ML20212J347

Proposed Tech Spec 3/4 4.6 Re Reactor Pressure Vessel pressure-temp Curve,Deleting Ref to NEDO-21778-A

ML20212J347
Person / Time
Site:	Grand Gulf
Issue date:	01/20/1987
From:	SYSTEM ENERGY RESOURCES, INC.
To:
Shared Package
ML20212J338	List: ML20212J334 ML20212J347
References
TAC-63558, NUDOCS 8701280146
Download: ML20212J347 (1)

Text

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REACTOR COOLANT SYSTEM t

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BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The actual shift in RT f the vessel material will be established period-NDT ically during operation by removing and evaluating in accordance with ASTM E185-73 and 10 CFR 50, Appendix H, irradiated reactor vessel material sp~ecimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the specimen data and recom-mendations of Regulatory Guide 1.99, Revision 1.

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C, C', and A, 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 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.,with t;. muuiiiu.auvo uv Fo.usraph IV.A.2.

p:r CE """ Lf::.:i.g T;pical %;;rt 9E00-21770 A.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment.

The surveillance requirements are based on the operating

. history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

~/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

l Components of the reactor coolant system were designed to provide access l

to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1977 Edition, and Addenda through Summer 1978.

l The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55afg) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication; however, single failure considerations require that two loops l

be OPERABLE or that alternate methods capable.of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

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8701280146 870120 PDR ADOCK 05000416 P

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