ML20056G814

Forwards Revised mark-ups for Issues A10,A17,A47,B5,C8,C17, 25,51,82,89,113,143,153 & Revised Index (Table 19B.1-1) in Response to NRC 930804 Comments for Review & Resolution

ML20056G814
Person / Time
Site:	05200001
Issue date:	09/01/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9309070196
Download: ML20056G814 (28)

Text

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1MCar:wrkenn 599Juse CA 95125 September 1,1993 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Dr. rectorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Schedule - USIs/GSIs

Reference:

1. NRC comments dated August 4,1993

2. Letter, R.W. Borchardt to P.W. Marriott, " Staff Review Comments on Generic Issues Related to the Advanced Boiling Water Reactor (ABWR) Design", August 12.1993.

Enclosed are 13 revised mark-ups (for Issues A10, A17, A47 B5, C8, C17,25,51,82,89,113, 143,153) and a revised index (Table 19B.1-1) and a revised index (Table 19B.1-1)in response i to Reference 1 for your review and resolution.

l 1ssue 73, " Detached Thermal Sleeves" has been resolved for BWRs, and GE has no cause to address this issue for the ABWR. Therefore this issue is withdrawn.

Mbeellaneous Safety issues The NRC Priority corrections for issues B17, B55, B56,57,78,83,120 have been made as directed. Where " Resolution Available" is indicated (Issues 145,155.1) we have used

" Resolved" for simplicity.

For the changes in issue titles (Issues 143,145,151, A17) we have used the title as shown on the description of the issue in NUREG-0933, July 1991. Note that Issue 145 is " Actions to Reduce Common Cause Failures" rather than " Improve Surveillance and Startup Testing Programs" 1

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Issues (120,121,151) that are resolved with no new requirements have been moved to the i section of the table " Issues Resolved With No New Requirements" as recommended.  !

Miscellaneous TMI Action Plan items f Paragraph 1.* The TMI Action Plan items listed in the NRC Reference 2 and identified in f~

the Table 19B.1-1 as for

• COL Applicant" have been discussed in Appendix 19B.1.1 and 19B3. ,

i Paragraph 2. The missing TMI items [1.A.I.4, l.C.1(1).1.C.1(2), l.C.1(3), l.D.1, ll.E.13] have  !

been added to Table 19B.1-1. The items assigned to " COL Applicana" [1.C.1(1), l.C.1(2),  ;

ll.E.1.3] are discussed in Appendix 198.1.1 and 19B3.

Paragraph 6. The correct reference for ll.K3(27) is Subsection 1 A.2.21. Table 19B.1-1 has been restructured and 19B 2.71 replaced.

Paragraph 7. The correct reference for lil.D33(1) is Subsection 19A.239. Table 19B.1-1 has l been restructured and 19B.2.72 replaced. ,

Paragraph 8. The correct reference for Ill.D33(2) is Subsection 19A.239. Table 19B.1-1 has been restructured and 19B.2.73 replaced. l Please provide a copy of this transmittal to Melinda Malloy.

• Refers to the Paragraphs of Page 9 and 10 of Enclosure 2 to Reference 2. ,

Sincerely,  !

SY ek Fox 1 dvanced Reactor Programs 1 cc: Alan Beard (GL) l Norman Fletcher (DOE)  !

Bernie Genetti (GE)

Carl Szybalski(GE) ,

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-I. ABWR 23A61cDAS <

Starwl'arti Pfarif REV.A Table 19B.1-1 SAFETY ISSUES INDEX NRC SSAR ,

Title Priority Subsection Generic Issues A-1 Water Hammer Resolved 19B.2.2 A-7 Mark I Long-Term Program Resolved 19B.2.3 A-8 Mark 1 Containment Paol Dynamic Loads - Long Term Program Resolved 19B.2.4 A-9 ATWS Resolved 19B.2.5 A-10 BWR Feedwater Nozzle Cracking Resolved 19B.2.6 A-13 Snubber Operability Assurance Resolved 19B.2.7 A-24 Qualification of Class IE Safety Related Equipment Resolv-d 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 leads Near Spent Fuel Resolved 19B.2.12 A-39 Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits Resolved 19B.2.13 A-40 Seismic 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-48 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment Resolved 19B.2.18 B-10 Behavior of BWR Mark III Containments Resolved 19B.2.19 B-17 Criteria for Safety-Related Operator Actions Medium Appendix 18A B-36 Develop Design, Testing and Maintenance Criteria for Atmospheric Cleanup System Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Normal Ventilation Systems Resolved 19B.2.21  :

B-55 Improved Reliability of TarFet Rock Safety Relief Valves Medium 19B.2.22 ,

B-56 Diesel Reliability High 19B.2.23 l B-61 Allowable ECCS Equipment Outage Periods Resolved 19B.2.24 i B-63 Isolation of low Pressure Systems Connected to the j Reactor Coolant pressure Boundary Resolved 19B.2.25 i B-66 Control Room Infiltration Measurements Resolved 19B.2.26 1 C-1 Assurance of Continuous Long Term Capability of Hermetic Seals on instrumentation and Electrical Equipment Resolved 19B.2.27 C-10 Effective Operation of Containment Sprays in a LOCA Resolved 19B.2.28 C-17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes Resolved 19B.2.29 New Generic Issues 15 Radiation Effects on Reactor Vessel Supports High 19B.2.30 23 Reactor Coolant Pump Seal Failures High 19B.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 Scram System Resolved 19B.2.33 i

July 22,1993 Amendment 19B. t-2 J

i

. ABWR l Starwbrd Plant '*$,^) l l

Table 19B.1-1 SAFETY ISSUES INDEX (Continued) i NRC SSAR Title Priority Subsection New Generic issues (Continued) 45 Inoperability of lastrumentation Due to Extreme Cold Weather Resolved 19B.2.34 51 Proposed Requirements for Improving the Reliability of Open Cycle Service Water Systems Resolved 19B.2.35 57 Effects of Fire Protection System Actuation on Safety-Related Equipment Meduim 19B.2.36 67.3.3 Improved Accident Monitoring Resolved 19B.2.37 75 Generic Implications of ATWS Events at the Salem Nuclear Plant Resolved 19B.2.38 78 Monitoring of Fatigue Transient Limits for Reactor Coolant System Meduim 19B.2.39 83 Control Room Habitability Possible Res. 19B 2.40 86 Ieng Range Plan for Dealing with Stress Corrosion Cracking in BWR Piping Resolved 19B.2.41 87 Failure of HPCI Steam Line Without Isolation Resolved 19B.2.42 89 Stiff Pipe Clamps Medium 19B.2.43 103 Design for Probable Maximtn Precipitation Resolved 19B.2.44 105 Interfacing Systems LOCA at BWRs High 19B.2.45 106 Piping and Use of Highly Combustible Gases in Vital Areas Medium 198.2.46 118 Tendon Anchorage Failure Resolved 19B.2.48 124 Auxiliary Feedwater System Reliability Resolved 19B.2.51 128 Electrical Power Reliability Resolved 19B.2.52 142 Leakage Through Electrica.1 Isolators in Instrumentation Circuits Medium 19B.2.53 143 Availability of Chilled Water Systems and Room Cooling High 19B.2.54 145 Actions to Reduce Common Cause Failures Resolved 19B.2.55 153 Losc of Essential Service Water in LWRs High 19B.2.57 155.1 More Realistic Source Term Assumptions Resolved 19B.2.58 liuman Factors issues IIF.1.1 Shift Staffing Resolved 18.8.2 HF.4.4 Guidelines for Upgrading Other Procedures High 18.8.1 18E.1.7 HF.5.1 Local Control Stari. ins High 18.8.11 liF.5.2 Review Criteria lor Human Factars Aspects of Advanced Controls and Instrumentation High 18.8.9 1ssues Resolved With No New Requirements A-17 Systems Interaction in Nuclear Power Plants Resolved 198.2.59 A-29 Nuclear Power Plant Design for Reduction of Vulnerabilhe lon 160 to Industrial Sabotage Resolved C4 ipp.

B-5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containments Resolved 19B -i C-8 Main Steamline Leakage Control Systems Resolved 19B.2.01.1 29 Bolting Degradation or Failure in Nuclear Power Plants Resolved 19B.2.62 l

July 22.1993 Anwrdnwm 19B.1-3 i

ABM 23A6100AS Starvlarvi Plant nev. A Table 19B.1-1 SAFETY ISSUES INDEX (Continued)

NRC SSAR Title Priority Subsection IMun.Resobed With No New Requirements LCuntinued) 82 Beyond Design Basis Accidents in Spent Fuel Pools Resolved 19B.2.63 113 Dynamic Qualification Testing of Large Bore flydraulic Snubbers Resolved 19B.2.64 120 On-Line Testability of Protection Systems Resolved 19B.2.49 121 ilydrogen Control for Large, Dry PWR Containments Resolved 19B.2.50 151 Reliability of Anticipated Transient without Scram Recirculation Pump Trip in BWRs Resolved 19B.2.56 Dil Issun I.A.I.1 Shift Technical Advisor Resolved COL App.

l.A.I.2 Shift Supervisor Administrative Duties Resolved COL App.

l.A.13 Shift Manning Resolved COL App.

l.A.I.4 Long-Term Upgrading Resolved Appendix 18E I.A.2.l(l) Qualifications-Experience Resolved COL App.

I.A.2.l(2) Training Resolved COL App.

I. A.2.l(3) Facility Certification of Competence and Fitness of Applicants for Operator and Senior Operator Licenses Resolved COL App.

l. A.2.3 Admimstration of Training Programs Resolved COL App.

l.A.2.6(1) Revise Regulatory Guide 1.8 Resolved COL App.

l.A.3.1 Revise Scope of Criteria for Licensing Examinations Resolved COL App.

l. A.4.l(2) Interim Changes in Training Simulators Resolved COL App.

l.A.4.2(1) Research on Training Simulators Resolved 19A.3.1

1. A.4.2(2 ) Upgrade Training Simulator Standards Resolved 19A.3.1 1.A.4.2(3) Regulatory Guide on Training Simulators Resolved 19 A .3.1 1.A.4.2(4) Review Simulators for Conformance to Criteria Resolved 19A.3.1 I.C.l(l) Small-Break LOCAs Resolvec. COL App. l I.C.l(2) Inadequate Core Cooling Resolved COL App.

1.C.l(3) Transients and Accidents Resolved 1A.2.1 1.C.2 Shift and Relief Turnover Procedures Resolved COL App.

1.C.3 Shift Supervisor Responsibili ties Resolved COL App. I 1.C.4 Control Room Access Resolved COL App.

I.C.5 Procedures for Feedback of Operating Experience to Plant Staff Resolved 19A.3.6 1.C.6 Procedures for Verification of Correct Performance of Operating Activities Resolved COL App.

I.C.7 NSSS Vendor Review of Procedures Resolved COL App.

I.C.8 Pilot-Monitoring of Selected Emergency Procedures for Near-Term Operating License Applicants Resolved COL App. j l.D.1 Control Room Design Reviews Resolved IA.2.2 3

!.D.2 Plant Safety Parameter Display Console Resolved 1A.2.3 '

I.D.3 Safety System Status Monitoring Medium 19A.2.17 1.D.5(2) Plant Status and Post-Accident Monitoring Resolved 19B.2.65 I.D.5(3) On-Line Reactor Surveillance System Near Res. 19B.2.66  ;

I.F.2(2) Include QA Personnelin Review and Approval of Plant Procedures Resolved 19A.2.43 I.F.2(3) include QA Personnelin All Design, Construction.

Installation. Testing, and Operation Activities Resolved 19A.2.43 Ju!y 22,1993 Amerdmem 19Bf 4 -

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. * .ABWR 23A6t00AS

- Stabril Pbrit REV.A l

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Table 19I1.1-1 SAFETY ISSUES INDEX (Continued)

NRC SSAR Title Priority Subsection TMI Issues (Continued) 1.F.2(6) Increase the Size of Licensees' QA Staff Resolved 19A.2.43 1.F.2(9) Clarify Organizational Reporting Levels for the QA Organization Resolved 19A.2.43 1.G.1 Trainit.g Requirements Resolved 1A.2.4 1.G.2 Scope of Test Program Resolved 19B.2.67 l II.B.1 Reactor Coolant System Vents Resolved 1 A.2.5 l li.B.2 Plant Shielding to Provide Access to Vital Areas and Protect l Safety Equipment for Post-Accident Operation Resolved 1 A.2.6 l lI.B.3 Post-Accident Sampling Resolved 1A.2.7 II.B.4 Training for Mitigating Core Damage Resolved COL App.

II.B.8 Rulemaking Proceeding on Degladed Core Accidents Resolved 19A.2.1 II.D.1 Testing Requirements Resolved 1A.2.9 l lI.D.3 Relief and Safety Valve Position Indication Resolved I A.2.10  !

II.E.1.3 Update Standard Review Plan and Develop Regulatory Guide Resolved COL App. I II.E.4.1 Dedicated Penetrations Resolved 1A.2.13 II.E.4.2 Isolation Dependability Resolved l

1 A.2.14 j II.E.4.4 Purging Resolved 19A.2.27 II.E.6.1 Test Adequacy Study Resolved 19B.2.68 l II.F.1 Additional Accident Monitoring Instrumentation Resolved I A.2.15 ll.F.2 Identification of and Recovery from Conditions Leading to Inadequate Core Cooling Resolved I A.2.16 II.F.3 Instruments for Monitoring Accident Conditions Resolved 1A.2.17 l II.J.4.1 Revise Deficiency Reporting Requirements Resolved COL App.

II.K.l($) Safety-Related Valve Position Description Resolved 1 A.2.18 li.K.l(10) Review and Modify Procedures for Removing Safety- I Related Systems from Service Resolved 1A.3.2 l II.K.l(13) Propose TechrJeal Specifications Changes Reflectmg i tmplementation of All Bulletin Items Resolved 19B.2.69 II.K.l(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.K.l(23) Describe Uses and Types of RV LevelIndication for Automatic and Manual Initiation Safety Systems Resolved 1 A.2.21 II.K.3(3) Report Safety and Relief Valve Failures Promptly and Challenges Annually Resolved 1 A.3.4 li.K.3(11) Control Use of PORV Supplied by Control Comnponents, Inc. Until Further Review Complete Resolved 19B.2.70 .

II.K.3(13) Separation of HPCI and RCIC System Initiation Levels Resolved I A.2.22 l lI.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 Modificatics Resolved 1 A.2.24 II.K.3(17) Report and Outage of ECC Systems - Licensee Repon and Technical Specification Changes Resolved 1A.2.25 ILK.3(18) Modification of ADS Logic - Feasibility Study and Modification for Increased Diversity for Some Event Sequences Resolved 1 A.2.26 My 22.1993 Amerkiment 19B.1-5

i 23A610nAS StarvEurd Plant REV.A Table 198.1-1 SAFETY ISSUES INDEX (Continued)

NRC SSAR Title Priority Subsection TMI Issues II.K.3(21) Restart of Core Spray and LPCI Systems on Low Level -

Design and Modification Resolved 1A.2.27 II.K.3(22) Automatic Switchover of RCIC System Suction - Verify Procedures and Modify Design Resolved I A.2.28 li.K.3(24) Confirm Adequacy of Space Cooling for HPCI and RCIC Systems Resolved 1 A.2.29 II.K.3.(25) Effect of Loss of AC Power on Pump Seals Resolved 1 A.2.30 ll.K.3(27) Provide Common Reference Level for Vessel Level Instrumentation Resolved 1A. 21 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 II.K.3(44) Evaluation of Anticipated Transients with Single Failure to Verify No Significant Fuel Failure Resolved 1 A.2.33.1 II.K.3(45) Evaluate Depressurization with Other Than Full ADS Resolved 19A.2.11 II.K.3(46) Response to List of Concerns from ACRS Consultant Resolved 1 A.2.33.3 III.A.I l(l) Implement Action Plan Requirements for Promptly Improving Licensee Emergency Prepardness Resolved COL App.

Ill.A.I.2(1) Technical Support Center Resolved 19A. 3.4 Ill. A.I.2(2) On-Site Operational Support Center Resolved 19A.3.4 III.A.I.2(3) Near-Site Emergency Operations Facility Resolved 19A.3.4 Ill.A.2.l(l) Publish Proposed Admendments to the Rules Resolved COL App.

Ill.A.2.l(2) Conduct Public Regional Meetings Resolved COL App.

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

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

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

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

III. A.3.3(2) Obtain Dedicated, Short-Range Radio Communication *ystems Resolved COL App.

Ill.D.Ll(l) Review Information Submitted by Licensee Pertaining to Reducing trakage from Operating Systems Resolved 1 A.2.34 III.D.3.3(1) Issue Letter Requiring Improved Radiation Sampling Instrumentation Pesolved 19A.2.39 III.D.3.3(2) Set Criteria Requiring Licensees to Evaluate Need for Additional Survey Equipment Resolved 19A.2.39 Ill.D.3.3(3) Issue a Rule Change Providing Acceptable Methods for Calibration of Radiation-Monitoring Instruments Resolved 19A.3.5 III.D.3.3(4) Issue a Regulatory Guide Resolved 19A.3.5 Ill.D.3.4 Control Room Habitability Resolved I A.2.36 July 22.1993 Amem! ment 19B 1-6

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I 198.2.6 A- 10: 11WR ITFDWATER NO7.71 I' CRACKING i

ISXUE 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 range fron. u2 inch to 3/4 inch in depth (including cladding), one crack penetrated the cladding into the 'oase meta. for a total depth of approximately 1.5 inch.  ;

, t It was determined that cracking was to high cycle fatigue caused by fluctuations in water temperature within the vessel in the nozzle region. These fluctuations occurred during periods oflow feedwater temperature when flow is unsteady and intermittent. Once initiated, the cracks enlarge from high pressure and thermal cycling associated with startups and shutdowns. This item was originally identified ,

in NUREG-0371 and was latar determined to be an unresolved safety issue (USI) (References I and 2)

ACCEFTANCE CRITERIA The acceptance criteria are based on developing a design that provides protection to the feedwater nozzle from the water temperature fluctuations. The feedwater nozzles experience thermal stress because the incoming feedwater is colder than that in the vessel. It is much colder during startups before the i feedwater heaters are in service and during shutdown after the heaters have been taken out of service.

Turbulent mixing of the hot water returning from the steam separators and dryers and the incoming cold

, feedwater causes thermal stress cycling in the nozzle bore unless it is thoroughly protected by the sparger j thermal sleeve. '

l In previous designs bypass leakage past the junction of the thermal sleeve and nozzle safe end hss been the primary source of cold water impinging on 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.

RESOLUTION 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 fluctuating temperatures and, therefore, the issue of high cycle fatigue in the feedwater nozzle has been resolved for the AB%%.

The ABWR utilizes a double feedwater nozzle thermal sleeve as can be seen from the attached figure. 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 leakage of cold feedwater between the thermal sleeve and the safe end. A secondary thermal sleeve is placed concentrically in the annulus between the inner thermal sleeve and the nozzle bore to prevent cold water that may be shedding from the outside surface of the inner sleeve impinge on the nozzle bore and the inside nozzle corner.

l ne material of the nozzle forging is SA-508, Class 3 low alloy steel and that of the safe end is SA- J 508, Class 1 carbon steel. The carbon steel safe end is welded to the nozzle forging with a carbon steel j weld. The nozzle itself has no cladding. I Welded thermal sleeves have been successfully used in at least three domestic reactors and in BWR/5s in Japan since 1977. The welded double thermal sleeve with no cladding inside the nozzle is considered an improvement of the welded single sleeve design in that the outer thermal sleeve provides additional protection against high cycle fatigue in the nozzle bore and the inside nozzle corner. ne double thermal sleeve as applied to the ABWR has not been used in ea:!ier plants although Monticello and Tsuruga Oapan) are using similar designs.

i i

l 4

The ABWR feedwater nozzle and thermal sleeve design does not correspond to any design mentioned  !

in Table 2 of NUREG 0619. The closest design is considered to be " Welded, clad removed (spargers ,

have top mounted elbows)". Ilence, the proposed program for ISI of the ABWR feedwater nozzles and spargers is based on this c'esign. Based upon programs approved by the NRC allowing relief from i periodic PT inspections, the following program is proposed:

UT exas.ation from the external surface of the nozzle safe ends, nozzle bores and nozzle blend radius every second outage, if indications are found in the safe ends, evaluate per section Xi of the ASME Code. If recordable indications are interpreted as cracks in any nozzle, proceed with repair as outlined in NUREG 0619. Paragraph 4.3.2.3.  ;

Visual inspection of flow holes and welds in sparger arms and sparger tecs every fourth outage.

Visual inspection of accessible areas of the nozzles from the ID surface on the same schedule as core internal components.

r it is believed that UT examination of the nozzle bore using advanced techniques gue better results than PT inspection of accessible areas. This method has successfully been tried out on several domestic reactors. Depending upon actual operating experience, it may be possible to extend the period terween .;

UT examinations.

I RIFERENCE

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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19ft161 B-5* DUCTIT YTY OF TWO-WAY SLABS AND STIFT T 9 AND BUCKT rNG Brfl AVIOR OF STFEL CONTAINMENTS ISil!E Generic Safety Issue (GSI) B-5 in NUREG-0933 (Reference 1), identifies two concems relating to cordamment design. First that sufficient information is not available to predict the behavior of two-way reinforced oorerete slabs; and second, that the structural design of a steel containment vessel subjected to unsymmetrical dynamic loadings may be govemed by the instability of the stell.

(1) Ductility of Tuo-Way Slabs and Stclls The first concern was originally identified in NUREG 0471 (Reference 2) and involved concern over the lack ofinformation related to the behavior of two-way reinforced concrete slabs loaded dynamically in biaxiti membrare tension (resulting from in-plare loads), flexure, and slear. If structures (concrete slabs) were to fail (floor collapse or wall collapse) due to loadmg caused by a losssf-coolant-accident (IDCA) or high-energy-line break (HELB), there would be a possibility that other portions of the reactor coolant system or safety-related systems could be damaged. Such loads would be caused by very concentrated high-energy sources causing direct impact on tie structures of concern. "Ihc damage could lead to an accident sequence resulting in the release of radioactivity to the em-ironment.

Because of NRC and itx!ustry concern, the American Concrete Institute addressed these dyname loads by  !

establishmg the nethodology identified in tie Appendix C Commentary to ACI 349-85 (Reference 3).  ;

l (2) Buckling Behavior of Steel Contamments j The second corcern, also identified in Reference 2, involves concern over the lack of a uniform, well-defined approach for design evaluation of steel contamments. The structural design of a steel contammert vessel subjected to unsymmetrical dynamic pressure loadings may be govemed by the instability of the i shell. For this type of loading, tle current design verification methods, analytical techniques, and tie acceptance criteria mav not be as comprehensive as they could be.Section III of the ASME Code (Reference 4) does not provide detailed guidance on the treatment of buckling of steel contamment vessels for such loading conditions.

l Moreover, this Code does not address the asymmetrical nature of the conramment shcIl due to the presence I of equipment hatch openings and other penetrations. Regulatory Guide 1.57 recommerris a mimmmn factor of safety of two animt buckling for the worst loadmg condition prwided a detailed rigorous analysis, considering in clastic behavior, is performed.

On tie other hand, the 1977 Summer Addendurn of tie ASME Code permits three altemate methods, but requires a factor of safety between 2 and 3 against buckling, depending upon applicable service limits. ,

i Hewever, NUREG-0933 states that the issue was resolved and no rew requiremerns were established.

ACCEPTANCE CRITERLA I The acceptaree criteria for part 1 of this issue is that the design of safety-related concrete structures shall meet tie ductility requirements of ACI 349, as endorsed by RG 1.142 (Reference 5).

71e acceptance criteria for part 2 of this issue is that the buckhng design of steel contamment vessels shall meet provisions of NE-3222 or code case N-284 of the ASM5 code.

RESOLUTION Tie design of ABWR safety-related concrete structures (otier than contammern)is based on ACI 349 as endorsed by RG 1.142. Part 1 of this issue is thus resolved for the ABWR.

August 23,1993 8

0 19R161 B.5

%e ABWR contamment is a reinforced concrete structure and it is designed according to ASMC-III, Division 2, Subsection CC. The steel components (not backed by concrete) of tic containment vessel are designed in accordance with to ASME-III, Subsection NE including the buckling provisions as stated in the acceptance criteria above. Part 2 of this issue is thus resolved for the ABWR.

REFEREhCES

1. NUREG4933,"A Trioritization of Gereric Safety Issues", (with Supplements) U.S. NRC, July 1991.

2. NUREG-0471, " Generic Task Problem Descriptions (Categories B, C, and D)", U.S. NRC, June 1978.

3. ACI 349-85," Code Requirements for Nuc! car Safety Related Structures", American Concrete Institute,1985.

4. ASME Boiler and Pressure Vessel Code,Section III, Division I, Subsection hT, American Society of ,

Mechanical Engmeers,1986.

5. Regulatory Guide 1.142,

• Safety Related Concrete Structums for Nuclear Power Plants (Other than Reactor Vessels and Containments) U.S. NRC, October 1981, Revision 1.

4 i.

August 23,1993 9

i 19 R 7 61.1 C-8; MAIN STEAM LINE I F AKAGE CONTROL SYSTEMS l ISSUE Dose calculations inchcated that operation of the main steam isolation valve leakage control system (MSIVLCS) required for some BWRs could result in higler offsite accident doses than if the system were not used and the integrity of the steam lines and condenser was mamtained. The calculations for accidents with a TID-14844 ,

(Reference 2) source indicated a potential increase in ofTrite teleases of ioene by two to three orders of mapitude l for pmper operation of a MSIVLCS, wlen compared to the calculations of releases assummg the steam system )

intact and MSIV leakage is eventually released through the condenser. Therefore, use cf the h*SIVLCS recommW in Regulatory Guide 1.% (Reference 3) could increase the overall risk to the public. After an (

extensive evaluation of alternative solutions,it was decided that Regulatory Gmde 1.96 was acceptable, and the I issue was resolved with no new requirements. (Reference 1.) i l

ACCEPTANCE CRITERg j

~Ihe acceptance criteria forhe resolution ofissue C-8 is provided in RG 1.%. RG 1.% states that the isolation  !

function of the MSIVs should h supplemented by a leakage control system (LCS), or if an alternative method is used it must be approved by the NRC staff. RG 1.% states that, a leakage control system is not required if the main  !

steam line leakage path can be relied on to remain intact and capable of prmiding significant dose reduction factors l for postulated accident conditions. j RESOLUTION s

1 The ABWR main steam line leakage path is designed to remain intact and capable of providing significant dose l reduction factors for postulated accident conditions. The design of the ABWR main steam leakage path is described in Subsection 3.2.5.3, " Main Steam Line Leakage Path." The main steam lines and all branch lines 2-1/2 inches in diameter and larger are designed to withstand the safe shutdown earthquake; the main steam and bypass lines at the turbine that are not safety-related, are analyzed to demonstrate their structural integrity under the safe shutdown i earthquake loadmg. The condenser anchorage is seismically analyzed to demonstrate that it does not fail. The  !

ABWR ahemative design approach has been resiewed by tie NRC staff. Therefore, upon approval of the j alternative design appmach. this issue is resolved ,

I REFERENCES

1. NUREG-0933,"A Prioritization of Geretic Safety Issues l' (wi:h Supplements) U.S. NRC, July,1991, i

2. TID-14844," Calculation of Distance Factors for Power and Test Reactor Sites," U.S. Atomic Energy Commission, March 23,1%2.

3. Regulatory Guide 1.%, " Design of Main Steam Isolation Valve Leakage Contml Systems for Boiling Water Reactor Nuclear Poa er Plants," U.S. NRC, June 1976.

August 23,1993 10

9 19R164 113; DYNAMIC OUAI TFICATION TESTING OF I_ARGE BORE HYDRAULIC SNUBBERR ISSUE Issue 113 in NUREG-0933 (Reference 1), addresses the need for requirements for dynamic qualification testmg oflarge bore hydraulic snubbers (>50 kips load rating). Qualification tests oflarge bore hydraulic snubbers typically utilize a shutoff valve in place of tie snubber control valve. To assure operability of the snubber contml valves when subjected to dynamic loads, testing slould be perfonned to determine the operational characteristics of the snubber contml valve.

ACCEPTANCE CRITTRIA The accq44s criteria for the resolution ofIssue 113 for the ABWR design are the performance of dynamic tests in accordance with SSAR Section 3.9.3.4.1 (3). 'Ihe dynamic load tests identified specifically for large bore hydraulic snubbers (LBHS) are to be performed in addition to the dynamic tests reqmred for mechanical and hydraulic snubbers.

RESOLUTION Tle ABWR design will have less than one quarter of tle number of snubbers at a typical BWR operating plant.

Mechanical and hydraulic snubbers will only be used for piping systems when dynanuc supports are required at locations where large thermal displacements prohibit the use of rigid supports.

Large bore hydraulic snubbers (LBHS) will only be used as piping restramts, they will rot be used in applications other than piping restraints. Mechanical and hydraulic snubbers ielumng LBHS are tested to insure that they can perform as rtquired during seismic and other dynamic loading events. These tests are described in SSAR section 3.9.3.4.l(3). Additional dynanuc cyclic load tests are required for LBHS to assure operability of the snubber control valves when subjected to dyrr.mic loads. 'Ihis requirement is specified in SSAR Section 3.9.3.4.l(3)(C).

Tie acceptance criteria for this issue are met, therefore, the issue is resolved for the ABWR design.

REFERENCES

1. NUREG-0933, *A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, December 1992.

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I August 23,1993 13 ,

19B.2.17 A-47: SAFETY IMPLICATIONS OF CONTROL SYSTEMS ISSUE This issue A-47 (Reference 1) concerns the potential for accidents or transients (e.g., overpressure, overfiling, reactivity events) being made more severe as a result of control syz, tem failures including control and instrumentation power support faults. Rese failures or malfunctions may occur independently or as a result of an accident or transient and u ould be in addition to any control system failure that may have initiated the event.

Although it is generally believed that control system failures are not likely to result in loss of safety functions which could lead to serious events or result in conditions that safety systems are not able to cope with, in-depth studies have not been performed. The purpose is to define generic criteria that may be used for plant-specific reviews.

ACCEPTANCE CRITERIA ]

The acceptance criteria for resolution is that the plant shall provide automatic reactor vessel overfill protection, and that plant procedures and technical specifications shall include provisions to verify periodically the operability of the overfill protection to assure that automatic overfill protection is available to mitigate main feedwater overfeed events during reactor power operation. Also, the system design and setpoints shall be selected with the objective of minimizing inadvertent trips of the main feedwater system during plant stanup, normal operation, and protection system surveillance.  !

RESOLUTION The reactor vessel overfill protection is described in Subsection 7.7.1.4(9),"Feedwater Control System", and l Figure 7.7 8,"Feedwater Control System IED" Plant procedures will be developed by the COL applicant. As a  ;

matter of good design practice for maximum availability, tLe feedwater system design and setpoints will be selected I to minimize inadvertent trips for all modes of operation. This system, with fault tolerant digital controllers and self- _;

test and on-line diagnostics, is described in Subsection 7.7.1.4. De BWR Owners Group improved technical  ;

specification (Reference 4) submittals oflimiting conditions for operations and surveillance requirements are consistent with the NRC resolution (e.g., Subsection 3.3.2.2,"Feedwater and Main Turbine Trip Instrumstation").

The ABWR resolution will follow the NRC-approved Owners Group submittals.

Therefore this issue resolved for the ABWR. ,

i REFERENCES i

1. NUREG-0933,"A Prioritization of Generic Safety issues," U.S. NRC (including Supplement 15).

2. NUREG-1217," Evaluation of Safety implications of Control Systems in LWR Nuclear Power Plants", June 1989. <

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3. Generic Letter 89-19," Request for Actioa Related to Resolution of USI A-47, Pursuant to 10CRF50.54(f)" U.S.

NRC, September 20,1989 i 1

4. NUREG 1433,"BWR/4 Standard Technical Specifications", U.S. NRC, September 1992.  !

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August 31,1993. 29

191L2,32 25 AUTOMATIC AIR IIEADER DUMP ON BWR SCPAM SYSTEM ISSUE This issue corecrns 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 slightly, thus filling the scram discharge volume with water but allowing little or no control rod movemern. Eventually, the rods will try to scram but the scram will be imparted in a manner similar to w hat happened at Brow ns Ferry Unit 3 on June 28,1980 (Reference 1). Meanw hile, the dropping air pressure can cause a transient (e g., via feedwater controller lockup) u hich w ould normally call for a scram.

ACCEPTANCE CRITERLA The acceptance criteria for this issue is specific to the scram discharge volume design and is not applicable to de ABWR. See the resolution discussion that follows.

RESOLUTION For the ABWR fine motion control rod drive (FMCRD) design, scram water is discharged through de drive directly into the reactor vessel. Tiere is no scram discharge volurne as used in previous BWR designs employing tic locking piston control rod dnvc (LPCRD). Consequently, the common mode failure or impairment of scram associated with loss of control air pressure and filling of the scram discharge volume is not applicable to the ABWR.

The analogous concern for the ABWR design is that the slow loss of control air pressure in the scram air header can allow some of de scram accumulators to leak to the reactor. This action could deplete de accumulators

• charge and impair or prevent their capability to scram the connected control rods, unless specific design features are provided to prevent or mitigate its occurrence. The ABWR design does provide protection against this event by incorporating the following features:

1. A scram air header low pressure alarm to alen tic operator of a low pressure condnion in the header.

(Rcrer to Figun. 4.6-8, Sheet 2.) The setpoint value is chosen to be gicater than the pressure at u hich tic scram valves could start to open in order to allow the operator the opporturuty to take corrective action.

2. A scram initiated by low pressure in the common header supplying the charging water to the scram accumulators. All tie accumulators will have sufTicient water volume to scram their associated control rods as long as tic CRD S3stem pump maintains the pressure in the charging header above the minimum required accumulator charging pressure, even if multiple scram valves are leaking. The pressure in the header will drop only if the total scram valve leakage flow is greater than the capability of the charging pump to provide make-up and maintain system picssure. If this should occur, instrumentation located in de charging leader u ill sense the loss of pressure and signal the RPS to initiate an immediate scram. The setpoint 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 u hile adequate accumulator charge is still available. (Refer to Subsection 4 6.1.2.4.3.)

In summary, tic ABWR incorporates design features to prevent die loss or impairment of the scram function due to a slow loss of control air in the scram system The 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 dow n tic plant before the accumulators are depleted. Therefore, this issue is resolved for the ABWR design.

REFERENCES

1. ' Report on the Brow ns Ferry 3 Partial Failure to 5 cram Event on June 28,1980," U.S. NRC, July 30,1980.

August 25,1993 51

191L2.35 Sh PROPOSED REOUIRE31ENTS FOR IMPROVING TIIE RFII ABILITY OF SERVICE WATER SYSTEMS ISSUE

!ssue 51 in NUREG-0933 (Reference 1), k!cntifies the susceptibility of the Station Senice Water System (SSWS) to fouling which leads to plant shutdowns arxl reduced power operation for repairs.

Tle SSWS cools the Component Cooling Water System (CCWS) through the Component Cooling Water Heat Exchangers and rejects tle leat to the ultimate heat sink (UHS) during normal, tansient, and accident conditions.

The CCWS in turn prmides cooling water to those safety-related components necessary to achieve a safe reactor shutdown, as well as to various non-safety rextor auxiliary components.

ACCEPTANCE CRITERIA Elimination of the possible effects of fouling of tic service water system and ultimate heat sinks is a design goal of tic ABWR. The Plant Designer is given specific requirements and guidance on achieving this goal, ireluding instruction to consider designs and new requirements which further mitigate the fouling effects. Additionally, the Plant Designer is directed to investigate the pmblem with ice as a flow blockage meclnnism and to dispose of and/or dissolve such ice as required.

RESOLUTION A review of operating plant experience shows that the mist prevalent problems with plant cooling water systems are due to tic corrosion and fouling caused by poor gaality senice water. In spite of a variety of water ueatnent schemes and use of expensive material, the wide range of harsh chermstry, silt and biological content result in a need for continuous maintenance of equipment. In order to make a significant operational imprm ement in this area, tie ABWR requirements for plant cooling water systems will Irelude the following (see Reference 2): ,

I (a) Direct service water will not be used for component cooling. A closed loop component cooling water l system will be utilized to transfer feat from the comporent beat loads via a leat exchanger to tie service water system and ultimate test sink. This minimizes the number of pieces of equipment which are in contact with the problem-causing senice water and focuses the problem on the component cooling water heat exchanger.

(b) 11e COL applicant shall treat raw senice water as needed to reduce tie effect of mud, silt, or orgamsms.

(c) The COL applicant shall provide materials for piping, pumps, and heat exchangers that offer greater resistance to the range of probable water clemistry conditions.

(d) 11e COL applicant shall make provisions to facilitate tie inspection of senice water piping and replace sections of piping during plant life.

The COL applicant shall provide sufficient redundancy of makeup pumps for tie ultimMe heat sink so that unkcup capabilities are not unduly reduced w hen one pump malfuretions. Tie need for a safety grade makeup shall be established in corjunction with establishing UHS water volume, as specified in Regulatory Guide 1.27 (Reference 3).

Tie COL applicant shall provide the safety related portions of tiese systems to meet the design bases dunng a loss of offsite power. These systems shall be designed to perform their cooling function umming a single active failure in any mechanical or electrical system.

August 23,1993 56

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4 19B.2.35 51  ;

l Tierefore, tids issue 51 is resolved for ABWR.

REFERENCES

1. NUREG.0933, "A Prioritization of Geretic Safety Issues", (with Supplements) U.S. NRC, July 1991.

2. Advanced Light Water Reactor Utility Requirement Document (Volume II), EPRI.

3. Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants, Revision 2, January 1976.

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4. Enclosure 1, Generic Ixtter 89 13," Service Water System Problems Affecting Safety-Related Equipment".

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August 23,1993 57 i

19B 2.29 C-17; IN1 TRIM ACCEPTANCE CRITERIA FOR SOr TDIFICATION AGENTS FOR RADIOACTIVE SOT TD WASTES ISS1!E NUREG-0471 Item C-17 (Reference 1) discusses the Interim Acceptance Criteria for Solidificatinn agents for radioactive solid wastes.

ACCFPTANCE CRITERLA The acceptance criteria for the resolution of C-17 is under development. This hTREG-0471 (Reference 1) task .

involves the development of criteria for acceptability of radwaste solidification agents to properly implement a process contml program for the packaging of diverse plant wastes for sin 11ow 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 (Refererre 2). BTP ETSB 11-3, on waste form has been developed under TMI Action Plan item IV.C.I. The ABWR is committed to meetmg the requirements in 10 CFR Part 61, Reference 3, (Subsection 11.4.1.2).

The COL applicant shall demonstrate that the wet waste solidification processes and the spent resin and sludge dewatering processes will result in products that comply with 10CFR61.56. A process control program (PCP) shall be provided for the processes employed.

This procedure will encourage the development and tise of additional acceptable methods of solidifying radioactive waste solids in the future. Thus this item las been iesolved for the ABWR.

REFERENCES

1. NUREG4471, " Generic Task Problem Descriptions (Categories B, C, arx1 D)," U.S. NRC, June 1978.

2. Memorandum for T. Speis from J. Funches, "Prioritization of Generic issues - Emironmental and Licenstng Improvements " Febmary 24,1983.

3. 10 CFR 61.56, " Licensing Requirements for Land Disposal of Radioactive Waste".

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i August 23,1993 68 l

191L2.43 89

STIFF PIPE CLAMPS ISSUE Issue 89 in hUREG.0933 (Reference 1), addresses the corrern 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, tiere is a concern that certain piping system conditions coupled with specific stiff pipe clamp design requirements could result in interaction effects that should be evaluated in order to determine the significance of the irduced pipe stresses.

"Ile ASME Section 111 Code (Reference 2), requires that, the effects of attachments in producing thermal stresses, stress concentrations and restramts on pressure retatmng members be taken into account in checkmg for compliaxe with stress criteria. Attachments to piping are generally categorized as integral and non-integral attachments. Lugs welded to the pipe wall are an example ofintegral n" Aments. Clamps used for nnaching hangers, struts arxl snubbers to the pipe by bolting are non-integral anachments. Piping design reports specincany address local stresses at integral attachments, such as lugs. Any additional stresses induced in tic pipe by non-integral, clamp bolted attachments, are not included in the piping design report.

ACCEPTANCE CRITERLA He acceptarre criteria for the resolution ofissue 89 is that the effect of stiff pipe clamps on piping stresses should be considered in tie piping system design. For stiff pipe clamps installed on straight runs of pipe or on bends with a radius of at least five pipe diameters, the pipe clamp induced stresses can be considered negligible and explicit consideration is not required. This acceptarre criteria is based upon analysis perfortred by GE.

In the 1980's, GE performed calculations for typical stiff pipe clamps used on BWR Main Steam and Recirculation piping systems. For each system, the stiff pipe clamps were installed on straight pipe or on bends with a radius of at least five pipe diameters. T!c purpose of these calculations was to evaluate the additional stresses at clamp locations due to the following:

1) Differential tiermal expansion of the pipe and clamp,

2) Discon:inuity stress in the pipe from internal pressure restraint,

3) Tiermal gradient through the pipe wall in the vicinity of the pipe clamp, and

4) Extemal loads due to dynamic events such as earthquake.

Maximum ixremental primary stresses were less than 25% of the primary stress allowables, and manmum ircremental secondary stresses were less than 40% of secordary stress allowables. The stresses at the clamp locations excluding clamp induced stresses were less than 30% of the ASME Section III code allowables. Tic total primary and secondary stresses, including clamp induced stresses, were less than 70% of allowable stress.

The govermag stress locations occurred at piping branch connections, cibows and shear lugs, ticy did not occur

at stiff pipe clamp locations. Tle stress intensification that occurs at cibows, brarch connections and shear lugs is much greater tinn that which occurs at stiff pipe clamps. Tierefore, when the additional clamp induced stresses are irrluded, the peak piping system stresses do not occur at tie clamp locations. Based or. these calculations, it was concluded that exp'icit consideration of clamp induced piping stresses is not required when tie clamps are installed on straight pipe or on bends with a radius of at least five pipe diameters.

RESOLUTION For the ABWR, the following stiff pipe clamp parameters will be very strmlar to those for tie BWR stiff pipe clamps evaluated in tie calculations summarued above:

Stiff pipe clamp geometry and material properties August 23,1993 71

Pipe schedule and material propenics Support rated loads less than or equal to 23 Tonnes Piping system operating pressures and temperatures and operating transients l Piping stre. ses at branch conrections and elbows much greater than at stificlamp locations.

Derefore, it can be concluded that the governing ABWR piping stresses will not occur at stiff pipe clamp .

locations. For the ABWR, the piping design specifications shall require that stiff pipe clamps be installed on straight !

runs of pipe or on bends with a radius of at least five pipe diameters. He pipe clamp induced stresses can then bc considered negligible and do not warrant explicit consideration. His issue is resolved for the ABWR. l REFERENCES

1. NUREG 0933,"A Prioritization of Generic Safety Issues" (with supplements), U.S NRC, July 1991.

2. American Society of Mechanical Engineering Boiler and Pressure Vessel Code,Section III i

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l 193.2.54 143: AVAILABILITY OF CIIILLED WATER SYSTE31S AND ROON1 COOLING l ISSUE In recent years, several nuclear power plants have experienced problems with 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 exist because of the desire to provide increased fire protection and the need to i avoid severe temperature changes in equipment control circuits. Since the Browns Feny fire, considerable effort has been expended to improve the fire protection of equipment required for safe shutdown. Generally, this improvement has been made by enclosing the affected equipment in small, isolated 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 i 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 contribute significantly to core-melt probability in certain plants. Because corrective measures are often taken at the affected plants once such failures occur, the impact of these failures on the proper functioning of air cooling systems may not been considered. Thus, plants with similar  ;

inherent deficiencies may not be aware of these problems.

Operability of some safety-related components is dependent upon operation of HVAC 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. (Reference 1) ,

ACCEPTANCE CRITERIA l

The impact ofloss of room cooling is an important design consideration for the ABWR. Under these circumstances, a key design objective is to ensure that ABWR safety-related equipment will still operate reliably during the period ofloss of room cooling. The following criteria will establish an acceptable ABWR design.

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1. An evaluation of the dependencies or non-dependencies of safety-related equipment on HVAC cooling shall be l performed. This evaluation will include assessments of room heat load and heatup rates, and establish  !

equipment operating conditions. Equipment ability to withstand these conditions without loss of function shall be established.

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2. For equipment found to be significantly dependent on HVAC cooling, an assessment of the HVAC system reliability shall be performed. PRA analyses will be carried out to assess plant risk and determine whether any modifications are necessary.

3. Corrective design measures shall be identified where necessary to reduce plant risk.

RESOLUTION ABWR design features which address the acceptance criteria include the following:

RCIC pump and turbine are designed to operate for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> without room cooling. This system will provide core cooling during a prolonged loss of HVAC cooling.

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August 30,1993 98 1

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t Operation ofother injection systems (HPCF, LPFL, RHR) is more dependent on the availability of room cooling. However, these systems are designed to operate for at least 10 minutes without room cooling.

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- Detailed design specifications for ABWR safety-related equipment will specify the room conditions under  ;

which equipment must operate without room cooling. Room heap assessments will be performed to  !

establish environmental conditions for equipment specification.

- Potential modifications including procedure changes or hardware changes evaluated through PRA analyses  :

to ensure acceptable plant risk.  !

t Despite the few extreme events which would cause loss of room cooling, the ABWR design incorporates several Safety-Related ilVAC systems which provide room cooling under most circumstances. These systems include:

Secondary Containment Safety-Related HVAC - providing 14 fan-coil units for safety-related equipment rooms, including the 3 divisions of ECCS pump rooms  !

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PJB Safety-Related Electrical Equipment ilVAC - 3 divisions each with 2-100% supply / exhaust fans. I air  !

conditioning unit. 1

- R/B Safety-Related Diesel Generator HVAC - 2 supply fans per division.

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- HVAC Emergency Cooling Water 3 divisions - provides chilleo water to R/B electrical equipment HVAC, C/B HVAC, and C/R habitability IIVAC.

The reliability and availability of these safety-related HVAC systems will be specified in detailed design to ensure a controlled environment for operation of safety-related equipment.

REFERENCES

1. NUREG-0933 "A Prioritization of Generic Safety issues"(with supplements), U.S. NRC, July 1991.

August 30,1993 99 l I

19B.2.57 153 LOSS OF ESSENTIAL SERVICE WATER IN LIGHT-WATER REACTORS 1.SS UE The Essential Service Water (ESW) system at a nuclear power plant supplies cooling water to transfer heat from  ;

various safcty-related and non-safety-related systems and equipment to the ultimate feat sink of the plant. Urder  !

Issue 153, the staff will examme all potential causes for ESW system unavailability, except those that are considered to be resolved by implementing the resolutions addressed in (Generic Ixtter (GL) 89-13 (Refererce 1), such as biofouling, sediment, cormsion, and crosion (Issue 51). The safety concems of this issue include partial or complete loss of ESW system functions resulting from common causes (such as icing of the intake stmeture), degradation of the ESW system, design deficiencies, ard procedural or maintenance errors. A complete loss of the ESW system could lead to a core-melt accident, posing a significant risk to the public.

The NRC evaluation of this issue has not yet been completed.

ACCEPTANCE CRITERIA Tie ESW system is needed in every phase of plant operations ani under accident conditions, supplies adequate cooling water to systems and comporents that are important to safe shurdown or to mitigate the consequences of the ,

accident. Under rurmal operating condition, tic ESW system provides component and room cooling (mamly via t!e i component cooling water system). During shutdown it also ensures that the residual teat is removed from the reactor core. The ESW system may also supply makeup water to fire protection systems, cooling towers, and treatnent systems at a plant.

The design features for the essential service water (ESW) system are summarized as follows:

1. Performance Requirements

- The ESW system will be designed to meet the required leat loads. ,

- The ESW system will be pmvided with two pumps ard two heat exchangers per division.  ;

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- The plant designer will provide analyses for all potential operating corditions that properly account for uncertainties.

2. System Arrangement

- Tie ESW system will be divided into approximately three equal-sinxi divisions.

- A division will be made up of irdependent piping systems, each with pumps, heat exchangers, strainers, controls and instnimentation, power supplies, and associated equipment required for i regulating system flow. I In addition, the ESW design shall address partial or complete loss of ESW system furetions resulting from common causes, degradation of tie ESW system, design deficiercies, ard procedural or maintenance errors. The plant desigrer should provide an assessment of these potential failure modes and their associated contributiora to the core damage frequerry ard should identify dominant accident sequerces.  ;

RESOLUTION y 1

i The ABWR Reactor Service Water (RSW) system removes feat from the Reactor Building Cooling Water (RCW) system ard transfers that feat to the Ultimate Heat Sink (UHS). The RSW system is provided in three i divisions. Each division has two pumps which serd cooling water to three RCW heat exchangers. Normally one j pump and two heat exchangers are operating in each division. When leat removal seguirements increase, the i remaining pump and leat exchange are automatically put into operation. If additional heat removal capacity is needed, some of tie ron-safety-related cooling loads may be taken out of operation.

August 26,1993 101 l

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19B.2.57 153 In case of failure w hich disables any of the three RSW divisions, tie other two divisions meet plant safety shutdown requirements. (Subsections 9.2.15 and Table 9.2-5).

11e ABWR RSW system divisions are physically and electrically separated from each otter. This reduces the potential effects of common causes. Normally, each division is operating at all times with the capability to put into senice the remaining pump and heat exchanger at any time. Margin is pro ided in pump flow capacity (and in RCW leat exchanger heat removaP.apacity). Periodic testing of tiese comporents will be performed and corrective action taken when reeded (Subsections 9.2.11.4 and 9.2.15.1.4).

Several potential causes of RSW system degradation are site dependent. The RSW system is designed to prevent this degradation from occurring Additionally, the COL applicant will provide de following system design features for those portions of the system uhich are not the AB%R standard plant scope: adequate NPSH for the pumps at low UI16HS water levels, low point drains and high point vents, presention of organic fouling (using methods such as trash racks, biocide treatment or tiermal backwashing, as required), component material selection suited to site water conditions, and protection agaimt floodmg, spraying, steam impingement, pipe whip, jet forces, missiles, fire and de effect of failure of any non-Scismic Category I equipment. If required, recirculation of warm water duuugh de intake stmetures will be provided to reduce the likehhood that ice will block cooling water flow. 1 (Subsections 9.2.5.4 and 9.2.15.2). Also system degradation is minimi7ed by periodic testing and inspection to insure integrity and functional capability. (Subsections 9.2.111.4 and 9.2.15.1.6).

Tic RSW pumps and pump house will be designed by the COL applicant, w to will consider and reduce the efTects of procedural arx! maintenance errors.

When the future plant-specific design is prepared, anoder assessment will be made of potential failure modes and deir associated contributions to the core damage frequency and tie donunant accident sequences will be identified.

"Ilese issues are resolved for tie ABWR through the design features of de RSW system, the system design feannes, and de Operational Reliability Assurance Activities (Subsection 17.3.9) w hich will be provided by de COL applicant. ,

l REFERFNCES l

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1. Generic Letter 89 13, Sen ice Water System Problems Affecting Safety-Related Equipment, July 15,1989. l l

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l 19.B.2.59 A-17 SYSTEMS INTERACTIONS IN NUCI EAR POWER PLANTS ISS!!E Unresolved Safety issue (US1) A-17 in NUREG-0933 (Reference 1) addresses the concem that inconspicuous or unanticipated interdependencies may exist between systems arxl may result in a degradation of the predicted capabihty of saf tvstems in an accident or transient. in particular from flooding and water intrusion In its regulatory analysis in NUREG-1229 (Reference 4). the NRC corcluded that for future plants the existing Standard Review Plans (SRPs) (Reference 5) in general cover the Assessment of System Interactions (Asis ) of concern. cxcept for the areas of internal flooding 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 and references NRC Information Notices regarding operating plant experiences. The NRC plans to develop an SRP relative to flooding and water intrusion, but otherwise not issue new requirements. la the ,

meantime, the NRC recommends that plant designers keep current on lessons leaned from operating experience as I

<. ported in LERs. and that the Probabilistic Risk Assessment (PRA) required for a future plant be also considered as ,

a tool to help uncover fla9 ding and water intrusion ASIS.

l ACCEPTANCE CRITERIA Re acceptance entcQn for the resolution of USI A-17 u that attention shall be paid in the detailed plant design to detecting and minimizing ti potential for ASIS due to th effects of floofng and water intrusion from internal plant sources, such as the incidents at operating plants refer. ced in NUREG-1174. The objective is to prescive the means for reachir.g 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.

Tley may also be influerced by build 6g design features.

For the ABWR design consideration is given to identifying flooding and water intrusion possibilities u hich are not covered by current SRPs. as discussed in NUREG-1174. These es ents include water or moisture relcase from sources internal to plant stmetures and evaluations of flooding events, by floor and compartment, for the Reactor Building, Con:rol Building. Service Building. Turbine Building, and Radwaste Building (as described in Sectio i 3.4,

" Water Level (Flood) Design"). All piping and compotents with flooding potential are considered. Tir > daations also consiocr the design features that provide protection such as sump and flood level alarms, water tight doors, sump pumps. elevated mounting of conipment, leak descetion systems dnp-prooimotors, and NEMA Type 4 motor control centers.

This evaluation has been made for each of the interaction incidents resulting from water intrusion at operating l plants described in the NRC Information Notices referenced in NUREG-1174. to identify the featurcs of the ABWR l desigr' u hich should ensure prevention of a similar interaction. I The analytical models developed for the ABWR design PRA (Section 19 Appendix R "Probabilistic Flooding Anal3sis") cvahiates the irnpact of micrnal flooding or water intrusion and determines ile risk (low). This analysis af identiDes equ.pment important to reducing flood risk for tic Reliability Assurance Program (RAP).

1 To summarize, the design process for the ABWR design takes into account the possibility for interaction  !

between water flooding ana other systems to occur that may degrade plant safety but are not easily recognizable. To tic extent practicable, ticsc interactions arc analyzed. and their impact on safety is evaluaicd. This issue is t!erefore .

resolved for the ABWR design.

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19.B.2.59 A-17 REFERENCES ,

l. NUREG41933, "A Pnontuation of Generic Safety Issues" (uith Supplements), U. S. NRC, July 1991.

2. NUREG/CR-3922, "Suney and Evaluation of S 3stem Interacuon Events and Sources", U. S NRC,  ;

January 1985.

1

3. NUREG/CR-4261, " Assessment of System Interaction Experictce at Nuclear Powcr Plants". U. S. NRC, June 1986.

4. NUREG-1299, " Regulatory Analysis for Resolunon of USl A-17", U. S NRC, August 1989,

5. NUREG4Wxt. " Standard Res iew P!an for the Review of Safety Analysis Reports for Nucicar Power Plants -  ;

LWR Edition". U. S. NRC.  ;

) 6. NUREG-1174. " Evaluation of S3stems Interaction in Nuclear Power Plants - Technical Findings Related to Unresolved Safety Issue A-17" U S. NRC. May 1989.  ;

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l Augmt 25,1993 106 u .

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1 1911.2.63 82: BEYOND DESIGN BASIS ACCIDENTS IN SPENT FUEL POOLS ISSUE Issue 82 in NUREG4933 (Rebrence 1), addresses the potential for s beyond-design-basis accident in u hich the l u ater is drained out of the spent fuel pool. In such an es ent the disclurged fuel from the last two refuelings may have sufficient decay heat to melt, ignite the zircaloy cladding and release fission products to tie atmosphere.

ACCEPTANCE CRITERIA I 1

The acceptance criteria for the resolution ofissue 82 is that the design of the spent fuel pool, storage racks. fuel j pool cooling and cleanup system and the load hark!!ing equipment in tie spent fuct pool area shall meet applicable i current requirements, i.e., tic g16 dance of the Standard Review Plan (SRP) Sections 9.1.2 - 9.1.5 (Reference 2) and I Regulatory Guide 1.13 (Reference 3).

RESOLUTION  !

I The ABWR design includes a spent fuel storage facihty. a fuel pool cochng argl cleanup system and a fuel handling system that meets the intent of Regulatory Guide 1.13 and SRP 9.1.2 <> l.5 as desenbed in Section 9.1, Fuel Storage and liandling Since the acceptance cntena are met for the spent fuel storage facility, this issue is resched for the ABWR.

]

REFE RENCES )

1. NUREG4933. "A Pnontization of Generic Safety issues,"(and Supplements 1-12). July 1991.

l

2. NUREG-0800," Standard Res icw Plan for the Review of Safety Analysis Reports for Nuclear Pow er Plants. l LWR Edition." l 1

3. Regulator? Guide 1.13.

• Design Objectis e for Light-Water Reactor Spent Fuel Storage Facilitics at Nuclear

, Power St sions." Revision 2. December 1981.

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