ML20066B367
| ML20066B367 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/31/1990 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20066B370 | List: |
| References | |
| NUDOCS 9101070100 | |
| Download: ML20066B367 (8) | |
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UNITED STATES o
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I ALABAMA POWER COMPANY a
DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 81 License No. NPF-8 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Alabama Power Company (the licensee), dated August 27, 1990, 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 license 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 anendment is in accordance with 10 CFR Part 51 of the Comission'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 2.C.(2) of Facility Operating License No. NPF-8 is hereby amended to read as follows:
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(2) _ Technical Specifications The Technical Specifications contained in Appendicet A and B, as revised through Amendment No. 81
, 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 its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By:
l Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 31, 1990 1
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1 ATIACHMENT TO LICENSE AMENDMENT NO, B1
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Replace the following pages of the Appendix A Technical Specifications i
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i Rpmove Paces insert Pages 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 B 3/4 4-7 8 3/4 4-7 B 3/4 4-10 B 3/4 4-10 8 3/4 4-14 8 3/4 4-14 f
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} t-j l l -F TEMPERATURE (279'F) FOR THE t- ~T I ;i -tTti i i t i -~~iiit i ii,,ii,,, i ii - SERVICE PERIOD UP TO 14 EFPY t~ i, l -* : i i i i ii ~t T t-II i l' I! I!'IIIII!I !I ~ 0 O 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG, F) Figure 3,4-2 Ferley Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 14 EFPY, rarley-Unit 2 3/4 4 29 krendment No, if, B1
i-j 2500 - l ll li ll di ! ! i, r i; i 2t ! i 7 i! i i j 2250 RTNTOAFTER 14 EFPY I M 1/4T: 152'F l;I [/ l ili i i 2000 3'4T: 124'F ! l !!ii !ll d [! i -N !!! I i!Ii fi i il<. I iiI ii iiii i /i i !i ii 'I -~ 1750 I i i lii iiii Ii i ii i 9 '.';i - L-UN ACCEPTABLE Ii i i t I /i i y) 1 i + n. i W OPERATION ii riii i i ! ii i ' 'i' ~ t w 1500 i !ll /l i - ! ![!! !f/ f J p, ve nii i i i r i z -t ~ W 1250 I' il CC t ! Q4 t '/ Ii i1 ! I i c ! t i g ! li I! -.L l ! ! o !Ii w .. a ! ! ! t Q 1000 44'! I IfiY l Qi i I o Y
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5 ! i E ^ I I IIi 750 i il 1 i ii. ! i /i i i i i 'F/HR. !I li 1 i i F 0-Mg'ImTf l l [ i i iiii i i i 20M e 500 LM M i !! 40nijj f i l, 60 250 100 ! ! !!I i Th!,
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j i I! i i i ii!i i i i i i i !ii i iiii i I i i !I ! i! iii! I I II II ! 'i' I ' ' I 0 O 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG. F) Figure 3.4-3 Farley Unit 2 Reaci:r Cooling System Cooldown Limitations l Applicable for the First 14 EFPY, 3 Farley-Unit 2 3/4 t.-30 Amendment No. 55, 81
W! REACTOR COOLANT SYSTp MS!5.................................................................... 4) The pressurizer heatup and cooldovn rates shall not exceed 100'F/hr and 200'F/hr respectively. The spray shall not be used if the temperature dif ference between the pressurizer and the spray fluid is greater than 320'F. 5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI. The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with ASTM E185-82, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance vith Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in VCAP-7924-A, " Basis for Heatup and Cooldovn Limit Curves, April 1975." Heatup and cooldown limit curves are calculated uting the most limiting value of the nil-ductility reference temperature, RT,,,The 14 EFPY service life l , at the end of 14 effective full power years (EFPY) of service life. periodischosensuchthatthelimitingRT,Ig at the 1/4T location in the core region is greater than the RT o the limiting unirradiated material. TheselectionofsuchalimSIlngRT ass m s Gat aH components in the Reactor Coolant System vill S,, operated conservatively e in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RT,,t; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RT,,,. Therefore, an adjusted reference temperature, based upon the fluence and the nickel and copper content of the material in question, can be predicted using VCAP-12471 and the re-commendations of Regulatory Guide 1.99, Revision 2, " Radiation Embrittle-ment of Reactor Vessel Materials." The heatup and cooldovn limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shif t in RT,,, at the end of 14 EFPY. FARLEY-UNIT 2 B 3/4 4-7 AMENDMENT NO. 55, 81
M'-' i This page has been deleted. FARLEY-UNIT 2 B 3/4 4-10 AMENDMENT NO. 55, 81 j
REACTOR COOLANT SYSTEM "j im!....................................................................... The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition svitches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the 10CFR rart 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered. This Rule states that the minimum metal temperature of the closure flange regions be at least 120'F higher than the limiting RT for these regions g when the pressure exceeds 20 percent of the preservice bydrostatic test pressure (621 psig for Farley Unit 2). In addition, the nev 10CFR Part 50 Rule states that a plant specific fracture evaluation may be performed to justify less limiting requirements. Based upon such a fracture analysis for Farley Unit 2, the 14 EFPY heatup and cooldovn curves are impacted by the l - nev 10CFR Part 50 Rule as shown on Figures 3.4-2 and 3.4-3. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductible failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. TheOPERABILITYoftvoRHRreliefval$esoranRCSventopeningofgreater than or equal to 2.85 square inches ensures that the RCS vill be protected from pressure transients which could exceed the limits of Appendix G to 10CFR Part 50 when one or.more of the RCS cold legs are less than or equal to 310'F. Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP vith the secondary vate. temperature of the steam generator-less than or equal to 50'T above the RCS cold lag temperatures or (2) the start of 3 charging pumps and their injection into a vater solid RCS. 3/4.4.11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and' operational readiness of these components vill be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR Part 50.55a(g) except where specific vritten relief has been granted by the Commission pursuant to 10CFR Part 50.55a(g)(6)(1). 3/4.4.12 REACTOR VESSEL HEAD VE!US The OPERABILITY of the Reactor Head Vent System ensures that adequate core cooling can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel. This system is in accordance with 10CFR50.44(c)(3)(iii).' FARLEY-UNIT 2 B 3/4 4-14 AMENDMENT NO. 35, 55, 81 4 c. -e, -_.ar,- .n- ,e-- n --}}