ML20062A154

Application for Amends to Licenses DPR-24 & DPR-27, Incorporating Changes in Tech Specs to Clarify Part of Auxiliary Feedwater Section to Table 15.3.5-3

ML20062A154
Person / Time
Site:	Point Beach
Issue date:	03/30/1990
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20062A158	List: ML20062A154 ML20062A162
References
CON-NRC-90-033, CON-NRC-90-33 VPNPD-90-155, NUDOCS 9010190151
Download: ML20062A154 (5)

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VPNPD 15 5 10 CFR 50.90

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- March 30, 1990

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s U. S. NUCLEAR REGULATORY COMMISSION 4

~ Document Control' Desk. i Mail' Station P1-137.

Washington, D. C. 20555. 1 j Gentlemen: 1

' DOCKETS 50-266 AND 50-301

. TECHNICAL SPECIFICATION CHANGE REQUEST 138 l/ fINSTRUMENTATION SYSTEM OPERATING CONDITIONS ,

j, ' POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2' ,

I' In'accordance'with the requirements of 10 CFR 50.59(c), 50'90,. .

and 50.4, Wisconsin Electric Power Companys(Licensee) hereby

. requests amendments'to Facility Operating Licenses DPR-24 and u

-DPR-27:for-Point Beach Nuclear' Plants, Units 1 and 2,,

• respectively,Etol incorporate changes.inLthe plant Technical .

. Specifications. These proposed changes will: clarifyrpart of the  ;;

" Auxiliary Feedwater" section of Table-15.3.5-3; update Items 3 d L

4 1and: 4,." Safety Valve Position Indicator" and'" Reactor Coolant R System Subcooling" of Table 15.'3.5-5; correct a1 typographical-error in'15.3.0;-and revise the Unit 1'and. Unit 2' specification-

,, (pages 15.3.1-8a.. u o  : Item 3.b of Table 15.3.5-3lprovides(operating conditions for the. -

finstrument" channels which-start the motor-driven auxiliary. j L,j 'feedwater pumps upon a trip:of'both main feed pumps'.- This l

? chan6el'is. permitted to be bypassed when reactor power is-less J E than.40%.-'This permissible' bypass condition-was includedtin' d

[ , Amendments 119 and 122 to-Units 1 and,2,-respectively,;as a resultsof thetinstallation of:the ATWS mitigating, system R i actuation ! circuitry - ( AMSAC) . .

As.explainedAin our application for the AMSAC. amendment dated

' February 24, 1989, the signal which disables AMSAC is based'on

,00 . reactor-power as derived from turbine first-stage impulse J pressure. At normal operating power levels, turbine first-stage l . impulse pressure'has a near linear relationship with reactor-l 9010190151 900330 l' '

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. power based on' reactor thermal output (RTO). However, at lower

power levels (e.g., 40%), RTO will be 2% to 8% greater than that- I sensed by turbine first-stage impulse pressure. This difference results from thermal losses to the moisture separator reheaters andLateam. generator' blowdown. Thus, when AMSAC is disabled at a O turbine power of 40%, reactor thermal output may be greater than 40%. 1 The use.of the words " reactor power" in the Specification has  ;

resulted in some confusion. .The system was designed such that l

~AMSAC would be bypassed when reactor power derived from turbine ot .first-stage pressure was less than 40%. .This is clear in both'

'WCAP-10858, which dOwcribes the AMSAC design, and the Technical j specification Change Request which proposed the change to this ,

Specification. Recent confusion on the interpretation of this l Specification resulted in Point Beach operators entering a three- j hour'LCO during a power reduction when AMSAC' correctly disabled  ;

at 40% turbine first-stage power while reactor thermal newer

' indicated 42%. This was reported to the NRC in Licensee Event Report 89-008-01 dated December 11,. 1989.

j To clarify this Specification, we propose to modify the

" Permissible Bypass Condition" for Table 15.3.5-3, Item 3.b,  !

"TripioflBoth Main Feedpumps Starts Motor Driven Pumps" to read I

" Turbine power

• is less than 40%" and to add a footnote to'the  :

table, " Turbine power.as-derived from the turbine first-stage I impulse. pressure." l We also-propose to modify the " Operator Action" column for Item 3.b to'includez a Limiting condition for Operation-(LCO)-which will allow a 48-hour period to repair this functional unit should j both channels become inoperable. If, at'the-end-of the 48-hour' period, at least one channel is not restored to operable status, l C

tturbine power would be reduced to less than 40% within the next

six hours. This provides a reasonable time to effect repairs and l

-limits any. unnecessary plant cycling _which would be required by l J

Technical Specification 15.3.0.without'this LCO. This time' period'is also consistent'with the similar functional unit in NUREG-0852, Standard Technical' Specifications (STS) for ,

Westinghouse: Pressurized Water Reactors. You may note that the l

" :STS require a plant shutdown if at least one channel cannot.be l s, .~ restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.- We propose-a turbine power reduction toLless than 40%, since AMSAC is in a permissible bypass

~ 3 condition at-that power level.. In addition, to make the

" functional: unit" description of this item consistent with STS, and since the'PBNP AMSAC already provides for this function, we have included the-feature whereby the trip of both main feedwater i pumps also starts the turbine-driven auxiliary feedwater pump. l We have also clarified Item 3.b Columns 1, 2, and 3 to read 1

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'NkC Document l Control-Desk i .

. March 30, 1990

, Page 3 2/ Main Feed Pump, 1/ Main Feed Pump, and 1/ Main Feed Pump, respectively.

The second: set'of changes we are proposing involves Items 3 and 4 oof Table 15.3.5-5, " Safety Valve Position Indication" and

" Reactor Coolant System Subcooling", respectively. Modifications have been completed which provide two reactor coolant system safety valve position-indicator channels per valve and two channels of reactor coolant system subcooling. Column 1 of Table 15.3.5-5, Item 3, " Safety Valve Position Indicator", and Item 4,.

" Reactor Coolant System'Subcooling*'should, therefore, be updated

-to list two channels / valve and two channels,.respectively.

The third change we-are proposing corrects a typographic error on Specification.Page 15.3.0-3. The first sentence of the third paragraph presently reads, "For example, Specification 15.3.7.A.1.e allows a-7 day..." This sentence should actually reference Specification 15.3.7.B.1.e and should be changed accordingly.

The'last change we are proposing.will change' Specification page 15.'3.1-8a for both-units to make them the same. The Unit 2 page was modified in Unit ~2 Amendment 128 dated October 24', 1989 to acknowledge ourEparticipation in the, Babcock and Wilcox Master Integrated Reactor Vessel Surveillance Program andLto brieflyn provide a, basis for the reactor vessel surveillance capsule removal schedule. Since both units are included-in this; program and-to maintain common specifications betweenLunits,.we'are proposing;to modify the Unit 1-and Unit 2 specifications slightly to;again :make them' identical. . Please' note that the changes.

, discussed-in this and the immediate' preceding paragraph are changes to the' bases for:these specifications and do-not constitute Technical Specification changes.

As required'by 10 CFR 50.91(a), we have evaluated these changes- 1 in accordance with the. standards specified in 10 CFR 50.92 to "

idetermine'ifLthe. proposed change constitutes a significant hazards' consideration. A proposed amendment involves no significant hazards consideration if operation of.the facility in accordance with the proposed amendment would'not-(1) involve a lsignificant increase in the probability =or consequences of an

' accident previously evaluated, (2) create'the possibility of a new or different' kind of accident from any accident previously^

' evaluated, or (3): involve a significant reduction in'a margin of safety.

In Table 15.3.5-3, Item 3.b, we propose to include in the functional unit description the feature whereby the turbine-driven pump starts following a loss of both main feed pumps.

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i, LThis is a change that constitutes an additional control not L presently included in the Technical Specifications and, thus, is Q an< administrative change. We have also determined that the L' clarifications to Item 3.b Columns 1, 2, and 3, as well as the ,

. clarification which changes reactor power'to turbine power, are  !

l also' administrative changes. We have, therefore,. determined that  !

L these changes will not' involve a significant hazards.

consideration. These-changes are'similar to the examples of- 1 changes 1 considered not.likely to involve a significant hazards consideration as provided by the NRC at 51-FR 7751. We have also y determined that the allowance of a 46-hour period for_ testing or j for repairs, should both AMSAC channels be inoperable, will not L, involve a significant hazards consideration. The purpose of this L functional 1 unit is to help protect the reactor core from a loss -

D of heat-sink caused-by the tripping cf both main feedwater pump h breakers. However, as~can be seen in FSAR Section 14.1.10, " Loss of Normal Feedwater," even if this functional unit is unavailable 4

when both main feedwater pump breakers trip, the auxiliary feedwater pumps will still start on low-low level in either' steam ,

K generator and no adverse conditions will result in the reactor-core. In addition,'this functional unit would be tested during a relatively'short period of time only twice a year and the AMSAC bypassed condition is automatically.and continuously indicated in-the control room. The allowance of a test period where both'

l. . channels would be bypassed appears to have been considered and o found acceptable by'the NRC in the AMSAC Safety Evaluation Report ofESeptember 17, 1986. Similarly, the 48-hour period to return at least one channel to operable status is again a relatively short time} period and has also been previously considered _____

acceptable by inclusion in the .STS.crthus:; the probabl~1'ity or '

-consequences of'a'previously evaluated accident will~not,be

-increased;:no i new or different kind-of accident will be created;

-and,the. margin of safety, as shown in FSAR Section 14.1.10,'will 'j

.(not be reduced. ,. ~ ~ -

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ThA7chandeI proposed to Items.3 and 4 of' Table-15.3.5-5 result

, 'fromethe installation of new safety valve position indicators and reactoracoolant system subcooling monitors. This was done to DJ  ; meet'the requirements.of NUREG-0737 and Regulatory Guide 1.97.

The modifications which installed the new channels for subcooling-

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and safety valve position indication were previously evaluated under the requirements of 10 CFR 50.59. These evaluations provided reasonable assurance that the modifications did not

, involve an unreviewed safety question. This proposed Technical x Specification change simply acknowledges'the fact that there are now twofchannels available for each of these items; will not change the minimum number of operable channels required for operation; and, therefore, cannot create a new or different kind of accident. Since there are now two channels available for each  ;

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. NRC Document Control Desk 1 q March;30,41990 Page 5  ;

item, it is more likely that at least one will be available to

' plant operators-at any given time; and, thus, the probability or consequences'of an accident involving either item will not be l Lincreased-norza margin of safety reduced. Accordingly, we have 3 determined'that this change does not present:a significant  ;

hazards consideration.

=Regarding the changes to Specification PageL15.3.0-3 and  :

Specification 15.3.1.~8a, we have determined them to be administrative changes, since the former corrects a typographical i error and the latter provides consistency between the two' units.. ,

Thus, they are similar to the examples of purely administrative- i changes considered not likely.to involve a significant hazards Econsideration, as provided by the Commission =at 51 FR 7751. As Jc noted, these are changes to the bases and not to the ,

L Specification. On this basis, we.have determined that these changes:will also not result in a significant hazards '

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l. consideration.'

Attached to this application are proposed revised Technical. J

. Specification pages with the specific changes on each page r identified by margin ~ bars. These pages are intended to facilitate your review and approval of the amendment request.

- Please-contact us if_you.have any questions concerning this

-request.

.e Very truly yours, E

h"- -

l C. _ W. Fay J Vice President- '

, ' Nuclear Power

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Enclosures . m Copies to NRC Regional. Administrator, Region III NRC Resident Inspector Subscribed and' sworn to before me

.. this.:t Way of M ,~1990.

m ,

Notary Public, State'of Wisconsin  ;

My . Commission ' expires 5 2 7 'T C .

Blind copies to Boston, Gorske/Finke, Charnof f, Krieser, Lipke, Newton, Rodgers (OSRC), Schoon, Zach, File A.S.2 (138)

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