result in 100% relative humidity in the reactor building. The principal means of isolating this break would be to close the scram exhaust valves which are located on the hydraulic control units; however, this is dependent upon the ability to reset scram, which cannot be absolutely ensured immediately following the scram.
Therefore, a rupture of the SDV could result in the unisolable break outside of pnmary containment, which is postulated to threaten emergency core cooling equipment by flooding areas in which this equipment is located and by causing ambient temperature and relative htunidity conditions for which this equipment is not qualified.
ACCEITANCE CRITERIA NUREG-0803 (Reference 1) provides guidance to ensure SDV pipe integrity, detection capability, mitigation capability and qualification of the emergency equipment to the expected environment.
RESOLUTION For the ABWR fine motion control rod drive (FMCRD) design, semm water is discharged through the drive directly into the reactor vessel. There are no CRD withdraw lines or SDV as used in previous BWR designs employing the locking piston control rod drive (LPCRD). Consequentiv, the issue of SDV isolation provisions as addressed in NUREU-0803 (Reference 1)is not applicable to me ABWR design.
In addition, for protection acainst a scram insert line break, the ABWR FMCRD desien incorporates a ball-check valve located m the FMCRD flange housing at the point of connection of the insert'line with the drive scram port. In the event of a rupture of the insert line, the ball-check valve will close to prevent reactor vessel flow out of the break. This feature is the same as used by the LPCRD in previous BWR designs.
For these reasons, this issue is resolved for the ABWR design.
REFERENCES 1.
NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," U.S.
NRC, August 1981.
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U 19B.2.34 4A INOPFR ABILITY OF INSTRUMENTATION DUE TO EXTRFME COLD WEATHER ISSUE i
Generic Safety issue (GSI) 45 in NUREG-0933 (Reference 1), addresses the potential for safety-related equipment instrument lines to become inoperable as a result of freezing or reaching the precipitation (i.e.,
solidilication) point of the sensing fluids.
Typical safety-related systems employ pressure and level sensors which use small bore instrumentation lines.
Most operating plants contam safety-related equipment and systems, parts of which are exposed to the ambient environment. 'Ihese lines generally contain liquid (e.g., borated water) which is susceptible to freezing. Where systems or components and their associated instrumentation are exposed to sub-freezing temperatures, heat tracing and/or insulation is used to minimize the effects of cold temperatures.
i Rese sensing and instrumentation lines are of concem because, should they freeze, they may prevent a safety-related system or component from performing its safety function. For example, an incident occurred at a plant wherein the heat tracing system surrounding sensing lines and level transmitters for the Refueling Water Storage Tank (RWST) failed dunng sub-freezing weather. The failure of the heat tracing systems resulted in freezing of the sensing lines and associated level transmitters causing a loss of all four RWST instrumentation channels, which -
could have resulted in the failure of the Emergency Core Cooling System, thus jeopardizing plant safety.
P Because of the possibility of a safety-related system failure, the NRC issued additional guidance given in l
Regulatory Guide 1.151 (Reference 2), to supplement the existing guidance and requirements which include the Standard Review Plant (SRP) Section 7.1,10 CFR 50. Appendix A and industry standard IS A-67.02, (References 3, 4, and 5, respectively). Regulatory Guide 1.151 specifically addresses the prevention of safety d alarms.
related instrument sensing line freezing and includes design issues such as diversity, independence, monitoring an ACCEPTANCE CRITERIA i
De acceptance criterion for the resolution of GSI 45 is that the fluid in safety-related equipment instrument
'I sensing lines shall be protected from freezing and maintained above the precipitation point.
t De protection of safety-related equipment instrument sensing lines from freezin providing environmental control systems which meet the requirements of 10 Ap ),7.1 Appendix A,(Rev.
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1), 7.5, (Rev. 3) and 7.7, (Rev. 3).
Also, should environmental control prove to be limited, attemative forms of sensing line protection such as heat tracing and/or insulation can be used. [ _t he use of heat tracing and/or insulation is not anticipated for the ABWR Standard Plant Design, however it is an acceptable alternate to environmental control.]
RESOLUTION j
.j De ABWR Standard Plant incorr.. instrument sensing lines in safety-related systems and components. All '
{
safety-related systems and compone-ed > ; the ABWR Standard Plant design, including instrument sensing lines,.
. aents which are maintained above the freezing (or precipitation) point i
are located in temyerature controlle of the contained fluid. The temperauf these environments are not expected to be less than 10 degrees C, as i
- shown in Appendix 31. In addition, the ABWR is committed to meet the requuements of Regulatory Guide 1.151 (see Table 1.8-20), which endorses and augments ISA S67.02. Therefore, this issue is resolved for the ABWR i
Standard Plant Design.
I REFERENCES i
1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, April 1989.
l 2.
Regulatory Guide 1.151 " Instrument Sensing Lines," U.S. NRC, July 1983.
I 3.
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plantrr-LWR Edition," U.S. NRC.
i 4.
10 CFR 50 Appendix A. " General Design Criteria for Nuclear Power Plants," Code of Federal Regulations, Office of the Federal Register, National Archives and Records Administration.
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ISA-S67.02, " Nuclear-Safety-Related Insuument Sensing Line Piping and Tubing Standards for Use in Nuclear Power Plants," Instrument Society of America,1980.
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1 19B.2.35 51 PROPOSED REOUIREMENTS FOR IMPROVING THE RELI ABII ITY OF OPEN CYCLE SERVICE WATER SYSTEMS ISSUE Generic Safety Issue (GSI) 51 in NUREG-0933 (Reference 1), identifies the susceptibility of the Station Senice Water System (SSWS) to fouling which leads to plant shutdowns and reduced power operation for repairs.
He SSWS cools the Component Cooling Water System (CCWS) through the Component Cooling Water Heat Exchancers and rejects the heat to the ultimate heat sink (UHS) during normal transient, and accident conditions.
De CCWS in turn provides cooling water to those safety-related components necessary to achieve a safe reactor shutdown, as well as to various non-safety reactor auxiliary components.
ACCEPTANCE CRITERIA Elimination of the possible effects of fouling of the service water system and eliminate heat sinks is a design goal of the ABWR. De Plant Designer is given specific requirements and guiding on achieving this goal, including mstruction to consider designs and new requirements which further mitigate the fouling effects. Additionally, the Plant Designer is directed to investigate the problem with ice as a flow blockage meclianism and to dispose of and/or dissolve such ice as required.
Finally, the SWS design of ABWR units at multiplant sites avoids the reliability problems described in this issue be reguiring the SWS to have two or more pumps, so that the loss of one pump will not prevent adequate SWS 110w.
He nnal design of the ABWR ultimate heat sinks and water flow systems will avoid or mmimize as achievablethe problems desenbed in this issue.
RESOLUTION A review of operating plant experience shows that the most prevalent pmblems with plant cooling water systems are due to the cormsion and fouling caused by poor quality service water. In spite of a variety of water treatment schemes and use of expensive material, the wide range of harsh chemistry, silt and biological content result in a need for continuous maintenance of equipment. In order to make a sigmficant Prational improvement in this area. the ABWR requirements for plant coohng) water systems will include the followihg (see Referen Section 19B.2.10, Service Water System Reliability :
(a) Direct mice water will not be used for component cooling. A closed loop component cooling water system will be utilized to transfer heat from the component heat loads via a heat exchanger to the senice water system and heat sink. This minimizes the number of pieces of equipment which are in contact with the problem-causing service water and fa:uses the problem on the component cooling water heat exchanger.
w) Raw service water will be filtered and treated to reduce the effect of mud, silt, or organisms.
(c) Materials for piping, pumps, and beat exchangers are specified to offer greater resistance to the range of probable water chemistry conditions.
(d) Provisions will be made to facilitate the inspection of service water piping and replace sections of piping during plant life.
Sufficient redundancy of makeup pumps shall be provided so that makeup capabilities are not unduly reduced when one pump malfunctions. The need for a safety grade makeup shall be estabitshed in conjunction wnh -
L establishmg UHS water volume, as specified in Reg. Guide 1.27 (Reference 3).
De safety related portions of these systems shall be designed to meet the design bases during a loss of offsite power. These systems shall be designed to perfonn their coolmg function assuming a single active failure in any.
mechanical or electrical system.
However, the NRC staff has a concern regarding the non-inclusion of provisions for: periodic analyses ofintake water and substrate, full flow testing of infrequently used loops, biocide treatment of the reactor service water and other systems susceptible to biofoulmg (Reference 4). Resolution of GSI $1 will require addressing these items in addition to the critena of 19B.2.10 for a site-specific application.
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REFERENCES 1.
NUREG-0933,"A Status Report on Unresolved Safety Issues", U.S. NRC, April 1989.
2.
Advanced Light Water Reactor Utility Requirement Document (Volume 10. EPRI.
3.
Regulatory Guide 1.27. Ultimate llent Sink for Nuclear Power Plants, Revision 2 January 1976.
- 4.. Generic Letter 89-13. " Service Water System Problems Affecting Safety-Related Equipment".
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19B.2.36 057; EFFECTS OF FIRE PROTECTION SYSTEMS ACTUATION ON SAFETY-RFI ATED EOUIPMENT ISSUE Generic Safety issue (GSI) 057 in NUREG-0933 (Reference 1), addresses the potential for safety equipment to become inoperable because of water spray from the fire protection system. IE Informau.-related on Notice 83-41 (Reference 2) identified experiences in which actuauon of fire suppression systems caused damage to safety-related equipment.
ACCEPTANCE CRITERIA He acceptance criteria for the resolution of GSI 057 is that the fire protection system be designed to preclude damaging safety-related equipment and rendering the equipment inoperable. In addiuon, the fue protection system shall be designed to meet 10 CFR 50 Appendix A (GDC 3) (Reference 3); which states in part: " Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse ettects of fires on structures, systems, and components important to safety. Fire fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of those structures, systems, and components."
RESOLUTION i
ne ABWR is designed to preclude water spray from the fire protection system onto safety-related equipment.
The sprinkler systems protecting the safety-related equipment is of the automane sprinkler type. Actuauon of these spnnkler systems requires the opening of the fusible link sprinkler heads and detection by combustible-products and/or heat detectors. In addition, the operator has the capability of isolating flow locally by manual isolation valves.
In order to prevent damage due to flooding, upon actuation of sprinkler systems, floor drains are provided and equipment is located to preclude the flooding of the equipment.
He basic layout of an ABWR and the choice of systems is such as to enhance the tolerance of the ABWR plant to fire. The systems are designed such that there are three independent safety-related divisions, any one of which is capable of providing safe shutdown of the reactor, it is assumed that a fire m any location in a divisional fire area results in an immediate loss of function of the entire division. The remaining two independent safety-related divisions are capable of performing the safe shutdown function.
Since the Fire Protection systems are designed to preclude inadvenent actuation and thus minimize damage to safety-related equipment and because these systems are designed in accotdance with 10 CFR 50 Appendix A (GDC 3), this issue is resolved for the ABWR (See Section 9.5 Appendix 9A)
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety issues"(with supplements), U.S. NRC, April 1989.
2.
IE Information Notice 83-41; Actuation of Fire Suppression System Causing Inoperability of Safety-Related Equipment"; SSINS No. 6835.
3.
10 CFR 50 Appendix A. " General Design Criteria," Office of the Federal Register, National Archives and Records Admmistration.
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19B.2.37 6733-IMPROVED ACCIDENT MONITORING ISSUE nis Generic Safety issue addresses the weaknesses in the accident monitoring of the type observed at the Ginna steam generator event (steam generator isolation, reactor coclant pumps trip, thennat shock from cold high pressure injection water). He weaknesses identified were: (1) non-redundant momtoring of RCS pressure;(2) fadure of the position indication for the steam generator relief and safety valves; and (3) the limited range of the charping pump Ilow indicator for monitoring charging flow during accidents. These conditions make it more difficult for correct operator action in response to such events. Subsequently, the NRC Staff prepared and issued RG 1.97 Rev. 2 (Reference 1) which was implemented at Ginna.
ACCEI'rANCE CRITERIA Re acceptance criteria for the resolution of this item is based on the full implementation of the post accident monitoring requirements of RG 1.97 and NUREG-0737 TMI Action Plans into the design of the ABWR.
RESOllrTION De ABWR has implemented into its basic design RG 1.97 requirements and the TMI action plan requirements of NUREG-0737 and NUREG-037, Supplement 1. Refer to the applicable Sections of 7.5,7.6.2.2,7.6.2.6,11.5, and 12.3.4. The ABWR certification Program is in full compliance with the latest issue of RG 1.97 (Ref. 3).
REFERENCES I
1.
NUREG-0737, NUREG-0737," Clarification of TMI Action Plan Requirements", U.S. NRC, November 1980.
i 2.
NUREG-0933,"A Prioritization of Generic Safety Issues"(with supplements) U.S. NRC, April 1989.
3.
US/ABWR SS AR Doc. # 23 A6100AF Amendment 26.
4.
Regulatory Guide 1.97 Revision 3. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident", U.S. NRC, May,1983.
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19B.2.38 7E GENERIC IMPLICATIONS OF ATWS EVENTS AT S AI E%I NUCI E AR PLANT ISSUE On two occasions, Salem Unit I failed to scram automatically due to failure of both reactor tri on receipt of an actuation signal. In both cases the unit was successfully tripped by manual action.p bre The failure of the breakers has been attributed to excessive wear due to improper inaintenance of the undervoltage relays which receive the trip signal from the protection system and cause mechamcal action to open the breakers.
Failure to scram (also commonly referred to as anticipated transient without scram, ATWS) could result in unacceptable consequences (Reference 1).
ACCEIrrANCE CRITERIA i
Re acceptance criteria for the resolution of this issue is that the plant must have a program for a post-trip review of unscheduled reactor shutdowns, the plant must have a program for safety-related equipment classification and vendor interface, the plant must have a program for post-maintenance operability testing, the plant must have a pro ram to control vendor-related modifications, preventative maintenance and -
surveillance for reactor breakers.
Rese acceptance criteria are described in Generic Letter 83-28 (Reference 2) and NUREG-1000 (Reference 3).
RESOLUTION requirements indicated in Genenc Letter 83-28 and m NUREG-1000.pability for the ABWR to He reactor protection (trip) system (RPS) design provides the ca He ABWR design also addresses and fulfills the ATWS rule of 10CFR50.62 as described in 19B.2.5, A-19 ATWS and Section 15.8.
Herefore this issue 75 is resolved for ABWR.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety issues" (with Supplements 1-12), July 1991, 2.
Generic Letter No. 83-28 " Required Actions Based on Generic implication of Salem A1WS Events". July 8, 1983.
3.
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," Volumes 1,2, April 1983 August 1983.
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19B.2.39 78: MONITORING OF FATIGUE TR ANSIENT I IMITS FOR REACTOR COOL ANT SYSTEM ISSUE Generic Safety issue (GSI) 78 in NUREG-9033 (Reference 1), addresses the concern that for Operating Plants, environmental effects were not taken into account in the design bases for Reactor Coolant Pressure Boundary (RCPB) components. Environmental effects on fatigue resistance of material are not explicitly addressed in the Frequency (CDF) due to environmenjn Fatigue curves. Therefore, an assessment of the in ASME Section III(Reference 2). Des effects on fatigue resistance of material should be perfonned ACCEPTANCE CRITERI A t
he acceptance criteria for the irsolution of GSI 78 are that environmental effects on the fatigue life of ASME Ill Class I carbon stect piping should be considered in accordance with GE Document number 40EllA414 (Reference 3). In addition, for Operating Plants an assessment of the increase in CDF due to environmental effects on fatigue resistance of material should be performed.
Environmental effects are considered by) increasing the local peak stress through four factors used as mu to the stress indices. The four factors are: (1 the notch factor, (2) the mean stress factor,(3) the environmental correction factor, and (4) the butt weld strength reduction factor. The faticue cumulative usage factors are calculated using the calculated local peak stresses and the ASME Section !!! Design Patigue curves.
RESOLUTION For the ABWR, environmental effects are included in the design bases for RCPB components. He calculated CDF includes the environmental effects on fatigue resistance of materials. Herefore, this issue is resolved for tle ABWR Standard design.
REFERENCES 1.
NUREG-0933. "A Prioritization of Generi.; Safety Issues"(with supplements) U.S. NRC, April 1989.
2.
American Society of Mechanical Engineering Boiler and Pressure Vessel Code,Section III 3.
G.E. Document No. 40811A414, Revision 1, " Plain Caroon Steels," General Electric Company L
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e 198.2.40 83-CONTROL ROO51 H ABITABil ITY ISSUE Safety Issue 83 in NUREG-0933 (Reference 1)is a concem over the loss of control room habitability following an accident release of extemal airbome toxic or radioactive material or smoke. Such a loss could impair or cause loss of the control room operators' capability to safely control the reactor and could lead to a core damaging accident.
ACCElrrANCE CRITERIA De acceptance criteria for the resolution of issue 83 is to verify that the control rootn design is adequate to prevent the loss of habitability of the control room during an accident. He desien must meet the guidance given in Standard Review Plan (SRP) Sections 6.4, Section 9.4.1, and Section 15.6.5.5,(Reference 2). De design must be in accordance with 10CFR50, Appendix A, General Design Criteria (GDC) 2,4, and 19 (Reference 3) and ASME AG-1 and AG-la (Reference 5). -
RESOLUTION structure which is important to safety and is des, system is described in Sections 9.4.1 and 6.4. He control Re ABWR main control room habitability igned to withstand the effects of natural phenomena, missiles and postulated accidents in accordance with GDC 2 and 4. The design of the control room (and its heating, ventilation and air condiuoning, llVAC, system) permits safe occupancy dunng abnormal conditions. Radiation exposure of control room habitability area personnel through the duration of any one of the postulated design basis accidents -
. does not exceed the guidelines set by GDC 19,i.e.,5 rem whole body radiation exposure. Smoke and toxic gas :
protection is provided as described m Subsection 6.4.4.2 by the use of non-combustible materials, purging by the HVAC, individual respirators, and site-specific considemuons of potential chemical releases. De control room Enginected Safety Feature filter trains shall be designed. tested and maintained to comply with the applicable provisions of Regulatory Guide 1.52 (Reference 4), as described in Subsection 9.4.1.1.7. Fire protection is provided by alarm systems, fire hoses and portable fire extinguishers, Sections 9.5.1,9A.4.2. Testing and inspection requirements are identified in Section 6.4.5.
Since the control room design parvents the loss of control room habitability during accident conditions, and since all of the NRC requirements and guidance are met, this issue is resolved for the ABWR.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" U.S. NRC,Iuly 1991 (and Supplements 1 12).
2.
NUREG-0800," Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - -
LWR Edition". U.S. NRC.
3.
10CFR50 Appendix A " General Design Criteria for Nuclear Power Plants". Office of the Federal Register, National Arcfiives and Records Admimstration.
4.
Regulatory Guide 1.52,"Desi, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup ystem Air Filtration and Adsorption Units of Light-Water. Cooled Nuclear Power-Plants", Revision 2, March 19 8.
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a 19B.2.42 87 FAILURE OF IIPCI STEAM LINE WITIIOUT ISOLATION ISSUE his issue concerns a postulated break in the Iligh Pressure Coolant Injection System (HPCI) steam supply line and the uncertainty regardmg the operability of the llPCI steam supply line isolation valves under the postulated e
conditions (Reference 1). A similar situation can occur in the Reactor Water Cleanup (RWCU) system.
De HPCI steam supply line has two containment isolation valves (MOV's) in series: one on the inside and one on the ou' side of the contamment. Both are normally open in order for the 11PCI system to perform its function.
The RWCU also has two normally open containment isolation valves (MOV's) which must remain open if the system is to perform its function.
The operation of the valves is tested periodically without steam. Also, due to Dow limitations at the valve manufacturer's facilities, only the opening charactenstics are tested under operating conditions Therefore, according to the NRC, the capability of the valves to close when exposed to the forces created by the flow resulting from a break downstream has not demonstrated.
Furthermore, NRC sponsored testing has increased the concern over whether MOV's can reliably be expected to operate under design basis (i.e., pipe break) conditions.
Under a contract from the NRC, Idaho National Engineering Laboratory (INEL) conducted tests on six MOV's.
The tests showed that all six valves required more force to open and close at the line break flow rates than was predicted. Two of the conditions tested were full guillotine breaks in the RWCU and HPIC systems. Rese test.
results were reported at an NRC sponsored meeting on April 18,1990 which prompted the NRC to issue Generic Letter 89-10 (Reference 2).
ACCEPTANCE CRITERIA The acceptance criteria for the resolution of Generic Safety Issue 87 is defined in Generic Letter 89-10 which requires adequately sized actuators for MOV's, verification of correct thrust and torque settings, and a pmgram for testing, inspection and maintenance of MOV's under differential pressure, temperature and flow condiuons so as to provide assurance that they will function when subjected to design basis condiuons.
RESOLUTION he ABWR does not have an HPCI system. It does have an RWCU system and a Reactor Core Isolation Cooling (RCIC) system which may fall under this issue.
He ABWR addresses the concems and issues identified in GL 89-10 specifically the method of assessment of the loads, the method of sizing the actuators, and the setting of the tor d limit switches) in Sections 3.9.3.2 Testmg of Valves. perability Assurance. 3.9.6.2.2 In-Service Testingque an Pump and Valve O Motor Operated valves and 19B.2.1 In-Situ Rus, compliance with GL 89-10 resolves concems on GSI-87 for the ABWR design.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Saiety Issues" (with supplements), U.S. NRC, April 1989, 2.
Generic Letter No. 89-10. " Safety-related Motor-operated Valve Testing and Surveillance" h
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198.2.43 89: STIFF PIPE CL AMPS ISSUE Generic Safety Issue (GSI) 89 in NUREG-9033 (Reference 1), add esses the concern that for Operating Plants, the effects of stiff pipe clamps were assumed to be negligible and were not explicitly considered in the piping design.
For some applications, there is a concern that certain piping system conditions coupled with specific still pipe clamp design requirements could result in internction effects that should be evaluated in order to determine the significance of the induced pipe stresses.
He ASME Section III Code (Reference 2), requires that, the effects of attachments in producing thermal stresses, stress concentrations and restraints on pressure retaining members be taken into account in checking for compliance with stress criteria. Attachments to piping are generally categorized as integral and non-integraf attachments. Lugs welded to the pipe wall are an example of integral attachments. Clamps used for attaching hangers, struts and snubbers to the pme by bolting are non-integral attachments. Piping design reports specifically address local stresses at integral attachments, such as lugs. An additional stresses induced m the pipe by non-integral, clamp bolted attachments, are not included in the Piping design report.
ACCEPTANCE CRITERIA The acceptance criteria for the irsolution of GSI 89 are that pipe clamps shall only be installed on straight runs of pipe or on bends with a radius of at least five pipe diameters. Smee the clamps will only be installed on straight or very nearly straight runs of pipe, evaluations previously performed show that the peak piping system stresses will not occur at the clamp locations. The stress intensification that occurs at elbows, branch connecuons and lugs is much greater than that which occurs at pipe clamps on essentially straight runs of pipe.
Sample calculations were made for typical clamps used on BWR Main Steam and Recirculation piping sy(stems, to evaluate stresses due to the following loads: (1) Differential thermal expansion of the pipe and the clamp; 2) ine from intemal pressure restraint; (3) Thermal gradient through the pipe wall in the Discontinuity stress in the p(4) Extemal loads produced by dynamic events such as earthquake and tnermo-hy vicinity of the pipe clamp; loads.
Approximate pipe stress distributions were calculated for these loads and conservatively combined to obtain the incremental primary and secondary stresses. Maximum incremental primary stresses were less than 25% of primary stress allowables, and maximum meremental secondary stresses were less than 40% of seconda' ry stress allowables.
The stresses at the clamp locations caleclated without considering local clamp effects typically fell between 15% and 30% of the ASME Section Ill Code allowables. De totalpiping primary and secondary stresses, including the clamp induced local stresses, were less than 70% of the ASME Section III Code allowables. The goveming stress locations occurred at the piping branch connections, lugs, elbows and transitions, they did not occur at the clamp locauons.
RESOLUTION For the ABWR, pipe clamps will only be installed on straight runs of ipe or on bends with radius of at least f
five pipe diameters. Based on the analyses summarized above, the total *npmg stress including the additional stresses induced by the pipe clamp will be less than the governing stresses that occur at piping branch connections, elbows, lugs and transitions. Therefore, the pipe clamp mduced stresses can be considered negligible and do not warrant explicit consideration.
t REFERENCF3 i
1.
NUREG-0M3, "A Fric,ridmoon of Generic Safety issues" (with supplements), U.S. NRC, April 1989.
2.
American Society of Mechanical Engiacering Boiler and Pressure Vessel Code,Section III l
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Issue 105-Page 1 4r 19B.2.15 105: Interfacing System LOCA at ABWRs
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In all currently operating light water reactors, there are a number of high/ low pressure interfaces between the reactor coolant pressure boundary (RCPB) and connected systems.
This leads to the situation that systems in BWRs are designed for a pressure lower than that of the primary system.
For example, the BWR primary system operates an about 2
70 kg/cm g, while the Residual Heat Removal (RHR) System can operate at 2
pressures up to 35 kg/cm g, and the pump suction lines are designed for 2
14 kg/cm.
Isolation valves, at least two, and piping to the primary 2
system are designed for about 88 kg/cm g.
The discharge of the BWR RHR System, which also functions as a low pressure injection system, passes through testable check valves prior to returning-to the reactor coolant system.
The common concern in the above issue is that either tests that require valve actuation, valve leakage, or multiple valve failures could result in a system pressure that exceeds the design pressure of the low pressure emergency cooling systems or other systems interfacing with the RCPB, causing them to fail from overpressure.
Risk calculations on existing plants suggest there may be a need for improved protection against the potential for overpressurization of some emergency cooling and decay heat removal systems (Reference 1).
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Acceptance Criteria Reference 2, indicated that future ALWR designs like the ABWR should reduce the possibility of a LOCA outside containment by designing to the extent practicable all systems and subsystems connected to the reactor coolant system (RCS) to an ultimate rupture strength (URS) at least equal to full RCS pressure.
Reference 3 found that for the ABWR the design pressure for the low-pressure piping systems that interface with the RCPB should have the following criteria to achieve satisfactory retention of the full 1040 psia reactor pressure on an ultimate rupture strength basis.
i 1.
The design pressure for the low-pressure piping systems that-l interface with the RCPB pressure boundary should be equal to I
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Issue 105 Page 2 0.4 times the normal operating RCPB pressure of 1025 psig (i.e.,
2 410 psig). 410 psig = 28.8 atg, where 1 at = 1 kg/cm and atg is gage.
2.
The minimum wall thickness of the low-pressure piping should be no less than that of a standard weight pipe.
i 3.
The remaining components in the low-pressure systems should also be designed to a design pressure of 0.4 times the normal i
operating reactor pressure (i.e., 410 psig).
This is accomplished in the SSAR by the revised boundary symbols of the P& ids to.
the 28.8 atg design pressure, which includes all the piping and components associated with the boundary symbols.
4.
A Class 300 valve is adequate for ensuring the pressure of the low-pressure piping system under full reactor pressure.
5.
The design is to be in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Subarticle NC/ND-3600.
p 6.
Periodic surveillance and leak rate testing are required of the pressure isolation valves per Technical Specification requirements as a part of the ISI program.
t Resolution The ABWR design was evaluated and upgraded to comply with the above criteria.
Criteria numbered 1 through 4 were accepted and irnplemented
-l in the SSAR documentation primarily by indicating the design pressure and design features on the system P& ids (Piping and' Instrument Diagrarns).
Criteria 5 and 6 were originally part of the ABWR design, and no upgrade was required to comply.
The increased design pressure was extended, forming an URS region extending outward from~ the RCPB, to the extent practicable.
The following items form the basis of what constitutes practicality and set forth the test
.j of practicality used to establish the boundary limits of URS for the ABWR:
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1.
It is impractical to design large tank structures to the URS design pressure that are vented to atmosphere and have a. low design l
pressure.
Tanks included in this category are:
Issue 105 Page 3 Condensate storage tank, SLC main tank, LCW receiving tank, HCW receiving tank, FPC skimmer surge tank, and i
FPC spent fuel storage pool and cask pit.
i These are termed low pressure sinks for the purposes of this discussion.
The suppression pool is also a low pressure sink that is impractical to upgrade its pressure since it -is part of the containment structure, which is designed to contain the most i
severe LOCA.
2.
It is impractical to consider a disruptive open flow path from reactor pressure to a low pressure sink.
As a consequence, the furthest downstream valve in such a path is assumed closed (with nominal leakage) so that essentially all of the static reactor pressure is contained by the URS upgrade.
3.
It is impractical to design piping systems that are connected to low pressure sink features to URS design pressure when the piping is always locked open to a low pressure sink by locked open valves.
Nominal leakage past the last closed valve is the only pressure source that could pressurize the line, and that line is locked open to the low pressure sink vented to atmosphere.
L As implied above, boundary limits of the URS design are established assuming slow rates of leakage between high and low pressure regions.
A key assumption to understanding the establishment of the boundary limits from the above practicality basis is that only static pressure conditions are considered.
Static conditions result by assuming that the last valve in tne URS region adjacent to a low pressure sink remains closed, yet considering a slow leak rate that would not over pressurize the down stream piping and components.
Thus, the dynamic pressurization effects, violent high flow ~ transients, and temperature escalations are precluded that would i
occur if the full RCPB pressure was connected directly to the low pressure sink. This means only static pressurization with small leak flow needs to be considered, and low pressure sinks do not pressurize because _ the path to
~
the sink is open.
The following twelve systems, interfacing directly or indirectly with the RCPB, were evaluated and upgraded.
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Issue 105-Page 4 1.
Residual Heat Removal (RHR) System 2.
High Pressure Core Flooder (HPCF) System 3.
Reactor Core Isolation Cooling (RCIC) System 4.
Control Rod Drive (CRD) System 5.
Standby Liquid Control (SLC) System 6.
Reactor Water Cleanup (CUW) System 7.
Fuel Pool Cooling Cleanup (FPC) System 8.
Nuclear Boiler (NB) System 9.
Reactor Recirculation (RRS) System 10.
Makeup Water (Condensate) (MUWC) System 11.
Makeup Water (Purified) (MUWP) System 12.
Radwaste System (LCW Receiving Tank, HCW Receiving Tank).
The low pressure piping boundaries were upgraded to URS pressures and extend to the last closed valve connected to piping interfacing a low pressure sink, such as the suppression pool, condensate storage - tank or other open configuration (identified pool or tank).
The lines from the URS i
boundary to the low pressure sink do not pressurize because the path to the sink is open.
Each interfacing system's piping was reviewed for upgrading.
For some systems, with low pressure piping and normally open valves, the valves were changed to lock open valves to insure an open piping pathway from the last URS boundary to.the tank or low pressure sink.
In addition to the above 12 systems, the following two systems were identified as requiring an ISLOCA evaluation.
Condensate, Feedwater and Condensate Air Extraction (C,FDW,AO)
System Sampling (SAM) System However, these two ' systems are not in sufficient detail to perform an ISLOCA evaluation.
Therefore, an ISLOCA evaluation for these two systems--
is cited in the SSAR as requirements for the COL applicant.
The periodic surveillance testing of the ECCS injection valves that interface L
with the reactor coolant system might lead' to ISLOCA conditions if their i
associated testable check valve was stuck open.
To avoid this occurrence, l
the RHR, HPCF, and RCIC motor operated injection valves will only be 1
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Issue 105 Page 5 i
tested during low pressure shutdown operation.
This practice follows from the guidance given by Reference 4, page 8, paragraph 7.
Although the following is not a new design feature, the RHR shutdown cooling suction line containment isolation valves are also only tested during shutdown operation.
These valves are interlocked against opening for reactor pressure greater than the shutdown cooling setpoint l
approximately 9.49 kg/cm2 gage (135 psig).
In summary, based on the NRC staff's new guidance cited in References 2 through 5, the ABWR is in full compliance. For ISLOCA considerations, a design pressure of 28.8 atg or (410 psig) and pipe having a minimum wall thickness equal to standard grade has been provided as an adequate margin with respect to the full reactor operating pressure of 72.1 atg (1025 psig) by applying the guidance recommended by Reference 2.
This design pressure was applied to the low pressure piping at their_ boundary.
symbols on the P& ids, and therefore, impose the requirement on the associated piping, valves, pumps, tanks, instrumentation and all other equipment shown between boundary symbols.
A note was added to each URS upgraded P&lD requiring pipe to have a minimum wall thickness equal to standard grade.
Upgrading revisions were made to 12 systems.
References 1.
NUREG-0933, "A Prioritization of Generic Safety Issues," (and Supplements 1 through 12), July 1991 2.
Dino Scaletti, NRC, to Patrick Marriott, GE, " Identification of New Issues for the General Electric Company Advanced Boiling Water Reactor Review," September 6,1991 3.
Chester Poslusny, NRC, to Patrick Marriott, GE, " Preliminary f
Evaluation of the Resolution of the Intersystem Loss-of-Coolant i
Accident (ISLOCA) Issue for the Advanced Boiling ~ Water Reactor
_(ABWR) - Design Pressure for Low-Pressure Systems,"
December 2, 1992, Docket No.52-001 4.
James M. Taylor, NRC, to The Commissioners, SECY-90-016,
" Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements,"
Jan. 12, j
1990
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5.
James M. Taylor, NRC, to The Commissioners, SECY-93-087,
" Policy, Technical, and Licensing Issues Pertaining _ to Evolutionary and Advanced Light-Water Reactor (ALWR)
Designs," April 2,1993 L
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19B.2.46 10& ' PIPING AND THE USE OF HIGHI Y COMBUSTIBI E C ASES IN VITAL AREAS r
ISSUE I
i Combustible gases such as hydrogen, propane and acetylene are used during normal operation of nuclear power i
plants and in plant laboratories. Most gases are used in limited quantities and for relatively shortperiods of time.
Ilydrogen, the most prevalent combustible gas used in nuclear power plants, is used as a coolant tor electric.
I generators. liydrogen also is used in the volume control tank (VCF), hcited in the auxiliary systems building of PWRs. De concern is that a hydrogen leakage and accumulation in this building could igmte and disable safety-
)
related equipment.
ACCEfrrANCE CRITERIA
[
The acceptance criteria for the resolution of issue 106,is that the hydrogen and other combustible gas piping be designed to preclude large releases and accumulation of combustible mxtures in buildings which enclose safety-.
- l related eq(uipment. Furthermore, the designer shall follow the guidance described in SRP 9.5.1, Fire Protecti{
~
Program Reference 2).
?
RESOLUTION De ABWR design incorporates various compressed gas systems for plant operating applistions such as the hydrogen water chemistry (HWC) system and the main generator hydrogen system. Both of these systems are non-nuclear, non-safety-related and are required to be safe and reliable, consistent with the requirements for using hydrogen gas as described in Subsection 93.9, Ilydrogen Water Chemistry, and Subsection 10.2 Turbine-Generator.-
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- 1 Since the combustible gas systems are designed in compliance with SRP 9.5.1, so that their failure will not
- i jeopardize safety related equipment this issue is resolved for the ABWR.
REFERENCES j
1.
NUREG-0933,"A Prioritization of Generic Safety Issues" (and Supplements 1 12), July 1991.
2.
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -
1 LWR Edition."
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19B.2.48 118: TENDON ANCHOR AGE FAII UHE ISSUE Generic Safety Issue (GSI) 118 in NUREG-0933 (Reference 1), addresses the failure of lower vertical tendon anchor heads in a PWR prestn:ssed concrete containment structure. The failures appear to have been caused by hydrogen suess cracking. The hydrogen is liberated by zine in the presence of water. Quantities of water rangmg from a few ounces to about 1.5 gallons were found in the grease caps.
ACCEPTANCE CRITERIA For the ABWR Standard design, the primary containment structure consists of a reinforced concrete design.
Since the prestresse J concrete containment design is not used in the ABWR Standard design, the tendon anchorage failure issue is not applicable, therefore, no acceptance criteria are needed.
RESOLUTION For the ABWR Standard design, the primary containment structure is of a reinforced concrete design, therefore GSI 118 is not applicable.
+
REFERENCES 1.
NUREG-0933. "A Prioritization of Generic Safety Issues" (with supplements). U.S. NRC, April 1989.
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Y 19B.2.50 121 IIYDROGEN CONTROL FOR I ARGE. DRY PWR CONTAINMENTS t
ISSUE nis issue 121 concerns the control of hydrogen concentrations in large, dry PWR containments during and after a degraded core accident. In December 19&l, the staff reconunended that rulemaking with regard to this issue could be safety deferred due to the greater inherent capability of these containments to accommodate large quantities of hydrogen. Ongoing NRC experimental and analytical procrams are intended to provide data to support a final recommendation on whether safe shutdown equipment is likely to survive a hydrogen burn (Reference 1).
ACCEL'TANCE CRITERIA Re acceptance criteria for the resolution of issue 121 is that the control of hydrogenjenerated in the containment m a degraded core accident studi meet the requirements of 10CFR50.34(0 (Reference 2) on limiting the distributed hydrogen concentration to 10 percent, on limiting combustible concentrations, and on maintaining safe shutdown equipment and contamment integnty.
RESOLUTION This issue does not apply to BWRs and pressure suppression contair. ment. Also the ABWR primary containment is inened and is, therefore, protected from hydrogen combustion regardless of the amount or rate of hydrogen generation.
REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, July 1991.
National Archives and Records Admim,n Criteria for Nuclear Power Plants" Office of the Federal Register 10CFR50 Appendix A."GeneralDesig 2,
stration.
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19B.2.51 124: AUYll I ARY FEEDWATER SYSTEM REI I ABILITY ISSUE Issue 124 in NUREG-0933 (Reference 1), addresses Auxiliary Feedwater System reliability and availability and its impact on reducing core-melt frequency in PWRs.
M5ElrrANCE CRITERIA The acceptance criteria for the resolution of issue 124 is that the Auxiliary Feedwater System shall be designed for a high degree of reliability (i.e., using reliability analyses the system shall attain 0.0001 to 0.00001 unavailability per demand)
RESOLUTION This issue 124 is not applicable to BWRs and is therefore resolved for ABWR.
REFERENCES 1.
NUREG-0933. "A Prioritization of Generic Safety Issues" (w ith supplements). U.S. NRC, July 1991.
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19B.2.52 128: FI FCTRICAL POWER RFI I ABILITY ISSIJ.E Generic Safety Issue (GSI) 128 in NUREG-0933 (Reference 1), addresses the reliability of on-site electrical systems.
De minimum acceptable de power system is comprised of two physically independent divisions widch supply de power for control and actuation of redundant safety f the singic lailure criterion for assuring a reliable de po position of regulatory staff, including the application o supply. These concerns stem from me dependence on de power of the decay heat removal systems required for long-i-
tenn heat removal. Failure of one de division would generally result in a reactor scram which then would require removal of decay heat. The frequency of reported single de division failures gives rise to the concem that the second de division may not be available.
Two of the specific reasons for the concern that safety-related power may be unreliable are also addressed by l
this issue. One is that some operating nuclear power plants do not have technical specifications or adnunistrative controls governing operational restrictions for Class IE 120 Vac vital instrument buses and associated inverters.
Without such restnctions these power sources could be out of service indefinitely and thereby may place certain safety systems in a situation where they could not meet the single failure criterion. The other is that the design of some plants do not provide interlocks to prevent the inadvenent closure of the single tie breaker between the 4160 V Class IE buses.
ACCFFTANCE CRITERIA ne NRC perfonned a generic evaluation of me reliability of safety-related de power and published the resUlts in NUREG-0305 (Reference 2) and NUREG-0666 (Reference 3).
NUREG-0666 provided recommendations and supporting tecimical bases for augmenting the minimum design criteria and procedural requirements which will provide greater assurance of de power supply reliability, These recommendations for augmenting the minimum requirements for de power systems are: (I) prohibiting certain design and operation features of the de power systems, such as use of a bus t e breaker, which could compromise division independence; (2) augmenting the test and maintenance activities presently r:, quired for battery operability to also include preventive mamtenance on bus connections, procedures to demonstrate de power availabdity from the battery to the bus, and administrative controls to reduce the likelihood of battery damage during testing, he maintenance, and charging activities; (3) requiring staggered test and maintenance activities to minimize t potential for human error-related common cause tailure associated with these operations; and (4) requiring design and operation features adequate to maintain reactor core cooling in the hot standby condition followmg the loss of any other system required for shutdown cooling.
For plants not yet built, the NRC is considering further enhancing the reliability of the de power supplies by (1) placing non-safety-related loads on completely separate de power supplies (i.e., non-safety-related balance-of-plant and switchyard batteries), and (2) dividing the de power supplies which are safety-related or essential into separate systems to reduce the probability of a reactor trip m the event of the loss of a single de bus.
Also under consideration is NRC endorsement of IEEE Standards 603 (Reference 4) and 308 (Reference 5) with possible revisions to the related Regulatory Guides.
RESOLUTION ne resolution for GSI 128 as stated above, suggests elements which are applicable only to the design or the administative operation of operating plants, and are not applicable to the design of the ABW R. For the ABWR, the problems described in this issue are completely avoided by the fo!!owing inherent design features (which are described in detailin Section 83):
1.
The ABWR utilizes four completely independent Class IE de divisions which power two-out-of-four logic to actuate safety systems. If a division is taken out of service, the logic reverts to two-out-of-three.
Because of this level of redundant trip channels, no single power supply failure results in a reactor scram, even when a division is out of service.
2.
Dere are no bus tie breakers between divisions. Ilowever,it is possible, through special administrative controls and key interlocks, to manu:dly power one division's de hiads from a d fferent division through the spam charger (see Figum 8.3-4, and Subsection 83.4.18).
3.
All non-Class IE de loads are powered from non-Class IE de sources with only one exception. His spe_ ial c
case is the Alternate Rod Insertion (ARI) function utilizing the Fine-Motion Control Rod Drive (FMCRD) 66
motors. For ATWS considerations, the reliability of this subsystem is enhanced by using Class IE power for the drive motors. This p(ower mierface exists only on Division I, and is isolated by zone-selective interlocked circuit breakers see 8.3.1.1.1).
i 4.
nree of the four de divisions are backed by independent Class IE diesel generators. (The fourth division batterv charger is supplied power from Division 11, and hence, is backed by the Division 11 diesel.) The non-dass IE plant investment protection (PIP) loads are backed by an on-site combustion turbine generator (CTG).
i There are two separate and independent connections from the off-site sources to each of the three Class IE buses, and to each of the three PIP buses.
6.
De ABWR fully complies with IEEEs 308 and 603.
In summary, the ABWR design for the electrical power system avoids the problems described in this issue.
Each division of the engineered safety systems has emergency on-site sources of ac and de power, and at least two connections for off-site power, all of which are separate and mdependent. There are three divisions of decay heat removal, each with its own emergency ac and de power source. This issue is considered resolved for the ABWR.
REFERENCES t
1.
NUREG-0933,"A Prioritization of Generic Safety issues"(with supplements) U.S. NRC, April 1989, 2.
NUREG-0305, ' Technical Report on DC Power Supplies in Nuclear Power Plants ** U.S. NRC, July 1977.
3.
NUREG-06% "A Probabilistic Safety Analysis of DC Power Supply Requirements for Nuclear Power Plants "
U.S. NRC, April 1981.
4.
IEEE Standards 603-1980. " Standard Criteria for Safety Systems for Nuclear Power Generating Stations," The Inshtute of Electrical and Electronics Engineers, Inc.
5.
IEEE Standard 308-1980. " Criteria for Class IE Electric Systems for Nuclear Power Generating Stations," He i
institute of Electrical and Electronic Engineers, Inc.
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1911.2.53 14h I E AKAGE TilROUGli El ECTRICAI ISOI ATORS IN INSTRUMENT CIRCUITS ISSUE Electronic isolators are used to maintain electrical separation between safety and non-safetv-related electrical systems in nuclear power plants, preventing malfunctions in the non-safety systems from degra' ding performance of safety-related circuits. Isolators are primanly used where signals from Class-lE safety-mlated systems are trans<nitted to non-Class IE control or display equipment.
'Ihere are a number of devices which may qualify as electrical isolators in a nuclear power plant including fiber optic and photo-electric couplers, transfonner-modulated isolators, current transfonners, amplifiers, circuit breakers, and relays. These isolators are designed and tested to pmvent the maximum credible fault applied in the transverse mode on the non-Class IE side of the isolator from degrading the performance of the safety-related circuits (Class-IE side) below an acceptable level.
This issue was identified by the staff in June 1987 and arose from observations made during SPDS evaluation tests that, for electrical transients below the maximum cmdible level, a relatively high level of noise could pass through cenaan types of isolation devices and be transmitted to safety-related circuitry. In some cases, the amount of energy that can pass through the isolator may be sufficient to damage or seriously degmde the performance of Class IE components, while, in other cases, electncally-generated noise on the circuit may cause the ssolation device to give a false output.
Due to the fact that there are a great number of each type of isolator in the field, this issue would require the staff to determine the extent to which potentially susceptible isolators are used in nuclear power plants and to identify the systems in which they are used. An NRC bulletin to all licensees to provide input on these questions would be necessary.
ACCElrrANCE CRITERI A Assuming that the staff detennines from the licensee responses to the proposed bulletin that a potential problem exists, a research program consisting of two major objectives would have to be initiated to develop the soluuon to this issue. The first objective would be to develop test procedures and acceptance criteria for isolators that licensees could use to determine the adequacy of installed isolators. The second objective would involve development of appropriate hardware fixes that could resolve the issue.
Therefore, with a reliable data base the final ste in the solution to this issue would be the issuance of a generic letter to licensees with the following guidelines for: p(1) inspection and testing of all electrical isolation devices between Class IE and non-Class 1E systems; (2) repair / replacement of isolators that fail the tests, including description of acceptable hardware fixes to the isolators; and (3) implementation of an annual program to inspect and test all electronic isolators between Class IE and non-Class IE systems.
RESOLUTION Fiber optic data links are the only type of isolation device used for electrical isolation oflogic level and analog signals between protection divisions and from protection divisions to non-safety-telated equipment.
Maximum credible electrical faults aPf ed at the outputs of isolation devices do not appiv to fiber optic li systems. The maximum credible fault is cable breakage causing loss of signal transmission. Paults cannot cause propagation of electrical voltages and currents into other electrical circuitry at the transmitting or receiving ends.
Conversely, electrical faults onginating at the input to the liber optic tmnsmitter can only damage the local circuitry and cause loss or corruption of data transmission; damaging voltages and currents will not propagate to the receiving end.
Fiber optic isolation devices are expected to have less difficulty than previous isolation devices in complying with all qualification requirements due to their small size, low mass, and simple electronic interfaces. The basic.
materials and components, except for the fiber optic cause itself, are the same as those used in existing, qualified isolation devices.
When using fiber optic devices as Class IE isolation devices, only the input side of the transmitting device and out side of the receiving device use electncal power. The low voltace power supplies for these devices use the same power source as the logic that drives the isolatmg device. For ABWR safety systems, this power is:
1.
Divisional 120 Volt Vital AC (UPS)- For Reactor Protection System (RPS) logic and Main Steam Isolation Valve (MSIV) logic.
2.
125 Volt Plant DC Power Supply - For ECCS logic and Leak Detection and Isolation System (LDS) logic.
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r 19B.2.54 14h AVAII ABil ITY OF CHII LED WATER SYSTEMS AND ROOM COOLING ISSUE in recent years, several nuclear power plants have experienced problems wid1 safety system components and control systems that were caused by a partial or total loss of heating, ventilating, and air conditioning (HVAC) systems. Many of these problems eust because of the desire to provide increased fire protection and the need to avoid severe temperature changes in equipment control circuits. Since the Browns Ferry fire, considerable effort has been expended to improve the fire protection of equipment reguired for safe shutdown. Generally, this improvement has been made by enclosing the affected equipment m small. asolated rooms. The result has been a significant increase in the impact of the loss of room cooling Plant control and safety have improved with the introduction of electronic integrated circuits; however, these circuits are more susceptible to damage from severe changes in temperature caused by the loss of room cooling.
It is believed that failures of air cooling systems for areas housing key components, such as residual heat removal pumps, switchgear, and diesel generators, could conuibute sigmficandy to core-melt probability in certain plants. Because correcuve measures are often taken at the affected plants once such failures occur, the tmpact of these failures on the proper functioning of air cooling systems has not been considered. Thus, plants with similar inherent deficiencies may not be aware of these problems.
Operability of some safety-related components is dependent unon operation of IIVAC and chilled water systems to remove heat from the rooms containing the components. If chilled water and HVAC systems are unavailable to remove heat, the ability of the equipment within the rooms to operate as intended cannot be assured.
ACCEI'TANCILCPITERI A A possible resolution to this issue would require a reevaluation of exh plant's room heat load and heat-up rate in order to identify areas in which a reduction in the dependence of equipment operability on HVAC and room cooling may be implemented. While the total elimination of this dependence may not be postible at all plants, this analysts would identify areas in which this dependence is critical. After the critical dependencies are identified, each plant would implement procedural changes (to provide alternate cooling) to climinate or reduce the dependencies where possible. Hardware modifications may be needed for situations m which a procedure change cannot be implemented to reduce a critical dependency.
He next step in the possible solution to this issue would be the issuance of a generic letter that would require cooling;(2) identify areas in which this dependence is cr.y systems and equipment operability on H licensees to: (1) evaluate the dependencies of plant safe <
tical; (3) identily appropnate procedure changes and hardware modificauons to minimize the etTects of the dependencies on pfant nsk; (4) submit this evaluation to the NRC for review and approval of the proposed modifications; and (5) implement the approved proposed procedural changes and hardware modifications. The generic letter would include guidance on acceptable procedures licensees could use to evaluate the potential dependencies in the designs of these systems. The generic letter would also -
include alternative solutions for improving the independence of systems that are critical to plant risk. It is assumed that a research project would form the basis for a more fully-developed solution and for the guidance in the generic letter.
RESOLUTION Re safety related equipment areas housing key components such as residual heat removal pumps, switchgear, and diesel generators shall be p(rovided with calibrated pressure and temperature monitors which can be c site. Pressure and temperature ambient and ditferential temperature) along with fiow requirements can be used to monitor and diagnose the applicable equipments perfonnances. This implementation will assure the total control of loss of HVAC systems, and protects the systems against fire. Herefore, this issue is resolved for the ABWR (Reference 1).
REFERENCES 1.
Advanced Light Water Reactor Utility Requirement Document (Volume 11), EPRI.
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19B.2.56 151: RFI I ABII ITY OF ANTICIPATED TRANSIENT WIT 110DT SCRAM RECIRCUI ATION PUMP TRIP IN BWRs ISS11 Generic Safety issue (GSI) 151 in NUREG-0933 (Reference 1), addresses the issue of the reliability of the ATWS RI'T in BWRs. GSI-151 specifically identifies a reliabilitv problem with GE's type AKF-25 circuit breaker and trip hardware. (actually a type AKF-2-25 breaker, per NRC's'IE Notice 87-12, Reference 2).
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI-151 is the use of reactor recirculation system pump trip hardware or method that is more reliable then the previously used AKF-2-25 breaker hardware or method.
RESOLUTION 7he design for the ABWR reactor recirculation system and RIYT method and hardware is completely different then the previously designed BWR reactor recirculation systems and RITT trip methods. The design is more diverse and redundantly reliable. Rather than using only two recirculation pumps and the associated single RPT breakers, the ABWR will use ten pumps and multiple pump and RPT tnn logic. circuits and hardware. Adjustable speed drive (ASD), recirculation incore mternal pumps (RIPS) are used. The ABWR R17f trip hardware (not yet specifically identified) will be completely different: Instead of using AKF-2-25 breaker switching hardware to provide a RPT, 7.7.1.3(7) and 7.7.1.3(8)g and ASD gate inverter turn-off circuit hardware provides the RIrf. See Subsect RFC controller switchin REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues"(with supplements) U.S. NRC.
2.
IE Information Notice 87-12," Potential Problems with Metal Clad Circuit Breakers, General Electric Type AKF-2-25", U.S. NRC, February 13,1987.
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19B,2.57 153: LOSS OF ESSENTI AL SERVICE WATER IN I IGliT-W ATER REACTQRS ISSI'E This issue addresses the potential unavailability of the essential service water (ESW) system for all LWRs except those seven multiplant sites addressed under issue 130. De ESW system at a nuclear power plant supplies related and non-safety-related systems and equipment to the coolmg water to transfer neat from various safetyferent names at vanous types of plants. The design and o ultimate heat sink of the plant. It is known by dif cturacteristics of the ESW system are different for PWRs and BWRs. In addition, these charactenstics may differ significandy in each of these reactor types.
Under issue 153, the staff will examine all potential causes for ESW system unavailability, except those that are considered to be resolved by implementing the resolutions addressed in GL 89-13 (Reference 1), such as biofouling, sediment, corrosion, and crusion (Issue 51). The safety concerns of this i-sue include partial or complete loss of ESW system functions resulting from common causes (such as icing of the intake structure), depradation of the ESW system, design deficicacies, and procedural or maintenance errors. A complete loss of the ESW system could lead to a core-melt accident, posing a signincant risk to the public.
De NRC evaluation of this issue has not yet been completed.
ACCEPTANCE CRITERI A De ESW system is needed in every phase of plant operations and, under accident conditions, supplies adequate cooling water to systems and components that are important to safe shutdown or to mitigate the consequences of the accident. Under normal operating condition, the ESW system provides component and room cooling (mainly via the component cooling _ water system). During shutdown it also ensures that the residual heat is removed from the reactor core. The LSW system may also supply makeup water to Gre protection systems, cooling towers, and treatment systems at a plant.
The design features for the safety service water (SSW) system are summarized as follows:
1.
Performance Requirements
- De SSW system will be designed to meet the required heat loads.
- The SSW system will be provided with two pumps and two heat exchangers per division.
- De plant designer will provide analyses for all potential operating conditions that properly account for uncertamues.
2.
System Arrangement
- The SSW system will be divided into approximately equal-sized divisions, three for the BWR and ABWR.
- A division will be made up of independent piping svstems, each with pumps, heat exchangers, strainers, controls and instrumentanon, power suppfies, and associated equipment required for regulating system flow.
address the safety concerns of Issue 153.ystem arrangement for the SSW system indicated above De performance requirements and s resulting from common causes, degradation of the SSW system, design deficiencies, and procedural or maintenance errors. The plant designer should provide an assessment of these potential failure modes and their associated contributions to the core damage frequency and should identify dominant accident sequences.
on the NSSS.gn of the ESW system varies substantiady from plant to plant and the ESW system He desi solutions are: (1) installation of a redundant intake structure including a service water pump; (2) hardware changes of the ESW system: (3) installation of a dedicated RCP seal cooling system: or (4) changes to TS or operational procedures.
RESOLUTION Re ABWR Reactor Service Water (RSW) system removes heat from the Reactor Building Cooling Water (FCW) system and transfers that heat to the Ultimate llcal Sink (UllS). The RSW system is provided in three 72
i divisions. Each division has two pumps which send cooling water to three RCW heat exchangers. Normally one pump and two heat exchangers are operating in each division. When heat removal requirements increase, the remaining pump and heat exchange are automatically put into operadon. If additional heat removal capacity is needed, some of the non-safety-related cooling loads may be taken out of operation.
shutdown requirements. (Subsection 9..g of the three RSW divisions, the other two divisions meet pla In case of failure which disables an
.11.)
The ABWR RSW system divisions are physically and electrically separated from each other. This reduces the potential effects of common causes. Nonnally, each division is operaung at all times with the capability (to put into service the remaining pump and heat exch er at any time. Margm is provided in pump flow capacity and in action taken when needed. ' Subsections 9.11.4 and 9.2.15.g of these components willbe performed and correc RCW beat exchanger heat removal capacit. Periothe testin 1.4.)
Several potential causes of RSW system degradation are site dependent. He RSW system is designed to orevent this degradation from occurring. Additionally, the COL applicant will provide the following system design l'eatures for those portions of the system which are not the ABWR standard plant scope: adequate NPSll for the pumps at low UllS water levels, low point drains and high point vents, prevention of organic fouling (using methods such as trash racks, biocide treaunent or thennal backwashmg, or required), component material selection suited to steam impingement, pipe whip, jet forces, missiles, site water conditions, and protection against flooding, spraying,ipment. If required, recirculation of warm water fire and the effect of failure of any non-Seismic Category I equ through the intake structures will be provided to reduce the likelihood that ice will block cooling water flow.
(Subsections 9.2.5.4 and 9.2.15.2.)
The RSW pumps and pump house will be designed by the COL applicant, who will consider and reduce the cffects of procedural and mamtenance errors.
When the future plant-specific design is prepared, another assessment will be made of potential failure modes and their associated contribuuons to the core damage frequency and the dominant accident sequences will be identified.
Rese issues are resolved for the ABWR through the design features of the RSW system and the system design features which will be provided by the COL applicant.
REFERENCES i
1.
GL89-13, Service Water System Problems Affecting Safety Related Equipment, July 18,1989.
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Advanced Light Water Reactor Utility Requirement Document (Volume II) EPRI.
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19B.2.58 155.1 MORE REAI ISTIC SOURCE TERM ASSUMPTIONS ISSUE Cunent siting regulations (10 CFR Part 100) require that an accidental fission product release from the core into containment be assumed and that its of fsite radiological consequences be evaluated against guideline doses given in Pan 100. The postulated source term is derived from TID-14844 (Reference 1) and is contamed in Regulatory Guides 1.3 and 1.4. The regulatory guides specify a release into containment of 100 percent of the core inventory of noble gases and 50 percent of the fodine fission puxlucts. lialf of the iodine is assumed to deposit on intenor surfaces assuming mstantaneous appearance within contairunent and that the iodine is predominately in elemental form (1 ).
2 Use of the TID-14844 source tenn has not been restricted to evaluation of plant midgation features and site suitability. Regulatory applications of the source tenn are broad, including use as the basis for (a) the post-accident environment for which safety-related equipment should be qualified (b) post-accident habitability requirements for the control room, and (c) post-accident sampling systems and accessibility.
A substantial amount of infonnation has been developed to update knowledge about LWR severe accidents and behavior of fission products that could be released into containment. Studies have confumed that, although the TID-14844 source term is substantial and diat its use has resulted in a high level of plant capability, the present recipe can be substantially improved.
In their staff reguirements memorandum (SRM) dated January 25,1991, the Conunission approved the tan proposed by the staf f to revise Part 100 to delete the source tenn and dose calculations and to directly specif site enteria: to issue (in parallel) an interim revision to Part 50 to retain the present source term and dose calcula on (but not for siting purposes); to update the TID-14844 source tenn; and, in a second-rule making phase, to incorporate severe accident and revised source tenn insights for future plants. In their SRM dated Apnl 11,1991, the Commission requested the staff to make recommendations on the values of irleases into containment (to update TID-14844), to provide a discussion of the status of EPRI's comparable values, and to discuss the use of the updated source term in evalua' ions of existing and future plants.
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ACCEI'FANCE CRITERI A l
The acceptance criteria for GSI 155.1 is that the plant shall be designed to ensure that the dose commitment to the public in the event of a licensing design basis accident shall be withm those limits prescribed by existing regulations based upon the limitations of 10CFR100.
RESOLUTION The ABWR is currently licensed to and analyzed to the existing Regulatory Guides, Standard Review Plans. and General Design Criteria which are based upon TfD-14844. 'Ihe use of revised source tenns based upon NUREG-1465 (Reference 2)is premature for the ADWR based upon the lack of clarification of what is a design basis event under the revised source tenns and lacking adequate guidance from the Commission as to acceptable methods and -
conditions, i.e., revised regulatory guides and standard review plans.
REFERENCES l.
DiNunno, JJ. et al. " Calculation of Distance Factors for Power and Test Reactor Sites", Technical Information Document 14844, March 23,1962.
2.
Soffer, L. et al, " Accident Source Terms for Light-Water Nuclear Power Plants", NUREG-1465 USNRC, Draft Report for Comment, June 1992.
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4 19B.2.59 A 17: SYSTE5fS INTER ACTION ISSUE Issue A-17 in NUREG-0933 (Reference 1) nddresses the concern that unintended or unrecognized dependencies may exist between systems that could lead to safety-significant events.
Nuclear plant design includes interdisciplinary reviews to assure the functional compatibility of the plant structures, systems and components and compliance with licensing requirements. Safety reviews and accident analyses provide further assurance that system functional and licensing requirements will be met. Thus the d and analysis of the plant take into account systems interactions, Nevertheless, the process may not consider all tne interactions of various plant systems. Based on this possibility, adverse systems interaction is defined as actions or consequences in one system tnat could adversely affect die redundance or independence of safety systems in another system or systems. Tlie issue involves preventing any adverse systems interactions that affect plant functions or regulatory requirements.
ACCEPTANCE CRITERIA The acceptance criteria for the resolution ofissue A-17 is that attention shall be given in the detailed plant design to detecting and minimizing the potential for adverse system interactions using the guidance of NLREG-1174 (Reference 2).
RESOI_UTION To respond to issue A-17 the detailed design of ABWR shall provide the following elements:
Design requirements that ensure separation and isolation of electrical power systems to preclude interactions that could adversely safety-related power as described in Chapter 8, Electric Power.
Design requirements that ensure spatial separation of systems and equipment to prevent interaction between redundant safety, grade equipment and systems or adverse interaction of non. afetygrade equipment with safety grade equipment. Considemtion shall include seismically coupled and floodmg spatial mteractions.
For example refer to Subsection 6.3.1, ECCS - Design Bases and Summary Descripuon.
Plant pmbabilistic evaluations to detect potential systems interactions (e.g., to ensure that redundant safety grade systems and equipment are not installed in the same room or fire area and to assess the im i
high energy line breaks and flooding) as described in Appendix 19D, Probabilistic Evaluations. pacts of REEERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues", July,1991 (and Supplements 1 12).
2.
NUREG-1174, " Evaluation of Systems Interactions in Nuclear Power Plants." May 1989.
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19B.2.60 A-29: NUCI EAR POWER PL ANT DESIGN FOR THE REDUCTION OF VULNFR ABILITY TO INDUSTRIAL SABOTAGE i
ISSUE j
Issue A-29 in NUREG-0933 (Reference 1), addresses the susceptibility of nuclear power plants to industrial sabotage, the resulting risk to plant safety, and the countermeasures to assure an acceptable level of protection.
Consideration should be given to sabotage during the design phase of theplant. De goal would be to achieve.
an acceptable level of protection of a plant to industnal sabotage b{emphasizmg design features which reduce the '
likelihood of the plant incurring damage from industrial sabotage, ith mternal and external.
ACCEFTANCE CRITERIA l
The acceptance criteria for the resolution of issue A-29, is that plants shall be designed to be msistant to the effects of internal and external sabotage through pmvention, deterrence and mitigation.
1 Specifically, plant safety-related systems and components required for the safe operation and shutdown of the l
plant shall be designed for protection against and mitigation of sabotage.
RESOLUTION I
ne ABWR design will mitigate the acts of sabotage through physical separations in the plant arrangementof independent, engineered safety systems, and the desien and location of barriers to resist threats. Refer to Section 9.5, Fire Protecuan, Section 3.4,Ploods and Section 3.6, Pipe Whip Protection.
Appendix 19C, Design Consideration Reducing Sabotage Risk, describes and analyzes the ABWR design I
features that reduce the nsk from postulated insider sabotage.
In addition, the ABWR design includes various methods of access control to pmvent intrusion as well as provide detection during a breach of the system. Specifically,10CFR73.55 (Reference 2). Subsection 13.6.3, Physical Sec' protection systems and controls for compliance with against torngo missiles, wmds, earthquakes and oods,The des n of the decay heat removal system rovide e
In summary, the ABWR design is highly resistant to sabotage, because of the feature described which protect against internal and external sabotage. Therefcre, this issue is resolved for the ABWR.
l REFERENCES 1.
NUREG-0933, "A Prioritization of Generic Safety Issues" (and Supplements 1-12).
2.
10CFR73.55, " Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage, Office of the Federal Register, National Archives Records Administration.
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198.2.61 B-05: DUCTII ITY OF TWO WAY SL ABS AND SilEl I S AND BUCKI ING BEIIAVIOR OF STFFI. CONTAINM ENTS 1
ISSUE Generic Safety Issue (GSI) B-05 in NUREG-0933 (Reference 1), identifies two concems relating to containment design. First that sufficient information is not available to predict the behavior of two-way reinforced concrete slabs; and second, that the structural design of a steel containment vessel subjected to unsymmetncal dynamic loadings may be governed by the instability of the shell.
(1) Ductility ot Two-Way Slabs and Shells he first concern was originally identified in NUREG-N71 (Reference 2) and involved concern over the lack of information related to the behavior of two-way reinforced concrete slabs loaded dynamically in biaxial membrane tension (resulting from in-plane loads), flexure, and shear. If structures (concrete slabs) were to fail (floor collap(llELB), there would be a possibihty that other portions of the reactor coolantse high-energy-line break system or safety-related systems could be damaged. Such loads would be caused by very concentrated ingh-energy sources causmg direct impact on the structures of concern. The damage could lead to an accident sequence resulting in the release of radioactivity to the envirotunent.
,1 Because of NRC and industry concem, the American Concrete Institute addressed these dynamic loads by establishing the methodology identified in the Appendix C Commentary to ACI 349-85 (Reference 3).
(2) Buckling Behavior of Steel Containments The second concern, also identified in Reference 2, involves concern over the lack of a uniform, well-defined approach for design evaluation of steel containments. The structural desi n of a steel containment vessel subjected to unsymmetrical dynamic pressure loadings may be governed theinstabilityof the shell. For this type of loading, the cum:nt design verification metinh, analvtica techniques, and the acceptance critena may not be as comprehensive as they could be. Section ill of the ASME Code (Reference 4) does not provide detailed guidance on the treatment of buckling of steel containment vessels for such loading condiuons.
Moreover, this Code does not address the asymmetrical nature of the containment shell due to the presence of equipment hatch openings and other penetrations. Regulatory Guide 1.57 recommends a minunum analysis, considenng in gainst buckling for the worst loading condition provided a detailed rigorou factor of safety of two a clastic behavior,is perfonned.
On the other hand, the 1977 Surraner Addendum of the ASME Code permits three alternate methods, but requires a factor of safety between 2 and 3 against buckling, depending upon applicable service limits.
ACCEIFTANCE CRITERIA The acceptance criterion for the first concem is that analysis methods used for two-way reinforced concrete slabs adequately address dynamic loading in biaxial membrane tension, flexure, and shear that occur due to a llELB or LOCA.
He acceptance criterion for the second concem is that all applied loads must be adequately addressed by the steel containment vessel design.
RESOLUTION Since the ABWR containment design is based upon ACI 349-85. which establishes methods by which the above loading conditions and the latest codes for the first concern of this issue are addressed, and the steel containment design meets the requirements of the ASME Code for te second concern of this issue, both concerns are fully resolved for the ABWR design, Reference 5.
REFERENCES 1.
NUREG-0933. "A Status Report on Unresolved Safety Issues". U.S. NRC December 1989.
2.
NUREG-N71, " Generic Task Problem Descriptions (Categories B, C, and D)". U.S. NRC, June 1978.
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l 3.
ACI 349 85," Code Requirements for Nuclear Safety Related Structures", American Concrete Institute,1985.
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ASME Boiler and Pressure Vessel Code,Section III, Division 1. Subsection NE, American Society of Mechanical Engineers,1986.
5.
Regulatory Guide 1.142," Safety Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments) U.S. NRC, October 1981, Revision 1.
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'i 19R,2.62 029: BOLTING DEGRADATION OR FAILURE IN NUCI E AR POWER PLANTS I
ISSUE j
issue 029 in NUREG-0933 (Reference 1), addresses botting d gradation within safety-related components and l
suppon structures and its impact on the integrity of the reactor con ant pressure toundary i
ne most crucialIniting applications and those constituting an integral part of the primary pressure boundary such as closure studs and bolts on reactor vessels and reactor coolant pumps. Degradauon of these bolts or studs
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could result in the loss of teactor coolant. Other botting applications such as component suppon and embedment anchor bolts or studs are essential for withstanding transient loads created during abnonnal or accident conditions.
ACCEPTANCE CRITERI A
'I The acceptance criteria for the resolution of issue 029 are that proven boldngfesigns, materials, and fabrication techniques shall be employed, in addition, reactor coolant pn:ssure boundary (RCPB) bolting shall meet the requirements of ASME Code, Section 111 (Reference 2). Also, for RCPB bolting the owner-operator shall use i
established industry practice in developing maintenance, assembly, and disassembly procedures. Furthermore, for RCPB and its support bolting, inservice mspection shall meet the requirements of ASME,Section XI (Reference 2).
li RESOLUTION resulted from poor maintenance practices. primarily an operating plant issue since most of the de Botting degradation of RCPB tolts is
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Bolting integnty is assured by the designer through the initial specification o,f proven botting materials and installation requirements, and by the owner-operator through the use of -
acceptable mamtenance and mspection practices.
For the ABWR design, only proven materials for the specific application and environment are employed, having
-i been selected after evaluation of the po,tential for corrosion wastage and intergranular stress corrosion cracking.
are fabricated, tested,ponents and their integral bolts, including the reactor vessel, reactor coolant Also, the RCPB com and installed in accordance with ASME Code, Sections 111 and XI. Finally, the owner operator must perform oeriodic inservice inspection in accordance with ASME Code Section XI. In addition, for critical.
pressure boundary applications such as the reactor vessel head closure, redundant seals and leak monitoring further.
assure the integrity of the RCPB. Rerefore, this issue is resolved for the ABWR Standard Design.-
4 REFERENCES 1.
NUREG-0933, "A Status Report on Unresolved Safety Issues", U.S. NRC, April 1989.
2.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III (Nuclear) and Section XI American Society of Mechanical Engineers.
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19B.2.63 82: BEYOND DESIGN B ASIS ACCIDENTS IN SPENT FUEL POOI S ISSUE Issue 82 in NUREG-0933 (Reference 1), addresses the potential for a beyond-design-basis accident in which the water is drained out of the spent fuel pool. In such an event the discharged fuel from the last two refuelings may have sufficient decay heat to melt, igmte the zircaloy cladding and release fission products to the atmosphere.
ACCElrrANCE CRITERI A The acceptance criteria for the resolution ofissue 82 is that the design of the spent fuel pool, storage racks, fuel pool cooling and cleanup system and the kiad handling equipment in the spent fuel pool area shall meet applicable current requirements, i.e., the guidance of the Standard Review plan (SRP) Sections 9.1.2 - 9.1.5 (Reference 2) and Regulatory Guide 1.13 (Reference 3).
RESOLUTION The ABWR design includes a spent fuel storage facilitv a fuel pml coolin handling system that meets the intent of Regulatory Guide 1.,13 and SRP 9.1.2 g and cleanup syste 9.1.5 as described in Section 9.1, Fuel Storage and llandling. Since the acceptance criteria are met for the spent fuel storage facility, this issue is resolved for the ABWR REFERENCES 1.
NUREG-0933. "A Prioritization of Generic Safety Issues,"(and Supplements 1 12), July 1991.
2.
NUREG-0800. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition."
3.
Regulatory Guide 1.13. "Desien Objective for Light. Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations," Revision 2. December 1981.
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ABWR n^uoors Standard Plant Rev A 19B3 COL LICENSE INFORMATION B.3.8 Interdisciplinary Design Reviews ms2RT I
j9ggj 19BJ.1 Quality Assurance Program COL applicants referencing the ABWR design shall establish an interdisciplinary design review COL applicants referencing the ABWR design group and direct reviews for site specific design and shall have a Quality Assurane "rogram satisfying construction work as required by Subsection the requirements of E'esection 19B.2.1(2) including 19B.2.25(4).
the right to impore additional quality assurance requirements.
19B.3.9 Sabotage Vulnerability During Plant Shutdown 19B.3.2 Prevent on of Core Damage i
The sabotage vulnerability analysis required by COL applicants ieferencing the ABWR design Subsection 19B.2.4(10) has been performed for the shall approve applienble design deviations in ABWR and is contained in Appendix 19C. However, divisional total indept ndence and separation both applicants referencing the ABWR design shallinclude mechanically and electrically as required by provision in the plant start.up procedures to inspect Subsections 19B.2.3(3),19B.2.4(2),19B.2.5(3), and critical safety equipment within the containment for 19B.2.11(2).
possible tampering just prior to scaling the containment in preparation for start up. Such 19B.3.3 Protection from External Threats equipment includes the ADS /SRV valves and associated accumulators and their charging lines and COL applicants referencing the ABWR design the inboard valves associated with the emergency core shall evaluate listed man-made hazards except cooling systems (i.e., HPCF, RHR and RCIC). '
sabotage on a site unique basis as required by Subsection 19B.2A(6).
19B.3.10 Impact of Security System on Plant Operation, Testing and Maintenance 19B.3.4 Ultimate Heat Sink Models In the design of the security system, applicants COL applicants referencing the ABWR design referencing the ABWR design shallinclude an shallimplement the development of predictive evaluation of its impact on plant operation, testing analytical models as required by Subsection and maintenance. This evaluation shall be conducted 19B.2.9(1) through 19B.2.9(4).
as required by Subsections 19B.2.4(12) and 9.5.13.11.
This analysis should include consideration of an 19B3.5 Ultimate Heat Sink Reliability emergency requiring evacuation of the control room in the control building to the remote shutdown panci COL applicants referencing the ABWR design in the reactor building.
shall have an ultimate heat sink design goal for the service water flow as required by Subsection 19BJ.11 Security Plan Compatibility with 19B.2.10(1) tbrough 19B.2.10(13).
ALWR Requirements
, 19B.3.6 Main Transformer Design The ABWR security plan will comply with the
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ALWR requirements as defined in 19B.2.4. Future COL applicants referencing the ABWR design amendments of the ALWR Requirements Document shall provide main transformer fire protection as must be reviewed for ABWR compliance by the required by Subsection 19B.2.18.
applicants referencing the ABWR Standard Plant.
19B3.7 Plant Siting 1983.12 Plant Security Systems Electrical Requirements COL applicants referencing the ABWR design shall approve in writing the listed final design COL applicants will provide non-Class IE vital parameters to be used at the plant site as required by (uninterruptible) ac power for the site security on 19B.2.19(3).
system. [See Subsection 19B.2.4(20)).
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iment 19 e.7 1 I 9 rt al col AfrHeariffsfety ILrves re)oIv/ ions & Ne 19B M The CM aff ocast Sha// pmic'e l
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J2-l I19B.3.13 Bolting Degradation or Failure -
f COL applicants shall provide the bolting information detailed in Subsection 19B.2.12(6).
198.3.14 Outside Sabotage i
COL applicants shall provide sufficient analyses to ensure that the plant is adequately protected from acts of outsider sabotage.
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- 19D.3-2 Amendment 25
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ABWR 2346 ooxs Standard Plant Rev A 19B3 COL LICENSE INFORMATION[B.3.8 Interdisciplinary Design Reviews nistR T T
Iggyi 19B.3.1 Quality Assurance Program COL applicants referencing the ABWR design shall establish an interdisciplinary design review COL applicants referencing the ABWR design group and direct reviews for site specific design and 1
shall have a Qu:lity Assurance Program satisfying construction work as required by Subsection the requirements of Subsection 19B.2.1(2) including 19B.2,25(4).
the right to impose additional quality assurance requirements.
1983.9 Sabotage Vulnerability During Plant Shutdown 19B.3.2 Prevention of Core Damage The sabotage vulnerability analysis required by COL applicants referencing the ABWR design Subsection 19B.2.4(10) has been performed for the shall approve applicable design deviations in ABWR and is contained in Appendix 19C. However, divisional totalindependence and separation both applicants referencing the ABWR design shallinclude mechanically and electrically as required by provision in the plant start-up procedures to inspect i
Subsections 19B.2.3(3),19B.2.4(2),19B.2.5(3), and critical safety equipment within the containment for 19B.2.11(2).
possible tampering just prior to sealing the containment in preparation for start-up. Such 19B.33 Protection from ExternalThreats equipment includes the ADS /SRV valves and associated accumulators and their charging lines and COL applicants referencing the ABWR design the inboard valves associated with the emergency core shall evaluate listed man-made hazards except cooling systems (i.e., HPCF, RHR and RCIC). '
sabotage on a site unique basis as required by Subsection 19B.2.4(6).
19B.3.10 Impact of Security System on Plant Operation, Testing and Maintenance 19B.3.4 Ultimate Heat Sink Models in the design of the security system, applicants COL applicants referencing the ABWR design referencing the ABWR design shallinclude an shallimplement the development of predictive evaluation of its impact on plant operation, testing analytical models as required by Subsection and maintenance. This evaluation shall be conducted i
19B.2.9(1) through 19B.2.9(4).
as required by Subsections 19B.2.4(12) and 9.5.13.11.
This analysis should include consideration of an 19B.3.5 Ultimate Heat Sink Reliability emergency requiring evacuation of the control room in the control building to the remote shutdown panel COL applicants referencing the ABWR design in the reactor building.
i shall have an ultimate heat sink design goal for the service water flow as required by Subsection 19B.3.11 Security Plan Compatibility with 19B.2.10(1) through 19B.2.10(13).
ALWR Requirements 19B.3.6 Main Transformer Design The ABWR security plan will comply with the ALWR requirements as defined in 19B.2.4. Future COL applicants referencing the ABWR design amendments of the ALWR Requirements Document shall provide main transformer fire protection as must be reviewed for ABWR compliance by the required by Subsection 19B.2.18.
applicants referencing the ABWR Standard Plant.
19B.3.7 Plant Siting 19B3.12 Plant Security Systems Electrical Requirements COL applicants referencing the ABWR design shall approve in writing the listed final design COL applicants will provide non-Class IE vital parameters to be used at the plant site as required by (uninterruptible) ac power for the site security on 19B.2.19(3).
system. [See Subsection 19B.2.4(20)].
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inunt ty 11 i9a.3./ COL AppthairfSglety'ZLive) rej0la N w.r & Ne 19B 3-1 licar f Sha// pn.dde The (01 aff r
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ABWR 23asicoas Standard Plant sev A F
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i I19B.3.13 Bolting Degradation or Failure COL applicants shall provide the bolting
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information detailed in Subsection 19B.2.12(6).
l 19B.3.14 Outside Sabotage l
COL applicants shall provide sufficient analyses to ensure that the plant is adequately protected from Qcts of outsider sabotage.
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