ML20134A517
| ML20134A517 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/22/1985 |
| From: | Butcher E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20134A518 | List: |
| References | |
| TAC-59060, TAC-59061, NUDOCS 8511070178 | |
| Download: ML20134A517 (11) | |
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UNITED STATES j'
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NUCLEAR REGULATORY COMMISSION
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 License No. DPR-24 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated June 17, 1985 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 application, the provisions of the Act, and the rules and regulations of the Comission; 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 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.
O B511070170 B51022 PDR ADOCK O*000266 P
, 2.
Accordingly, the license is acreded by changes to the Technical Specifications as indicated in t6e attachment to this license amendment, and paragraph 3.8 of Facility Operating License No..
DPR-24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 98
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORI COMMISSION i
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Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing
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Attachment:
Changes to the Technical c
Specifications i
Date of Issuance: October 22, 1985 i
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UNITED STATES y
',g NUCLEAR REGULATORY COMMISSION WASHINGT ON, D. C. 20555 WIFCONSIN ELECTRIC POWER COMPANY DOCKET N0. 50-301 POINT BEACH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
102 License Nc. DPR-27 1.
The Nuclear Regulatory Comission (the Comission) has found thet:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated June 17, 1985 complies with the standards ait' 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 application, the provisions of the Act, and the rules and regulations of the Comission; 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 activ'> ties will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 'cen satisfied.
f
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to +his license amendment, and paragraph 3.B of Facility Operating License No.
DPR-27 is hereby amended to read as follows:
B.
Technical Specifications j
TheTechnicalSpecificationscontqgedinAppendicesAandB, as revised through Amendment No. 1
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Edward J. Butcher. Acting Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 22, 1985 t
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ATTACHMErlT T0 tICENSE AMENDMENTS NOS. 98 AND 102 i
i TO FACILITY OPERATING LICtM5E N05. UVM-24 AND UFM-27 1
DOCKET N05. 50-266 AND 50-301 l
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1 Revise Appendix A as follows:
l Remove Partes Insert Pages 15.3.1-5 15.3.1-5 i
15.3.*-7 15.3.1-7 i
15.3.1-8 15.3.1-8 15.3.1-8a 15.3.1-8a Table 13.3.1-1(Unit 1only) 15.3.1-1(Unit 1only) j' Table 15.3.1-2 (Unit 2 only) 15.3.1-2 (Unit 2 only) i i
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Basis:
All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to reactor system temperature and pressure changes.0) These cyclic loads are introduced by normal unit load transients, reactor trips, and startup and shutdown operation. The rumber of thermal and loading cycles used for desig'. purposes are shown in Table 4.1-8 of the FSAR.
l During unit startup and shutdown, the rates of tempenture and pressure changes are Itmited. The maximum plant heatup and cooldown rate of 100*F per hour is consistent with the design number of cycles and satisfies stress limits for cyclic operation.
The ASME Ccdc.Section III, Non-mandatory Appendix G contains procedures for j
the development of heatup and cooldew.; curves for protection against non-ductile failure. The ASME Coce requires that a 1/4 wall thickness flaw, either on the inside or outside dependin] upon the location of concern, be assumed to exist i. the structure. As the Code of Federal Regulations.
i Title 10, Chapter 50, Appendix G invokes the ASME Code, Appendix G, the t
ASME Code procedures are utilized in developing the heatup and cooldown limitation curves.
i During bratt.p. the themal gradients in the reactor vessel wall produce themal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive stresses tend to alleviate the tensile stresses in6ced by the internal pressure. Therefore.
apressure-temperaturecurvebasedonsteadystateconditions(i.e.,no thermal stresses) sepresents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
i The heatup analysis also covers the determination of pressure-temperature l
limitations for the case in wh!ch the outer wall of the vessel becomes l
'he '.ontrollin, location. The thermal gradients established during heatup
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A endment No. 2A, i 15.3.1-5 Unit No. 1-98 Unit No. 2-102 4
19 2
neutron exposure of the vessel is computed to be 3.5 x 10 neutrons /cm for l
40 years of operation at 1518 MWt and 80 percent load factor.(
This is the exposure expected at the inner reactor vessel wall. However, the acutron fluence used to predict the ART shift is the one-quarter shell talckness NDT neucron exposure. The relationship between fluence at the vessel 10 wall and the fluence at the one-quarter and three-quarter shell thicknews 1cc ations has been calculated and is presented in References 3 and 4 as a fur.ction of Effective full Power Years. These curves are used to determine the fluence at the location of interest when the heatup and cooldova curves are to be revised.
Once the fluence is determined, the temperature shift used in revising the heatup anc cooldown curves is obtained from the temperature versas fluence curves (the 0.25% Copper Base, 0.20% Weld lire for Unit 1 and the 0.30% Copper base. 0.25% k' eld line for Unit 2) also contained in Ref erences 3 and 4.
These curves are used because they are based upon a substantial amount cf experimental data and rcpresent the results of the chemical analysis of the veld metal in the reactor vessels.
The heatup and cooldown curves presented in Figures 15.3.1-1 and 15.3.1-2 (Unit 1) and 15.3.1-3 and 15.3.1-4 (Unit 2) were calculated based on the above information and the methods of ASME Code Section III (1974 Cdit t an) l Appendix G, " Protection Against Nonductile Failure", and are applicable up to the operational exposure indicated on the figures. Corrections for pcastble instrumentation inaccuracies have bcen incorpora:ed into these curve =. The temperature correction is mAde by adding the temperature erTOr (24*F) to the required temperature and the pressure correction is made by subtracting the pressure error (64 psi) from the required pressure. 1hese corrections adjust the curves in the conservative direction.
Unit No. 2 - Amendment No. 31, p. 9d Unit No. 1 - Amennment No. 24, s/P10{5.3.1-7 i
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Le act.ui tenpr;ture ef t c; '.'u.'estet neerial will be established periedicall; :.,rf rg 4:ntit t tf reroving and evaluating reactor vessel na!.riai 'rreatattan surv. U Urce we:imc'- installed near the inside wall of :te ra: tor mt el ir the tc e area. Sir :e the neutron spectra at the ircactc tion sa c.es cnd r.;5 sel irside radius are identified by a specified le.: 'ac;cr, tqe rea.:urec. terxr:ture shift for a sample is an ekcellent
~ ;t :Mc c' tv LU ets of p.a.er o;,eration on the adjacent section of the t
t e acht, t s <, e l.
It the experwntal temperature shift (at the 30 ft-lb le c1) nn vat tottantiate tre predicted shift, new prediction curves and
- n..n ar d cu l.cvm ett.es rLst be deve10 red.
Tre pn m o terpren re limit lir.es shown on Figures 15.3.1-1 (Unit 1) and
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.n t:. 2) for reacttr criticality and for inservice leak and hydro-
- tati: terting hs.e beta trM.e3 to assure compliance with the minimum te cerature riq;ircunts of.hpendix G to '.0 CFR 50 for reactor criticality a*d f:r inservice ien and hydror, tar.ic testing.
Trs saray :t.:;1: *0. be used if t% temperature difference between the
- ims. truer anf s;raf fluid is smter than 320 F.
This limit is imposed ta s tatsir. t% t'ettui streites nt the pressurizer spray line nozzle below t u O!si;r I-M t.
Th3 itgeratu e rewirvier.:s for the steam generator correspond with the l
l enured hDI far the trell.
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It. re ctar sessel mattrf als surveillance capsule removal schedules are preser,t"1 in T ele 15. !.1-1 fr.r Unit I and Table 15.3.1-2 for Unit 2.
These sched.les bive b+.ea dfveloped based upon the requirements of the Code tsf Nder.11 Rogulation'.. Titir 10, Chapter 50, Appendix H and with consideration of M.M Standard E-185-E2. When the capsule lead factors are considered, the l
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']8 Unit No. 2 - 10' 1
scheduled removal dates will provide materials data representative of about' 10%, 20%, 50%, 90% and 110% of the actual reactor vessel exposure anticipated l
during the vessel life.
References (1) FSAR Section 4.1.5 (2) Westinghouse Electric Corporation, WCAP-10638
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(3) Westinghouse Electric Corporation, WCAP-8743 (4) Westinghouse Electric Corporation, WCAP-8738 I
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Unit 1 - Amendment No. 51, 98 Unit 2 - Amendment No. 57, 102
- 15. 3.1. -8a
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TABLE 15.3.1-1 I
PCINT PEAQt NUCLEAR 7: ANT, UNIT NO. 1 REACTOR VESSEL SURVEIILANCE CAPSULE REMOVAL SCHE:1ILE Capsule Approximate h tter g'soval Date*
v sep* h 1972 (actual)
S December 1975 (actual)
R October 1977 (actual)
T March 1984 (actual)
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Spring 1994 N
Stand 2r/
- he actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reacta plant shutdown.
Unit 1 - Amendment No. W, 98
TAR E 15.3.1-2 POINT BEAQi NUCLEAR FIANT, UNIT NO. 2 REACTCR VESSEL SURVCLIANOE CAPStJLE RD80 VAL SCHEDUI.E Capsule Approximate Intter assaoval Date*
V Mcwamber 1974 (actual)
T March 1977 (actual)
R April 1979 (actual)
P fall 1989 s
Fall 1995 m
standby
- The actual rauncwal dates will be adjusted to coincide with the closest scheduled plant refueling cutage or major reactor plant shutdown.
1 Unit 2 - Amendment No. J51/,102
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