ML20112F729

TS Change Request 188 to Licenses DPR-24 & DPR-27,modifying TS 15.2.3, Limiting Safety Sys Settings & Protective Instrumentation & 15.5.3, Design Features - Reactor

ML20112F729
Person / Time
Site:	Point Beach
Issue date:	06/04/1996
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20112F732	List: ML20112F729 ML20112F734
References
VPNPD-96-035, VPNPD-96-35, NUDOCS 9606100081
Download: ML20112F729 (7)

Text

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Wisconsin

!Electnc POWER COMPANY j 231 W Michigan. PO Box 2046 Milwaukee. WI 53201 2046 (414)221-2345

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VPNPD-96-035 10 CFR 50.4 10 CFR 50.90 June 4,1996 Document Control Desk US NUCLEAR REGULATORY COMMISSION

Mail Station PI-137

Washington, DC 20555 Gentlemen

DO( f L FS 50-266 AND 50-301 TECHNICAL SPECIFICATIONS CHANGE REOUEST 188 -

MODIFICATIONS TO TS SECTIONS 15.2.3. " LIMITING SAFETY SYSTEM SETTINGS AND PROTECTIVE INSTRUMENTATION."

l AND 15.5.3. " DESIGN FEATURES - REACTOR."

l POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 l

In accordance with the requirements of 10 CFR 50.4 and 50.90, Wisconsin Electric Power Company l (Licensee) hereby requests amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP) Units 1 and 2, respectively, to incorporate changes to the plant Technical Specifications. The proposed revisions will modify Technical Specification Section 15.2.3," Limiting l Safety System Settings and Protective Instrumentation," and Section 15.5.3, " Design Features -

Reactor," to incorporate changes associated with the operation of PBNP, Unit 2 with replacement steam generators.

i Marked-up Technical Specifications pages, a safety evaluation, and the no significant hazards consideration are enclosed.

DESCRIPTION OF CURRENT LICENSE CONDITION Technical Specification Section 15.2.3, Specification B(2) lists the setpoint for the 1.igh pressurizer pressure reactor trip function as s 2385 psig.

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. Technical Specification Section 15.2.3, Specification B(3) lists the setpoint for the low pressurizer pressure reactor trip function as 21865 psig for operation at 2250 psia and 21790 for operation at 2000 psia.

Technical Specification Section 15.5.3.B.3 states that the nominal liquid volume of the Reactor Coolant System (RCS) at rated operating conditions is 6040 cubic feet.

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Document Control Desk j June 4,1996 Page 2 QESCRIPTION OF PROPOSED CHANGES j Specification 15.2.3.1.B(2) is being modified to read as follows:

"(2) High pressurizer pressure * - s 2385 psig for operation at 2250 psia primary system pressure s 2210 psig for operation at 2000 psia primary system pressure"

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In order to eliminate the possibility of an inadvertent reactor trip while adjusting these setpoints, we desire to make the adjustments during the next scheduled refueling outage for each unit. As j such, these proposed setpoints will be implemented for Unit 2 during the Fall,1996 refueling outage (U2R22) and for Unit I during the Spring,1997 refueling outage (UIR24). If approved, this TSCR will be implemented during U2R22 and the changes to the Unit 2 setpoints can be l made immediately. From the time this TSCR is implemented until UIR24, however, Unit I will be operating with the existing setpoints. Therefore, the following footnote is being added to the bottom of the page to clarify this situation:

- These values apply to Unit 2 following U2R22 and to Unit I following UlR24. Prior to UlR24, the high pressurizer pressure reactor trip setpoint for Unit 1 is s 2385 psig."

Specification 15.2.3.1.B(3)is being modified to read as follows:

"(3) Low pressurizer pressure * - 21905 psig for operation at 2250 psia primary system pressure 21800 psig for operation at 2000 psia primary system pressure" For the same reasons mentioned above, a footnote is being added to the bottom of the page to  ;

read as follows: l i

- These values apply to Unit 2 following U2R22 and to Unit I following UlR24. Prior to  ;

UlR24, the low pressurizer pressure reactor trip setpoint for Unit 1 is'21790 psig."

Specification 15.5.3.B.3 is being modified to read as follows:

"3. The nominal Reactor Coolant System volume (both liquid and steam) at rated operating conditions and zero percent steam generator tube plugging is:

Unit 1 - 6500 ft' 3

Unit 2 - 6643 ft "

l Document Control Desk June 4,1996 Page 3 ]

BASIS AND JUSTIFICATION We are planning to replace the steam generators in PBNP Urut 2 during the Fall,1996 refueling outage.

Westinghouse performed the safety analyses for operation of Unit 2 with the replacement steam generators. The analytical limit for reactor coolant system pressure was revised to implement an average coolant temperature (T.v,) window and increase operating margin to the overpower delta T (OPAT) reactor trip setpoint. As such, the values for the high and low pressurizer pressure reactor trip setpoints have been revised. The safety analyses for the new pressurizer pressure trip setpoints were performed to l cover both units. Thus, the proposed setpoints will apply to the operation of Unit I as well as Unit 2.

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Installation of the replacement steam generators changes the nominal volume of the reactor coolant i system as described in TS section 15.5.3. The proposed changes more accurately describe the nominal I total (liquid and steam) volume of the reactor coolant system for both units.

These changes are necessary to provide for the safe operation of PBNP Unit 2 with the replacement steam generators. Therefore we request approval of these changes by September 1,1996.

We have determined that the proposed amendments do not involve a significant hazards consideration, authorize a significant change in the types or total amounts of any effluent release, or result in any I significant increase in individual or cumulative occupational exposure.' Therefore, we conclude that the - l proposed amendments meet the requirements of 10 CFR 51.22(c)(9) and that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared. l 1

If you require additional information, please contact us.

Sincerely, Bob L nk Vice President Nuclear Power l

l. KVA l cc: NRC Resident Inspector j NRC Regional Administrator PSCW i

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TECIINICAL SPECIFICATIONS CIIANGE REOUEST 188 SAFETY EVALUATION INTRODUCTION Wisconsin Electric Power Company (Licensee) has applied for amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant Units 1 and 2. The proposed revisions will modify Technical Specification (TS) Section 15.2.3, " Limiting Safety System Settings and Protective Instrumentation," and Section 15.5.3, " Design Features -

Reactor," to incorporate changes associated with the operation of the PBNP, Unit 2 replacement steam generators.

EVALUATION We are planning to replace the steam generators in PBNP Unit 2 during the Fall,1996 refueling outage. Westinghouse performed the safety analyses and evaluations for operation of Unh 2 with the replacement steam generators including; large and small break LOCA, LOCA forces, non-LOCA events, LOCA long and short term mass and energy release, steam generator tube j rupture, and systems and components. The safety analyses and evaluations examined the operating conditions with the new steam generators including an average coolant temperature (Tog) window, an operating pressure of either 2000 or 2250 psia, a reduced thermal design flow, a slightly larger primary volume, and a slightly smaller secondary volume. All analyses and I evaluations were performed in accordance with NRC approved methodologies.  !

The analytical limit for reactor coolant system pressure was revised to implement the average coolant temperature window and increase operating margin to the overpower delta T (OPAT) reactor trip setpoint. As such, the values for the high and low pressurizer pressure reactor trip setpoints have been revised.

The safety analyses and evaluations reached the following conclusions: I e The changes associated with the replacement steam generators do not result in exceeding any LOCA design or regulatory limits. l

. Non-LOCA acceptance criteria continue to be met.

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. The containment design pressure limit will continue to be met.

• Steam generator tube rupture doses meet the acceptance criteria of a "small fraction"  :

of 10 CFR 100 limits.  ;

1 In order to maintain one set of safety analyses for both units, the analyses for operation of Unit 2 I with the replacement steam generators were performed to encompass the operation of Unit 1. l Therefore, the proposed changes apply to the operation of both units and maintain the margin of safety for each.

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1 l The proposed changes to the description of nominal liquid and steam volume for the reactor coolant systems will more accurately describe the total (liquid and steam) volume for each unit.

The existing wording describes only an approximate liquid volume of the RCS for both units. '

CONCLUSION I l

1 The proposed changes were arrived at through safety analyses and evaluations performed in  ;

accordance with NRC approved methodologies. The proposed changes will ensure the safe operation of both Unit 2 with replacement steam generators and Unit I with its existing steam generators. Therefore, approval of these changes will ensure and enhance the continued safe operation of Point Beach Nuclear Plant.

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9 TECHNICAL SPECIFICATION CHANGE REOUEST 188 NO SIGNIFICANT HAZARDS CONSIDERATION 1 1

In accordance with the requirements of 10 CFR 50.91(a), Wisconsin Electric Power Company (Licensee) has evaluated the proposed changes against the standards of 10 CFR 50.92 and has determined that the operation of Point Beach Nuclear Plant, Units 1 and 2, in accordance with the proposed amendments does not present a significant hazards consideration. The analysis of the requirements of 10 CFR 50.92 and the basis for this conclusion are as follows:

1. Operation of this facility under the proposed Technical Specifications will not create a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not involve a change to structures, systems, or components which would affect the probability or consequences of an accident previously evaluated in the PBNP Final Safety Analyses Report (FSAR). The proposed setpoints maintain the  ;

margin to safe operation of Unit 2 with the replacement steam generators. In order to i maintain one set of safety analyses for both units, the analyses for operation of Unit 2 with the replacement steam generators were performed to encompass the operation of Unit 1.

Therefore, the proposed changes apply to the operation of both units and maintain the margin of safety for each. The proposed change to the description of nominal RCS  !

volume is an administrative change and has no effect on plant operation. Therefore, the probability or consequences of a previously evaluated accident are not significantly increased as a result of these changes.  !

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2. Operation of this facility under the proposed Technical Specifications change will not  !

- create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve a change to the plant design. The proposed setpoints maintain the margin to safe operation of Unit 2 with the replacement steam generators. In order to maintain one set of safety analyses for both units, the analyses for operation of Unit 2 with the replacement steam generators were performed to encompass the operation of Unit 1. Therefore, the proposed changes apply to the operation of both units and maintain the margin of safety for each. These changes do not affect any of the parameters or corditions that contribute to initiation of any accidents. The proposed change to the description of nominal RCS volume is an administrative change and has no effect on plant operation or initiation of any accidents. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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3. Operation of this facility under the proposed Technical Specifications change will not create a significant reduction in a margin of safety. l The proposed setpoints maintain the margin to safe operation of Unit 2 with the replacement steam generators. In order to maintain one set of safety analyses for both ,

units, the analyses for operation of Unit 2 with the replacement steam generators were i performed to encompass the operation of Unit 1. Therefore, the proposed changes apply I to the operation of both units and maintain the margin of safety for each. The proposed change to the description of nominal RCS volume is an administrative change and has no effect on plant operation. Therefore, the proposed changes will not create a significant reduction in a margin of safety.

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