ML20237G788

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Safety Evaluation Accepting Util 870731 Proposed Change to Bases Section of Tech Specs,Reflecting Commitment to 1982 Rev of ASTM E 185 Re Reactor Vessel Surveillance Program Required by 10CFR50,App H.Rev to Page B 3/4 4-7 Encl
ML20237G788
Person / Time
Site: Waterford Entergy icon.png
Issue date: 08/20/1987
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20237G779 List:
References
NUDOCS 8709020412
Download: ML20237G788 (3)
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\,.m./ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING CHANGE TO BASES SECTION OF TECHNICAL SPECIFICATIONS FACILITY OPERATIONS LICENSE NO. NPF-38 LOUISIANA POWER AND LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION n UNIT 3

_ DOCKET NO. 50-382

1. 0 INTRODUCTION By letter dated July 31, 1987, Louisiana Power and Light Company (LP&L, the licensee) submitted a proposed change to the Bases for the Appendix A Technical Specifications for their Waterford Steam Electric Sation, Unit 3. The proposed change to the Bases for Section 3/4.4.8, " Pressure Temperature Limits", is being made to reflect a change to Appendix H to 10 CFR Part 50.

2.0 EVALUATION LP&L has proposed a change to the Bases Section of the Technical Specifica-tions for Waterford 3 which deal with the reactor vessel surveillance program to reflect their commitment to the 1982 revision of ASTM E 185 in accordance with 10 CFR 50, Appendix H, Item II.B.1 which states in part:

"For each capsule withdrawal after July 26, 1983, the test procedure and reporting requirements must meet the requirements of ASTM E 185-82 to the extent practical for the configuration of the specimens in the capsule."

The Bases previously cited the 1973 version of ASTM E 185 as the methodology by which the specimens were to be evaluated.

3.0 CONCLUSION

Based on its review of the licensee's July 31, 1987 submittal, the staff concludes that the proposed change to the Bases of the Appendix A Technical Specification for Waterford 3, which reflect a commitment to the 1982 revision of ASTM E 185 regarding the reactor vessel surveillance program as required by 10 CFR 50, Appendix H, is acceptable.

Principal contributor: J. Wilson Dated: August 20, 1987 0709020412OghG2 l PDR ADOCK O PDR l P )

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. REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to 100 F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figures 3.4-2 and 3.4-3. The limitations on the Reactor Coolant System heatup and cooldown rates are further restricted due to stress limitations in the Reactor Coolant Pump. As part of the LOCA support scheme, the Reactor Coolant Pump has a ring around the suction nozzle of the pump. The support skirt is welded to the ring. Due to this design, the heatup and cooldown rates must be limited to maintain acceptable thermal stresses.

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these test are shown in Table B 3/,4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using FSAR Table 5.3-1 and the recommendations of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT NDT at the end of the applicable service period, as well as adjustments for possible errors in the pressure and temperature sensing instruments.

The actual shift in RT NDT f the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82 and 10 CFR Part 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in <

Table 4.4-5. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the delta RT NDT determined from the surveillance cc.psule is different from the calculated delta RT NDT f r the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figures 3.4-2 and 3.4-3 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.

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