ML20149D969
| ML20149D969 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/20/1996 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20149D086 | List: |
| References | |
| NUDOCS 9707180050 | |
| Download: ML20149D969 (1) | |
Text
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- JUL-14-97 MON 13:53 11S00NSIN ELECTRIC NPBU FAX NO. 4142212010 P.02/02 l*
r The actual temperature shift of the vessel material will be established
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periodically during operation by removing and evaluating reactor vesnt 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 operation on the adjacent section of the reactor vessel.
If the experimental i
,terparature shift (at the 30 ft-lb level) does not substantiate the predicted shift, new prediction curves and heatup and cooldown curves must be developed.
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 requirement.s of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and & drostatic testing.
4 Thespra)shouldnotbeusedifthetemperaturedifferencebetweenthepressur-izer and spray fluid is greater than 320F*. This limit is imposed to maintain the thermal stresses at" the pressurizer spray line arozzle below the ' design limit.
The temperature [signirements for_the steam generator correspond with the measured j NOT for the shell.
1 The reactor vessei materials kurveil ance capsule removal schedules have been developed based upon the requirements of the Code of Federal Rcquiations, 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 j
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Additionally, the Master Integrated Reactor Vessel Surveillance Program.
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
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9707180050 97o714 PDR ADOCK 05000266 P
PDR 2
agferences (1)
FSAR, Sectier, 4.1.5 (2) Westinghouse Electric Corporation, WCAP-12794, Rev.
12795, Rev.
3 (3) Westinghouse Electric Corporation, WCAP-8743 (4)
Wtstinghouse Electric Corporation, WCAP-8738
,V (5)
Babcock & Wilcox, BAW 1803 (6). Regulatory Guide 1.99, Revision 2
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Unit 1 - Amendment No. 168 Unit 2 - Amendment No. 172 15.3.1-8 March 20, 1996 4