ML20197J380

Forwards Suggested Response to DW Mazur 850215 Question 2 Re Future Licensability of Facility.Response Items Based on Commission 1985 Policy & Planning Guidance.Nj Palladino 840420 Ltr Also Forwarded

ML20197J380
Person / Time
Site:	Columbia, Washington Public Power Supply System
Issue date:	03/19/1985
From:	Minners W Office of Nuclear Reactor Regulation
To:	Adensam E Office of Nuclear Reactor Regulation
References
CON-WNP-0792, CON-WNP-792, TASK-***, TASK-OR, TASK-TM NUDOCS 8503260594
Download: ML20197J380 (18)

Text

MAR 191905 NOTE TO: Elinor Adensam, Chief, LB-4/DL FROM: Warren Minners, Chief, SPEB/ DST

SUBJECT:

CONTRIBUTION TO RESPONSE TO WPPSS We offer the enclosed write-up as a suggested response to Question 2 in the letter from D. W. Mazur, Washington Public Power Supply System, to W. J. Dircks, NRC, dated February 15, 1985, on the subject of " Future Licensability of WNP-1 and WNP-2."

The five numbered items in our suggested response are based on the Commission's 1985 Policy and Planning Guidance.

The generic safety issue status tables (Tables 3 and 4A) come from the

" Generic Issues Management Control System: First Quarter FY 85 Update" (memo for H. Denton from T. Speis, 1/18/85).

Some of the columns on Tables 3 and 4A are not needed for the present purpose and could be deleted. These are (a) the third and fifth columns in Table 3 (" Priority" and "Date Approved for Resolution") and (b) the third, fifth, and sixth columns of Table 4A (" Lead Division / Branch," "Date Approved for Resolution," and " Resolution Date as of June FY-84").

We request that DST be on concurrence for the completed reply.

Please let George Sege (ext. 24609) or me know if we can be of further assistance.

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Enclosure:

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Suggested Response to WPPS Question 2. (Includes a main write-up and its enclosures.)

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SUGGESTED RESP 0flSE TO WPPSS QUESTI0tl 2 Question 2:

Is there concern on the part of the NRC that future regulatory changes could seriously challenge the licensability of the plants (B&W 205 and CE System 80) prior to their completion (assuming completion is not until early 1990s)?

Response

We cannot at this time foresee any radical changes in regulatory requirements over the next few years that would in themselves seriously challenge the licensability of WNP-1 and WNP-2. It should be recognized, however, that changes in requirerrents, potentially affecting WNP-1 and WNP-3 along with other reactors, may be developed in the intervening years in the normal course of regulatory activity. Examples of possible sources of future regulatory change include the following:

1. Development of procedures for proceeding with projects should a utility desire to reactivate a project after construction and licensing have stopped.

The Commission has directed the staff to develop such procedures by the end of 1985, and to be prepared to deal with possible restoration of construction on deferred plants by FY 1987.

2. Implementation of a safety goal, upon completion of current evaluation efforts.

3. Resolution of current unresolved generic safety issues and implementation of the resolution of ones recently resolved. A current list of such issues that may apply to WNP-1 and WNP-3 is enclosed (Enclosure 1).

/

2 Resolution of a generic issue does not, of course, necessarily mean imposition of a new requirements. Indeed, in some cases, reductions in requirements may result.

4. Implementation of a policy an severe accidents, upon completion of development.

5. Improvements in construction quality assurance policy.

A report bearing on this subject (NUREG-1055) was recently submitted to the Congress and issued for public ccmment. We enclose the overview section of this lengthy report, along with Chairman Palladino's trans-mittal letter (Enclosure 2). The public comments have now been received.

The staff has analyzed the comments and prepared a detailed quality assurance implementation plan, taking the comments into account. The plan is currently before the Commission for con' sideration.

Should any new requirements arise. before completion of licensing of WNP-1 and WNP-3, the granting of operating licenses for those plants would be subject to selective application of such new requirements in accordance with the backfit rule, or amended or successor rules ar.d policies regarding backfit of plant-specific or generic requirements.

Enclosures:

1. Generic Safety Issues Recently Resolved and Scheduled for Resolution (First Quarter FY-85)

2. Excerpt from NUREG-1055 (May 1984):

title sheet to page x

---W

N ENCLOSURE 1 GENERIC SAFETY ISSUES RECENTLY RESOLVED AND SCHEDULED FOR RESOLUTION First Quarter, FY-85

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TABLE 3 GENERIC SAFETY ISSUES RESOLVED BY FISCAL YEAR Issue Type of Date Approved Date Number Title Priority; Resolution for Resolution Resolved GENERIC SAFETY ISSUES RESOLVED FY-83

~

A-11 Reactor Vessel Materials USI GL 82-26 01/79 10/82 Toughness A-16 Steam Effects on BWR Core Spray HR MPA-012 NRR-0P FY83 03/29/83 Distribution A-39 Determination of' Safety USI SRP Revision 01/79 10/82 Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containment B-53 Load Break Switch NR SRP Revision NRR-0P FY83 07/28/83 II.E.5.1 (B&W) Design Evaluation NR No NRR-OP FY83 03/21/83 Requirement ,

IV.C.1 Extend Lessons Learned From NR No NRR-0P FY83 04/15/83 TMI to Other NRC' Programs Requirement GENERIC' SAFETY ISSUES RESOLVED FY-84 12 BWR Jet Pump Integrity Medium No NRR-0P FY83 09/25/84 Requirement 20 Effects of Electromagnetic NR NUREG/ NRR-OP FY83 11/15/83 Pulse on Nuclear Plant Systems CR-3069 40^^^ Safety Concerns Associated NR MPA-865 07/18/83 12/27/83 With Breaks in the BWR Scram System As of First Quarter FY-85

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Ttbla 3 (Crntinuid)

GENERIC SAFETY ISSUES RESOLVED BY FISCAL YEAR Issue Type of Oate Approved Date Number Title Priority Resolution for Resolution Resolved 45*** Inoperability of Instruments NR SRP Revision 09'/08/83 03/23/84 Oue to Extreme Cold Weather. ,

50*** Reactor Vessel Level NR MPA-F26 07/28/83 09/06/84 Instrumentation in BWRs 69** Make-Up Nozzle Cracking in NR MPA- 12/06/83 09/27/84 B&W Plants A-1 Water Hammer USI SRP Revision 01/79 03/15/84 A-12 Steam Generator and Reactor USI SRP Revision 01/79 10/83 Coolant Pump Supports B-10 Behavior of BWR Mark III High SRP Revision NRR-OP FY83 09/10/84 Containments B-26 Structural Integrity of Medium No NRR-OP FY83 09/27/84 Containment Penetrations Requirement B-60 Loose Parts Monitoring NR GL NRR-0P FY83 09/25/84 Systems (LPMS)

I.A.1.4 Long-Term Upgrading of NR New Rule NRR-OP FY83 01/01/84 Operating Personnel II.A.1 Siting Policy Reformulation Medium No NRR-OP FY83 09/20/84 Requirement II.E.5.2 (B&W) Reactor Transient NR NUREG-0667 HRR-OP FY83 09/28/84 Response Task Force -

III.D.2.5 Offsite Dose Calculation Manual NR NUREG/ NRR-0P FY83 01/17/84

-CR-3332 As of First Quarter FY-85 l .

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Trbla 3 (Continued)

GENERIC SAFETY ISSUES RESOLVED BY FISCAL YEAR Issue Type of Date Approved Date Number Title Priority Resolution for Resolution Resolved GENERIC SAFETY ISSUES RESOLVED 1st QUARTER FY-85 22 Inadvertent Boron Dilution NR GL 11/05/82 10/15/84 Events A-41 Long Term Seismic Program Medium No NRR-OP FY83 10/10/84 Requirement B-54 Ice Condenser Containments Medium NUREG/ NRR-OP FY83 10/22/84 CR-4001 1.G.2 Scope of Test Program Medium No NRR-OP FY83 10/05/84 Requirement

~

As of First Quarter FY-85

Trbis 4A NRR GENERIC SAFETY ISSUES ScilEDULED FOR RESOLUTION Lead Date Resolution Current Issue Division / Approved for Date as of Number Resolution Title Branch Priority Resolution June FY-84 Date 14 PWR Pipe Cracks DE/MTEB NR NRR-OP 12/84+

, 06/85+

23 Reactor Coolant Pump Seal Failures DE/EQB 11 07/86 07/06

.NRR-0P 29 Bolting Degradation or Failures in DE/MTEB 11 HRR-OP 09/85 11/85 Nuclear Power Plants 36** Loss of Service Water (Calvert DSI/ASB NR 02/15/84 03/85 05/86 Cliffs Unit 1) .

48** LCO for Class 1E Vital Instrument DSI/ICSB NR 10/25/83 02/86 02/86 Buses in Operating Reactors 49** Interlocks and LCOs for Class IE DSI/PSB M 07/31/84 TBD 09/88 Tie Breakers 51*** Proposed Requirements for Improving DSI/ASB M 06/03/83 10/86 09/88 Reliability of Open Cycle Service Water Systems __

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61** SRV Discharge Line Break Inside The DSI/CSB M 11/30/83 03/87 03/87 Wetwell Airspace of BWR Mark I &

Mark II Containments 65*** Component Cooling Water System DE/EQB 11 .07/28/83 07/86 07/86 Failure 66** Steam Generator Requirements DL/0RAB NR 09/85 11/16/83, 09/85 68** Loss of AFWS Due to AFW Steam IIELB DSI/ASB ll 04/24/84 10/87 10/87 70** PORV and Block Valve Reliability DSI/RSB M 05/14/84 TBD 03/85+

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As of First Quarter FY-85

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• Lead Date Resolution Current Issue Division / Approved for Date as of Resolution Number Title Branch Priority Resolution June .5Y-84 Date 77*** Flooding of Safety Equipment DSI/ASB H 09/08/83 ~08/86 08/86 Compartments by Back-Flow Through Floor Drains

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79*** Unanalyzed Reactor Vessel Thermal DE/MEB M 07/25/83 02/86 03/85+

Stress During Natural Convection Cooldown 82** Beyond Design Bases Accidents in DSI/AEB M 12/07/83 09/86 09/86 Spent Fuel Pools 83** Control Room Habitability (CRH) DSI/MTEB NR 08/83 '10/84+ 07/85+

86* Long Range Plan for Dealing With DE/MTEB NR 10/01/84 TB0 05/85+

Stress Corrosion Cracking in BWR ~

Piping 93* Steam Binding of Auxiliary DSI/ASB H 10/19/84 TBD 08/87 Feedwater Pumps A-3,'4 Steam Generator Tube Integrity DL/0RAB USI 01/79 12/84 06/85 i and 5 A-17 Systems Interaction DST /GIB USI 01/79 03/86 03/86 A-29 Nuclear Power Plant Design for the DSI/ASB M NRR-0P 06/86 06/86 Reduction of Vulnerability to Industrial Sabotage -

A-30 Adequacy of Safety Related DC Power DSI/PSB H NRR-0P' 07/85 07/85+

Supplies A-40 Seismic Design Criteria DE/SGEB USI 01/79 04/85 07/85 As of First Quarter FY-85 I

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Tebla 4A (ctntinuid)

• Lead Date Resolution Current Issue Division / Approved for Date as of Resolution Number Title Branch Priority Resolution June FY-84 Date A-43 Containment Emergency Sump DST /GIB USI 01/79 09/84 05/85 Performance p

A-44 Station Blackout ,

DST /GIB USI 01/79 02/85 01/86 A-45 Shutdown Decay Heat Removal DST /GIB USI 02/81 02/86 02/86 Requirements A-46 Seismic Qualification of Equipment DST /GIB USI 02/81 02/85 08/85 in Operating Plants A-47 Safety Implications of Control DST /GIB USI 02/81 04/86 06/86 Systems A-48 Hydrogen Control Mease.;s and DST /CIB USI 02/81 06/86 06/86 Effects of Hydrogen w as 'n Safety Equipment A-49 Pressurized Thermal Shock DST /GIB USI 12/81 12/85 03/86 B-5 Ductility of Two Way Slabs and DE/SGED M NRR-0P 09/86 09/86+

Shells and Buckling Behavior of Steel Containments B-6 Loads, Load Combinations, Stress DE/MEB H NRR-0P 10/84+ 04/85+

Limits -

B-55 Improve Reliability of Target Rock DE/MEB M NRR-0P 05/85 05/85 Safety Relief Valves B-56 Diesel Reliability DSI/PSB H NRR-0P 01/85 01/85+

B-58 Passive Mechanical Failures DE/EQB M NRR-0P 10/84+ 10

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As of First Quarter FY-85

Tcble 4A (centinued) .

Lead Date Resolution Current' Issue Division / Approved'for Date as of Resolution Number Title Branch Priority Resolution June FY-84 Date B-61 Allowable ECCS Equipment Outage DST /RRAB M NRR-OP 09/87 09/87 Periods ,

B-f54 Decommissioning of Nuclear Reactors DE/CMEB NR NRR-0P 09/85 09/86+

C-8 Main Steam Line Isolation Valve DSI/ASB .H NRR-0P 12/86 12/86 Leakage Control Systems

.C-11 Assessment of Failure and .DE/MEB. M NRR-OP 10/84+

Reliability of Pumps and Valves As of First Quarter FY-85 a

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Table 48 -

NRR TMI GENERIC SAFETY ISSUES SCHEDULED FOR RESOLUTION Lead Date Resolution Current Issue Division / Approved for Date as of Resolution ,

Number Title Branch Priority Resolution June FY-84 Date I.A.2.2 Training and Qualifications of DHFS/LQB H NRR-0P 04/85 04/85+

Operations Personnel I.A.2.6 Long-Term Upgrading of Training DHFS/LQB H NRR-0P 04/85 08/85+

1 (1) and Qualifications - Revise Regulatory Guide 1.8 I.A.2.6 Operator Workshops DHFS/LQB M M9R-OP 09/84 03/85 (4)

I.A.2.7 Accreditation of Training DHFS/LQB M NRR-0P 01/85 01/87

-Institutions I.A.3.4 Licensing of Additional Operations DHFS/LQB M NRR-0P 09/84 03/85 Personnel I.A.4.2 Review Simulators for Conformange DHFS/0LB H NRR-OP 04/85 04/85+

(4)

I.B.I.1 Organization and Management of # # # # #

Long Term Improvements I.B.1.1 Prepare Draft Criteria DHFS/LQB (1) .

I.B.1.1 Prepare Commission Paper DHFS/LQB (2)

1. Human Factors Program Plan Issues No. 6.1 and 6.2 have replaced TMI Issues 1.B.1.1 (1 through 4) ,

As of First Quarter FY-85 '

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Lead Date Resolution Current Issue Division / Approved for Date as of Number Resolution Title Branch Priority Resolution __; June FY-84 Date I . B.1.1 Issue Requirements for the DilFS/LQB - -

.(3) Upgrading Management and '

Technical Resources I.B.I.1 Review Responses t'o Determine DHFS/LQB (4) Acceptability llFPP Establish NRC Position on Manage- DilFS/LQB M llFPP . 11/85+ 04/86+-

6.1 ment and Organization at Operating Nuclear Power Plants liFPP Development of HRC Assessment Mate- DilFS/LQB H HFPP 02/86+ 08/86

6. 2 rials for Evaluating Management and Organization at Nuclear Power Pl. ants I.C.9 Long-Term Program Plan for Upgrad- DilFS/PSRB M NRR-0P 05/85+ 07/85 ing of Procedures I.D.3 Safety System Status Monitoring DHFS/HFEB H NRR-0P 09/84+ 03/85+

II.B.6 Risk Reduction for Operating DST /RRAB 11 NRR-0P 11/84 ACA Reactors at Sites with liigh Population Densities .

II.C.2 Continuation of Interim Reliability DST /RRAB H -

ACA 03/85 Evaluation Program II.E.4.3 (Containment) Integrity Check DSI/CSB H NRR-0P 07/84 03/85+

II.E.6.1 Test Adequacy Study DE/MEB M NRR-0P 11/85 , 11/86

1. fluman Factors Program P.lan Issues No. 6.1 and 6.2 have replaced THI Issuss 1.B.1.1 (1 through 4)

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A's of First Quarter FY-85 ce

Table 4B (c:ntinuid) .

Lead Date Resolution Current Issue Division / ' Approved for Date as of Resolution Number Title Branch Priority Resolution June FY-84 Date III.D.2.3 Develop Procedures to Discriminate '

DE/EHEB N1 NRR-0P 09/84 07/85 (1) Between Sites / Plants III.D.2.3 Discriminate Between Sites and DE/EHEB N1 NRR-OP 09/84 07/85 (2) Plants that Require Consideration of Liquid Pathway Interdiction

. Techniques III.D.2.3 Establish Feasible Method of DE/EHEB N1 NRR-0P 09/84 07/85 (3) Pathway Interdiction III.D.2.3 Prepare a Summary Assessment DE/EHEB .N1 NRR-OP 09/84 07/85 (4) .

III.D.3.1 Radiation Protection Plans DSI/RAB H NRR-OP 03/85 09/85 .

IV.E.5 Assess Currently Operating ' Plants DL/SEPB H NRR-0P 07/85 03/85

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l ENCLOSURE 2 NUREG-1055 For Comment Improving Quality anc t7e Assurance of Quality in the Design and Construction of Nuclear Power Plants A Report to Congress ate u hed ay1b W.'Altman, T. Ankrum, W. Brach Division of Quality Assurance, Safeguards, and inspection Programs

, Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555

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r ABSTRACT At the request of Congress, NRC conducted a study of existing and alternative programs for improving quality and the assurance of quality in the design and construction of commercial nuclear power plants. A primary focus of the study was to determine the underlying causes of major quality-related problems in the construction of some nuclear power plants and the untimely detection and correction of these problems. The study concluded that the root cause for major quality-related problems was the failure or inability of some utility managements to effectively implement a management system that ensured adequate control over all aspects of the project. These management shortcoming arose in part from inexperience on the part of some project teams in the construction of nuclear power plants. NRC's past licensing and inspection practices did not adequately screen construction permit applicants for overall capability to manage or. provide effective management oversight over the construction project.

The study recommends a number of improvements in' industry and NRC programs.

For industry, the study recommends self-imposed rising standards of excellence, treatment of quality assurance as a management tool, not a substitute for management, improved trend analysis and identification of root causes of quality problems, and a program of comprehensive third party audits of present and future construction projects. To improve NRC programs, the study recom-mends a heavier emphasis on team inspections and resident inspectors, an enhanced review of new applicant's capabilities to construct commercial nuclear power plants, more attention to management issues, improved diagnostic and trending capabilities, improved quality and quality assurance for operating reactors, and development of guidance to facilitate the prioritization of quality assurance measures commensurate with the importance of plant struc-tures, systems, and components to the achievement of safety.

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g UNITED STATES NUCLEAR REGULATORY COMMISSION 3 .' ( .. j WASHINGTON,0, C. 20555 v/

MAR 21 1985 Docket No. 50-397 MEMORANDUM FOR: A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing FROM: John 0. Bradfute, Project Manager Licensing Branch No. 2, Division of Licensing

SUBJECT:

WNP-2 PUMP AND VALVE INSERVICE TESTING DATE & TIME: April 16 - 17,1985 7:30 a.m. - 5:00 p.m.

LOCATION: Room P-114 Phillips Building Bethesda, Maryland PURPOSE: To discuss preparation of Pumps and Valves Inservice Testing and Safety Evaluation Report for Washington Nuclear Project No. 2 (WNP-2). The agenda is enclosed.

PARTICIPANTS *: NRC WNP-2 EG&G F. C. Cherny P. L. Powell H. Rockhold J. D. Page T. F. Hoyle C. Ranson R. Auluck M. T. Reis R. Lyon J. O. Bradfute R. P. Wolfgram.

f John 0. Bradfute, Project Manager Licensing Branch No. 2 Division of Licensing

Enclosure:

As stated cc: See next page

• Meetings between NRC technical staff and applicants for licenses are open for interested members of the public, petitioners, intervenors, or other parties to

' attend as observers pursuant to "Open Meeting Statement of NRC Staff folicy",

43 Federal Reaister 28058, 6/28/78. ,

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'E I 46 CK 05 66597 I7ff - - -

P PDR

WNP-2 Mr. G. C. Sorensen, Manager Regulatory Programs Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 cc: Nicholas Reynolds, Esquire Bishop, Cook, Liberran, Purcell & Reynolds 1200 Seventeenth Street, N. W.

Washington, D. C. 20036 Mr. G. E. Doupe, Esquire Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99532 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council Mail Stop PY-11 Olympia, Washington 98504 P. L. Powell, Licensing Manager Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Mr. W. G. Conn, SR. N/M Group Supervisor Burns and Roe, Incorporated 601 Williams Boulevard Richland, Washington 99352 R. B. Glasscock, Director Licensing and Assurance Washington Public Power Supply System P. O. Box 968, MD 650 Richland, Washington 99352 M. J. D. Martin WNP-2 Plant Manager Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352

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ATTACHl4E NT WASHINGTON PUBLIC POWER SUPPLY SYSTEM PROJECT, UNIT 2 PUMP AND VALVE INSERVICE TESTING PROGRAM

, QUESTIONS AND COMMENTS

1. VALVE TESTING PROGRAM 1 A. General Questions and Comments

1. The cur rent NRC position is that rapid-acting valves are identified as those valves with a limiting value of full-sticke time of i seconds or less, and the requirements of IWV-3417(a) need not be applied to these valves. However, these valves must be verified to stroke in less than .

2 seconds or the corrective actions of IWV-3417(b) must be met. WNP-2 relief request No. RV-1 is affected by this NRC staff position.

, 2. The limiting value of full-stroke time should be provided in the WNP-2 IST program for all active power operated valves included in the IST program. -

3. Simple check valves should be categorized "C" instead of "B-C", since category "B" applies only'to valves that can be mechanically cperated.

4. What plant conditions could preclude quarterly testing of valves identified to be tested during "ALL" plant operating modes as defined on page 4-6 of the WNP-2 IST program?

5. The NRC has concluded that the applicable leak test procedures and l requirements for containment isolation valves are determined by 4

10CFR50, Appendix J. Relief from Paragraphs IWV-3421 through -3425 for containment isol~ation valves presents no safety problem since the

! intent of IWV-3421 through -3425 is met by Appendix J requirements, however, the licensee shall comply with Paragraphs IWV-3426 and -3427.

This NRC position affects WNP-2 relief request No. RV-4.

6. Provide clarification of the Code requirements for which relief is being requested in relief request No. RV-5. Is relief being requested from all of paragraph IWV-3427 or just for IWV-3427(b)?

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7. Review any function important to safety for the excess flow check valves utiljzed by WNP-2 to determine if they should be included in the IST program'and be categorized A-C.

8. All valves that are Appendix J leak tested to verify their containment isolation function should be included in the WNP-2 IST program and be categorized A, A-C, A passive, or A-C passive as applicable.

9. What cooling systems are used to cool the reactor control room in order to meet the control room habitability requirements? Are the applicable system pumps and valves included in the WNP-2 IST program?

10. Is credit taken at WNP-2 for a post accident sampling system? If so, all of the appropriate system valves should be included in the IST program.

11. Provide the P& ids that show the neutron monitoring system at .;NP-2 (specifically the traversing in-core prebe system ball and shear valves).

At r/p B. Reactor Core Isolation Cooling System

1. How does opening valve RCIC-V-13 during power operation result in tripping the main turbine generator (refer to note 1.k on page 4-43).

2. Should valve RCIC-V-13 be leak tested per Section XI and Appendix J to verify both it. pressure boundary isolation function and its containment isolation function while valve RCIC-V-19 receives only an Appendix J leak test?

3. How is check valve RCIC V-30 full-strcke exercised during the quarterly valve testing?

4. Review the safety function of valve RCIC-V-64 to determine if it should be included in the IST program and be categorized A passive.

5. How are valves RCIC-V-65 and 66 full-stroke exercised during the cold shutdown valve testing?

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6. How is check valve RCIC-V-086 full-stroke exercised open quarterly during power operation?

7. Do valves RCIC-V-lli and 112 perform a function important to safety in the closed position? If so, how are these valves verified closed during the quarterly valve testing?

8. Review any function important to safety for the following valves to determine if they should be included in the WNP-2 IST program.

Valve P&ID Coordinates RCIC-V-25 0-9 RCIC-V-26 D-9 RCIC-V-54 E-9 RCIC-V-61 C-13

/nfs20 C. Low Pressure Core Spray System

1. Does the test flow path permit design accident ficw to pass through check valve LPCS-V-3 during the quarterly valve test?

2. Does the air operator on LPCS-V-6 fully open the valve disc daring cold shutdown testing of the valve? If not, how is this valve full-stroke exercised?

3. Do valves LPCS-V-33 and 34 perform any function important to safety in the closed position? If so, how is check valve LPCS-V-33 verified closed quarterly?

/d52o D. High Pressure Core Spray System

1. Does the test ficw path' permit design accident flew to pass through check valves HPCS-V-2 and 24 during quarterly testing?

2. How is valve HPCS-V-16 full-stroke exercised during quarterly valve testing?

3. Does the air operator en HPCS-V-5 fully open the valve disc during cold shutdown testing of the valve? If not, how is this valve full-strcke exercised?

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4. Do valves HPCS-V-6 and 7 perform any function important to safety in the closed position? If so, how is check valve HPCS-V-7 verified closed quarterly?

5. Does valve HPCS-V-23 perform any function important to safety in the open position?

yy,a/ E. Residual Heat Removal System

1. Will the air operators on the following valves fully open the valves?

If not, how are these valves full-stroke exercised?

RHR-V-41A, B, and C RHR-V-89 RHR-V-50A and B

2. Do valves RHR-V-84A, 848, 84C, 85A, 85B, and SSC perform any function

, important to safety in the closed position? If so, hcw are cSeck valves RHR-V-84A, 848, and 84C verified closed quarterly?

3. What is the safety function of the following valves?

RHR-V-101A and B RHR-V-103A and B

4. Review any safety function for the following valves to determine if they should be included in the IST program.

RHR-V-19 RHR-V-105 '

RHR-V-122A and B Ms24 F. Service Water Systems

1. Why was note 3 referenced for valves HPCS-V-28, SW-V-1 A, and SW-V-1B in the' valve test tables when tile valves were not specifically addressed by the note,7

2. What is the safety function of valves SW-V-69A , 698, 70A, and 70B?

3. Review the function important to safety for valves SW-V-38A and 38B to determine if they should be included in the IST program as

. Category B valves.

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4. How are check valves SW-V-203 and 208 verified closed, their position important to safety, quarterly?

gyff G. Reactor Closed Cooling System

1. What risks associated with non-safety grade equipment outweigh the benefits of performing the Section XI testing of safety grade valves quarterl'y (refer to note 1.d)?

2. g' hat is the function important to safety of the following valves?

,,RCC-V-129 RCC-RV-34A RCC-V-130 RCC-RV-348 RCC-V-131

/vf524, H. Fuel Pool Cooling System

1. Is credit taken for the fuel pool cooling system at WliP-2? If so, what is the safety grade cooling water supply to the fuel pool heat exchangers?

2. Review any function important to safety for check valves FPC-V-146A and 8 (P&ID coordinates J-10) and FPC-V-149 (coordinates C-9) to determine if.they should be included in the IST program as Category C valves.

3. P&ID :1526 indicates that the actuators for the folicwing valves are inoperable, how are these valves exercised quarterly as indicated in the IST program?

FPC-V-172 FPC-V-175 FPC-V-181A FPC-V-173 FPC-V-184 FPC-V-1818

4. Does the condition described in note 3 affect the ability of the referenced valves to perform their functions important to safety?

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I. Control Rod' Drive System M523 1. Review the function important to safety of the folicwing hydraulic control unit valves to determine if they should be included in the IST program and be categorized as indicated.

P&ID M528 Valve Category Coordinates HCU-il4 (185 valves) C C-2 HCU-115 (185 valves) C C-5 HCU-126 (185 valves) B C-4 HCU-127 (185 valves) B C-3 HCU-138 (185 valves) C C-4

/vfs/s J. Main Steam System

1. Do the main steam relief valve downcomer vacuum breakers (MS-y-37

, series and 38 series) perform any function important to safety in the closed position?

2. When and how is operability of the actuator verified for the ADS valves (MS-RV-3D, 4A, 4B, 4C, 40, SB, and SC)?

MSfo K. Reactor Recirculation System

1. What risks associated with non-safety grade equipment are far greater than the benefits of performing Section XI testing of safety grade valves quarterly (refer to note 1.j)?

/d543 L. Primary Containment Cooling and Purge System

1. Provide a more detailed technical justification for not meetirg the corrective action requirements of !WV-3427 for the sup;cessien chamber to drywell vacuum breaker valves (refer to relief request No. RV-6).

2. There is an apparent conflict between the frequency of performing the drywell to suppression chamber bypass leak test indicated in relief request No. RV-6 and WNP-2 Technical Specification 4.6.2.1.d.2.

The relief request indicates that the test will be performed "at least once per 18 months", while the Technical Specification permits the frequency to increase to 40 10 months (see h*.P-2 Technical Specification 4.6.1.2.a).

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" 3. hrethefollowingvalvesAppendixJ1eaktested?

P&ID M543 P&lD M543 Valves Coordinates Valves Coordinates PI-VX-262 E-13 PI-VX-266 F-7 PI-VX-263 E-13 PI-VX-267 E-7 PI-VX-264 E-13 PI-VX-268 E-7 PI-VX-265 B-14 PI-VX-269 C-6 Ms54 M. Containment Atmosphere Control

1. Review any function important to safety for valves CAC-TCV-4A and B (P&ID M554 coordinates 0-12 and 0-4 respectively) to determine if they should be included in the IST program as category B valves.

/vffy(, N. Containment Instrument Air System

1. Review the function important to safety for the following valves to determine if they should be categorized "C" instead of "A-C".

CIA-V-36M, P, R, S. U, and V CIA-V-4CM, P, R, S, U, and V

0. Diescl Generator Systems

1. The current NRC position is that the emergency diesel generator and HPCS diesel generator air start systems perform a function irportant to safety and the appropriate system valves should be included in the IST program and be tested in accordance with the Code.

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2. PUMP TESTir4G PP.0 GRAM

1. ,How is pump flowrate measured for the following pumps?

Standby liquid control pumps SLC-P-1A and 18 Diesel fuel oil transfer pumps 00-P-1A, IB, and 2

2. How is pump discharge pressure measured on the following pumps?

Standby liquid control pumps ~SLC-P-1A and IB Diesel fuel oil transfer pump 00-P-2 o

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WAR 21 1985 Docket No. 50-397 MEMORANDUM FOR: A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing FROM: John 0. Bradfute, Project Manager Licensing Branch No. 2, Division of Licensing

SUBJECT:

WNP-2 PUMP AND VALVE INSERVICE TESTING DATE & TIME: April 16 - 17, 1985 7:30 a.m. - 5:00 p.m.

LCCATION: Room P-114 Phillips Building Bethesda, Maryland PURPOSE: To discuss preparation of Pumps and Valves Inservice Testing and Safety Evaluation Report for Washington Nuclear Project No. 2 (WNP-2). The agenda is enclosed.

PARTICIPANTS *: NRC WNP-2 EG&G F. C. Cherny P. L. Powell H. Rockhold J. D. Page T. F. Hoyle C. Ranson R. Auluck M. T. Reis R. Lyon J. O. Bradfute R. P. Wolfgramm McInal cirned by:

John 0. Bradfute, Project Manager Licensing Branch No. 2 Division of Licensing

Enclosure:

As stated cc: See next page

' Meetings between fiRC technical staff and applicants for licenses are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy",

43 Federal Register 28058,6/28/78.

LBf2/DL/PM(/ LB)2/0L BC JBradfute:lb ASchwen(cer 03/p /85 03/2(./85

Meeting Notices en 1 NRC Participants -

NR F. C. Cherny Local PDR J. D. Page PRC System R. Auluck NSIC J. O. Bradfute LBf2 Reading HDenton/DEisenhut HThompson/FMiraglia TNovak BYoungblood ASchwencer GKnighton EAdensam J7wolinski CGrimes bec: Applicant & Service List GHolahan CThomas Glainas SVarga DVassallo JMiller JStolz JKnight, Acting Dir., DE WJohnston RBosnak', Acting, AD/CSE, DE RBernero LRubenstein WHouston DMuller TSpets FSchroeder -

FRowsome TWRussell, Acting Dir., DHFS ACRS-16 OELD, Attorney Paton JPartlow BGrimes EJordan Fingram, PA Receptionist (Only if meeting is held in Bethesda)

Project Manager J. Bradfute E. Hylton Region 5 s

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