ML19332D335

Application for Amend to License NPF-67,adding Note to Applicability Statements of Tech Specs 3.7.1.2 Auxiliary Feedwater Sys & 3.7.1.6 Atmospheric Relief Valves to Allow Entry Into Mode 3 to Perform post-maint & Mod Testing

ML19332D335
Person / Time
Site:	Seabrook
Issue date:	11/22/1989
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19332D336	List: ML19332D335 ML19332E100
References
NYN-89150, NUDOCS 8912010031
Download: ML19332D335 (4)

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L New Hampshire Ted C..._. _

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Senior Vice President and Chief Operating C4ficer NYN- 89150 November 22, 1989 1

United States' Nuclear Regulatory Commission i Washington, DC 20555 J

^ ' Attention: Document Control Desh

References:

Facility Operating License NPF-67. Docket No. 50-443

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Sr.bjec t : Request for License Amendment: Applicability for Auxiliary Feedwater System and Atmospheric Relief Valves Gentlemen:

Pursuant to 10 CFR 50.90 New Hampshire Yankee (NHY) hereby proposes to amend the Seabrook Station Operating License (Facility Operating License  !

hPF 67) hy incorporating the proposed changes, provided herein as Enclosure l 1, into the Seabrook Station Technical Specifications. The proposed changes  ;

iL involve the addition of a note to the Applicability Stataments of Technical i Specifications 3.7.1.2 ' Auxiliary Feedwater System," and 3.7.1.6 i

" Atmospheric Relief Valves " to allow entry into MODE 5 to perform necessary *

. post-maintenance or post-modification testing.

The basis for these proposed changes is provided in Enclosure 2 which includes a safety evaluation. Based upon the information contained in  !

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Enclosure 2. NHY has concluded that the proposed changes do not involve an '

Unreviewed Safety Question pursuant to 10 CFR 50.59, nor do they involve a i Significant Hazard Consideration pursuant to 10 CFR 50.92. '

New Hampshire Yankee has reviewed the proposed changes utilizing the criteria specified in 10CFR 50.92 and has determined that the proposed ,

i changes will not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The change involves clarifications to the administrative rules concerning entry into '

MODE 4 or MODE 3 following a cold shutdown during which maintenance '

or modification work was performed on the turbine-driven Emergency

• Feedwater (EW) pump or the atmospheric steam dump valves (ASDVs).

There are no physical changes to the facility.

O The failure of an ASDV to close is a condition analyzed in FSAR 15 .1. . Post maintenance / modification retests would normally involve an exercise test during which time there is a potential for the valve to fail to close following an open stroke. The internal 15 pilot arrangement and internal steam leakage characteristics of i .DC

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these valves make it necessary to perform retests at normal

! -ac operating pressure and temperature (NOP/NOT) with steam present rather than during cold plant conditions. This test would be h.

. New Hampshire Yonkee Division of Public Service Company of New Hampshire y < EE . P.O. Box 300

• Seabrook, NH 03874

• Telephone (603) 474 9521 pon.n.;

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- United States Nuclear Regulatory Commission November 22, 1989 l

Attention Document Control Desk Page 2 ,

J performed while the plant is in a shutdown condition with all  ;

rcontrol rods inserted and the suberiticality margin greater than or equal to the MODE.3 Technical Specification requirement. In.

addition, the routine surveillance test which would be used as the l retest method for almost all maintenance / repair situations '

involves stroking one ASDV at a time with the block valve closed.

This would eliminate the potential for any cooldown effect.

i The failure of the turbine-driven EFW pump to operate when required .}

could involve a decrease in secondary heat removal capability. '

However, the proposed change applies only to stable shutdown conditions where the two motor-driven auxiliary feedwater (AFW) pumps (the startup feed pump (SUFP) and the motor-driven EFW  !

pump), are OPERABLE and the plant is preparing for a startup rather than performing a cooldown ur#tt emergency conditions.

The proposed change will not increase the consequences of an accident previously evaluated in the FSAR. The proposed change applies to MODE 3 therefore the reactor will be subcritical under the conditions stated above. These tests would not be performed '

with steam generator tube leakage or primary / secondary coolant activities in excess of technical specification limits.

i ASDV or turbine-driven EFW pump post maintenance / modification  ;

retests involve brief time intervals in stable shutdown plant-conditions as described above. Multiple failures in both fuel '

integrity and primary to secondary boundary integrity (unrelated to a failure in either the ASDV or turbine-driven EFW pump) would have to occur to cause an increase in the dose consequences.

2. Create the possibility of a new or different kind of accident from -

L any previously evaluated. The proposed change does not alter l- physical components and the failure of these components has been previously analyzed. The test methods used for the ASDVs will only involve the test of a single valve at a time. In the case of the turbine-driven EFW pump, the SUFP and the motor-driven EFW pump are required to be operable. '

L L The possibility of a malfunction of a different type than any l previously evaluated in the FSAR will not be increased. The failure of an ASDV to open when demanded or fail to close when

required are the worst case failure modes for these valves and have L been previously evaluated.

3. Involve a significant reduction in a margin of safety. The proposed changes apply to the administrative rules concerning entry into MODES 3 and 4 under conditions where the components involved ar expected to be capable of operation but have not undergone a retest to actually demonstrate OPERABILITY. The ASDVs are not L

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".: o United States Nuclear Regulatory Commission November 22, 1989 Attentjon Document Control Desk Page 3 specifically mentioned in the Technical Specification Bases.

'However, their primary purpose is to facilitate a controlled plant cooldown if the condenser is not available as a heat sink. The AFW pumps are mentioned in the bases for Technical Specification 3.7.1.2 as being required to support a'cooldown of the reactor coolant' system to less than 350 degrees with a loss of offsite power present.

If you have any questions regarding this request, please contact Mr. Richard R. Belanger at (603) 474-9521 extension 4048.

Very truly yours.

Y.hb .L Ted C. Feigenbaum Enclosures cci Mr. William T. Russell Regional Administrator United States Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia PA 19406 Mr. Victor Nerses. Project Manager Project Directorate I-3 United States Nuclear Regulatory Commission Division of Reactor Projects Washington, DC 20555 Mr. George L. Iverson, Director Office of Emergency Management State Office Park south 107 Pleasant Street Concord, NH 03301 1 Mr. Antone C. Cerne NRC Senior Resident Inspector P.O. Box 1149 Seabrook, NH 03874 1

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L New Hampshire Yankee November 22, 1989 L

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6 ENCLOSURE 1 TO NYN-89130 PROPOSED TECHNICAL SPECIFICATION CHANGES

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