ML20198B270

Informs That Staff Has Incorporated Rev of Bases for TS 3/4.7.1.2, Afs Into WCGS Tss,Per 981108 Request.Rev Specifies Essential SWS Requirements for turbine-driven Afs. Overleaf Pages Provided to Maintain Document Completeness

ML20198B270
Person / Time
Site:	Wolf Creek
Issue date:	12/16/1998
From:	Poslusny C NRC (Affiliation Not Assigned)
To:	Maynard O WOLF CREEK NUCLEAR OPERATING CORP.
References
TAC-MA4196, NUDOCS 9812180127
Download: ML20198B270 (4)

Text

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• December. 16, 1998-

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' 1

' Mr. Otto L' Maynard '

i President and Chief Executive Officer ,

Wolf Creek Nuclear Operating Corporation Post Office Box 411'

. Burlington, Kansas 66839 e>

SUBJECT:

WOLF CREEK NUCLEAR GENERATING STATION - TECHNICAL SPECIFICATION BASES CHANGE, AUXILIARY FEEDWATER SYSTEM

. (TAC NO. MA4196)

Dear Mr. Maynard- I The staff has incorporated the revision of the Bases for Technical Specification 3/4.7.1.2

Auxiliary Feedwater System" requested by your letter dated November 6,1998, into the Wolf l Creek Nuclear Generating Station Technical Specifications. The revision specifies the essential -

l service water system requirements for the turbine-driven auxiliary feedwater pump.

l This letter acknowledges the revision. The corresponding overleaf page has been provided to maintain document completeness.

Sincerely, Original Signt By .

Chet Posiusny, Prg 9t Manager Project Directorate IV l

Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation l.

Docket No. 50-482 DISTRIBUTION Docket File

Enclosure:

Bases Pages PUBLIC PDIV-2 Reading cc w/ encl: See next page EAdensam (EGA1)

OGC WBateman EPeyton KThomas ACRS PGwynn, RIV WJohnson, RIV 1800W

^

TMarsh GHill(2)

I DOCUMENT NAME: WCA4196.LTR *See Previous Sheet for Concurrence j l OFC PDIV-h PDIV-2/LA NRR:SPLB l l: NAME CPoslNny EM TMarsh*

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- DATE 12/l% 98 12/cl/98 12/01/98 OFFICIAL RECORD COPY 9812190127 981216

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Mr. Otto L. Maynard -2 December 16, 1998 cc w/enci:

Jay Silberg, Esq. _ Chief Operating Officer ~

Shaw, Pittman, Potts & Trowbridge Wolf Creek Nuclear Operating Corporation l 2300 N Street, NW P. O. Box 411 -

Washington, D.C. 20037 Burlington, Kansas 66839 Regional Administrator. Region IV Supervisor Licensing U.S. Nuclear Regulatory Commission Wolf Creek Nuclear Operating Corporation 611 Ryan Plaza Drive, Suite 1000 P.O. Box 411 L

Arlington, Texas 76011 Burlington, Kansas 66839 '

- Senior Resident inspector U.S. Nuclear Regulatory Commission .

U.S. Nuclear Regulatory Commission Resident inspectors Office 1 P. O. Box 311 8201 NRC Road Burlington, Kansas 66839 Steedman, Missouri 65077-1032 l

Chief Engineer -

Utilities Division l

Kansas Corporation Commission 1 1500 SW Arrowhead Road I Topeka, Kansas 66604-4027 1 Office of the Govemor State of Kansas Topeka, Kansas 66612 Attorney General Judicial Center 301 S.W.10th 2nd Floor Topeka, Kansas 66612 County Clerk Coffey County Courthouse Burlington, Kansas 66839 Vick L. Cooper, Chief Radiation Control Program Kansas Department of Health and Environment Bureau of Air and Radiation Forbes Field Building 283 Topeka, Kansas 66620

! l i

i

. 3/4.7 PLANT SYSTEMS I '

BASES 3/4.7.1 TURBINE CYCLE d

3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% (1320 psia) j of its design pressure of 1200 psia during the most severe anticipated system l operational transient. The mar.imum relieving capacity is associated with a turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, (1971 Edition). The total relieving capacity for all valves on all of the steam lines is 18.23 x 108 lbs/h which is 115% of the total secondary steam flow of 15.85 x 108 lbs/h at 102% RATED THERMAL POWER. A minimum of two OPERABLE l safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

l STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases:

For four loop operation:

SP = I ) X~ I II ) x (109).

Where:

SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, X = Total relieving capacity of all safety valves per steam line in 1bs/ hour.

WOLF CREEK - UNIT 1 B 3/4 7-1

PLANT SYSTEMS BASES SAFETY VALVES (Continued)

Y = Maximum relieving capacity of any one safety valve in lbs/ hour.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The Auxiliary Feedwater System (AFW) system is configured into three inde)endent AFW pumps and associated flow 23ths. An AFW pump and associated disclarge flow path are considered OPERABLE when the components and flow 3aths regired to provide redundant AFW flow to the steam generators are OPERABLE.

This requires that the two motor-driven AFW (MDAFW) pumps be OPERABLE in two diverse paths, each ca)able of automatically transferring the suction from the condensate storage tant to an Essential Service Water (ESW) supply and supplying AFW to two steam generators. This requires the turbine-driven AFW ,

(TDAFW) pump to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MY Vs. and shall be capable of automatically transferring the suction from the condensate storage tank to an ESW supply and supplying AFW to the steam generators. The piaing, valves, instrumentation, and controls that are in the required flow pat 1 and are required for the train to perform its specified function (s), are also required to be OPERABLE. -

Because each ESW supply flow path to the TDAFW pump provides 100% ca the " Required ESW Supply" to the TDAFW pump is provided ,

by a single.p OPERABLE, supply flow Jath (the suction flow path begins at the point where the ESW piping brancies into two lines, one supplying the MDAFW pump and one supplying the TDAFW pump, and ends at the suction of the TDAFW pump) and associated OPERABLE suction isolation valve. '

The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power.

Testing of each electric motor-driven auxiliary feedwater pump on a fixed orifice recirculation flow and ensuring a discharge )ressure of greater than or equal to 1535 psig verifies the capability of eac1 pump to deliver a total feedwater flow at the pum) discharge of 575 gpm and creating pressure of 1221 psig to the entrance of t1e steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow at the pump discharge of 1145 gpm and creating a pressure of 1221 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the RHR System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss-of-offsite power and then a cooldown to 350 F at 5'O F per hour. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 SPECIFIC ACTIVITY-The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

WOLF CREEK - UNIT 1 B 3/4 7-2 December 16, 1998

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