ML20149D969

Proposed Tech Specs,Providing Corrected Page to Bases for TS 15.3.1

ML20149D969
Person / Time
Site:	Point Beach
Issue date:	03/20/1996
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20149D086	List: ML20149D969
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