ML20149K029
| ML20149K029 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/19/1988 |
| From: | Kintner L Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8802230330 | |
| Download: ML20149K029 (28) | |
Text
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION 3,
e W ASHING TON. D. C. 20655 FEB 19 my Oceket No. 50-416 LICENSEE:
System Energy Resources, Inc. (SERI)
FACILITY: Grand Gulf Nuclear Station Unit 1
SUBJECT:
$UKVARY OF NOVEMBER 20, 1987 MEETING REGARDING REACTOR COOLANT PRESSURE BOUNDARY FOR THE REACTOR WATER CLEANUP SYS1EM AND THE STANDBY LIQJID CONTROL SYSTEM The purpose of the meeting was to discuss the interpretation of NRC require-rents in the determination of the reactor coolant pressure bourdary (RCPB) for tne reactor water cleanup system (RWCU) and the stancey liquid control system (SLCS). is a list of meeting participants. is a hand-out prepared by the licensee to cescribe its interpretation of NRC requirements for the RWCU system and the SLCS. is a copy of slides prepared by the licensee to describe an alternate proposed resolution of the RCPB issue for these two systems.
The chronology for the review of modifications to the RWCU system and the SLCS is listed on page 2 of Enclosure 2.
The RWCU system (Sheet 7 of Enclosure 2) was found by the licensee to have a cesign error, in that an inboard contain-rnent isolation valve (F252) had the same power supply (Division A) as the out-board containment isolation valve (F004) in that line.
Therefore, failure of the containrent isolation function of this line was subject to a single power supply failure. Tne licensee's investigation showec that the initial plant cesign was changed in Au;ust 1979 (LER $7-Cll-CO, Aupust 28, 1957). Tne initial design proviced Division 8 power for valve F252 and Division A power for valve F253, with the RCPB piping (ASME Code, Class 1) extending through valve F253, which is outside the drywell. Thus, the initial design had correct power sup:'ies te oerfere the c:*tain ent 1sclatier functicn (valves FCC4 and r252),
the drywell isolation function (valves F252 and F253), and the reactor coolant pressure boundary isolation function (valves F001 and F253).
De cesign was changed by designating the ASME Coce Class 1 piping to end at valve F252 (inside the drywell) and interchanging the power supplies to valves F252 and F253, resulting in the incorrect as-built system.
The licensee proposed to inter-change the power supplies for valves F252 and F253 and to allow the piping between valves F252 and F253 to remain designated ASME Code Class 2.
The staff expressed concern that this as-built RWCU piping configuration did not meet NRC requirements for the RCPB in 10 CFR 50.2, 10 CFR 50.55a(c), and General Design Criterion 55 in Appendix A to 10 CFR 50. The staff's position was that these regulations require that the RCPB include all pressure contain-ing components up to anc including the cuttoard crywell isolation valve and that the RCpB components meet ASWE Coce Class 1 requirerents unless the provi-siens of paragraph 10 CFR 50.55a(c)(2) are met.
For GGNS Unit 1, wnten is a h
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2 EWR-6, Mark III plant the drywell is considered by the staff and the licensee to be the primary reactor containment structure for the purpose of designating RCPB piping.
The licensee stated that the section of RWCU piping between valves F252 and F253 could be made to meet the provisions of 10 CFR 50.55a(c)(2)(ii) and there-fore could remain as ASVE Code Class 2 piping. A comparison of proposed design changes with the regulatiers is given on pages B and 9 of Enclosure 2.
In response to staff questions, tne licensee said the normal reactor coolant system makeup was considered to be the RCIC pump; however, staff questioned its appli-cability because it cannot be used to cool dcwn the reactor to a cold condition.
The control rod drive pump, the normal eakeup system, delivers about 75 spm to the reactor, which is not adequate rnakeup for a rupture of the 6-inch diameter RWCU pipe.
The licensee said tne isolation valves are automatically closed by a signal sensing a differential flow in the RWCU system greater snan 75 spm but did not knov if the autcmatic closure instrumentation was redurdant and safety grade.
Valves F252 and F253 are ircluded in ne A$vE Section XI IST valve operacility test program out are not requirec to be leak tested per ASME Sec-tion XI because they are not pressure isolation valves.
Local leak rate test-ing is required for valve F252 in accordance with Appendix J to 10 CFR 50 because it also serves as a containment isolation valve.
As an alternate to demonstrating conformance to the reovirements of 10 CFR 50.55a(c)(2)(ii), the licensee proposed to retain the ASME Code Class 2 designa-tion for the RWCU piping between valves F252 and F253 but to perform periodic augmented inservice inspections consistent with the ASME Code Class 1. ISI Program and demonstrate by analysis, that the piping would teet ASME Code Class I stress criteria (pages 1 and 2 of Enclosure 3).
The Licensee proposed to reouest an exemption to 10 CFR 50.55a, based en the above reans of reetin; the uncerlying purcose of the rule, i.e.,
to assure reactor ecolant pressure councary integrity.
The proposed modifications to the stancby licuid centrol system (SLCS) raised sitailar staff concerns regarding the designation of ASME Code Class 2 piping f: certain portions of the modified system. The existieg S'.C5 is sho.n ci page 12 of Enclosure 2, where the Class 1-to-Class 2 interface is at the exple-sive valves (F004A and F004B).
The proposed SLCS Class 1-to-Class 2 interface would be inside the drywell at isolation valve F006 which would be m.oved from outside the drywell to inside the drywell.
The staff expressed concern that this configuration did not meet the explicit requirements for isolation valves in General Design Criterion 55 to nave one isolation valve outboard of contain-nent (in this case the drywell) and one isolation valve inboard of containment and the proposed arrangement was not justified on ancther defined easis.
The licensee statec that one reason for moving valve FCC6 insice the crywell was to provide acced assurance that the pipe would not be pressuriced and neated in the event valve FC07 leaked curing operation.
As an alternate to moving isolation valve F006 and tre ASVE Coce Class 1-to-Class 2 interface insice the cry ell, the licensee proposed to retair. the out-board isolation valve F006 in its present location and move the Class 1-to-Class 2 interface from the explosive valves to isolation valve F006 (pages 3
3 and 4 of Enclosure 3).
A new check valve would be added inside the drywell to reduce potential for pressuri:ing and heating the piping outside the drywell.
This arrangement could meet regulatory requirements for the RCPB piping in the SLCS.
The licensee said it would modify its license amendment applications for changes to the Technical Specifications related to the RWCU and SLCS rodifications con-i sistent with Enclosure 3 ar.d would submit a request for an exemption to 10 CFR 50.55a for the RWCU system.
/ub L. L. Kintner Senior Project Manager Project Directorate II-1 Division of Reactor Projects I/II c,
Jeenextpage Ph:PD21:DRPR D
MR LKintner EAdensam l
02/ [SS 02/jf//S3 Distribution See attached sheet e
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Mr. Oliver D. Kingsley, Jr.
System Energy Resources Inc.
Grand Gulf Nuclear Station (GGh5)
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Mr. Ted H. Cloninger Mr. C. R. Hutchinson Vice President, Nuclear Engineering GGNS General Manager and Support System Energy Resources. Inc.
System Energy Resources, Inc.
Post Office Box 756 Post Office Box 23054 Port Gibson, Mississippi 39150 Jackson, Mississippi 39205 Robert B. McGehee, Esquire The Nonorable William J. Guste, Jr.
Wise Carter, Child Steen and Caraway Attorney General P.O. Box 651 Department of Justice Jackson, Mississippi 39205 State of Louisiana Nicholas S. Reynolds. Esquire Bishop. Libeman, Cook, Purcell Office of the Governor Snd Reynolds State of Mississippi t
j 1700 17th Street, N.W.
Jackson, Mississippi 39201 Washington, D. C.
20036 Attorney General Mr. Ralph T. Lally Gartin Building Manager of Quality Assurance Jackson, Mississippi 39205 Middle South Utilities System Services. Inc.
P.O. Box 61000 Mr. Jack McMillan, Director New Orleans, Louisiana 70161 Division of Solid Waste Manager 4nt i
Mississippi Department of Natural Mr. John G. Cesare Resources Director, Nuclear Licensing System Energy Resources, Inc.
Post Office Box 10385 P.O. Box 23054 Jackson, Mississippi 39209 l
Jackson, Mississippi 39205 Alton B. Cobb, M.D.
l Mr. R. W. Jackson, Project Engineer State Health Officer Bechtel Power Corporation State Board of Health 15740 Shady Grove Road P.O. Box 1700 Gaithersburg, Maryland 20877-1454 Jackson, Mississippi 39205 l
Mr. Ross C. Butcher President Senior Resident Inspector Claiborne County Board of Supervisors U.S. Nuclear Regulatory Cceanission Port Gibson, Mississippi 39150 Route 2, Box 399 l
Port Gibson, Mississippi 39150 i
Regional Administrator, Region 11 U.S. Nuclear Regulatory Coescission i
101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 Mr. James E. Cross l
GGNS Site Director Systes Energy Resources, Inc.
P.O. Box 756 Port Gibson, Mississippi 39150 i
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O ENCLOSURE 1 Participants in NRC-SERI Meeting, November 20, 1937 Nu: lear Regulatory Commission (NRC}
P. T. Kuo Chief System and Component Integrity Section Mechanical Engineering Branch, DEST K. Dempsey Mechanical Engineer Mechanical Engineering Branch, DEST J. Kudrick Chief. EWR Systems Section Plant Systems Bran:h, DEST L. Kintner Senior Project Manager, Project Dire:torate II-I, ORP I/II System Enercy Resources, Inc. (SERI)
Ted H. Cloninger Vice President, Nuclear Engineering and support l
Fred Titus Dire: tor, Nuclear Plant Engineering i
Steven E. Themas Prin:1 pal Mechanical Engineer l
Mike Meier N;: lear Piar. Engineering Ce;t.
J. G. Cesare Dire:ter, Nuclear Licensieg 1
J.
O.,
Fowler Nuclear Licensing Cept.
4 Walter D'Ardenne GE Nuclear Energy, Consultant to SERI i
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1 NRC-SERI E ETING NO W BER 20, 1987 BETHESDA l
PROPOSED CHANGES TO RWCU/SLCS SYST B 'S AND 1
ASSOCIATE TECHNICAL SPECIFICATION CHANGES l
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INTR 00VCTORY RD%RKS L. L. K!rnNER (NRC)
T. H. Cl.WINGER (SERI) 1
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!!, OVERVIEW 0F ISSUES J. G. CESARE (SERI) 111. COPPLIANCE WITH ICCcR50,55A S. E. Tecm s (SERI) i IV. NRC COPMNTS/ DISCUSSION V.
ACTION AGREENNTS/l.ICENSING 1
SCHEDULE COPNITMNTS/00NCLUS10N l
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4 CFROPOLOGY - KEY EVENTS ATWS/SLCS 6/84 10CFR50.62 - ATWS RULEMAK!NG 4/16/85 NRC GL85-06 ISSUED 10/14/85 SERI RESPONSE TO 10CFR50.62 AND GL85-06.
COMMITTED TO IMPLEMENTATION OF SLCS MODIFICATION BY SECOND REFUELINr. OUTAGE 2/11/87 NRC REQUEST FOR ADDITIONAL INFORMATION 4/3/87 SERI RESPONSE 8/13/87 SERI SUBMITTED TECH SPEC CHANGES TO SUPPORT MODIFICATION (It!CLUDES DISCUS $!ON OF SLCS CHECK VALVE MOVEMENT) 11/3/87 CONFERENCE CALL.
NRC IDENTIFIED CONCERNS REGARDING CHECK VALVE RELOCATION, RPCB, AND ASME CODE CLASSIFICATION 11/10/87 CONFERENCE CALL.
FURTHER DISCUSSION.
SERI REQUESTED MEETING.
RWCU 8/28/87 SERI SUBMITTED LER REGARDING IDENTIFICATION OF INAPCROPRIATE POWER SUPPLY TG CONTAINMENT ISOLATION VALVE.
9/15/87 MEETING WITH NRC (ATLANTA) TO DtSCUSS ISSUE, INVESTIGATION, CORRECTIVE ACTIONC.
10/28/87 SERI REQUESTED TECH SPEC CHANGE TO SUPPORT CORRECTIVE ACTIONS (REVERSAL OF DIVISIONAL POWER SUPPLY) 11/J6f87 CONFERENCE CALL.
NRC CONCERNS EXPRESSED REGARDING jh RCPB SELECTION AND ASME CODE CLASS!FICATION OF P!P!NG AND VALVE EXTERIOR TO DRYWELL 11/10/87 CONFERENCE CALL.
FURTHER DISCUSSION.
SERI REQUESTED MEETING.
FORMAL NRC ADVISORY THAT THE SELECTION OF RCPB WAS NOT ADEQUATE.
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If a Loss-of-Coolea*, Accident should occur, the drywell channels released steam around the weir wall and through the horizontal vents and into the suppression pool.
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0 10CFR50.55a REQUIREMENTS 10CFR50.55a fc) Reactor Coolant Pressure Boundarv (1)
Components whict. are part of the reactor coolant prossure boundary must meet the requirements for class 1 components in Section III of the ASME Boiler and Pressure Vessel
- Code, except as provided in paragraphs (c) (2),
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(c) (2)
Components which are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in 50.2 need not meet the requirements
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of paragraph (c) (1) of this section, Provided (ii) The component is or can be isolated from the reactor coolant system by two 1)
Two valve isolation fron valves in s6 ries (both closed, the reactor coolant system.
both open, or one closer. and the other open).
Each open 2)
Open valves must be valve must be capable of capable of automatic closure, automatic actuation
- and, assuming the other valve is closure time must be such that open, its closure time must be in the event of a failure of such that, in the event of a the isolated component postulated failure of the component during normal 3)
Each valve remains reactor operation, each valve operable, and remains operable and the reactor can be shut down and 4) the reactor can be shut cooled down in an orderly down and cooled down assuming
- manner, assuming makeup is normal reactor coolant system provided by the reactor makeup only (i.e.,
not coolant makeup system only, feedwater or ECCS).
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PROPOSED RWCU_ DESIGN _ CHANGES
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REQUIREMENTS l
1)
Two valve isolation from The proposed RWCU design meets the reactor coolant system.
these requirements, in that 2) open valves must be 1) two Class 1 valves are capable of automatic closure.
provided (T001 and T252):
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closure time must be such that 2) both valves are capable in the avant of a failure of of automatic closure (one i
the isolated components of the valves is normally closed at all times.
3)
Each valve remains closure of either T001 or operable, and T252 is assured by administrative controls 4) the reactor can be shut and an interlock.),
i down and cooled down assuming normal reactor coolant system 3) both valves are protected makeup only (i.e., not from a failure of the i
isolated components to ensure operability, and 4
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of either interface isolation valve, system losses are limited to the i
leakage of the other interface velve.
Closure time is not a consideration, since one of the two interface isolation valves is always closed, (E) - b *(9)
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ANSI /ANS-52.1-1983 Case 3 SC-1 or -2 piping of the RCPB penetrating primary containment:
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From SC-1 to SC-2, the interface is at the outermo[st connection of two valves, each of which is capable of automatic closure; from SC-2 to any less stringent class, the interface is at the outermost connection of a third remote manual valve.
l Closure time requirements of interface valves shall meet the requirements of 3.3.2.1.2.
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Two velve isolation from The proposed SLCS design meets the reactor coolant system.
the 10CFR50.55a class 1 exception criteria, in that 2) open valves must be capable of automatic closure.
1) two class i valves are
- provided, closure time must be such that in the event of a failure of 2) both valves are normally the isolated components closed check valves on an influent line, 3)
Each valva remains operable, and 3) both valves are unaffected by a failure 4) the reactor can be shut of the isolated down and cooled down assuming components, and is normal reactor coolant system ensured, and makeup only (i.e., not faedwater or Eccs).
4) assuming a single failure of either interface isolation valvn, systen losses are limited to the leakage of the other interface valve.
Closure time is not a consideration, since both of the interface isolation valves are normally closed.
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Fron.9C-1 to SC-2, the interface is at the outermost connection of two check valves, both of which are inside primary containment, where the flow path is into the reactor coolant system only from SC-2 to SC-3, the interface is at the outernost connection of one automatic valve located outside primary containment, meeting the closure time requirements of 3.3.2.1.2.
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ENCLOSURE 3 I
PROPOSED MODIFICATIONS I
TO REACTOR WATER CLEANUP SYSTEM AND STANDBY LIQUID CONTROL SYSTEM l
i PREPAREC EY l
SYSTEM ENERGY RE500.;:. INC.
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EWCU Pt WOSED ACTION I
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PERFORM ASME III, class 1 ANALYSTS FOR PIPING IMAOUGH F253
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s MEETING SUFFARY DISTRIBUTION:
eDocket3 NRC POR Local PDR P021 r/f E. Adensarn Project Manager Lester L. Kintner 0GC-8 E. Jordan J. Partlow NRC Participants ACRS(10)
NRC PARTICIPANTS P. T. Kuo K. Dempsey J. Kudrick L. Kintner I
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