ML20094F436

Application for Amend to License NPF-5,revising Tech Specs to Change Isolation Actuation Setpoint for Six Valves.Fee Encl

ML20094F436
Person / Time
Site:	Hatch
Issue date:	08/06/1984
From:	Beckham J GEORGIA POWER CO.
To:	Stolz J Office of Nuclear Reactor Regulation
Shared Package
ML20094F440	List: ML20094F436 ML20094F446
References
NED-84-413, TAC-55555, NUDOCS 8408100045
Download: ML20094F436 (6)

Text

1. ) Georgia Power Company 333 Piedmont Avenue Atlanta, Georgia 30308 Telephone 404 526 7020 Mati ng Address-Post Office Box 4545 Atlanta, Georgia 30302 g Georgia Power J. T. Beckham, Jr. the southern erectnc systern Vice Pres: dent and General Manager Nuclear Operations NED-84-413 August 6, 1984 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Gief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Camission Washington, D. C. 20555 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EIMIN I. HATCH NUCIEAR PLAN 1' UNIT 2 REQUEST TO CHANGE ISOIATION ACIUATION SEI' POINT IN THE TECHNICAL SPECIFICATIONS FOR SIX VALVES Gentleen:

As required by 10 CFR 50.59(c) (1), and in accordance with the provisions of 10 CFR 50.90, Georgia Power Cmpany proposes mendments to the Plant Hatch Unit 2 Technical Specifications (Appendix A to the Operating License) . The proposed changes would provide relief for Unit 2, which is in the last stages of preparation for startup following an extended shutdown for refueling and major plant modifications.

On May 4,1984, plant personnel identified that an apparent discrepancy existed betwt.an the installed actuation setpoints for certain RHR systs and Core Spray systs isolation valves and the inferred actuation setpoint in the Technical Specifications. Each of the ten valves in the attached Table l 1 is identified in the TMhnical Specifications as isolating on a Group 2 signal. 7be Group 2 isolation signals are either Reactor Vessel Water Ievel

-Iow (RPV level 3) or Drywell Pressure -High. However, the original installed plant design calls for valve isolation to occur at Reactor vessel Water Level -Low Iow Iow (RPV level 1) instead of RPV level 3.

The architect / engineer was requested to investigate the apparent discrepancy to identify whether a design error existed or whether the 1bchnical Specification valve group identification was incorrect. The A/E conclusion, received by Georgia Power Campany on August 2,1984, is based on a thorough review of pertinent A/E and USSS vendor design drawings. The drawings indicate that the valves listed in Table 1 should, in fact, isolate at RPV level 1. Electrical elmentary drawings indicate that the subject valves should receive a close signal frm switch 1 of instrments 2B21-N031 A-D.

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4 GeorgiaPower d Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief

' Operating Reactors Branch No. 4 August 6, 1984 Page W o i

< Each instrment has four switches, two of which are labelled " low" and two of which are labelled "high". %e instrment data sheet for 2B21-N031 A-D indicates that switches 1 and 2 should be set at a value consistent with RPV level 1: and switches 3 and 4 should be set at a value consistent with RPV level 2. Neither of these level settings is consistent with a Group 2

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isolation signal (RPV level 3) . We therefore conclude that a change to the ,

%chnical Specifications is necessary in order to make that docment i consistent with the original design drawings.

! %e following discussion provides a licensing basis to support the

, proposed changes. All of the valves listed in Table 1 are associated with ,

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her the RHR or Core Spray systes. Both of these systes are designed to autmatically actuate at RPV level 1. All of the valves in Table 1 are i normally closed. i 4

i Each of the subject valves receives an actuation signal fra ATIS l Transitter Trip Units 2B21-N691 A-D, which replaced switches 2B21-NO31 A-D l 2- to provide the RPV level 1 ECCS actuation signals. %is present

configuration is consistent with the original design of the plant. l Valves 2 Ell-F0ll MB and 2 Ell-F026 MB are RHR heat exchanger drains to i the Torus and the RCIC syste respectively. %ese valves cannot impact the ,

ability of the RHR syste to autmatically supply cooling water to the reactor until .such time as the syste is actuated, which would occur at the RPV level 1 trip' point.

Valves 2 Ell-F016 MB .and 2 Ell-F028 MB are Contalment Spray Isolation valves which could be used to spray-the drywell or torus in the event of a l IOCh. % ese valves, again, will not impact the ability of the RHR systs to autmatically supply cooling water to the reactor until such time as the RHR syste is actuated, which would occur at the RPV level 1 trip point. 'In addition, once closed these' valves cannot be easily opened until RPV water level has recovered to above 2/3 core height.

J Valves 2E21-F015 MB are the Core Spray syste full flow test valves to the wrus._ %ese valves are-normally closed, and would be open only during a'_systs test. %ese valves cannot impact the Core Spray syste's ability  ;

• to autrnatically provide water to the core until such time as the Core Spray i systa.. is actuated. %is would occur at the RPV level 1 trip setpoint.

%e design actuation point of each of the subject valvec is consistent with the design ' actuation point of its syste. Se accident analyses, as reported in the ESAR, assmes that the Core Spray and RHR systes would be i actuated at a RPV level 1. trip point. %erefore, the actuation of the valves at RPV level 1 is consistent with the original design of the plant as reported in the FSAR.

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1GeorgiaPoWerI Director of Nuclear Reactor. Regulation Attention: Mr. John F. Stolz, Chief

Operating Reactors Branch No. 4 '

August 6, 1984 Page %ree

% e Plant' Review Board has reviewed the proposed changest .to.the Technical Specifications, and has concluded that.the proposed changes do not constitute an unreviewed safety question, because 1) Se probability of occurrence or the consequences of an accident or malfunction of equipnent important to safety is not increased above those evaluated in the PSAR due s . to this change, because the original accident analysis as presented in the PSAR . assmes . that ' valves 2 Ell-F0ll MB, 2 Ell-PT)l6 MB , 2 Ell-F026 MB ,

2 Ell-F028 MB, and 2E21-F015 MB would receive their actuation signal at the RPV level 1 trip setpoint. 2) %e possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR does not

- result fra this change because the design is consistent with the design-considered - in the original accident analysis. 3) %e margin of safety as defined in the basis for the Technical Specifications is not reduced due to this change in that the.. safety . analysis was based on the original design

'which assmed that the subject valves close at the RPV level 1 trip setpoint.

In the brief period of. time since the discrepancy was resolved as incorrect isolation setpoints in the Technical Specifications, we have evaluated the potential scope and expected duration for a plant - design change and modification. We difficulty in raising the isolation setpoint for the affected valves fra RPV level 1 to RPV level 3 arises fra the absence of sufficient EOCS instrment channels at RPV level 3 to provide the necessary neber of channels for divisional redundancy of the isolation signal. Two new instrment channels would be required at RPV level 3, 3 ceplete with level transmitters, trip units, relays, reset switches, .

tubing, cables, conduit, and wiring. Design engineering and installation j are each estimated to require three weeks to.ceplete. - In addition, plant- r procedures would need revision, and functional ~ tests would need to be (

written and impleented. . Of the abcve -listed items, procurenent of level i transmitters is the limiting factor, with an estimated 20 weeks delivery time. . Although most work could be performed prior to receipt of the level transnitters, installation and functional testing of the transmitters would

. bring the estimated critical path schedule to see 22 weeks.

Due to the relatively short period of time remaining before the planned Unit .2 'startup, Georgia Power Capany finds it necessary to request +

expedited N K review of our proposed changes, and relief fra the obligatory 30 day notice period required by 10 CER 50.92, in order to enable Unit 2 to

.resm e operations on achedule.

. Instructions for ' incorporation of these changes, along with copies of the affected Technical Specification pages are enclosed as Attachnent 2.

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> r Georgia Power b Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 August 6, 1984 Page Ebur In accordance with the revised fee schedule, a check fer $150.00 is enciesed.

As required by 10 CFR 50.92, an analysis of the proposed changes to the Technical Specifications is enclosed with this subnittal.

Pursuant to the raluirenents of 10 CER 50.92, J. L. Ledbetter of the Georgia Department of Natural Resources will be sent a copy of this letter and all applicable attachnents.

J._ T. Beckhan, Jr. states that he is Vice President of Georgia Power Ctropany and is authoris.ed to execute this oath on behalf of Georgia Power Ccepany, and that to the best of his knowledge and belief the facts set forth in this letter are true.

GBORGIA PWER O@iPANY By: f[I- --- __%_

J. T. Beckhan, Jr. '

Sworn to and subscribed before me this 6th day of August,1984.

h 3-: [ .n--a Notary Public Notary PuSk %, g

'msson boires Aus26,1986 EncloE,ure xc: H. C. Nix, Jr.

Senior Resident Inspector J. P. O'Reilly, (NRC-Region II)

J. L. Imdbetter k T '"

i e GeorgiaPower d AmCmmr 1 NRC DOCKET 50-366 EIMIN I. HA101 NUCLEAR PLANP UNIT 2 TABLE 1 Contairunent Isolation Valve / Function 2 Ell-F0ll MB RHR Heat Exchanger Drain Isolation Valves 2 Ell-F026 MB RHR Heat Exchanger Drain Isolation Valves 2 Ell-F016 MB Contairnent Spray Isolation Valves 2 Ell-F028 MB Contairnent Spray Isolation Valves 2E21-F015 A&B Core Spray Flow Test Line Isolation Valves t

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i GeorgiaPower d i

t-ATTACMENT 2 NRC DOO TT 50-366 OPERATING LICENSE NPF-5 EIMIN I. HA'ICH NUCEAR PIANr LNIT 2 IROIOSED CHA!GES 'IO TECHNICAL SPECIFICATIONS The proposed change to 'Itchnical Specifications (Appendix A to Operating License NPF-5) wculd be incorporated as follows:

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