ML20094L006

Responds to 840806 Request for Addl Info Re Change to Isolation Actuation Setpoint in Tech Specs for Six Valves. Proposed Amend Revises Tech Spec Isolation Setpoint for Each Valve Making Setpoint Consistent W/Original Design of Plant

ML20094L006
Person / Time
Site:	Hatch
Issue date:	08/10/1984
From:	Gucwa L GEORGIA POWER CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
NED-84-420, TAC-55555, NUDOCS 8408150242
Download: ML20094L006 (5)

Text

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N' Gzorgia Power Company 333 Piedmont Avenue Atlanta. Georgia 20308 Telephone 404 526-6526 -

Maihng Address' Post Off;ce Box 4545 Atlanta, Georgia 30302 h Georgia Power L T. Gucwa the southem eW2nc s ystem Manager Nuclear Eng:neenng and Chief Nuciear Encineer NED-84-420 August 10, 1984 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4

, Division of Licensing h U. S. Nuclear Regulatory Ccmnission Washington, D. C. 20555 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN I. HA7CH NUCLEAR PIANT UNIT 2 ADDITIONAL INEDBMATION - REQUEST 'IO CHANGE TECHNICAL SPECIFICATIONS CONTAI?NENT ISOIATION %LVES Gentimen:

Our letter of August 6,1984, requested expedited NRC action regarding proposed changes to the Unit 2 Technical Specifications in order to enable Unit 2 to resme operation on schedule. The present status of Unit 2 is as follows: Fuel loading is cmplete. RPV reassenbly, hydrostatic testing, and final valve aligment prior to pulling control rods is scheduled for empletion by about August 14, 1984.

Discussions with the NSSS vendor, General Electric Cmpany, have lead to the following points:

1. According to General Electric Ompany, with regard to the subject valves' isolation signal, the Hatch 2 as-built design is the see as that of all dmestic BWR-4 through IER-6 plants.

2. The current Technical Specification (i.e., Table 3.6.3-1) is incorrect.

3. The probability of the subject valves being open (i.e., the valves are normally closed) coincident with a IOCA large enough to require a rapid EOCS response is low enough to be beyond the BWR design I basis. I

4. The ECCS analysis is not affected since the actuation signal assmed in the analysis is unchanged and the valves are assmed to be closed.

l C408150242 840810 PDR ADOCK 05000 gg P

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' Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 August 10, 1984 Page 'IWo

%e August 6,1984, letter concluded that the actuation of the (ten) subject contalment isolation valves at RPV level 1 is consistent with the L

original design of the plant as reported in the FSAR. %is conclusion was based on information found in Chapter 7 of the Unit 2 FSAR as discussed below.

Chapter 7.3 of the FSAR identifies the low water level initiation signal for the RHR and Core Spray systes as having a trip setpoint of -146.5 inches. %is is referred to in our August 6,1984 letter as RPV level 1.

Section 7.3.1.2.3.2 of the FSAR addresses the logic and sequencing for initiation of the Core Spray system. %is section states that the Core Spray test bypass valves are closed and interlocked to prevent opening following the receipt of a Core Spray initiation signal (RPV level 1) .

Section 7.3.1.2.3.4 of the FSAR addresses actuated devices in the Core Spray system. % is section also states that:

"Upon receipt of an initiation signal, the test bypass valve is interlocked shut... %e signal received upon autmatic Core Spray initiation overrides all other signals."

Similarly, Section 7.3.1.2.4.2 of the FSAR addresses logic and sequencing for initiation of the LFCI mode of the RER systs. %is section also states that following receipt of a LECI initiation signal, " valves in other systems (contaiment spray and RHR) are autmatically positioned so that the water pmped frm the suppression chmber is routed correctly."

Section 7.3.1.2.4.3 states that "the valves that divert water for contaiment cooling are signaled closed on receipt of a LFCI system initiation signal."

As shown above, one of the ECCS initiation signals is RPV water level 1 In addition, ECXL (i.e., Reactor Vessel Water level-Iow Low Low) .

initiation will occur upon receipt of a high drywell pressure (2 psig) signal. Either ECCS initiation signal will cause the subject Core Spray and RHR system valves to align to their proper positions.

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i GeorgiaPowed Director lof Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Brand No. 4 August-10, 1984

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%e contaiment isolation function of the' valves is provided by the sane signals that initiate the EOCS systens. Either RPV low water level or high drywell pressure will ' initiate closure of the subject valves. Figures

. 6.2-25 and'6.2-30.of the-Unit 2 FSAR show the contaiment pressure responses

to a postulated recirculation systen line break and a 0.1 sguare foot liquid line = break. -In .both cases, - the high drywell pressure trip setpoint is reached in less than 10 seconds. Figures.6.3-13 and 6.3-22 show the reactor

- water level -inside -the shroud following a recirculation systen discharge line break and a -1.0 sguare foot line break. %ese two figures typify the initial -water ~ level changes for a range of break sizes prior .to the ,

injection of water into the vessel by the EOCS systens. These figures show l

. water ' level renaining above- both the RPV level 1 and RPV level 3 trip 1 setpoints in excess' of' twenty seconds for the wide spectra of postulated l break sizes. The contalment isolation function is provided first by a high drywell pressure signal, with the low reactor water level signal being received after the high drywell pressure signal. We proposed change to the wchnical Specifications leaves the contaiment isolation performance undanged as a result of the order in which isolation signals would be

%is is due to contaiment pressure causing EOCS initiation and

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received.

. contaiment isolation for the subject valves, prior to receipt ' of an EOCS initiation and contaiment isolation on low reactor _ water level. Although Unit 2 was not originally evaluated for conformance to the Standard Review Plan, the described actuation signal logic remains consistent with the acceptance criteria stated in Section 6.2.4 of the Standard Review Plan.

We have reviewed Technical Specification Table 3.6.3-1 to assure that no I

- other Group 2 isolation valves have an incorrectly identified RPV water l level initiation' signal. %e only renaining Group 2 valves 'in Table 3.6.3-1 are four radwaste systen valves. %ese valves are designed to actuate on RPV level 3 or Drywell Pressure high signals, and are thus consistent with the Group 2 isolation actuation nmenclature in the table.

Also enclosed is a revised discussion of the no significant hazards determination originally subnitted as Attachment 3 to our August 6, 1984 raguest.

Very truly yours, f X g+<

L. T. Gucwa t

RDB/rb

~Attachnent xc: J. T. Beckhan, Jr.

H..C. Nix, Jr.

J. P. O'Reilly (NRC- Region II)

Senior Resident Inspector J. L. IAubetter

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l GeorgiaPower d A CNN 3 NRC DOCKET NLMBER 50-366 l OPERATING LICENSE NPF-5 ENIN I. HA'IGI MJCLEAR PIANT UNIT 2 10CFR50.92 EVAIUATION FOR REQudr 'IO CHANGE ISOIATION ACTUATION SETPOINT IN 'INE 'IEQNICAL SPECIFICATIONS FOR SIX VALVES

%e proposed men &ent would revise the Technical Specification isolation setpoint for each of the vcives of Table 1 to make the .setpoint consistent with the original design of the plant. Contalment isolation valves listed in Table 1 (Attachment 1 to this letter) are associated with the RHR and Core Spray systes. %ese valves are normally closed and are designed to go closed on receipt of an isolation signal.

The present isolation value found in the Technical Specifications for the subject valves is Reactor Pressure Vessel (RPV) level 3. RPV level 3 is one of two isolation signals which is associated with Group 2 isolation.

'Ibe existing Technical Specification for the subject valves erroneously states that the valves go closed on a Group 2 isolation. %e original design drawings for the plant, however, state that the valves in question should go closed on a RPV level 1 signal.

We proposed change would replace "(Group) 2" with an asterisk and a footnote which reads " Closes upon actuation of the LPCI mode of RHR via a Im Iow Inw (level 1) signal frm 2B21-N691A, B, C, D. Refer to item 2.b of Table 3.3.3-1" for the RHR systs valves, and " Closes upon actuation of Core Spray via a Im Iow low (Ievel 1) signal frm 2B21-N691 A, B, C,D. Refer to its 1.a of Table 3.3.3-1" for the Core Spray syste valves.

BASIS:

his change is to make the Technical Specifications consistent with the original design basis, as identified by vendor drawings and instrment data sheets, and with the licensing basis provided in the FSAR. %e design actuation point of each of the subject valves is consistent with the overall syste design. %e accident analyses, as reported in the FSAR, assmes that the Core Spray and RHR systes would be actuated at a RPV level 1 trip point. Actuation of the subject valves at RPV level 1 is consistent with the original design of the plant as reported in the FSAR. The probability of the normally closed valves being open coincident with a postulated IOCA is small enough to be beyond the design basis of the BNR and is therefore not considered in the EOCS analysis. The ECG analysis and conformance with 10 CER 50 Appendix K criteria are not affected by postulating closure of the valves at RPV level 1. Although Plant Hatch Unit 2 was not originally evaluated for conformance to the Standard Review Plan, the described actuation signal logic remains consistent with the acceptance criteria stated in Section 6.2.4. of the Standard Review Plan.

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- l GeorgiaPowerd Amcanwr 3 (Continued)

NRC DOCIET NLNBER 50-366 OPERATING LICENSE NPF-5 ENIN I. HA'IUI NUMEAR PUWr UNIT 2 10CFR50.92 EVAWATICN FOR REQUEST 'IO CHANGE ISOLATION ACIUATION SEITOINT IN 'IEE TECINICAL SPIEIFICATIONS E0R SIX VALVES The change in isolation signal fran RPV water level 3 to RPV water level 1 represents a decrease in margins for that isolation signal, however, because the proposed change continues to meet the acceptance criteria of 10 CFR 50 Appendix K and Standard Review Plan Section 6.2.4, the proposed change is consistent with Iten (vi) of the "Exanples of Anendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6, 1983, issue of the Federal Register.

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