ML20246P093
| ML20246P093 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/28/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246P090 | List: |
| References | |
| NUDOCS 8909110039 | |
| Download: ML20246P093 (4) | |
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION 5
ij WASHINGTON. D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS'.~89 AND 82 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT N05. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
In response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations," the Northern States Power Company (the licensee) requested to revise the pressure / temperature (P/T) limits in the Prairie Island Nuclear Generating Plant (PINGP) Unit Nos. I and 2 Technical Specification Section 3.1.
The request was documented in a letter from the licensee dated January 12, 1989.
ne purpose of the revision is to change the effectiveness of the P/T limits from 15 to 20 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 the data from actual surveillance capsules. 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 cuidance: Appendices G arc H to 10 CFR Part 50; the ASTM Standards and ASME Code, which are refercreed in Appendices G and H; 10 CFR 50.36(c)(2); Regulatory Guide 1.99, Revision 2; Standard Review Plan (SRP) Section 5.3.2 ano Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the Technical Specifications. The P/T limits are among the limiting conditions of operation in the Technical Specifications for all commercial nuclear plants in the U.S.
Appendices G and H to 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance. These must be considered in setting P/T limits. An acceptable method in constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G to 10 CFR Part 50 specifies fracture toughness and testing require-nents for reactor vessel naterials in accordance with the ASME Code and, in particular, to test the beltline materials in the surveillance capsules in accordance with Appendix H to 10 CFR Part 50. Appendix H, in turn, refers to the ASTM Standards. These tests define the condition of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference j
temperature. Appendix G also requires the licensee to predict the effects j
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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 licensees and permittees use the methods in Regulatory Guide (RG)1.99,' Revision 2topredicttheeffectofneutronirradiationon reactor vessel materials. This guide defines the ART as the sum of the uni
- radiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and margin to account for uncertainties in the predic'. ion method.
Appendix H to 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, requires that the capsules be installed in the vessel before startup and that they contain test specimens-that are made from plate, weld and heat-affected-zone materials of the reactor beltline.
2.0 EVALUATION We have evaluated the effect of neutron irradiation embrittlement on each beltline material in PINGP-1 and PINGP-2 reactor vessels. The amount of neutron irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
We have determined that the material with the highest ART (most embrittled) at 20 EFPY for both units was the circumferential weld between the intermediate and lower shells in Unit 2.
The licensee has removed three surveillance capsules from PINGP-1 and three capsules from PINGP-2. The results.from capsules V, P, and R in Unit I were pu]lished in Westinghouse Reports WCAP-8916, WCAP-10102, and WCAP-11006, respectively. The results from capsules V, T, and R in Unit 2 were published in Westinghouse Reports WCAP-9212, WCAP-9877 and WCAP-11343, respectively.
All surveillance capsules contained Charpy impact specimens and tensile specimens which were made from base metal, weld metal, and heat affected zone (HAZ) metal.
For the limiting beltline material, the weld between the intermediate and lower shells in Unit 2, we have calculated the amount of neutron irradiation embrittlement in accordance witn RG 1.99, Rev.2. The ART at 20 EFPY at AT was calculated to be 126*F. The ART was determined by the least squares extrapolation method of the Unit 2 surveillance data. The least squares method is described in Section 2.1 of RG 1.99, Rev. 2.
The licensee used the method in RG '. 99, Rev. 2., to calculate an ART of 144.8"F for the liuiting weld metal in the beltline of Unit 2.
We performed a similar calculation and verified the licensee's ART value to be conservative (see Table 1). Substituting the ART of 144.8'F into equations in SRP 5.3.2, we verified that the proposed P/T limits for heatup, cooldown, criticality, and hydrotest meet the beltline material requirements in Appendix G to 10 CFR Part 50.
- 4 In addition to beltline materials, Appendix G to 10 CFR Part 50 also imposes l
P/T limits based on the reference temperature for the reactor vessel closure l
flange materials.Section IV.2 of Appendix G states that when pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the
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temperature of the closure flange regions that are 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 4*F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires the predicted Charpy USE at end of life to be above 50 ft-lb. Based on data from a surveillance capsule withdrawn at 8.56 EFPY, the measured Charpy USE is 75 ft-lb for the intermediate to lower shell weld metal. This is a 4.5% reduction from the unirradiated value of 78.5 ft-lb. Using the method in RG 1.99, Revision 2, the predicted Charpy USE of the weld metal at end of life will be below 50 ft-lb. The staff will monitor the weld metal Charpy USE from future surveillance capsules. The surveillance capsule data will provide early warning of the decrease in Charpy USE, because the surveillance capsule lead factors are greater than 1.0.
Furthermore, since capsule R was withdrawn after 8.56 EFPY and had a lead factor of 2.93, the Charpy USE will be greater than 75 ft-lb for about 25 years of reactor operation.
Based on the above evaluation, the staff concludes that the proposed P/T limits on the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 20 EFPY, because the limits conform to requirements of Appendices G and H to 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.
Hence, the proposed P/T limits that would be incorporated into the PINGP-1 and PINGP-2 Technical Specifications are acceptable.
The licensee has also proposed several miscellaneous administrative changes that serve to clarify the TS requirements. The evaluation of these proposed changes are as follows:
1.
The TS 3.1.B.3 deals with the pressurization requirements of the steam generator when the temperature is below 70 F.
The licensee proposes the replace word " vessel" with the phrase " steam generator shell." This change serves to clarify the requirement and does not affect the intent of the requirement. The staff finds the proposed change acceptable.
2.
The description of the steam generator pressure / temperature limitation and the pressurizer temperature limits are being reinserted in the bases section. The descriptions were inadvertently omitted during the previous revision to the heatup and cooldown curve (i.e.,
Amendments Nos. 80 and 73 dated November 14,1966). This is considered an editorial change having r.o affect on the heatup and cooldown requirement and therefore is acceptable.
3.
The licensee is proposing several changes to TS.3.1.G which serve to enhance the requirements concerning the RCS temperature below the minimum presspirzation temperature (MPT). The proposed changes consist of specifying the temperature value (310'F) for MPT.
Other proposed changes involve actions to assure that the safety injection pumps are not operated at RCS temperature below 200'F and would be operated for maintaining cooling capability and inventory control only when residual heat removal system is inoperable. ' These proposed changes are considered administrative in nature, clarifying the requirement and therefore are acceptable.
3.0 ENVIRONMENTAL' CONSIDERATION Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on August 28, 1989 (54 FR 35542). Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, and (2) public 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.
Principal Contributors:
B. Elliot D. DIanni Dated:* August 28, 1989
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