ML20005G037
| ML20005G037 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 01/10/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20005G036 | List: |
| References | |
| NUDOCS 9001180025 | |
| Download: ML20005G037 (5) | |
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![smrg'g UNITED STATES NUCLEAR REGULATORY COMMISSION L
W ASHINGTON, D. C. 20665 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N05.125 AND 129 TO FACILITY OPEkATING
.ICENSE N05. DPR-Za AND DPR-27 WISCONSIN U.ECTRIC POW x COMPANY POINT BEALH NUC MR PLANT..U(IT MQ5..). AND - 2 D90KET 4)5. 50-266 AND bD-301
1.0 INTRODUCTION
In response to Generic Letter 88-11, "NRC Position on Radiation Embrittle-ment of Reactor Vesset Materials and Its Effect on Plant Operatbns,- the WisconsinElectricPowerCompany(P/TJlimitsinthePointBeachNuclear (the licensee) requested permission to revise the pressure / temperature Plant Units 1 and 2. Technical Specifications Section 15.3.1. The request was documented in two letters from the licensee dated August 3 and October 3, 1989. This revision also changes the effectiveness of tie F/T limits to 18.1 effective full power years (EFPY). The licensee proposed to use one set of P/T limits for both units. The proposed P/T limits were developed based' on Regulatory Guide (RG) 1.99, Revision 2.
The proposed revision provides up-to-date P/T limits for the operation of the reactor coolant system during heatup, cooldown, criticality, and hydrotest.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance: A)pendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME code, w11ch are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG1.99,Rev.2;StandardReviewPlan(SRP)Section5.3.2;andGeneric Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CtR 50.36 to provide Technical Sp(ecifications for the operation of theIn particular,10 C plant.
of operation be included in the Technical Specifications. The P/T limits are among the limiting conditions of operation in the Technical Specifi-cations for all commercial nuclear plants in the U.S.
Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appenaix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveiliance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards.
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.a 2-These tests' define the extent of vessel embrittlement at the time of capsule 1
withdrawal in terms of the increase in reference temperature.
Appendix G i
r also requires the licer.see to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that i
licensees and permittees use the methods in RG 1.99, Rev. 2, to predict the effect of. neutron irradiation on reactor vessel materials.
This guide 3
defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to acccunt for uncertainties in the prediction method.
1 Appendix H of 10 CFR Part 50 requires the licensee to establish a surveil-lance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards.which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in Point Beach I and 2 reactor vessels.
The amount of irradiation embrittlement was calculated in accordance with Section 2 of RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 18.1 EFPY for both units was the circumferential weld
'8579/72442 between the intermediate and lower shells in Unft 2 with 0.26%
copper (Cu), 0.60% nickel (Ni), and an initial RT f -6 F.
ndt The licensee has removed four surveillance capsules from Point Beach Unit I and three capsules from Point Beach Unit 2.
The results from capsule V were published in a Battelle-Columbus Laboratories report dated June 15, 1973; and the results from capsules 5, R and T in Unit I were published in Westinghouse reports WCAP-8739, WCAP-9357, and WCAP-10736, respectively.
The results from capsule V were published in a Battelle-Columbus Labcra-tories report dated June 10, 1975; and the results from capsules T and R in Unit 2 were published in Westinghouse Reports WCAP-9331 and WCAP-9635, respectively.
All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material, weld 8579/72442, the staff calculated the ART to be 258.30F at 1/4T (T = reactor vessel beltline thickness) for 18.1 EFPY.
The staff used a neutron fluence of 2.05E19 n/cm2 at 1/4T.
The ART was determined by the least squares extrapolation method using the Point Beach Unit 2 surveillance data.
The least squares method is described in Section'2.1 of RG 1.99, Rev. 2.
i The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 258.40F at 18.1 EFPY at 1/4T for the same limiting weld metal.
Substituting the ART of 258.40F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
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!..* In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T linits based on the reference temperature for the reactor vessel closure flange materials.Section IV.2 of Appendix G states that when the pressure 3
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 flange reference temperature of 60*F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of. Appendix G recuires that the predicted Charpy USE at end of life be above 50 ft-lb.
Based on data from surveillance capsule R 2
withdrawn from Unit 2 with a fluence of 2.01E19 n/cm, the measured Charpy i
USE.is 47 ft-lb for the intermediate to lower shell weld metal 8579/72442.
l This is a 27.7% reduction from the unirradiated value of 65 ft-lb. Using the method in RG 1.99, Rev. 2, the predicted Charpy USE of the weld metal at the end of life will be below 50 ft-1b.
Based on data fron surveillance capsule R, the staff has also determined that the USE for weld 8579/72442 in Point Beach Unit 2 will be at 50 ft-lb in 8.1 EFPY, Based on data from surveillance capsule R withdrawn from Unit I with a fluence of ?.22E19 n/cm2, the measured Charpy USE is 51 f t-lb for the intermediate to lower shell weld metal 8350/6178?. This is a 21.5%
reduction from-the unirradiated value of 65 ft-lb. The staff will monitor the weld metal Charpy USE from future surveillance capsules.
The staff and the licensee know that the USEs of the Point Beach Units 1 and 2 reactor vessels will be below 50 ft-lb. The licensee has joined the Integrated Reactor Vessel Material Surveillance Progran sponsored by the B&W Owners Group to resolve the issue of low USE in the near future. The staff is also monitoring the progress of the program.
3.0 FINDINGS The staff concludes that the proposed P/T limits for the reactor coolant systen for heatup, cooldown, leak test, and criticality are valid through 18.1 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Rev. 2 to calculate the ART.
Further, the licensee has simplified the TS by making the most limiting set of curves applicable to both Units 1 and 2.
.Hence, the proposed changes may be incorporated into the Point Beach Units 1 and 2 I:._
Technical Specifications.
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4.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change in the insta11etion or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes an inspection or surveillance requirement. The staff has determined that the amendments involve no significant increase in the z
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amounts.-and no significant change in the types, of any effluents that may k
be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has pre-l viously published a proposed finding that these amendments involve no i
significant hazards consideration and there has been no public comment on L
L such finding.
Accordingly, these amendments meet the 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 these amendments.
5.0 ~ CONCLUSION The staff 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, and (2) such activities will be conducted in compliance with the Commission's regula-tions, 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.
Principal' Contributor:
J. Tsao Dated: January 10, 1990
6.0 REFERENCES
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1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2 Pressure-Temperature Limits 3.
R. W. Britt (Wisconsic Electric) letter to USNRC, August 3, 1989 4.
Final Safety Analysis Report, Point Beach Nuclear Plant, Units 1 and 2 5.
J. Wetmore (USNRC) letter to licensee, March 17, 1976 6.
J. S. Perrin, et al., " Final Report on Point Beach Unit No. 1 Pressure Vessel Surveillance Program Evaluation of Capsule V " Battelle-Columbus Laboratories, June 15, 1973 7.
S. E. Yanichko and S. L. Anderson, " Analysis of Capsule 5 from the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program, WCAP-8739," Westinghouse Electric Corporation, November 1976
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S. E. Yanichko, et al., " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program, WCAP-10736," Westinghouse Electric Corporation, December 1984 9.
S. E. Yanichko and S. L. Anderson, " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program, WCAP-9537," Westinghouse Electric Corporation, August 1978 1
- 10..J. S. Perrin, et al., " Point Beach Nuclear Plant Unit No. 2 Pressure Vessel Surveillance Program:
Evaluation of Capsule V," Battelle-i Columbus Laboratories, June 10, 1975 11.
J. A. Davidson, et al., " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program, WCAP-9331," Westinghouse Electric Corporation, August 1978 12.
S. E. Yanichko.and S. L. Anderson, " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program, WCAP-9635," Westinghouse Electric Corporation, December 1979 13.
R. A. Abdoo (Wisconsic Electric) letter to'USNRC, October 3, 1989 14.
C. W. Fay (Wisconsic Electric) letter to USNRC, November 29, 1988 l
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