ML20080C770

Application for Amend to License NPF-42,revising TS 4.7.1.2.1.b.1 & 4.7.1.2.1.b.2 to Clarify SR for Verifying Correct Required Position for Valves in Afs

ML20080C770
Person / Time
Site:	Wolf Creek
Issue date:	12/12/1994
From:	Johannes R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20080C772	List: ML20080C770 ML20080C774
References
CO-94-0032, CO-94-32, NUDOCS 9412200139
Download: ML20080C770 (11)

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. NUCLEAR OPERATING CORPORATION Rchard N. Johannes '

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December 12, 1994 Co 94-0032 U.'S. Nuclear Regulatory Commission ATIN i - Document Control Desk Mail Station P1-137 Washington, D. C.

20555

Subject:

Docket No. 50-482: Revision to Technical Specification 4.7.1.2.1 Gentlemen:

This. letter transmits an application for amendment to Facill'ty Operating License No. NPP-42 for Wolf Creek Generating Station (WCGS). _ This license amendment request proposes revising Technical Specifications 4.7.1.2.1.b.1 and 4.7.1.2.1.b.2 to clarify the surveillance requirements for verifying the -

correct required position for the valves in the auxiliary'feedwater system.

Attachment I provides a safety evaluation including a description' of the proposed change.

Attachment II provides a

no significant hazards consideration determination and Attachment III provides an environmental-impact determination.

The specific change to the technical specification proposed by this request is provided in Attachment IV.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Kansas state official.

This proposed revision to the WCGS technical specifications will be fully implemented within 30 days of formal Nuclear Regulatory Commission approval.

9412200139 941212

4. B x 411/ Burtingt n, KS 66839 / Phone:(316) 364-8831 l

PDR ADOCK 05000482 An Equal opportundy Employer M,T/HcNET P

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y%, NIfTyou : havel any iguestions.(concerning i this matter, ; please" contact [ medat;-

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Attachments I - Safety Evaluation II'- No Significant Hazards Consideration. Determination o'

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III - Environmental Impact Determination'.

IV - Proposed Technical Specification Change ccs' G..W. Allen (KDHE), w/a' L. J.' Callan 1(NRC),, w/a D. D. Chamberlain (NRC) ', w/a-J J. P. Ringwald (NRC), w/a a

J. C. Stone (NRC), w/a-r I

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ISTATE OF.. KANSAS.

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cocarrY OF COFFEY.

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Richard.N. Johannes,' of lawful. age,, being first duly! sworn upon' oath says. that$ #

he; is... Chief. Administrative officer of. ~ Wolf.l Creek. Nuclear. ; Operating '

I.,'l Ecorporationi. that he has( ' read the - foregoing document and knows ?the ; content 1

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thereof;. that'he has executed that same.for.and on behalf,of said corporation:.

with. full. power ~and authority to do'so; and:that;theifacts:therein stated are:<.

4 true and correct to the best of.his knowledge, information'and belief.

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~1 3 NotaryPublic Stato cf Kansae

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SUBSCRIBED and swor.1 to before me this 5

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p; fM Attachment I to CO 94-0032 Page 1.of 4 ATTACHMENT I SAFETY EVALUATION i

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Attachment"I:to Co.94-0032'

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safety Evaluation p

Proposed Channe This license' amendment request proposes revising' Technical' Specification l

Surveillance Requirements 4.7.1.2.1.b.1 and 4.7.1.2.1.b.2 to clarify the

- i re,, irements for verifying the correct required position for the-valves in the.

. auxiliary feedwater system.

Specifically, this.-proposed change' would-stipulate that the discharge control valves for the auxiliary feedwater pumps' must be fully open when the reactor is'above 10 percent rated thermal power'or-whenever the auxiliary feedwater system is placed;in standby.

Also, it'would specify that all other valves except for the discharge control valves in the auxiliary feedwater system flowpath that are not locked, sealed,.or otherwise

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. secured in position, be in the correct position.

l Evaluation Currently, Technical Specification Surveillance Requirement 4.7.1.2.1.b.1L j

requires the verification that each non-automatic valve in the ' auxiliary feedwater system flow path that is not locked, sealed,'or otherwise secured.in' d

position is in its correct position.

Also, Technical Specification'

'j Surveillance Requirement 4.7.1.2.1.b.2 requires the verification ' that each automatic valve in the auxiliary feedwater system flowpath is. in the fully' 1

open' position whenever the.nuxiliary feedwater' system is placed in automatic."

control'or when the reactor is above 10' percent rated thermal power.

These-surveillance requirements are applicable in Modes 1, 2, and 3.

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Each motor-driven auxiliary feedwater ' pump discharges through' a nonreturn H

valve and a locked-open isolation valve to feed two steam generators through individual sets.of a locked open isolation valve, a normally.open, motor-operated control valve, a check valve followed by a flow restriction orifice, I

and a locked-open globe valve.

The turbine-driven ' auxiliary feedwater pump discharges through'a nonreturn valve, a locked-open gate valve to each of:the four steam generators through ' individual. sets of a locked-open - isolation

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valve, a normally. open air-operated control valve,. followed by a nonreturn

'l valve,,a flow restriction orifice, and a locked-open globe valve.

(Reference Figure 1).

'During emergency conditions, the motor-driven auxiliary feedwater pump normally open control valves are automatically operated to. limit runout flow'

. under all secondary side pressure conditions. ' This is. required to prevent pump' suction cavitation at ' high flow rates.

The turbine-driven auxiliary.

feedwater pump design includes a lower net positive suction head requirement.

.Therefore, the turbine-driven auxiliary feedwater pump normally open control valves are remote manually operated.

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Attachment'I to CO 94-0032 Page-3.of 4 Currently, the turbine-driven auxiliary feedwater pump remote manual discharge control valves are administrative 1y controlled as automatically' operated valves, although the Updated Safety Analysis Report states that the. valves are remote manually operated.

Administrative 1y controlling these valves as automatic instead of non-automatic allows ~all of the auxiliary feedwater pumps discharge control valves-to be treated identically under Surveillance Requirement'4.7.1.2.1.b.2.

If the remote manual discharge control valves were considered non-automatic, Surveillance Requirement. 4.7.1.2.1.b.1 would be-

' applicable.

The application of this surveillance requirement would require entry into the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Limiting Condition for.. Operation wher;ever the turbine-driven auxiliary feedwater pump remote manual discharge control valves are throttled following a reactor trip.

This requirement would appear to be excessive since the discharge control valves are required to be throttled following a reactor trip since decay heat drops off rapidly after the reactor

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trip.

It is also being proposed that " automatic control" be changed to " standby" in Technical Specification 4 7.1.2.1.b.2.

Currently, the Operations department does not utilize the' terminology

• automatic control," but instead uses

• standby."

This proposed change would not alter when the auxiliary feedwater pump discharge control valves are required to be in.. the fully open position from the current technical specification requirements.

The proposed changes to the technical specification'curveillance' requirements

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described above are being requested to clarify that the discharge centrol valves for the motor-driven and turbine driven auxiliary feedwater' pumps are required to be in the ' fully open position whenever the auxiliary feedwater system is placed in standby or when the reactor is above 10 percent rated thermal power. The change is administrative in nature since the change merely clarifies the demonstration of operability required. in the surveillance requirements.. Also, the proposed change would alleviate confusion during the performance of the surveillance requirements.

l Based on the above discussions and the no significant hazards consideration determination presented in Attachment II, the proposed changes do not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis reports or create a possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis i

reports or reduce the margin of safety as defined in the basis for any technical specification.

Therefore, the proposed changes do not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.

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ai Attachment II: to CO 94-0032-PacJe 1. of. 2.

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ATTACHMENT'II NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

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?Attachmcht II to CO 94-0032-

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Page 2'of12 No Significant'Masards Consideration Determination a

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This license, amendment request proposes revising Technical -Specification-r Surveillance : Requirements. 4.7.1.2.1.b.1 - and ' 4.7.1.2.1.b.2 to stipulate that' the discharge control valves for'the auxiliary feedwater pumps must be-fully-open when the reactor is above 10 percent rated thermal power or whenever the auxiliary feedwater system is placed in standby.

Also, it would specify that all other valves except for the discharge control' valves. in the auxiliary feedwater system flowpath that are not locked, sealed, or,otherwise secured in-

. position, be in the. correct position.

Standard I - Involve

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Significant Increase

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Consequences of an Accident Previously Evaluated The proposed changes.do not affect the ability of the - auxiliary. feedwater system to perform 'its intended safety. function.

The changes are-administrative in nature since they',merely clarify - the demonstration of operability required,in'the surveillance requirements.

Standard II - Create the Possibility of a New or Different Kind of Accident

,from any Previously Evaluated There are '. no.new failure modes or mechanisms associated with. the proposed changes.

The changes are administrative changes to remove confusion when performing surveillance requirements to demonstrate operability.

Standard III - Involve a Significant Reduction in the Margin of. Safety These ' proposed - changes ' do not effect any technical : specification margin' of safety..

The changes only provide clarification for. performance. 'of' surveillance requirements.

Based on the above discussions, it has been determinedthat u the - requested technical specification. revision' does ' not involve ' a significant increase: int the probability or consequences of an accident or other adverse condition over' previous evaluations; or. create'the possibility of an new or different kind of accident or condition over previous evaluations;. or. involve a,significant reduction.in a margin of ' safety.'

The requested license amendment does not

, involve a significant hazards considtration.

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,.Attac ment III to CO 94-0032 h

, Page.-l.of 2, ATTACHMENT III ENVIRONMENTAL IMPACT DETERMINATION J

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t Attachment III to.CO 94-0032 j

Page 2'of:2i Environmental-Impact Determination b,

10 CFR L 51.22 (b). specifies the criteria for categorical. exclusions from the t;

requirement'. for 'a specific. environmental assecament " per 10.CPR 51.21.. 'This amendment request meets' the criteria specified in 10 CFR 51. 22 (c) (9) as specified below:

r (1).

the amendment involves no significant hazards consideration As demonstrateu in Attachment II, the proposed change does not. involve any significant hazards consideration.

(ii) there is.no significant change in the types or significant increase in the amounts of any effluents that may be released offsite The proposed change does not ' involve a change to the. f acility or operating procedures which would cause an increase in the amounts of effluents or create new types of effluents.

(iii) there is no.significant. increase in individual

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cumulative.

occupational radiation exposure The proposed change does not create. additional exposure to personnel nor affect levels of radiation present. Also,'the proposed change does not result; j.

in any increase in individual or cumulative occupational radiation exposure.

jj Based on the above. it is concluded' that there will. be no impact ~ on the environment' resulting from this. change and the ' change meets the criteria specified in 10 CFR 51.22 for a categorical 1 exclusion from the. requirements of.

10 CFR 51.21 relative to requiring a specific. environmental assessment'by the Commission, f

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