ML15112B104
| ML15112B104 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/15/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112B103 | List: |
| References | |
| NUDOCS 8305090124 | |
| Download: ML15112B104 (2) | |
Text
o.UNITED STATES#
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDENT NO.
120TO FACILITY OPERATING LICENSE NO.
DPR-38 AMENDMENT NO.120 TO FACILITY OPERATING LICENSE NO.
DPR-47 AMENDMENT NO.1 17 TO FACILITY OPERATING LICENSE NO.
DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS.
1, 2 AND 3 DOCKETS NOS.
50-269, 50-270 AND 50-287 Introductich By letter dated February 23, 1983, Duke Power Company (DPC or the licensee) proposed changes to the Oconee Station's common Technical Specifications (TSs) to revise the withdrawal schedule for the reactor vessel surveillance capsules. The licensee requests that Capsule OCI-A, which is being irradiated in the reactor vessel of Crystal River, Unit No. 3, be removed and be tested at the end of the-fourth fuel cycle--instead of the seventh fuel cycle. The licensee estimates that the capsule's accumulated neutron fluence at the en-d-of the fourth fuel cycle would-be 7.1 x 1018 n/c2 - 2 (E>lMev) and at the end of the seventh fuel cycle would be 1.4 x 10 n/cm
Background
DPC is participating in the Babcock & Wilcox (B&W) Owners Group Integrated Reactor Vessel Material Surveillance Program (B&W Report BAW-1543).
The weld material in Capsule OCI-A is identified as WF-112. This weld material is the limiting material in the reactor vessel of Arkansas Nuclear One, Unit 1 (ANO-1),
whose owners are also participating in the integrated program. WF-112 weld material was welded using weld wire and flux which is similar to the limiting weld material in the Oconee reactor vessels. Hence, the results of the tests on irradiated WF-112 weld material will be useful for determining the material fracture toughness of the Oconee and ANO-1 reactor vessels.
Discussion and Evaluation The 1/4 thickness and inside surface reactor vessel beltline mqterial properties are the critical properties for determining the operating conditions and accident limitations for a reactor vessel. In accordance with Appendi-x G, 10 CFR 50, pressure-temperature limits for normal opera tion and anticipated upset conditions are evaluated based on the 1/4 thickness reactor vessel beltline material properties. In accordance with Commission Report SECY-82-465, the accident conditions of pressurized thermal shock are limited by the inside surface reactor vessel beltline 8305090124 830415,1 PDR ADOCK 05000269" P
PDR"
-2 material properties.
The maximum end of life (EOL) neutron fluence at the 1/4 thickness location for the Oconee and ANO-1 react r vessel beltlines is estimated as 7.2 x 118 n/cm 2 and 7 x 1018 n/cm, respectively.
The maximum EOL neutron fluence at the inside surface for the Qconee and ANO-1 reap or vessel beltlines is estimated as 1.3 x 1019 n/c-and 1.27 x 10' n/cm2, respectively. Thus, Capsule OCI-A at th 8end o the fourth fuel cycle will accumulate neutron fluence (7.1 x 10 n/cm ),
which is equivalent to the maximum EOL neutron fluence at the 1/4 thickness beltline locations for the Oconee and ANO-1 reactor vessels, but less than the maximum EOL neutron fluence at the inside beltline surface for the Oconee and ANO-1 reactor vessels.
The proposed withdrawal schedule for Capsule OCI-A will provide material properties which will be useful for determining the EOL pressure temperature limits for the Oconee and ANO-1 reactor vessels, but will not be irradiated for a sufficient amount of time to provide material properties for EOL accident limits for the Oconee and ANO-1 reactor vessels.
- Hence, we consider the proposed withdrawal schedule for Capsule OCI-A acceptable provided the neutron irradiation received by Capsule OCI-C, which contains the same weld material as Capsule OCI-A, exceeds the anticipated maximum EOL neutron fluence at the inside surface for the Oconee and ANO-1 reactor vessel beltlines.
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 insignificant from the standpoint of environmental impact and, pursuant to 10 CFR §51.5(d)(4), that an environmental impact statement, or negative declaration and environ mental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We 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 an accident previously evaluated, do not create the possibility of an accident of a type different from any evaluated previously, and do not involve a significant reduction in a margin of safety, 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 tn 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 safety of the public.
Dated: April 15, 1983 The following NRC personnel have contributed to this Safety Evaluation:
B. Elliot,-E, Ctonner.