ML20211N506

Forwards TSs Bases Pages B 3/4 10-2 & B 3/4 2-4 for LGS, Units 1 & 2,being Issued to Assure Distribution of Revised Bases Pages to All Holders of TSs

ML20211N506
Person / Time
Site:	Limerick
Issue date:	09/09/1999
From:	Buckley B NRC (Affiliation Not Assigned)
To:	Hutton J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
References
TAC-MA5924, TAC-MA5925, NUDOCS 9909130008
Download: ML20211N506 (8)

Text

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Mr. Jim:s A. Hutton September 9, 1999 Director-Licen:ing, MC 62A-1 PECO Energy Company 1 Nuclear Group Headquarters Correspondence Control Desk P.O. Box No.195 -

Wayne, PA 19087-0195

SUBJECT:

UPDATED TECHNICAL SPECIFICATIONS (TS) BASES PAGES - LIMERICK GENERATING STATION (LGS), UNITS 1 AND 2 (TAC NOS. MA5924 AND MA5925)

Dear Mr. Hutton:

By letter dated June 9,1999, the PECO Energy Company submitted a change to TS Bases Section 3/4.10.8," Inservice Leak and Hydrostatic Testing," and TS Bases Section 3/4.2.3," Minimum Critical Power Ratio." Bases Section 3/4.10.8 was revised to clarify that the f requirements for the Reactor Enclosure Secondary Containment apply to an extended area I encompassing both the Reactor Enclosure and the Refueling Area during inservice Leak and Hydrostatic Testing. Bases Section 3/4.2.3 was revised to remove the operating limit dependence on the recirculation system motor generator set stop limiter settings. We have no objection to the wording. The enclosed revised TS Bases pages B 3/410-2 and B 3/4 s-4 for LGS, Units 1 and 2, are being issued to assure distribution of the revised bases pages to all holders of the TS.

I Sincerely, Bartholomew C. u ley Sr rofect%linager, Section 2 Project Directorate l Division of Licensing Project Management j Office of Nuclear Reactor Regulation Docket Nos. 50-352 and 50-353

Enclosure:

As stated cc w/ encl: See next page o\

DISTRIBUTION:

• Docketfilesg; 1IC B. Buckle'y PUBLIC OGC 9

PDl-2 r/f ACRS M. O'Brien DOCUMENT NAME: G:\PDI-2\ Limerick \LIA5924.MEM *See previous concurrence  ;

- To receive a copy of this document, indicate in the box: "C" = Copy without I attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy  !

0FFICE PM:PDI-2 @[] l E LA:PDI-2 e, 1 SRxB* l SPLB* l SCfjf-2 l  !

NAME BBuckley ' M0'Brien (UlOj/O JWermlel JHannon JClifford I DATE 09/'7/99 09/ff/99 08/06/99 08/31/99 d{/1/99 9909130008 990909 cial Record Copy PDR ADOCK 05000352 P PDR

v. .

Mr. James A. Hutton September 9, 1999 Dir:ctor-Licensing, MC 62A-1

' PECO Energy Company Nuclear Group Headquarters 1 Correspondence Control Desk i P.O. Box No.195 '

Wayne, PA 19087-0195 i

SUBJECT:

. UPDATED TECHNICAL SPECIFICATIONS (TS) BASES PAGES - LIMERICK GENERATING STATION (LGS), UNITS 1 AND 2 (TAC NOS. MA5924 AND MA5925) {

l

Dear Mr. Hutton:

By letter dated June 9,1999, the PECO Energy Company submitted a change to TS

{

Bases Section 3/4.10.8, " Inservice Leak and Hydrostatic Testing," and TS Bases Section 3/4.2.3, " Minimum Critical Power Ratio." Bases Section 3/4.10.8 was revised to clarify that the 1 requirements for the Reactor Enclosure Secondary Containment apply to an extended area encompassing both the Reactor Enclosure and the Refueling Area during inservice Leak and Hydrostatic Testing. Bases Section 3/4.2.3 was revised to remove the operating limit dependence on the recirculation system motor generator set stop limiter settings. We have no

' objection to the wording. The enclosed revised TS Bases pages B 3/410-2 and B 3/4 2-4 for LGS, Units 1 and 2, are being issued to assure distribution of the revised bases pages to all holders of the TS.

Sincerely, Bartholomew C. uble"y Sr rofect%linager, Section 2 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation .

Docket Nos. 50-352 and 50-353

Enclosure:

As stated cc w/ encl:- See next page plSTRIBUTION:

Docket Files B. Buckley PUBLIC OGC PDI-2 r/f ACRS M. O'Brien DOCUMENT NAME: G:\PDl-2\ Limerick \LIA5924.MEM *See previous concurrence To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy

{

OFFICE PM:PDl-2 % lE LA:PDI-2 .lL SRXB* l SPLB* l SCfjf-2 l I NAME BBuckley H0'Brien JWermiel JHafinon JClifford DATE 09/'T/99 09/F/99 08/06/99 08/31/99 L9/1/99 Official Record Copy i l

ps>* *'% 9 i

[L 4 UNITED STATES

{ jt NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September.9, 1999 Mr. James A. Hutton Director-Licensing, MC 62A-1 PECO Energy Company Nuclear Group Headquarters

= Correspondence Control Desk P.O. Box No.195 Wayne, PA 19087-0195-

SUBJECT:

UPDATED TECHNICAL SPECIFICATIONS (TS) BASES PAGES - LIMERICK GENERATING STATION (LGS), UNITS 1 AND 2 (TAC NOS. MA5924 AND MA5925)

Dear Mr. Hutton:

By letter dated June 9,1999, the PECO Energy Company submitted a change to TS Bases Section 3/4.10.8," Inservice Leak and Hydrostatic Testing," and TS Bases Section 3/4.2.3, " Minimum Critical Power Ratio." Bases Sectior. 3/4.10.8 was revised to clarify that the requirements for the Reactor Enclosure Secondary Containment apply to an extended area encompassing both the Reactor Enclosure and the Refueling Area during inservice Leak and Hydrostatic Testing. Bases Section 3/4.2.3 was revised to remove the operating limit dependence on the recirculation system motor generator set stop limiter settings. We have no objection to the wording. The enclosed revised TS Bases pages B 3/410-2 and B 3/4 2-4 for LGS, Units 1 and 2, are being issued to assure distribution of the revised bases pages to all holders of trw TS.

Sincerely, e s C Bartholomew C. Buckley, Sr. Project Manager, Section 2 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation

- Docket Nos. 50-352 and 50-353

Enclosure:

As stated cc w/ encl: See next page

r. .

-1

~

Limerick Generating Station, Units 1 & 2 .,

cc:

J. W.- Durham, Sr., Esquire Chief-Division of Nuclear Safety Sr. V.P. & General Counsel PA Dept. of Environmental Resources PECO Energy Cornpany P.O. Box 8469 2301 Market Street Harrisburg, PA 17105-8469 Philadelphia, PA 19101 Manager-Limerick Licensing,62A-1 Director-Site Engineering PECO Energy Company Limerick Generating Station 965 Chesterbrook Boulevard P.O. Box 2300 Wayne, PA 19087-5691 Sanatoga, PA 19464 Mr. James D. von Suskil, Vice President  !

Limerick Generating Station Manager-Experience Assessment Post Office Box 2300 Limerick Generating Station Sanatoga, PA 19464 P.O. Box 2300 Sanatoga, PA 19464 Plant Manager Limerick Generating Station Library P.O. Box 2300 U.S. Nuclear Regulatory Commission Sanatoga, PA 19464 Region i 475 Allendale Road Regional Administrator, Region i King of Prussia, PA 19406 U.S. Nuclear Regulatory Commission 475 Allendale Road -

Senior Manager-Operations King of Prussia, PA 19406 Limerick Generating Station P.O. Box 2300 Senior Resident inspector Sanatoga, PA 19464 U.S. Nuclear Regulatory Commission Limerick Generating Station Dr. Judith Johnsrud P.O. Box 596 National Energy Committee Pottstown, PA 19464 ' Sierra Club 433 Orlando Avenue Chairman State College, PA 16803 J Board of Supervisors of L.imerick Township 646 West Ridge Pike Linfield, PA 19468 l

m . .

W

?!4.10 SPECIAL TEST EXCEP C NS BASES 3/4.10.8 INSERVICE LEAK AND WYDR0 STATIC TESTING

- This special test exception permits certain reactor coolant pressure tests to be performed in OPERATIONAL CONDITION 4 when the metallurgical characteristics of th~e reactor pressure vessel (RPV) or plant temperature control capabilities during these tests require the pressure testing at temperatures greater than 200*F and less than or equal to 212*F (normally corresponding to OPERATIONAL CONDITION 3). The additionally imposed OPERATIONAL CONDITION 3 requirements for SECONDARY CONTAINMENT INTEGRITY provide conservatism in response to an operational event.

Invoking the requirement for Refueling Area Secondary Containment Integrity along with the requirement for Reactor Enclosure Secondary Containment Integrity applies the requirements for Reactor Enclosure Secondary Containment Integrity to an extended area encompassing Zones 1 and 3. Operations with the Potential for Draining the Vessel, Core alterations, and fuel handling are prohibited in this secondary containment configuration. Drawdown and inleakage testing performed for the combined zone system alignment, shall be considered adequate to demonstrate integrity of the combined zones.

Inservice hydrostatic testing and inservice leak pressure tests required by Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code are performed prior to the reactor going critical after a refueling outage. The minimum temperatures (at the required pressures) allowed for these tests are determined from the RPV pressure.and temperature (P/T) limits required by LCO 3.4.6, Reactor Coolant System Pressure / Temperature Limits. These limits are conservatively based on the fracture toughness of the reactor vessel, taking into accounta anticipated vessel neutron fluence. With increased reactcIr fluence ovir tMie, the minimum allowable vessel temperature increases at a given pressure. Periodic updates to the RCS P/T limit curves are performed as necessary, based upon the results of analysis of irradiated surveillance specimens removed from the vessel.

/

MAY 0 61999 LIMERICK - UNIT 1 B 3/4 10-2 Amendment No. 188 ECR 99-00864 Revised by NRC letter dated

r. .

3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING This special test exception permits certain reactor coolant pressure tests to be performed in OPERATIONAL CONDITION 4 when the metallurgical characteristics of the reactor pressure vessel (RPV) or plant temperature control capabilities during these tests require the pressure testing at temperatures greater than 200*F and less than or equal to 212*F (normally corresponding to OPERATIONAL CONDITION 3). The additionally imposed OPERATIONAL CONDITION 3 requirements for SECONDARY CONTAINMENT INTEGRITY provide conservatism in response to an operational event.

Invoking the requirement for Refueling Area Secondary Containment Integrity along with the requirement for Reactor Enclosure Secondary Containment Integrity applies the requirements'for Reactor Enclosure Secondary Containment Integrity to an extended area encompassing Zones 2 and 3. Operations with the Potential for Draining the Vessel, Core alterations, and fuel handling are prohibited in this secondary containment configuration. Drawdown and inleakage testing performed for  !

the combined zone system alignment shall be considered adequate to demonstrate i integrity of the combined zones. 1 Inservice hydrostatic testing and inservice leak pressure tests required by Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code are perfor:r.cd prior to the reactor going critical after a refuel 1r.; outage. The minimum temperatures (at the required pressures) allowed for these tests are determined from the RPV pressure and temperature (P/T) limits required by LC0 3.4.6, Reactor Coolant System Pressure / Temperature Limits. These limits are conserwahvely based on the fracture toughness of the reactor vessel, takinq into account anticipated vessel neutron fluence. With increased reactor f' uence over time, the minimum allowable vessel temperature increases at a given pressure. Periodic updates to the RCS P/T limit curves are performed as necessary, based upon the results of analysis of irradiated surveillance specimens removed from the vessel.

I l

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MAY 0 61999 LIMERICK - UNIT 2 B 3/4 10-2 Amendment No. 95 ECR 99-00864 Revised by NRC letter dated

F. .

POWER DISTRIBUTION llMITS BASES

.3/4.2,3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions

> as specified in Specification 3.2.3 are derived from the established fuel cladding transients.integrity Safety Limit MCPR, and an analysis of abnormal operational For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit

'in Specification 2.2.MCPR at any time during the transient assuming-instrument trip To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in.the largest reduction in CRITICAL POWER RATIO CPR . The t {

flow, increase in pressur(e an)d power,ype of transients positive reactivity evaluated insertion, and were loss of coolant temperature decrease.

The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate transients are discussed j in Reference 2.

The MCPR operating limits derived from the transient analysis are dependest on the operating core flow and power state (MCPR(F , and MCPR(P), respectively 10 ensure adherence to fuel design limits during the w)orst transient that occurs e).f with moderate frequency (Ref. 6). Flow dependent MCPR limits determined by steady state thermal hydraulic methods with key p(MCPR(F))

hysics response are 4

analyze slow flow runout transients. inputs benchmarked using- the three d dimensional transient code (Ref. 8). Power dependent MCPR limits (MCP Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrans are between 25% RTP and 30% RTP. bypassed, high and low flow MCPR(P), op The MCPR operating limits specified in the COLR are the result of).ho-ro.-

Design Basis Accident (DBA) and transient analysis.

is determined by the larger of the MCPR(F), and MCPR(P) limits.The operating limit MCP JUN 0 31999 LIMERICK - UNIT 1 B 3/4 2-4 Amendment No. 7 M, M, W, #

ECRLG99-0113d i Revised by NRC letter dated

)

POWER DISTRIBUTION LIMITS BA3ES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Satety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate transients are discussed in Reference 2.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR(F), and MCPR(P), respective 4pttet ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Ref. 6). Flow dependent MCPR limits (MCPR(F)) are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 7) to analyze slow flow runout transients.

Power dependent MCPR limits (MCPR(P)) are determined mainly by the one dimensional transient code (Ref. 8). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPR(P), operating limits are provided for operating between 25% RTP and 30% RTP.

The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPR(F), and MCPR(P) limits.

JUN 0 31999 LIMERICK - UNIT 2 B 3/4 2-4 Amendment No. +, 40-ECR LG 99-01138 Revised by NRC letter dated

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