ML20111A950
| ML20111A950 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/19/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20111A947 | List: |
| References | |
| NUDOCS 8501080303 | |
| Download: ML20111A950 (3) | |
Text
{{#Wiki_filter:_ _.._ _ _ _. _ /",n one\\ UNITED STATES 8 NUCLEAR REGULATORY COMMISSION h WASHINGTON, D. C. 20565 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 86 TO FACILITY OPERATING LICENSING NO. DPR-66 i DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 DOCKET NO. 50-334 INTRODUCTION i By letter dated May 21, 1984, Duquesne Light Company (the licensee) proposed 3 i administrative changes to the Technical Specifications (TS) set forth in Appendix A to tne license for the purpose of 1) revising the Reactor Pressure Vessel Material Irradiation Surveillance Schedule shown in Table 4.4-3; 2) referencing 10 CFR 50 Appendix H rather than ASTM E185-70 for surveillance specimens in bases 3/4.4.9 on page B 3/4 4-9 (Appendix H now endorses ASTM E185-73, -79 and -82); and
- 3) eliminating surveillance requirements for the first three refueling outages that are no longer applicable for inspection of the reactor vessel nozzles (4.4.10 a,b, and c on page 3/4 4-29).
{ No physical changes to the facility or equipment will be made as a result of this revision. The revision represents administrative changes required to up date the Technical Specification to 10 CFR 50 Appendix H and updated ASTM E185-82 requirements. EVALUATION AND DISCUSSION The function of the Reactor Pressure Vessel Material Irradiation Program is te evaluate the toughness changes of the vessel belt line materials and weld metal when exposed to fast neutron fluence. The surveillance program is designed to show that the reactor pressure vessel coolant boundary is designed with sufficient margin to ensure that, when stressed under operating, maintenance, l testing, and postulated accident conditions: (1) the boundary behaves in a nonbrittle manner, and (2) the probability of rapidly propagating fracture is l minimized as required by general design criterion 31 of Appendix A. This is accomplished throughout the service life of the vessel by testing in-vessel samples and calculation of changes in fracture toughness of the reactor vessel L materials caused by neutron radiation and the thermal environment. The licensee's request for a change in the surveillance schedule is based on l
- 1) analysis of the first capsule (V) removed from the reactor vessel and l-l 8501080303 841219 l
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2 reported in " Analysis of Capsule V from the Duquesne Light Company, Beaver Valley Unit No. 1 Reactor Vessel Radiation Surveillance Program", WCAP-9860, and 2) changes in the standard practice for conducting surveillance tests as indicated in ASTM E185-82. WCAP-9860 indicates that the capsule received a fast-neutron fluence of 2.55 x 10" n/cm2 compared to the calculated value of 2.58 x 10" n/cin2 and that based upon these fluence measurements the vessel thickness fl ue r.r.3 after 1.02 effective-full power years (EFPY) is 1.14 x 10" n/cm compared to a calculated 2 fluence of 1.15 x 10" n/cm. The agreement between fluence calculations and 2 fluence measurements is excellent which provides reliability for analytical projections of future vessel toughness. Revision 2 of Regulatory Guide 1.99 indicates that the fluence factor trend curve, considering the effects of nickel, should be steeper at lower fluences and flatter at the higher fluences. The staff's review of the acceptability of the licensee's proposed changes to Table 4.4-3 utilized applicable portions of WCAP-9860, but does not constitute a technical review of WCAP-9860 in its entirety. The change requested on page B 3/4 4-9 is in referencing 10 CFR 50, Appendix H, rather than ASTM E185-70 as the applicable document for removing and evaluating irradiation surveillance specimens. This change constitutes a clarification of the requirements. The changes to TS Surveillance requirement 4.4.10 constitute an administrative change in the elimination of inspection requirements for the first three refueling outages which have been completed. The indications detected during the inspection of the subject reactor vessel nozzle attachments, following the third refueling outage and reported in Westinghouse Electric Corporation letter to DLC IS-GCE-022, were evaluated as being acceptable with no corrective action mandated. Based on review of appiicable documents, the staff finds that the proposed amendment to the DLC Technical Specification represents administrative changes required to update the Technical Specifications consistent with 10 CFR 50 code of Federal Regulations and involves no physical changes in the plant safety related systems, structures or components. The staff therefore finds the proposed amendment to be acceptable. ENVIRONMENTAL CONSIDERATION This amendment involves a change in administrative procedure and requirements. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22 (c)(10) pursuant to.10 CFR 51.22(b) and no environmental impact statement or environmental assessm'ent need be prepared in connection with the issuance of this amendment.
a, 3 CONCLUSION We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endargered by operation in the proposed manner; and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Dated: December 19, 1984 Principal Contributor: Samuel D. Reynolds Jr. O O =-, --.-,. -}}