ML20044D739

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Forwards Listed Info to Support Accelerated Advanced BWR Review Schedule for USIs & Gsis,Including GI A-1,A-10 & A-17,based on 930506 Telcon
ML20044D739
Person / Time
Site: 05200001
Issue date: 05/18/1993
From: Fox J
GENERAL ELECTRIC CO.
To: Poslusny C
Office of Nuclear Reactor Regulation
References
REF-GTECI-A-01, REF-GTECI-A-10, REF-GTECI-A-17, REF-GTECI-PI, REF-GTECI-RV, REF-GTECI-SY, TASK-A-01, TASK-A-1, TASK-A-10, TASK-A-17, TASK-OR NUDOCS 9305200211
Download: ML20044D739 (22)
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GE Nuclzar Energy-

. t Genem! Ekee Com;;any 175 Cute kenue, Sun Jose. CA 95T2b ,

May 18,1993 Docket No. STN 52-001 t

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Chet Poslusny, Senior Project Manager l Standardization Project Directorate 'l Associate Directorate for Advanced Reactors and License Renewal .  !

Office of the Nuclear Reactor Regulation .I i

Subject:

Submittal Supporting Accelerated ABWR Review Schedule - USIs and GSIs ]

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Dear Chet:

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Enclosed are the following: 1

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1. Replacement
  • for Safety Issues Inder page 19B.1-4 (New information -- listing TMI Issues)-
2. New pages 19B.1-5 through 19B.1-7 (Balance of TMI Issues) ]

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3. Rework of Generic Issues A-1, A-10 and A-17 based on our May 6,1993 conference call. j
4. Resolution of those TM1 Issues that are not: -l i
a. Addressed in Appendix 1A I
b. Addressed in Appendix 19A
c. COL Action Item
5. Inclusion of New Generic Issue 145 previously listed as " awaiting NRC input".

We are currently updating Issues 113,120, and 151 and reviewing issue 73 for potential A' B WR '

application with the source information provided by Melinda Malloy.

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  • Replacement page(s) for my April 30,1993 letter, amn 200013 ~ -),

9305200211 930518 i

int ADOCK 05200001~ \/ " l i A- - PDR / - -l ,

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Chet Poslusny, Senior Project Manager May 18,1993

. Page?

I am assuming that the enclosed Safety Issues Index combined with that of my April 30 letter plus potentialIssue 73 constitute all of the required ABWR safety issues. I 1

Please provide a copy of this transmittal to Melinda Malloy.

Sincerely, i

  • t ,

Jack Fox Advanced Reactor Programs  :

cc: Bernie Genetti(GE)

Carl Szybalski(GE)

No-man Fletcher (DOE)  :

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l ABWR mems I Standard Plant i

Table 19B.1-1 SAFETY ISSUES INDEX (Continued)

L NRC SSAR Title Priority Subsection ,

New Generic Issues (Continued) ,

Iunes Resolved With No New Renuirements j (Continued) 29 Bolting Degradation or Failures in Nuclear Plants Resolved 19B.2.62 82 Beyond Design Bases Accidents in Spent Fuel Pools Resolved 19B163 l 113 Dynamic Qualification Testing of Large Bore Hydraulic Snubbers Resolved 19B164 i

TMY Tunes l 1.A.L1 ShiftTechnical Advisor Resolved COL App. l 1.A.I.2 Shift Supervisor Administrative Duties Resolved COL App.  !

LA.I.3 Shift Manning Resolved COL App.  ;

1.A.2.1(1) Qualifications-Experience Resolved COL App. l LA.2.1(2) Training Resolved COL App. l 1.A.2.1(3) Facility Certification of Competence and Fitness of Applicants for Operator and Senior Operator Licenses Resolved COL App. ,

I.A.2.3 Administration of Training Programs Resolved COL App. i LA.2.6(1) Revise Regulatory Guide 1.8 Resolved COL App. t l 1.A.3.1 P2 vise Scope of Criteria for Licensing Examinations Resolved COL App. 1 LA.4.1(2) Interim Changes in Training Simulators Resolved COL App. i LAA.2(1) Research on Training Simulators Resolved COL App. ,

l.AA.2(2) Upgrade Training Simulator Standards Resolved COL App. (

l.A.4.2(3) Regulatory Guide on Training Simulators Resolved COL App. l l LA.4.2(4) Review Simulators for Conformance to Criteria Resolved COL App.  ;

1.C.2 Shift and Relief Turnover Procedures Resolved COL App. j 1.C.3 Shift Supervisor Responsibilities Resolved COL App. j 1.C.4 Control Room Access Resolved COL App.  ;

1.C.5 Procedures for Feedback of Operating Experience to Plant Staff Resolved 19A.2.41  !

1.C.6 Procedures for Verification of Correct Performance of  ;

Operating Activities Resolved COL App.  ;

LC.7 NSSS Vendor Review of Procedures Resolved COL App.  ;

1.C.8 Pilot-Monitoring of Selected Emergency Procedures for  !

Near-Term Operating License Applicants . Resolved COL App.

1.D.2 Plant Safety Parameter Display Console Resolved 1A13 l LD.3 Safety System Status Monitoring Medium 19A117 l.D.5(2) Plant Status and Post-Accident Monitoring Resolved 19B.2.65 1.D.5(3) On-Line Reactor Surveillance System Near Res. 19B.246 LF.2(2) Include QA Personnel in Review and Approval of Plant Procedures Resolved COL App.

1.F.2(3) Include QA Personnel in all Design, Construction, Installation,

( Testing, and Operation Activities Resolved COL App.

LF.2(6) Increase the Size of Licensees' QA Staff Resolved COL App.

l LF.2(9) Clarify Organizational Reporting Levels for the QA Organizatioie Resolved COL App.

I 1.G.1 TrainingRequirements Resolved 1A.2A LG.2 Scope of Test Program Resolved 19B.2.67 l II.B.1 Reactor Coolant System Vents .

Resolved 1A.2.5 II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect l

l 19B.1-4 Amendment l

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ABWR 23xeioois

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Safety Equipment for Post-Accident Operation Resolved 1 A.2.6 II.B.3 Post-Accident Sampling Resolved 1 A.2.7 II.B.4 Training for Mitigating Core Damage Resolved COL App.

II.B.8 Rulemaking Proceeding on Degraded Core Accidents Resolved 19A.2.21 II.D.1 Testing Requirements Resolved IA.2.9 II.D.3 Relief and Safety Valve Position Indication Resolved 1 A.2.10 II.E.4.1 Dedicated Penetrations Resolved 1 A.2.13 II.E.4.2 Isolation Dependability Resolved 1 A.2.14 ll.E.6.1 Test Adequacy Study Resolved 19B.2.68 II.F.1 Additional Accident Monitoring Instrumentation Resolved 1 A.2.15 l

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Amendment 19B.1/

ABWR m6-s "A

Staridard Plant Table 19B.1-1 SAFETY ISSUES INDEX (Continued)

NRC SSAR Title Priority Subsection TMI Issues (Continued)

II.F.2 Identification of and Recovery from Conditions Leading to Inadequate Core Cooling Resolved 1 A.2.16 II.F.3 Instruments for Monitoring Accident Conditions Resolved 1 A.2.17 113.4.1 Revisc Deficiency Reporting Requirements Resolved COL App.

II.K.1(10) Review and Modify Procedures for Removing Safety-Related Systems from Service Resolved 1 A.2.19 II.K.1(13) Propose Technical Specification Changes Reficcting Implementation of All Bulletin Items Resolved 19B.2.69 II.K.1(22) Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems When FW System Not Operable Resolved 1 A.2.20 II.Kl(23) Describe Uses and Types of RV Level Indication for Automatic and Manual Initiation Safety Systems Resolved 1A2.21 II.K.3(3) Report Safety and Relief Valve Failures Promptly and Challenges Annually Resolved 1 A.2.21.1 II.K.3(11) Control Use of PORV Supplied by Control Components, Inc.

Until Further Review Complete Resolved 19B.2.70 ll.K.3(13) Separation of HPCI and RCIC System Initiation Levels Resolved 1 A.2.22 II.K.3(15) Modify Break Detection Logic to Prevent Spurious Isolation of HPCI and RCIC Systems Resolved 1 A.2.23 II.K.3(16) Reduction of Challenges and Failures of Relief Valves -

Feasibility Study and System Modification Resolved 1 A.2.24 II.K.3(17) Report on Outage of ECC Systems - Licensee Report and Technical Specification Changes Resolved 1 A.2.25 II.K.3(18) Modification of ADS Logic-Feasibility Study and Modification for increased Diversity for Some Event Sequences Resolved 1 A.2.26 II.K.3(21) Restart of Core Spray and LPCI Systems on Low Level - Design and Modification Resolved 1 A.2.27 II.K.3(22) Automatic Switchover of RCIC System Suction - Verify Procedures and Modify Design Resolved 1 A.2.28 II.K.3(24) Confirm Adequacy of Space Cooling for HPCI and RCIC Systems Resolved 1 A.2.29 II.K.3(25) Effect ofless of AC Power on Pump Seals Resolved 1 A.2.30 II.K.3(27) Provide Common Reference level for Vessel Level Instrumentation Resolved 19B.2.71 II.K.3(28) Study and Verify Qualification of Accumulators on ADS Valves Resolved 1 A.2.31 II.K.3(30) Revised Small-Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K Resolved 1 A.2.32

, II.K.3(31) Plant-Specific Calculations to Show Compliance with 10 CFR 50.46 Resolved 1 A.2.33 l II.K.3(44) Evaluation of Anticipated Transients with Single Failure to l Venfy No Significant Fuel Failure Resolved 1 A.2.33.1 II.K.3(45) Evaluate Depressurization with Other han Full ADS Resolved 1 A.2.33.2 II.K.3(46) Response to List of Concerns from ACRS Consultant Resolved 1 A.2.33.3 Ill.A.1.1(1) Implement Action Plan Requirements for Promptly Improving Licensee Emergency Preparedness Resolved COL App.

III.A.1.2(1) Technical Support Center Resolved COL App.

I III.A.1.2(2) On-Site Operational Support Center Resolved COL App.

Ill.A.1.2(3) Near-Site Emergency Operations Facility Resolved COL App.

III.A.2.l(1) Publish Proposed Amendments to the Rules Resolved COL App.

Amendment 19B.1[

ABWR mems REV ^

Staridard Plant Ill.A.2.1(2) Conduct Public Regional Meetings Resolved COL App.

III.A.2.l(3) Prepare Final Commission Paper Recommending Adoption of Rules Resolved COL App.

III.A.2.l(4) Revise Inspection Program to Cover Upgraded Requirements Resolved COL App.

III.A.2.2 Development of Guidance and Criteria Resolved COL App. ,

Ill.A.3.3(1) Install Direct Dedicated Telephone Lines Resolved COL App.  !

III.A.3.3(2) Obtain Dedicated. Short-Range Radio Communication Systems Resolved COL App. I Ill.D.I.l(l) Review Information Submitted by Licensees Pertaining to Reducing Leakage imm Operating Systems Resolved COL App.

Ill.D.3.3(1) Issue Letter Requiring Improved Radiation Sampling Instrumentation Resolved 19B.2.72 III.D.3.3(2) Set Criteria Requiring Licensees to Evaluate Need for Additional Survey Equipment Resolved 19B.2.73 Ill.D.3.3(3) Issue a Rule Change Providing Acceptable Methods for Calibration of Radiation-Monitoring Instruments Resolved COL App.

III.D.3.3(4) Issue a Regulatory Guide Resolved COL App.

III.D.3.4 Control Room Habitability Resolved I A.2.36 I

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Amendment 19B.1[1

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l 19B.2.2 A-1: WATER H AMMER ISSUE ,

Unresolved safety Issue (U3I) A-01 in NUREG-0933 (Reference 1) addresses identifying the probable causes of water hammer and minimizing the susceptibil,ity 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 vekicity of a fluid in a closed volume. There are various types of water hammer, including steam condensation-induced water hammer, which occurs in the secondary side of a PWR steam generator at the connection to the feedwater  :

line. His type of water hammer involves steam generator feettrings and piping. Water harnmer has been  !

l observed in may fluid systerns including residual heat removal, contaitunent spray, sersice water, feeds ater systems, and main steam lines. In addiuon to condensation-induced water hammer, other forms of initiating events which cause water hammer can occur, such as steam driven slugs of water, pump startup with partially empty lines, and nipid valve cycling.

Regardless of the initiating event, water hammer and the resulting fluid accelerations can cause damage to the af fected 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 compcments.

Accordin to NUREG-0927 (Reference 2), water hammer can be induced by operator / maintenance l actions and b design inadequacies. Experience has shown that water hammer events reported on LERs are about equalh divided between operator or maintenance actions and design deficiencies. The NRC i implemented SRP changes relative to the design, operation, and maintenance of new plants to minimize the <

probability and effects of water hammer, and issued a Branch Technical Position (BlP) for pre-operational  !

tests.

ACCFirI'ANCE CRITERI A Reference I concluded that USI A-1 was resolved by the publication of SRP sections 3.9.3 Rev.1, 9.2.1, Rev.3, 3.9.4 Rev. 2, 9.2.2, Rev. 2, 5.4.6 Rev.3, 10.3, Rev. 3, and 5.4.7 Rev.3, 10.4.7, Rev. 3.

l 6.3.1, Rev.' '

Compliance with these SRPs becomes the acceptance criteria for resolving this issue.

RESOI UTION l l

De ABWR deskn complies with the above listed SRPs and therefore the water hammer issue is i resolved. Of all the ABWR systems, the systems discussed below are the only systems considered as I having a potential for water hammer.

Potential water hammer conditions are prevented by implementation of the following analyses, design features, m.d pre-operational tests. SSAR section references are given.

1. Water hammer evaluation is required for specific piping regions as follows.

i A. Comiensate and Feedwater System. The feedwater lines wre demonstmted to have immunity to failure from water hammer effects. 10.4.7.3,3E.6.2.2,3E.6.2.7 20.310 Question-Response 430.89 B. Main steam lines are analvred for dynamic kudings due to fast closing of the turbine stop valves. 5.4.9.f(4),

C. All components of the main steam supply system are designed to accommodate the loads and stresses resulting from steam hammer.10.3.3 i

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19B 2.2 A 1

2. Water hammer evaluation is part of the leak Before Break (LBB) analysis consideration. )

(3.6.3.2(5))If the COL applicant applies the LBB analysis (3.6,8th paragraph), the systems identified by Table 3E.1-1 would be evaluated for water hammer. (3.6.3,3rd paragraph and Table 3E.1-1) These systems include rnain steam, feedwater RCIC,llPCF, RIIR, arxl CUw. Relative to LBB, feedwater lines were demonstrated to have immunity to failure from water hammer effects. i (3E.6.2.7) RCIC, llPCF, and RIIR Systems are precluded from water hammer by their keep filled j features and absence of fast acting valves. j

3. Applicable systems are filled with water, atxl kept filled with water, which prevents water hammer when pumps are started from a standy condition. Systems described in the SSAR are as follows.

A. Operatine procedures will be develoned so that all divisions of the Reactor Senice Water System (kSW) are maintained full of water to prevent water hammer. 9.2.15.1.1(5)

B. Operating procedures will be developed so that all components of the Turbine Senice W ater System (TSW) are maintained full to prevent water hammer. 9.2.16.2.2(5) l l C. ResidualIleat Removal (R)IR). 5.4.7.1.1.4,6.3.2.2.5, and 14.2.12.1.8(3)(m)

D. Iligh pressure Core Flooder ()IPCF). 6.3.2.2.5, and 14.2.12.1.10(3)(n)

E. Reactor Core Isolation Cooling (RCIC). 5.4.6.2.5.1,6.3.2.2.5, and 14.2.12.1.9(3)(n)

F. IIVAC Emergency Cooling Water System. 9.2.13.1.2(6) l 4. Condensation Induad Water llammer (CIWII) for the ECCS systems (RilR. IIPCF, and RCIC) t was evaluated for the ABWR (Reference 5). De conclusion was that the ECCS injection piping l configuration was not susceptible to CIWil.

t A. For the RIIR System low pressure fkoder (LPFL) mode, the water in the sloped, but nearly horizontal, injection line Hashes to steam during reactor depressurization. An ,

analysis was perfonned that indicated about 80Tr of the water remained in the pipe after i depressurization. Therchve, slow injection of cold water by the LPFL injection valve into l l the horizontal LPFL pipe partially filled with saturated water will not cause CIWil. '

B. For the IIPCF System, in the event of a LOCA, the hig,h pressure flooder spargers located i inside the RPV shroud are immersed in a two-phase mixture. During the flashine penod prior to IIPCF initiation, the RPV id depressunzing and water in the piping can flash. A steam bubble can fann at the piping's high point. For the IIPCF high pressure system, injection begins within a few secouds, and the entrance of subcooled water could cause decompression inside the pipe. Any water slug accelerated from the reactor side towards the upstream piping will flash into a two-phase mixture because the water is in a saturated condition. A slug of two-phase mixture, which is highly compressible, colliding with another surface has been analyzed and found to produce a pressure pulse of the order of 10-20 psi. His analysis was "donc carlier for typical BWR-5 and BWR-6 piping using TRACB01 computer code. 10-20 psi pressure pulses are not considered sigmficant. and it is concluded that CIWil is not a problem for the lilCF injection piping.

C. For RCIC System during system initiation, the water level in the reactor is a Level 2 or l

higher, which is higher than the feedwater nozzle height. He fluid condition at the j fcedwater sparger as water when RCIC water is pumped into the vessel. Derefore, CIWil  :

will not occur at the time of RCIC makeup water mjection into the reactor vessel. l l S. Pre-operatiomd tests are specified for the purpose of verifying the piping keep-fill methods are operational. Filled pipchnes preclude water nammer associated with pump startup.

A. RilR 14.2.12.1.8(3)(m)

B. IIPCF 14.2.12.1.10(3)(n) l C. RCIC 14.2.12.1.9(3)(n) l i J l

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4 19B.2.2 A-1 REFERENCES

1. NUREG-0933, "A Status Report on Unresolved Safety Issues", U.S. NRC, April 1989.
2. NUREG-0927. Revision 1, " Evaluation of Water llanuner Oecerrences in Nuclear Power Plants", U.S.

NRC, April 1984.

3. 10CFR50 Appendix A
  • General Design Criteria for Nuclear Power Plants". Code of Federal Regulations,' Of fiee of the Federal Repster, 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. ,
5. Jack Fox, GE, to Chet Poslusny, NRC, " Submittal Supporting Accelerated ABWR Review Schedule -

l Water llammer Evaluation," April 11,1993.

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19B.2.6 A-10: BWR FFFDWATER NO77I F CRACKING 15 SUE 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 depth (including claddmg), one crack penetrated the cladding into the base metal for a total depth of approximately 1.5 inch.

It was determined that cracking was due to high-cycle fatigue caused by fluctuations in water temperature within the vessel in the nozzle region. Dese fluctuations occurred during periods oflow feedwater temperature when flow is unsteady and intermittent Once initiated, the cracks enlarge Irom tu h pressure and thermal cycling associated with startups and shutdowns. This item was originally identified in h -0371 and was later determined to be a unresolved safety issue (USI) (References I and 2).

ACCEFTANCE CRITERIA The acceptance criteria is based on developing a design that provides protection to the feedwater nozzles from the water temperature fluctuations. De 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 minng of the hot water retuming from steam separators and dryers and the incc<ning cold feedwater causes thermal stress cycling of nozzle bore unless it is thoroughly protected by the sparger thennal sleeve.

Bypass leakage past the junction of the thermal sleeve and nozzle safe end is the pnmary source of cold water impingmg upon the nozzle bore. A secondar contact with"the outer surface of the sleeve. y source is the layer of water that sheds off after being cooled by resol _UTION

%c welded double sleeve design gives a low fatigue usage factor in the nozzle bore and at the inner nozzle corner. De design in the feedwater nozzleprotects theresolved has been nozzleforfrom the fluctuatigemperatures AB and, therefore, the issue of high cycle fatigue De ABWR utilizes a double feedwater nozrJe thennal sleeve. An inner thermal sleeve leading the cooler feedwater to the feedwater sparger is weldt-j to the nozzle safe end. De welded thermal sleeve design was adopted to assure that there is no lu- kage of cold feedwater between the thermal sleeve and the safe end. A secondary thermal sleeve is placed concentrically in the armulus between the inner thermal sleeve and the nozzle bore to .

prevent cooled water that ma and the inside nozzle corner.y be shedding from the outside surface of the inner sleeve impinge on the nozzle bore As a desiyn feature, the triple feedwater nozzle sleeve was considered less desirable than the double sleeve arrangement Rr ABWR because the triple sleeve allows more cold water to impinge on the nozzle by leakage past two sets of cylinder ring seals and there is also greater risk of complete failure of the cylinder ring seals. He greater leakage past these seals causes Freater thermal stress cycling of the nozzle.

REFERENCE

1. NUREG-0619 *BWR Feedwater Nozzle and Control Rod Drive Return 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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19.B.2.59 A-17: SYSTEMS INTER ACTIONS IN NL'Cf F AR POWER PL ANTS [

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i Unresolved Safety Issue (USI) A-17 in NUREG-0933 (Reference 1) addresses tie concem that inconspiceuous I

or unanticipated intenlependences may exist between s} stems and may result in a degradation of the predicted capability of safety systems in an accident or transient,in particular from ikxxiing and water intrusion.

In its regulator SRPs (Referencein 5)y analysis general in NUREG-1229 cover the (Reference Assessment of System 4), (Asis Interactions the NRC concluded

) of concern, exceptthat forareas f(r the future plants the e of intemal flocxling and water intrusion. A flooding event could cause a transient and also disable the equipment needed to mitigate the consequences of the event. NUREG-1174 (Reference 6) provided guidance in this area a:id references NRC Information Notices regarding operating plant experiences. The NRC plans to develop an SRP relative to ikxxiing and water intrusion, but otherwise not issue new requirements. In the meantime, the NRC recommends that plant designers keep current on lessons teamed from operating experience as repcried in LERs. and that the Probabilistic Risk Assessment (PRA) required for a future plantbe also considered as a tool to help uncover flooding and water intrusion ASIS.  ;

f ACCEPTANCE CRITERIA I 'Ibe acceptance criterion for the resolution of USI A-17 is that auention shall be paid in the detailed plant design l to detecting and minimizing the potential for ASIS due to the effects of ikxxlin  ;

plant sources, such as the incidents as operating plants referenced in NUREG g and water intrusion from intei means for reaching and maintaining a safe hot shutdown. ,

RESOLUTION ASIS are difficult to predict or detect, and are determined by the specific, detailed system designs and layouts. l They may also be influenced by building design features. j For the ABWR desien. consideration is given to identifying flooding and water intrusion possibilities which are not covered by current SkPs, as discussed in NUREG-1174. 'Ihese events include water or moisture release from sources intemal to plant structures, and may involve only small amounts of water and subtle c(unmunication paths to sensitive equipment such as electrical cabinets This evaluation has been made for each of the interaction incidents resulting imm water intrusion at operatiny plants described in the NRC Information Notices referenced in NUREG-1174, to identify the features of the ABWR design, which should ensure preventior, of a similar interaction.

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'Ibe analytical models developed for the ABWR design PRA (SSAR Chapter 19 Appendix R have the capability to evaluate the impact ofikxxiing or water intrusion which appears to be siemficant. As the detailed design is developed, these analytical models will be used to identify potential water nooding problems and provide guidance j on their clumnauon. '

To summarize, the design process for the ABWR design takes into account the possibility for interaction between water ikxxiing and other systems to occur that may degrade plant safe!" # but are not easily recognizable. To the extent evaluated, practicable, as the design progresses these interachons will be identifsed. Their impact on safet Design.

REFERENCES 1

1. NUREG-0933, "A Status Report on Unresolved Safety Issues". U. S. NRC, July 1991. i l l l 2. NiiREG/CR-3922, " Sun ey and Evaluation of System Interaction Events and Souras" U. S. NRC, January l

1985.

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3. NUREG/CR-4261,
  • Assessment of System Intemetion Experience at Nuclear Power Plants", U. S. NRC, June 1986.
4. NUREG-1299," Regulatory Analysis for Resolution of USI A-17",U.S.NRC, August 1989.

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19.B.2.59 A 17 I

5. NUREG-0800," Standard Review Phm for the Review of Safety Analysis Reports for Nuclear Ibwer Plants -

LWR Edition". U. S. NRC.

6. NUREG-1174, " Evaluation of Systems Interaction in Nuclear Power Phints - Technical Findings Related to Unresolval Safety issue A-17",13. S. NRC, May 1989.

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4 19I1.2.65 f.D.5(2) PI ANT STATUS AND POSTACCIDENT MONITORIM ISSUE Re issue addressed is documented in TMI Action Plan, and focuses on the need to improve the ability of nuclear power plant control room operators to prevent, diagnose, and pmperly respond to accidents by improving the informauon provided to them. (Refere' ice 1)

! ACCEPTANCE CRITERIA he acceptance criteria for the resolution of Issue I.D.5(2) is that plant status and postaccident monitoring is in l

full compliance with RG 1.97 (Reference 2) of NUREG.0660 (Reference 1).

i RESOI UTION i %e ABWR design ofits information systems (imponant to safety) provide information for manual initiation l and control of safety systems. These systems provide mdication to the control room that plant safety functions are

being accomplished and provide information from which appropriate actions can be taken to niitigate the r j consequences of anticipated operational occurrences and ecodents. It is designed to perform as described in

! Subsection 7.5, Information Systems Importml to Safety, and is in compliana: with RG 1.97 (Reference 2). ,

i REFERENCES i 1.

A 4 NUREG-0660, "NRD" Action Plan Developed as a Result of the Bil-2 Accident," U.S. NRC, May 1990.

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, 2. Regulatory Guide 1.97 Revision 3. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess -

Plant and Environs condition During and Following an Accident," U.S. NRC May 1983.

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l 19B.2.66 LD.50) ON-LINE RE ACTOR SURVEII I ANCE SYSTEM ISSUE j I

NUREG-0933 (Reference 1), Generic Safety Issue (GSI), item I.D.5(3) addresses the TMI issue of an "On-line Reactor Surveillance System". His issue specifically concems detecting abnormal reactor core intemal's noise l

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associated with on-line stactor operation, e.g., detecting loose internal reactor parts. s i

ACCElTANCE CRITERIA i i

The acceptance criteria for the resolution of GSI-1.D.5(3) is that, based on the on-going generic BWR programs, it is concluded that the technical resolution of this issue has been identified. See Reference 1.

RESOLUTION Re primary cause of core vibration is high and turbulent reactor water recirculation flow. To detect such vibration, the ADWR design incorporates a reactor vessel loose parts monitoring system (LPMS), that complies with NRC's Reg. Guide 1.133 requirements, dated Ma 1981. In addition, with the redesign for the ABWR reactor core r internals, i.e core fuel supports, fuel boxes and strument channel's etc., problem reoccurrence has essentially been I climinated. The LPMS and other ABWR instrumentation systems, will continue to monitor various reactor l operational parameters, e ., reactor core vibration, neutron flux tterns and stability; and thus, any problem recurrence would be ui detected rior to any adverse core e fects which might result. Furthermore, when  !

compared to most o r 's, the A WR design incorporates small, rather than two large reactor water  !

recirculation pumps and these are in-core type pumps. This arra is designed to more uniformly distribute  !

core flow and thus reduce any flow turbulence that might lead to cosening of react (r internal core parts. l i

REFERENCES

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1. NUREG-0933,"A Status Repon on Unresolved Safety Issues". U.S. NRC, April 1989.
2. 22A6100AB, Rev. C, Amendment 21, ABWR Standard Plant SS AR, Section 4.4.3,
  • loose Pans Monitoring i Systern. [

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19.B.2.67 I.G.2: SCOPE OF TEST PROGH AM ISSl'E The major thrust of TMI Action Plan I.G is to use the preoperational and startup test programs as a training exercise for the operating crews, la contrast to this, Item 1.G.2 calls for a more comprehensive test program to search for anomalies in a plant's response to a transient. The safety significance of this assue lies in the early discovery of anomatics of unanticipated plant behavior. When a plant resporxis to a transient in an anomalous or unanticipated manner, the result may be an acrident caused directly by the new phenomena. or by the surprise or confusion on the part of the operators. i (Reference 1.) t ACCEPTANCE CRITERI A

'Ihe acceptanct criteria for the resolution ofissue I.G.2 is compliance with Standard Review Plan (SRP) Chapter 14 (Reference 2), and Regulatory Guide 1.68 (Reference 3).

RESOlllTION The ABWR will have a test program to evaluate and demonstrate, to the extent possible, plant operating procedures to provide assurance that the operating group is knowledgeable about the plant and procedures and fully l prepared to operate the facility in a safe manner as described in Chapter 14. Subsection 14.2.7, Conformance of Test ,

program with Regulatory Guides, identifies Regulatory Guide 1.68 and other applicable regulatory guides used in the devchyment of test programs. Therefore this issue is resolved for ABWR.

REFERENCES l

1. NUREG-0933,"A Prioritization of Generie 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.
3. Regulatory Guide 1.68, " Initial Test Programs for Water-Cooled Nuclear Power Plants," U.S. NRC.

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19B.2.68 II.E.6.1: TEST ADEOUACY STUIW ISSUE The purpose of this TMl Action Plan (Reference 1) is to establish the adequacy of current requirements for safety-related valve testing. It recommends a study which wculd result in recommendations for attemate means of verifying performance requirernents. ,

ACCElTANCE CRITERI A -

he acceptance criteria for the resolution of Issue ILE.6.1 is in conformance to ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987, Parts 1,6, and 10 as required by NRC Generic Letter 89-10 (Reference 3).

kt' SOLUTION Valve performance is critical to the successful functioning of a large number of the plants

  • safety systems. In-senice tesung~ of safety-related valves will be performed in accordance with the requirements of ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987. Parts 1,6 and 10 as described in Reference 2. De ABWR standard plant (Reference 2) lists the in-senice testing parameters and frequencies for the safety-related valves. He reason for each code defined testing exception orjustification for each cale exemption request is noted in the description of the affected valve. Valves ving a containment isolation function are also noted in the listing (Reference 2)

Details of the in-senice testing program, including test schedules and frequencies will be reIorted in the in-senice inspection and testing plan which will be provided by the applicant referencing the ABWR design. De plan will integrate the applicable test requirements for safety-related vafves including those listed in the technical specifications (Chapter 16) and the containment isolauon system. His plan will include baseline pre-senice testing to sup[ ort the periodic in-senice testing of the components. Depending on the test results, the plan will provide a comrrutment to disassemble and inspect the safety related valves u hen limits of the OM Code are exacded. He primary elements of this  ;

delineated in Reference. .Herefore, pan including the regmrements this issue is resolved forof Generic the ABWR.letter 89-10 for motor operated valves, are REITRENCES i

1. NUREG-0660, NRC Action Plan Developed as a Result of the TMI-2 Accident," U.S. NRC, May 1980,  !

(Revision 1), August 1980. ,

2. Subsection 3.9.6: TestinF of Pumps and Valves. ,

t i l 3. NRC Letter to All Licensees of Operatine Power Plants and lloiders of Construction Permits for Nuclear Power  !

j Plants.

  • Safety-Related Motor-Operated Valve Testing and Surveillance (Generic Letter No. 89-10)- 10 CIR '
50.54(f)," June 28,1989, i i

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l 1911.2.69 II K.103) PROPOSE TECIINICAL SPECIFICATION CIIANGES REFI_ECTING IMPI EMENTATION OF ALL Il0L1 ETIN ITEMS ISSL E Issue ll.K (measures to mitigate small - break loss of coolant accidents and loss of feedwater accidents) has the objective ofimproving the capability to miticate the consequences of small-break accidents and loss of feedwater

, events. Nine Inspection and hnforcement (IE) bulletins were issued to operating plants with twenty-eight l requirements (Task II.K and Table C.1, NUREG-0660. Reference 1) for reviews of plant design and operation l Issue 11 K.l(13)is one of the twenty-eight requirements of the overallissue. It is directed at im l Technical Specification changes that would be required from other changes madebulletin to respond items. to all Ib emen ACCEPTANCE CRITERIA 1

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'Ihe acceptance criteria for the resolution ofissue II.K.l(13)is in compliance with 10CFR50.36, Technical Specifications (Reference 2) and the interim " Proposed Policy Statement on Technical Specification Improvements it Nuclear Power",52rR3788, February 1987.

RESOLUTION The ABWR demonstrates in Chapter 15, Accident Analysis, the capability to respond to the full spectrum of line breaks and loss-of-feedwater accidents without loss of contamment or significant core darnage. Chapter 16.

Technical Specifications sets forth the restrictions on plant operation required to control the transients and abnonrd events of Chapter 15 to ensure conformance with the NRC rules identified in the Acceptance Criteria for this issue.

l Accordingly, the analyses of Carter 15 and the operational conditions and limitations of Chapter 16 ensure that I the ABWR fulhlis the intent of issue II.K.l(13).

REITRENCES

1. NUREG-0660,*NRC Action Plan Developed as a Result of the TMI-2 Accident". U.S. NRC, May 1980.
2. 10CFR50.36, " Technical Specifications". Office of the Federal Register, Nation d Archives and Records
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19.B.2.70 II.K.N11) CONTROL USE OF PORY SUPPI IFD BY CONTROL COMPONENTS. INC. UNTIL FURTIIER REVISION COMPLETE ISSUE Issue II.K (Measures to Mitigate Small-Break less-of Coolant Accidents and Imssef-Feedwater Accidents) has the objective of improving the capability to miti events. For this issue, the Bulletms arxf Orders (gate the consequences ofreviews small-break accidents and loss of feedw 4

u&O) Task Force conducted generic of systems reliability, emergency procedures, and operator trainine as documented in NUREG-0626 (GE), (Reference 2), and the NRC issue'l some 32 recommendations for the BWR (Task II.K and Table C.3, NUREG-0660. Reference 1) for reviews of plant design and operations.

Issue ll.K.3(11) is one of the 32 BWR recommendations of the Bulletins and Orders Task Force. It re plants tojustify the use of PORVs (Power Operated Relief Valves) supplied by Cor. trol that Components, had Inc.

tailed dunng testmg.

ACCEPTANCE CRITERI A I

De acceptance criteria for the resolution of issue ll.K.3(11) is compliance with 10CFR50. Appendix A, j General Design Criterion 15, Reactor coolant system design, and the applicable etxles and standards governing safety / relief vahes (SRV)

! RESOLUTION De ABWR demonstrates in Chapter 15, Accident Analysis, the capability to respond to the full spectrum of line breaks and loss-of-feedwater accidents without loss of contamment or significant core damage. i 1

Subsection 5.2, Integrity of Reactor Coohmt Pressure Boundary, describes the overpressure protection provided by the SRVs performing an overpressure relief valve function, an overpressure safety valve functi(m or an

Automatic Depressunzation system ( ADS) function.

l The SRV for the ABWR is not a Power Operated Relief Valve by Control Components, Inc. It is a spnng-kuded safety valve for the safety valve function, witt a pneumatic cylinder / piston for power operation in the ADS arxl relief f unction.

I Subsection 3.9.3.2.4.2. Main Steam Safety / Relief Vahe describes the qualification by type test of the SRVs to ,

a IEEE 344. Recommended Practice for Seismic Qualification of Class IE Equipment for Sucfear Power Generating i Stations. (Reference 3), for operability dunng a dynarme event. I J l Herefore, this issue II.K.3(11) is resolved for the ABWR.

l REFERENCES

1. NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," U.S.NRC, May 1980.
2. NUREG-0626. " Staff Report on the Generic Assessment of Feedwater Transients and Small Break less-of-Coolant Accidents in Boding Water Reactors Designed by the General Electric Company," U.S.NRC, January

, 1980.

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3. IEEE Std 344-1987, Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations.

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4.B.2.71 II.K.3427)
Provide Common Reference Level for Vessel 4

Inst rument at i on 1

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ISSUE i

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The overall issue II.K. " Measures to Mitigate Small-Break Loss-of-Coolant Accidents and Loss-of-Feedwater Accidents." has the objectives to perf orm systems reliability and to ef fect changes in ,

emeroency operatina procedures and operator training to improve the  !

] capability to mitigate such accidents. [

The concern in issue II.K.3(27) is that different reference points of .
the various reactor vessel water level instruments could cause

, operator confusion. Either the bottom of the vessel or the active i

fuel were considered to be reasonable ref erence. points. (Reference 1).

ACCEPTANCE CRITERIA i

The acceptance criteria for the resolution of issue II.K.3(27) is to j confirm that the ABWR desion has a common zero reference for all water level indications.

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. RESOLUTION 1

The resolution of this issue II.K.3(27) for the ABWR is accomplished by settina a common reference for the reactor vessel water level at the top of the active fuel as shown on Fiqure 5.1-3, Nuclear Boiler System P&ID. and as described in Section 7.7. Control Systems Not  ;

, Required for Safety. Therefore, this issue II.K.3(27) is resolved for }

the ABWR.

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a F REFERENCES  :

1. NUREG-0933. " A Prioritization of Generic Saf ety Issues." (with 'I Supplements). July 1991.

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1 19B.2.72 III.D.3.3m- ISSUE LETTER REOUIRING IMPROVED RADI ATION SAMPLING l INSTRUMENTATION ISSUE:

his item was clarified in NUREG-0737, requirements were issued, and Multi-Plant Acticm F-69 was established for implementation purposes. Are these NUREG-0737 requirements depicted in the ABWR SSAR7 i i

ACCEPTANCE CRITERI A: -

De acceptance criteria for the resolution of issue item Ill.D.3.3(1) is that the ABWR SSAR is in full j compliance with tie requirements of NUREG-0737. ,

i RESOLUTION 3  !

Item Ill.D.3.3(1) which conarns inplant radiation monitoring is resolved in ABWR SSAR Section 12.3.4 (Area Radiation and Airbome Rulicactivity Monitoring Instmmentation) which also references each area detector kcation ,

on the plant layout drawings for each building (ABWR SSAR Figures 12.3-56 through 12.3-73) as well as the i specific area radiation channels for each building, tle detector map kration, the channel wnsitivity range, and the I heal alarm areas. (ABWR SS AR Tables 12.3-ithmugh 12.3-7).  !

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REFERENCES:

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1. NUREG-0660, "NRC Action Plan Devekiped as a Result of the TMI-2 Accident". U.S. NRC, May 1990.

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2. NUREG-0737," Clarification of TMI Action Plan Requirements" U.S. NRC, November 1980. j i

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19H.2.73 III.D.3.M2h SIT CRITERI A REOUIRING LICENSFFS TO EVAI UATE NFFD FOR ADDITIONAL SURVEY EOUIPMENT ISSUE:

his NUREGJXie item required the NRR to set criteria requiring licensees to evaluate in their plants the need for additional survey equipment and idiation monitors in vital areas and requiring, as necessary, installation of area monitors with remote readout. The Is RR was to evaluate the need to specify the minimum types and quantities of ponable monitoring instrumentation, including very high dcwe rate survey instruments. atmg reactors were to be review-d for conformance with Standard Review Plan (SRP) Section 123.4, " Area iation and Airborne Radioactivity Monitorir!g Instrumentation". The NRR was to revise SRP Sections 12.5 and 123.4 u incorporate additional snonitor reqmrement criteria.

ACCElrrANCE CRITERIA:

De axeptance criteria for the resolution of issue Item I!!.D33(2)is that the ABWR SSAR is in full compliance with the requirements of NUREG-0MO.

RESOLUTION:

Item III.D33(2) which concerns licenwes evaluate the need for additional radiation survey equipment is resolved in ABWR SSAR Section 123.4 (Area Radiation and Airborne Radioactivity Monitormg Instnimentation).

This item also concerned the need to specify the minimum types and quantities of portable monitoring instrumentation, including very high dose rate survey instmments. As noted in ADWR SSAR Sections 12.5.2, 19A.239 and 19A3.5, COL apphcants will provide the portable instruments in operating reactors that accumtely measure radio-iodine concentration in plant areas under accident conditions.

REFERENCES:

1. NUREG-0MO,"NRC Action Plan Developed as a Result of the TMI-2 Accident"_, U.S. NRC, May 1990. -
2. NUREG-0737, " Clarification of Bil Action Plan Requirements", U.S. NRC, November 1980. I 4

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4[ f 19.R.2J0C 145: ACTIONS TO REDUCE COMMON CAUSE FAILURES [

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ISSUE i Issue 145 is concerned that common cause failures can be a major cause of a system failure. De ' .!

hil-2 and David Besse incidents were examples of scenarios involving common cause failures. l (Reference 1.)

Effective maintenance is important to ensure that design assumptions and margins in the original design basis are maintained. In the design of nuclear power plants, an important safety margin is the redundancy of equipment to perform safety functions. His redundancy, however, can be degraded by common cause failures. Herefore, defense against such failures (by root cause analyses and investigations) over the life of the plant is an impor t part of the licensec's mamtenance program. j v i his issue is still under evolution within the NRC with a regulatory gupide urxler development to i supplement the maintenance rule,10CFR50.65, Reguirements for Monitanng the Effectiveness of i Mamtenance at Nuclear Power Plants. (Reference 2.) i i

ACCElrTANCE CRITERIA [

He acceptance criteria for the resolution ofIssue 145 is to demonstate compliance with the maintenance rule,10CFR50.65. l RESOLUTION f i

Compliance with 10CFR50.65 will be the responsibility of the COL applicant. l In addition, the ABWR design demonstrates in its Response to Severe Accidents. Chapter 19,its  !

capability to respond to system interactions and common cause failures, (Subsection 19.2.3.4).

Acrefore, this issue 145 is resolved for the ABWR. l i

EITERENCES l

1. NUREG-0933, "A Prioritization of Generic Safety Issues," (with Supplements), December 1992.

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2. 10CFR50.65, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power l Plants " Office of the Federal Register, National Archives and Records Administration. )

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