ML20035E542
| ML20035E542 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/07/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20035E541 | List: |
| References | |
| GL-88-11, NUDOCS 9304160153 | |
| Download: ML20035E542 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
l RELATED TO AMENDMENT N0. 19 TO FACILITY OPERATING LICENSE NO. NPF-86 NORTH ATLANTIC ENERGY SERVICE CORPORATION SEABROOK STATION. UNIT NO. 1 DOCKET NO. 50-443
1.0 INTRODUCTION
By letter dated August 17, 1992, the North Atlantic Energy Service Corporation (NAESC0/the licensee) proposed a revision to Section 3/4 4.9 of the Appendix A Technical Specifications (TS) for the Seabrook Station, Unit 1.
NAESCO l
proposed the following:
(1) revise the applicability of pressure / temperature (P/T) limits (Figures 3.4-2 and 3.4-3) to state that the curves are valid up to 11.1 effective full power years (EFPY) of operation vice the current 16 EFPY; (2) indicate the copper content of the limiting material, 0.06%; (3) revise the Reference Temperature for Nil Ductility Transition (RT,) in the y
l existing P/T limits from 110*F to 108'F and from 87'F to 86*F for the 1/4T (T = reactor vessel beltline thickness) and 3/4T reactor vessel locations, respectively; (4) revise the Bases for TS 3/4 4.9 to state that P/T curves are valid for 11.1 EFPY and that Regulatory Guide (RG) 1.99 (Rev. 2) was used to calculate RT,; and (5) delete Bases Figure B 3/4.4-2.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance: Appendices G and H of 10 CFR Part 50, the ASTM Standards and the ASME Code as referenced in Appendix G, 10 CFR 50.36(c)(2), RG 1.99 (Rev. 2),
Standard Review Plan (SRP) 5.3.2, and Generic Letter 88-11.
Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations," recommends RG 1.99 (Rev. 2) be used in calculating P/T limits, unless the use of different methods can be justified.
i Appendices G and H describe specific requirements for fracture toughness and l
reactor vessel material surveillance that must be considered in the P/T limits. An acceptable method for constructing the P/T limits is described in SRP 5.3.2.
Appendix G specifies fracture toughness and testing requirements for reactor vessel materials under the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested under Appendix H.
Appendix H, in turn, refers to ASTM Standards.
These tests define the extent i
of vessel embrittlement at the time of capsule withdrawal in terms of the l
increase in reference temperature.
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l 9304160153 930407 PDR ADOCK 05000443 P
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l Appendix G also requires a prediction of the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART).
Generic Letter 88-11 recommended use of the methods in RG 1.99 (Rev. 2), to predict the effect of neutron irradiation on reactor vessel materials.
RG 1.99 defines the ART as the sum of the unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
2.0 EVALUATION The staff has evaluated the effect of neutron irradiation embrittlement on each beltline material in the Seabrook reactor vessel under RG 1.99 (Rev. 2).
The staff determined that the limiting material with the highest ART at 11.1 EFPY is the lower shell plate, R1808-3, with 0.06% Cu, 0.57% Ni, and initial RTndt of 40*F.
At the 1/4T and 3/4T locations the staff calculated ART values of 107"F and 85'F, respectively, for 11.1 EFPY.
NAESCO used the method in RG 1.99 (Rev. 2), to calculate an ART of 108'F at 1/4T and 86*F at 3/4T for 11.1 EFPY.
The difference between licensee's ART and the staff's ART is due to round off error in the staff's calculation of 1
ART values.
Substituting the staff's ARTS into equations in SRP 5.3.2, the staff verified that the proposed P/T limits meet the beltline material requirements in Appendix G.
In addition to beltline materials, Appendix G also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120*F for normal operation and by 90*F for hydrostatic pressure tests and leak tests.
Based on the reference temperature of 10"F for the reactor closure flange at the Seabrook Station, Unit 1, the staff determined that the proposed P/T limits satisfiesSection IV.A.2 of Appendix G.
NAESCO removed the first surveillance capsules, designated capsule U, during the first refueling outage in August 1991 after 333.37 effective full power days (EFPD) of operation. The results from capsule U were published in Yankee Atomic Electric Company report YAEC-1853. The program utilizes six surveillance capsules.
Bases Section 3/4 4/9 has been revised to indicate that NAESCO is committed to removing the remaining surveillance capsules in accordance with the requirements of Appendix H and ASTM E185-73, and to evaluate the data obtained from the surveillance capsules in accordance with RG 1.99 (Rev. 2).
The staff finds that the proposed changes are based on applicable regulatory guidance and conform to the requirements of Appendix G of 10 CFR Part 50.
Therefore, the staff finds that the proposed changes are acceptable for incorporation into the Seabrook Unit 1 Technical Specifications.
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3.0 STATE CONSULTATION
i In accordance with the Commission's regulations, the New Hampshire and l
Massachusetts State officials were notified of the proposed issuance of the i
amendment. The State officials had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a I
facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no i
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 Commission has previously issued a proposed finding that the amendment
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involves no significant hazards consideration, and there has been no public comment on such finding (57 FR 58247). Accordingly, the amendment meets the
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eligibility criteria for categorical exclusion set forth in 10 CFR 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 the amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations 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, (2) such activities will be condur',J in compliance with the Commission's regulations, and (3) the issuance of tae amendment will not be inimical to the common i
defense and security or to the health and safety of the public.
6.0 REFERENCES
1.
Regulatory Guide 1.99 (Revision 2), Radiation Embrittlement of Reactor Vessel Materials, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2: Pressure-Temperature Limits 3.
Code of Federal Regulations, Title 10, Part 50, Appendix G, Fracture Toughness Requirements 4.
Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel l
Material Surveillance Program Requirements 5.
Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations, July 12, 1988 6.
E.C. Biemiller, R.J. Cacciapouti, Analysis of Seabrook Station Unit 1 Reactor Vessel Material and its Impact on Plant Operations, July 12, 1988
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l l i 7.
August 17, 1992, Letter from Ted C. Feigenbaum (NYN-92111) to USNRC Document Control Desk,
Subject:
License Amendment Request 92-06; Revised RCS Pressure / Temperature Limits
- 8. August 17, 1992, Letter from Ted C. Feigenbaum (NYN-92112) to USNRC Document Control Desk,
Subject:
Revised Reference Temperature Values for Pressurized Thermal Shock Events
- 9. August 17, 1992 Letter from Ted C. Feigenbaum (NYN-92113) to USNRC Document Control Desk,
Subject:
Reactor Vessel Surveillance Capsule Report Prir.cipal Contributor:
A. Wilford Date:
April 7, 1993 l
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