ML20213D697

Application for Amend to License NPF-42,changing Tech Spec 3/4.7.1.6 & Associated Bases to Incorporate Limiting Condition for Operation & Surveillance Requirements for Steam Generator Atmospheric Relief Valve.Fee Paid

ML20213D697
Person / Time
Site:	Wolf Creek
Issue date:	11/07/1986
From:	Koester G KANSAS GAS & ELECTRIC CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20213D700	List: ML20213D697 ML20213D737
References
KMLNRC-86-210, NUDOCS 8611120196
Download: ML20213D697 (8)

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KANSAS GAS AND ELECTRIC COMPANY THE ELECTRIC COWANY GLENN L MOESTER vict pensiopst. soucLEam November 7, 1986 Mr. B. R-. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Cmmission l Washington, D.C. 20555 l

l KMLNRC 86-210 Re: Docket No. STN 50-482 Subj: Proposed Technical Specification Supporting Steam Generator Tube Rupture Analysis

Dear Mr. Denton:

he purpose of this letter is to transmit three original and 40 conformed copies of an amendment to Facility Operating License No.

NPF-42 for Wolf Creek Generating Station, Unit No. 1. Sutmittal of this requested amendnent supports, in part, resolution of License Cbndition 2.c (11) of Operating License NPF-42.

We application for amendnent transmits a proposed Wolf Creek Generating Station, Unit No. 1, Technical Specification 3/4.7.1.6 and its associated Basas. The proposed changes to the Technical Specifications are provided in Attachment III. A complete Safety Evaluation and Significant Hazards Consideration are provided in Attachments I and II respectively.

In accordance with 10 CFR 50.91, a copy of the awlication, with attachments is being provided to the designated Kansas State Official.

Enclosed is a check for the $150.00 application fee required by 10 CFR 170.21. l The proposed modification of the Wolf Creek Generating Station Technical Specifications will be fully implenented within 30 days of formal Nuclear Regulatory Cm mission approval.

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201 N. Market - Wichita, Kansas - Mail Address: PO. Box 208 i Wchita, Kansas 67201 - Telephone: Area Code (316) 261-6451 tp

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Mr. H. R. Denton Novenber 7, 1986 10 G RC 86-210 Page 2 If you have any questions concerning this matter, please contact me or Mr. O. L. Maynard of my staff.

Very truly yours, M@ A Glenn L. Koester Vice President - Nuclear <

GLK:wbb Enclosure ,

Attachments: I-Safety Evaluation II-Significant Hazards Consideration Analysis III-Proposed Technical Specification Changes cc PO'Connor JCumins GAllen EJohnson

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STATE OF KANSAS )

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I, John A. Bailey, of lawful age, being first duly sworn upon oath, do depose, state and affirm that I am Director Engineering and Technical Services of Kansas Gas and Electric Company, Wichita, Kansas , that I have r signed the foregoing letter of transmittal for Glenn L. Koester, Vice President -

Nuclear of Kansas Gas and Electric Company, know the content thereof; and that all statements contained therein are true.

By John A. Bailey /

Director Engineering and 'Iechnical Services .

SUBSCRIBED and sworn to before me this 7 day of , 1986. , ','l. . .'.t .'. f,,

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Mr. H. R. Denton November 7,1986 o Attachment I to IDENRC 86-210 Page 1 SAFE 1Y EVAWATICE Introduction

' Wolf Creek License Condition 2.c (11) states: " Prior to restart following the first refueling outage, nE shall sutznit for NRC review and approval an analysis which demonstrates that the steam generator tube rupture (SGTR) analysis presented in the FSAR is the most severe case with respect to the release of fission products and calculated doses. Consistent with the analytical asstmiptions, the licensee shall propose all necessary changes to Appendix A to this license".

%e SGTR analysis has been ca pleted and subnitted to the NRC for review (References 1, 2, and 3).. h is analysis : takes credit for Atmosgheric - Relief Valve (ARV) operation for SGTR mitigation and states that DE will sutznit a Technical Specification amendnent request concerning the operability of the ARV's in a SGTR event.

Appropriate Technical Specification requirements have been developed.

Se proposed requirements state that three of the four steam generator atmospheric relief valves shall be operable in Modes 1, 2, and 3. If only two ARV's are operable, this specification allows 30 days to return one of the two inoperable ARV's to operable status. If only one or no ARV's are operable, this specification allows'only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to return to at least two operable ARV's. An action statanant is provided which allows the plant to start up if the Limiting Condition of ' Operation (ICO) is not met. % e proposed Technical Specification would not increase the surveillance requirements for the ARV's beyond those addressed in existing Specification 4.0.5. Finally, a footnote is included to specifically note that an ARV renains operable if its associated block valve is closed solely because of ARV leakage.

Evaluation Se purpose of the license ;-.c Ar.ad. request to incorporate technical

} specification ICO and surveillance requirements for the steam j generator ARV's into the Wolf Creek Operating License is'to assure the availability of mitigating equipnent assuned in the SGTR analysis.

Se technical specification requirements constitute additional limitations on facility operations and satisfy, in part, the specific requirements of License Condition 2.c (11) of the Operating License.

No requirements on ARV operability have been included in the existing

Wolf Creek Technical Specifications because the ARv's have not been i

required in the mitigation of postulated accidents and transients.

i ne Wolf Creek Generating Station design includes four ARV's, one ARV for each steem generator. S e ARV's are safety-grade and have Class 1E electrical control circuits (Reference 4).

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Mr. H.~R. Denton November 7, 1986 Att+ a. I to IG GIRC-210 Page 2 he operability of at least three of 'the four ARV's ensures that reactor decay heat can be dissipated to the atmosphere in the event of

'. a SGTR coincident with a loss of offsite power and that the Reactor Coolant Systen (RCS), can be cooled down to the point of Residual Heat i,

Removal System (RRR) operation. In the initial phases of a SGTR event, l the ARV's are relied upon for achieving proper RG subcooling i prior to equalizing pressures between the RG and the faulted steam generator. 'We proposed nunber of operable ARV's assures that subcooling can be adlieved within assuned time constraints' and in

> accordance with single failure assunptions used in the SGIR analysis.

4 For the RCS cooldown to RHR initiation, only one ARV is needed for the

[ heat removal required. Three operable ARV's are adequate, assuning that one of the operable valves is on the faulted steam generator, and r thus is not available for heat removal, and that one ARV fails to

function in accordance with accident assunptions.

. The proposed Technical' specifications are applicable in plant '

i operational Modes 1, 2, and 3. Since the purpose of the ARV's is to provide for removal of reactor decay heat during the initial @ases of the SGTR event and during cooldown to RHR initiation, and since the RHR system is available in Modes 4, 5, and 6 this Technical Specification is only applicable in Modes 1, 2, and 3.

i Each ARV is equipped with a manual block valve, located near the ARV in the auxiliary building, to provide a positive shutoff capability should an ARV develop leakage. Se ARV is considered operable if the block valve is closed solely because of leakage. Closure of the block

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valves of all ARV's because of leakage does not endanger the reactor

!.. core because decay heat can be dissipated with the steam line safety I

valves. Also, consistent with SGTR analysis assu@tions, a block valve can be used to control release of steam to the atmosphere.

%e restoration time periods provided in the proposed LCD Action statement are based on the low likelihood of having a SGTR event coincident with a loss of offsite power during the time period that

! one or more of the required ARV's is out of service.

period (30 days) for the case of one inoperable ARV isThe allowed time i longer than the time period (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) allowed when more than one ARV are inoperable.

%e surveillance requirements of the Wolf Creek inservice valve testing 1

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' program for ASME Code valves are considered appropriate and sufficient for the ARV's. %erefore, the surveillance requirements in the j proposed Specification Technical4.0.5. Specifications refer to the provisions of existing  !

Testing of ARV's at power which involves stroking i the valves is performed with the block valves in the closed position. '

h is precludes a plant transient associated with inadvertent failure of an ARV in the open position.

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Mr. H. R. Denton Novenber 7, 1986 Attachment I to RMDEC 86-210 Page 3 (bnclusions Based on the above evaluation, it is concluded that the proposed ARV Technical Specifications constitute new I00 and surveillance requirements which govern operability of existing equignent in the Wolf Creek design. It was also concluded that the proposed Technical Specifications would not adversely affect the conclusions in existing plant transient and accident analyses and that the proposed requirements provide acceptable operability limitations for ARV's consistent with the asstaptions used in the SMR anlysis. '1herefore, the level of plant safety described in the Wolf Creek FSAR is not adversely affected by the proposed amendnent.

References

1. SI2EC 86-01, dated 01/08/86, Steam Generator Tube Rupture Analysis - SNUPPS

2. SLNRC 86-03, dated 02/11/86, Steam Generator Tube Rupture Analysis - SNUPPS

3. SINRC 86-08, dated 04/01/86, Steam Generator Tube Rupture Analysis - SNUPPS

4. Wolf Creek Generating Station Final Safety Analysis Report, Section 10.3 1

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Attachment II to RMDEC'86-210 Page 1 STMTPICAMP HAEARDS (rmTr5MATION ANALYNTS h proposed Technical Specifications for steam generator Atmospheric Relief Valves (ARV's). incorporate new requirements into the Wolf Creek Operating -License. These requirements constitute additional limitations on facility operations to assure operability of the ARV's and to i==a operating restrictions if less than the required ntaber of ARV's is available.

h proposed specifications do not involve a significant hazards consideration because operation of Wolf Creek Generating Station

l. (WOGS) in accordance with the proposed requirements would not: '

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1. involve a significant increase in the probability or consequenses of an accident previously evaluated. 'the ARV's are not relied upon for mitigation of accidents or transients, other than the Steam Generator Tube Rupture (3GTR) event, previously evaluated in the WOGS Final Safety Analysis Report The Limiting Conditions for Operation (IOD) do not (FSAR) .  !

4 alter the manner in which ARV operation was considered in previous accident and transient analyses. Surveillance testing of ARV's, which involves stroke testing, is performed (in accordance with existing surveillance test program procedures) l  ;

with the block valves closed. Thus, a plant transient is '

precluded. To assure that the ARV's are available for

! mitigation of a postulated SGIR event, the proposed specifications establish surveillance requirements and restrictions on ARV operability concistent with the asstmptions used in the SGTR event analysis submitted to the NRC in accordance with WOGS License Condition 2.c (11) . Therefore, the

proposed Technical Specification requirements would neither increase the probability of an accident nor increase the ,

consequences of accidents previously evaluated. '

2. create the possibility of a new or different kind of accident

from any previously evaluated. Establishing new ICO and surveillance requirements for the existing NOGS ARV's does not result in the possibility of new or different types of I accidents. l

3. I involve a significant reduction in a margin of safety. 'the '

proposed Technical Specifications assure that the margin of {

' safety established in the current SGTR analysis is maintained. i As discussed in items 1 and 2 above, the proposed LOO and surveillance requirements do not alter the margins to safety established in previous accident and transient analysis or in j established surveillance test programs.  :

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Mr. H. R.. W November 7, 1986 Attachment II to IDENRC 86-210 Page 2 h NRC has established guidance concerning the determination of whether a significant hazards consideration exists by providing certain exanples (51 m 7751) of amendnents not likely to involve a significant hazards consideration. h proposed ARV Technical Specifications conform to NRC exanple (ii) "A change that constitutes an additional limitation, restriction or control not presently included in the technical specifications, e.g. a more stringent surveillance requirement."

Based on the above discussion, supplemented by the existing similarity of the proposed specification requirenents to an NRC example of a change not likely to involve a significant hazards consideration, it

~has been concluded that the proposed ARV Technical Specifications do not involve a significant hazards consideration.

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