ML20117K597
| ML20117K597 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/02/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20117K583 | List: |
| References | |
| TAC-54449, TAC-54894, NUDOCS 8505150502 | |
| Download: ML20117K597 (5) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 58 TO FACILITY OPERATING LICENSE N0. NPF-2 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-348 INTRODUCTION In letters from F. L. Clayton, Jr. to S. A. Varga dated March 1,1984 and April 20, 1984 the Alabama Power Company submitted reactor vessel material surveillance test data and changes to the Farley Unit 1 Technical Specifi-cations, respectively. The reactor vessel material surveillance test data were detailed in WCAP-10474, " Analysis of Capsule U from the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program." The proposed technical specifications were changes to the pressure-temperature limits and the reactor vessel material surveillance program withdrawal schedule.
DISCUSSION AND EVALUATION Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G,10 CFR 50, which became effective on July 26, 1983.
Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G,10 CFR 50 are dependent upon the initial RT f r the limiting materials in the beltline and closure flange regions NDT of the reactor vessel and the increase in RT resulting from neutron NDT irradiation damage to the limiting beltline material.
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The Farley Unit I reactor vessel was procured to ASME Code requirements, which did not specify fracture toughness testing to determine RT I"
NDT l
each reactor vessel material.
Hence, the initial RT f r some materials NDT in the closure flange and beltline regions of the reactor vessel could not be determined in accordance with the test requirements of the ASME Code.
The RT f r these materials were estimated using the method recommended NDT by the staff in Section 5.3.2 of the NRC Standard Review Plan.
The initial RTNDT values for the limiting materials in the beltline and closure flange regions are 0 F and 60 F, respectively.
The increase in RT resulting from neutron irradiation damage was esti-NDT mated by the licensee using the method recommended in Regulatory Guide 1.99, Rev. 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." In Table I we have compared the increase'in RTNDT predicted by the regulatory guide with that measured from the surveil-lance material, which was reported in WCAP-10474.
The method of predicting neutron irradiation damage in Reg. Guide 1.99, Rev. 1 provides conservative estimates, bec'ause the increase in RTNDT predicted by the regulatory guide exceeds that from the surveillance material.
The amount of time that pressure-temperature limits are effective depends upon the amount of neutron irradiation damage.
Utilizing the method recommended in Regulatory Guide 1.99, Rev.1 to predict the neutron irradiation damage, the neutron fluence estimates in Technical Specification Figure B 3/4.4-1 and initial RTHDT values for the limiting materials in the beltline and closure
s 4 flange region of 0 F and 60*F, respectively, we have determined that the proposeu pressure-temperature limits are acceptable for 7 effective full power years.
Appendix H, 10 CFR 50 contains the regulatory requirements for a reactor vessel materials surveillance program. Appendix H requires that the pro-posed withdrawal schedule be approved prior to implementation and references ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessel." For the Farley Unit I reactor vessel, ASTM E 185-82 recommends that there be minimum of 5 capsules and that these capsules be withdrawn when the neutron fluence received by the capsules corresponds to the amount identified in Table I of ASTM E 185-82.
This table recommends that the capsules be withdrawn at various neutron fluences throughout the plant's life and that the fifth (last) cap-sule be withdrawn at a neutron fluence not less than once or greater than
- twice the peak end-of-life (E0L) vessel fluence.
The peak neutron fluence to be received by the Farley Unit 1 reactor vessel 19 is estimated at 6.4 X 10 n/cm2 (E>1MeV). We have compared the expected neutron fluence to be received by each capsule to that required by ASTM E 185482 and conclude that the withdrawal schedule for the Farley Unit 1 capsules meets the intent of ASTM E 185-82.
Hence, we consider acceptable the proposed revision to the Farley Unit 1 Technical Specification.
-4 TABLE I
' COMPARISON OF REG. GUIDE 1.99 PREDICTION MODEL AND FARLEY UNIT 1 SURVEILLANCE TEST DATA FROM WCAP 10474 Material '
Increase in Reference Temperature, ARTNDT (F*)
From Surveillance Predicted by Test Data Reg. Guide 1.99, Rev. 1 Plate 6919-1 90 173 (transverse)
Plate 6919-1 105 173 (longitudinal)
Weld Metal 80 180
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. ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of a facility component located within the restricted area as-defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sec 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this anendment.
CONCLUSION We have concluded, based on the consideration discussed above, that: (1) there is reasonable assurance that the health and safety of the'public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated: May 2, 1985 Principal Contributor:
B. J. Elliot L