ML20101E178
| ML20101E178 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  (DPR-24-A-168, DPR-27-A-172) |
| Issue date: | 03/20/1996 |
| From: | Hansen A NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20101E184 | List: |
| References | |
| NUDOCS 9603220179 | |
| Download: ML20101E178 (9) | |
Text
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m2 RfC UNITED STATES s
,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2005MK91
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WISCONSIN ELECTRIC POWER CONPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT NO. 1 AMEnDriENT TO FACILITY OPERATING LICENSE '
~ Amendment No. 168 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 Nay 26, 1994, as supplemented January 5, April 25 and October 12, 1995, and February 2 and March 1, 1996, 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 application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) 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.
9603220179 960320 PDR ADOCK 05000266 P
l
. 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-24 is hereby amended to read as follows:
B.
Technical Snecifications l
The Technical Specifications contained in A>pendices A and B, as revised through Amendment No. 168, are tereby 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 45 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMIS'SION pfi~
~
j Allen G. Hansen, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical l
Specifications i
Date of issuance: March 20,1996
os atou y=
Jg UNITED STATES NUCLEAR REGULATORY COMMISSION E
f WASHINGTON, D.C. 2061 5 0001
%*****/
WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT NO. 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No.172 License No. DPR-27 1.
The Nuclear hgulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated May 26, 1994, as supplemented January 5, April 25 and October 12, 1995, and February 2 and March 1, 1996, 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 application, the previsions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) 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.
' 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 Soecifications 1
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.172, 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 45 days from the date of issuance.
FOR THE NUCLEAR REGULATORY C0m !SS10N
[f. --/.
Allen G. Hansen, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Attschment:
Changes to the Technical Specifications Date of issuance:
March 20, 1996
l ATTACHMENT TO LICENSE AMENDMENT NOS.168 AND 172 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified 1
below and inserting the enclosed pages..The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT TS 15.3.1-7 TS 15.3.1-7 TS 15.3.1-8 TS 15.3.1-8 TS Figure 15.3.1-1 TS Figure 15.3.1-1 TS Figure 15.3.1-2 TS Figure 15.3.1-2 I
I of the vessel is computed to be 2.5 x 10" neutrons /ce' for 40 years of operation l
at 1518 MWt and 80 percent load factor.(') This maximum fluence is.ne exposure expected at the inner reactor vessel wall. However, the neutron fluence used to l
predict the ART, shift is the one-quarter shell thickness neutron exposure.
The relationship between fluence at the vessel ID wall and the fluence at the one-(
quarter and three-quarter shell thickness locations is as presented in Regulatory Guide 1.99 Revision 2, " Radiation Damage to Reactor Vessel Materials."
(Reference 6)
Once the fluence is determined, the adjusted reference temperature used in revising the heatup and cooldown curves is obtained by utilizing the method in Section 1.1 of Regulatory Guide 1.99 Revision 2 (Reference 6) for the limiting weld material of both Unit I and Unit 2.
The heatup and cooldown curves presented in Figure 15.3.1-1 and 15.3.1-2 were calculated based on the above information and the methods of ASME Code Section III (1974 Edition), Appendix G, " Protection Against Nonductile Failure", and are applicable up to the operational exposure indicated on the figures.
The regulations governing the pressure-temperature limits (10 CFR 50 - Appendix G and ASME Code Section III - Appendix G) do not require additional margins for instrumentation uncertainties be added to the heatup and cooldown curves.
This is because the inclusion of instrumentation uncertainties, in addition to other conservatisms in the methods for calculating the pressure temperature limits, is at necessary to protect the vessel from damage.
l Unit 1 - Amendment No. 24,53,90,125,168 Unit 2 - Amendment No. 57,50,102,120,172 15.3.1-7
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l The actual temperature shift of the vessel material will be established periodi-cally during operation by removing and evaluating reactor vessel material irradia-l tion surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are identified by a specified lead factor, the measured temperature shift for a sample is an excellent indicator of the effects of power operation on the adjacent section of the reactor vessel.
If the experimental temperature shift
]
(at the 30 ft-lb level) does not substantiate the predicted shift, new prediction
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curves and heatup and cooldown curves must be developed.
The pressure-temperature limit lines shown on Figure 15.3.1-1 for reactor critical-l ity and for inservice leak and hydrostatic testing have been provided to assure j
compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.
The spray should not be used if the temperature difference between the pressurizer and spray fluid is greater than 320F'. This limit is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
The temperature requirements for the steam generator correspond with the measured NDT for the shell.
The reactor vessel materials surveillance capsule removal schedules have been developed based upon the requirements of the Code of Federal Regulations, Title 10, Part 50, Appendix H, and with consideration of ASTM Standard E-185-82.
When the capsule lead factors are considered, the scheduled removal dates accommodate the weld data needs of all the participants in the Babcock and Wilcox Master Integrated Reactor Vessel Surveillance Program. Additionally, the schedule will provide plate / forging material data as well as fluence data corresponding to the expiration of the current licenses and of any future license extensions.
References (1)
FSAR, Section 4.1.5 (2)
Westinghouse Electric Corporation, WCAP-12794, Rev. 2/12795, Rev. 2 (3)
Westinghouse Electric Corporation, WCAP-8743 (4)
Westinghouse Electric Corporation, WCAP-8738 (5)
Babcock & Wilcox, BAW 1803 (6)
Regulatory Guide 1.99, Revision 2 Unit 1 - Amendment No. 24,90,125,131,168 15.3.1-8 Unit 2 - Amendment No. 57,102,128,129,135,172
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