ML20082U693
| ML20082U693 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/13/1991 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20082U680 | List: |
| References | |
| NUDOCS 9109200303 | |
| Download: ML20082U693 (5) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ F. / / TABLE 15.3.1-1 POINT DEACH NUCLEAR PLANT UNIT NO. 1 REACTOR VBTEEERVIIITANCCTAFE,LT REMOVAL SCHEDULE Capsule Approximate Letter Removal Date, j V September 1972 (a val) S December 1975 4ctual) R October 1977 (actual) T Harch 1984 (actual) P Spring /1994 N Standby / t
- The actual retnoval dates will be sted to coincide with the closest b
scheduled plant refueling outag major reactor plant shutdown, s k N e ( 9109200303 91o913 PDR ADoct o5000266 P PDR Unit 1 Amendment 98 October 22,.1985
a, ' TABLE 15.3.1-2 POINT BEACH NUCLEAR PLANT UNIT NO. 2 REACTOR VE33EOURVE!IIARCETAF5U[DEROVAl~3CHEDULE Capsule Approximate letter Removal Date* V November 1974 (actual) T March 1977 (actual) R April 1979 (actu ) S Fall 1990 P Fall 1996 H Standby 'J/ The actual removal dates will be a. ed to coincide with the closest i N.. scheduled plant refueling outage - ajor reactor plant shutdown. ^ n C Y o October 24, 1989 Unit 2 Amendment 128 ---___m_. _ _ _.
m'- y g D. pressurelhapenture limits Specification: 1. The Reactor Coolant System temperature and pressure shall be limited in accordance with the limit lines shown in figure 15.3.1-1 and 15.3.1-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: a. A maximum heatup of 100'f in any one hour, b. A maximum couldown of 100*f in any one hour, and c. An average temperature change of $10'T per hour during inservice leak and hydrostatic testing operations. 2. The secondary side of the steam generator will not be pressurized above 200 psig if the temperature of the steem generator vessel shell is l'elow 70*f. 3. The pressurizer temperature shall be limited to: a. A maximum heatup of 100'r in any one hour and a maximum cooldown of 200*f in any one hour, and b. A maximum spray water temperature differential between the pres- ' surizer and spray fluid of not greater than 320'f. 4. T h ef re a c t o'r; fv e s s ch i r r ad i.a t i o~rilsuW e 111 anc e^J s pdc i me n s ye~ Fenio v ed jand 6xamine'dJadcordiji'gito NRCLappFbvedfschediales,it6(determitia Ehahges hiiniiiterlillpfoherties{ The-reactor-vessel-material--treadiat4en s u rv e ill a n c e-s peci men s-s h a ll-be-remo v ed-and-e x am i ned -in-ecto rd a n ce w i th-t he-s ched ul es - pre sen t ed-4 n-l abl es-1 Sr3 rl-1-{Un i t-1 )-a nd-15r3 rl-G lbn 4 t-2 )-t o-d e t e rm i ne-c h a n g e s-i n-ma t e ri a l-p r9 pe r t-l e sr The results of these examinations shall be considered in the evaluation of the prediction method to be used to update figures 15.3.1-1 and 15.3.1-2. Revised figures shall be provided to the Connission at least sixty (60) days before the calculated exposure of the applicable reactor vessel exceeds the exposure for which the figures apply, t Unit 1 - Amendment No. if5-15.3.1-4 Unit 2 - Amendment No. 449-4cnuary-40r-1990-
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f The actual temperature shift of the vessel material will be established periodically during operation by removing and evaluating reactor vessel material irradiation 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 ope'.astion 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 curves and heatup and cooldown curves must be developed. 5 The pressure-temperature limit lines shown on figure 15.3.1-1 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to e - 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 pressur-iter 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 ND1 for the shell. ]heireast4QWselimitsfialsysurveill_ ant;s cajiiulbRsmova11schsdul.s[hapsLbssh dspel o;ssdlbissd Tspanitheleq41 rement Csfj t he36dslo fi rsds rall Regdl a t i sn si TiEW?NW5.00 Appe6dMH,Tand31t[tsWsideFationIofJSTNLStahdahd[ER85E821 The-reaeter-ve+sel-mater 4ah-surve444ence-eepsule-remeval-schedules-are presented 4e-Table-1Mrl-1-fee-Un4 t-1-and-Tehle-1Mrt-2-fee 4ni1-2. These-+chedules-have been-devel oped-ba sed-upon-t h e-requ iremen t+-e f-t he-Code-o f4ede ral-Regul a t4 en sy T414e-Mr-Chaptee-50r-Appendix-H-and-with-considerat4en-ef-ASTH -s tendard E-18643r When the capsule lead factors are considered, the scheduled removal l Unit 1 - Amendment No. 125 15.3.1-8 Unit 2 - Amendment No. 129 January 10, 1990
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- 1-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, s References (1) FSAR, Section 4.1.5 (2) Westinghouse Electric Corporation, WCAP-10638 (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. Unit 2 Amendment No. 15.3.1-8a .}}