ML20071G789
| ML20071G789 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/04/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19305C648 | List: |
| References | |
| NUDOCS 8003310210 | |
| Download: ML20071G789 (3) | |
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UNITED STATES NUCLEAR PEGULATORY COMMISSION e
o tg 4a WASHINGTON. D. C. 20555 Q' Tl SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 57 AND 56 TO FACILITY OPERATING LICENSE NOS. DPR-32 AND DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 Int'roduction O
sy ietter dated Decem8er 15, 1978 virsinia Eiectric aad eower Company (VEPCO) submitted an application to amend the Technical Specifications appended to Facility Operating Licenses DPR-32 and DPR-37 for Surry Units 1 and 2.
The requested change would extead the acceptable operating period of the present operating limits from 3.C EFPY to 11 EFPY.
Discussion 10 CFR Part 50, Appendix G " Fracture Toughness Requirements", requires that pressure-temperature limits be established for reactor coolant sy: stem hectup and cooldown operations, inservice leak and hydrostatic tests, and reacter core operation. These limits are iequired to ensure that the stresses in the reactor vessel remain within acceptable limits. They are intended to provide adequate margins of safety during any condition _of normal operation, including anticipated operational occurrences.
i Q The pressure-temperature limits depend upon the metallurigical properties i
of the reactor vessel materials. The properties of materials in the vessel beltline region vary over the lifetime of the vessel because of the effects of neutron irradiation. One principal effect of the neutron irradiation The pressure-temperature operating limits must N) to is that it causes the vessel material nil-ductility temperature (RT N
increase with time.
modified periodically to account for.this radiation-induced increase in l
' by increasing the temperature required for a given pressure. The j
opekting limits for a particular operating period are based on the material N
properties at the end of the operating period. By periodically revising the pressure-temperature limits to account for radiation damage, the stresses and stress intensities in the reactor vessel are maintained within acceptable l
limits.
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The magnitude of the shift in RT is proportional to the neutron fluence towhichthematerialsaresubje5Nd. The shift in RT can be predicted from Regulatory Guide-1.99.
To check the validity of $kTe predicted shift in RT a reactor vessel material surveillance program is required.
SurveYNa,nce specimens are periodically removed from the vessel and tested.
The results of these tests are compared to the predicted shifts in RTN and the pressure-temperature operating limits are revised accordingly.DT' Evaluation The operating limits currently in the Technical Specifications are acceptable for operation to 3.8 EFPY. The same pressure-temperature operating limits are used for both Surry reactor vessels. The limits are based on the weld e
l metal in the Surry 1 vessel which is the most limiting material. To date one material surveillance capsule has been removed from each vessel and tested. The test results on these capsules. clearly show that the Unit 1 i
surveillance weld metal is the limiting material. However, this weld O
material is not identical to the weld metal used to fabricate the reactor vessel; i.e., these welds were not made from the same heats of weld 4
wire and flux. Therefore VEPC0 recalculated the amount of radiction damage l
using damage predictier. curves supplied by Westinghouse. This calculation showed that the present limit curves are acceptable drough 32 EFP7.
We have reviewed the proposed change to the operating limits and have perfomed independent calculations to verify compliance with Appendix G, i
10 CFR 50. We agree that the surveillance weld metal in Unit 1 is not identical to that used to fabricate the reactor versel. However, since it i
was made by the same type of. weld wire and flux, we consider it to be l
representative of the vessel seld. The NRC t,taff is currently studying i
the efiect.s of different heats of weld wire-flux combinations or, radiatien i
damage. Although this study is not complete, it indicates that the surveil-lance data is applicable for the determination of radiation damage on the Surry 1 reactor vessel.
' O The resuits of the tests s the te suramance capsuies -re reviewed again.
It is concluded gat thg limiting material is the Unit 1 weld metal.
At a fluence of 2.5 x 10 n/cm this weld metal shows an increase in RT of 165'F. This increase in RT is approximately on the upper limit lie of Regulatory Guide 1.99, Revi!Nn 1 and is consistent with data generated from other surveillance programs on similar weld metal samples. Using the above shift in RT and the procedures in Regulatory Guide 1.99, Revision 1, the staff find!D[ hat the operating limits currently in the Technical Specifications are acceptable for operation through 11 EFPY. We have discussed this with the licensee and he agrees with this change.
These i
operating limits are acceptable for both Surry 1 and 2 since the limiting material in the Surry 2 reactor vessel has a greater resistance to radiation damage than the limiting material in the Unit 1 vessel.
For this operating period these operating limits are in.accordance with Appendix G,10 CFR 50.
Conformance with Appendix G to 10 CFR Part 50 in establishing safe operating l
limitations will ensure adequate safety margins during operation, testing, l
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2-maintenance and postulated accident conditions and constitutes an acceptable basis for satisfying the requirements to NRC General Design Criterion 31, Appendix A,10 CFR Part 50.
VEPCO also requested a change to Technical Specification 4.2, Section G regarding materials irradiation testing. The proposed change would delete the requirement for conducting Charpy tests on irradiated surveillance specimens..
This change is not in accordance with Appendix H.10 CFR 50 and is therefore not acceptable. This specification should remain unchanged.
We have discussed this with the licensee and he agrees with not making this change.
Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is in-O significant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative delcaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion He have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and
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safety of the public.
Date: March 4, 1980 4
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