ML20058K984
| ML20058K984 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/06/1993 |
| From: | Gody A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058K989 | List: |
| References | |
| NUDOCS 9312160088 | |
| Download: ML20058K984 (9) | |
Text
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'o UNITED STATES 8
NUCLEAR REGULATORY COMMISSION
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.143 License No. DPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated February 21, 1992, as supplemented by letters dated April 16, 1992 and March 4, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9312160088 931206 PDR ADOCK 05000266 P
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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license i
amendment, and paragraph 3.B of Facility Operating License No.
DPR-24 is hereby amended to read as follows:
i B.
Technical Scecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
143, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective immediately upon issuance. The Technical Specifications are to be implemented within 20 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
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~b Anthony T. Gody, Jr., Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications 1
Date of issuance:
December 6, 1993 l
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(.f oq'9 UNITED STATES.
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NUCLEAR REGULATORY COMMISSION o
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WISCONSIN ELECTRIC POWER COMPANY r
DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.147 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated February 21, 1992, as supplemented by letters dated April 16, 1992 and March 4, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
t 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-27 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendi:es A and B, i
as revised through Amendment No.
147, are here!y incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective immediately upon issuance. The Technical Specifications are to be implemented within 20 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f/W Anthony T. Gody, Jr., Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation l
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Attachment-Changes to the Technical Specifications Date of issuance:
December 6, 1993 J
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ATTACHMENT TO LICENSE AMENDMENT NOS.143 AND 147 TO FACILITY OPERATING LICENSE N05. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. _ The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
REMOVE INSEPT 15.3.4-2a 15.3.4-2a 15.3.4-2b 15.4.7-1 15.4.7-1 15.4.7-la
2.
Single Unit Operation - One of the three operable auxiliary feedwater pumps associated with a unit may be out-of-service for the below specified times. The turbine driven auxiliary feedwater pump may be out-of-service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
If the turbine driven auxiliary feedwater pump cannot be restored to service within that 72-hour time period, the reactor shall be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Either one of the two motor driven auxiliary feedwater pumps may be out-of-service for up to 7 days.
If the motor driven auxiliary feedwater pump cannot be restored to service within that 7-day period, the operating unit shall be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D.
The main steam stop valves (MS-2017 and MS 2018) and the non-return check valves (MS-2017A and MS-2018A) shall be operable.
If one main steam stop valve or non-return check valve is inoperable but open, power operation may continue provided the inoperable valve is restored to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise the reactor shall be placed in a hot shutdown condition within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With one or more main steam stop valves or non-return check valves inoperable, subsequent operation in the hot shutdown condition may proceed provided the inoperable valve or valves are maintained closed. An inoperable main steam stop valve or non-return check valve may however, be opened in the hot shutdown condition to cool down the affected unit and to perform testing to confirm operability.
Basis A reactor shutdown from power requires removal of core decay heat.
Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condenser. Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as feedwater in the steam generator is converted to steam by heat absorption. Normally, the capability to return feedwater flow to the steam generators is provided by operation of the turbine cycle feedwater system.
I 15.3.4-2a Unit 1 - Amendment No. 7E,$7, 91,197,777,143 Unit 2 - Amendment No. U,57, 97,779,777,147
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i The eight main steam safety valves have a total combined rated capability of 6,664,000 lbs/hr. The total full power steam flow is 6,620,000 lbs/hr, therefore eight (8) main steam safety valves will be able to relieve the total full-power steam flow if necessary.
i In the unlikely event of complete loss of electrical power to the station, decay heat removal would continue to be assured for each unit by the availability of either the steam-driven auxiliary feedwater pump or one of the two motor-driven auxiliary steam generator feedwater pumps, and steam discharge to the atmosphere via the main steam safety valves or atmospheric relief valves. One motor-driven auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from a unit. The minimum amount of water in the condensate storage tanks ensures the ability-to maintain each unit in a hot shutdown condition for at least one hour
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concurrent with a loss of all AC power.
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15.3.4-2b Unit 1 - Amendment No.143 Unit 2 - Amendment No.147 1
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3 15.4.7 MAIN STEAM SYSTEM VALVES Applicability Applies to periodic testing and surveillance of the main steam stop valves (MS-2017 and MS-2018) and the non-return check valves (MS-2017A and MS-2018A).
i Obiective To verify the ability of the main steam stop valves to close upon signal and to verify that the non-return check valves are operable.
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Specification A.
Main Steam Stop Valves The main steam stop valves shall be tested under low flow conditions of 5% steam flow or less following plant shutdowns for major fuel j
reloading. The test shall be performed during the plant startup prior to admitting steam to the turbine. Closure time of five seconds or less shall be verified. The five seconds shall be measured from the time of signal initiation until the valve indicates closed.
B.
Non Return Check Valves The non-return check valves shall be tested for operability during shutdown for major fuel reloadings.
Basis I
The main steam stop valves serve to limit an excessive reactor coolant system
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cooldown rate and resultant reactivity insertion following a main steam break incident. Their ability to close upon signal should be verified at each scheduled refueling shutdown. A closure time of five seconds was selected as j
being consistent with the expected response time for instrumentation as i
detailed.in the steam line break incident analysis. The test procedure need not require steam to be flowing in the pipe. The purpose of the non-return 15.4.7-1 Unit 1 - Amendment No.
W.143 Unit 2.-AmendmentNo.Lp),147 i
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I check valves is to prevent the blowdown of both steam generators in the event i
of a main steam line piping break upstream of the main steam stop valves. The non-return check valves are swinging disc check valves which are opened by normal steam flow.
During no-flow conditions the non-return check valves are shut. The position of the non-return check valves, and thus the ability of the valves to close and perform their safety function, can be verified locally when no steam flow conditions are established.
t References FSAR - Section 10.4 FSAR - Section 14.2.5 I
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1 15.4.7-la Unit 1 - Amendment No.143-Unit 2 - Amendment No.147