ML17157B017
| ML17157B017 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/10/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17157B016 | List: |
| References | |
| NUDOCS 9201220291 | |
| Download: ML17157B017 (8) | |
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- 0 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 RELATED TO AMENDMENT N0.116TO FACILITY OPERATING LICENSE NO. NPF-14 AMENDMENT NO. 85 TO FACILITY OPERATING LICENSE NO. NPF-22 PENNSYLVANIA POWER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
~SUS UEHANNA STEAM ELECTRIC STATION UNITS 1
AND 2 DOCKET NOS. 50-387 AND 388-Vl O~ ~
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
1.0 INTRODUCTION
By letter dated April 18,
- 1991, as supplemented September 27, 1991 and January 3, 1992, the Pennsylvania Power and Light Company and Allegheny Electric Cooperative, Inc. (the licensees) submitted a request for changes to the Susquehanna Steam Electric Station, Units 1 and 2, Technical Specifications (TS).
The requested changes would make changes to the Susquehanna Steam Electric Station (SSES),
Unit 1 and Unit 2 Technical Specifications to revise the pressure-temperature (P/T) curves for compliance with 10 CFR Part 50, Appendix G, as requested in Generic Letter 91-01.
The proposed changes affect Technical Specification Section 3.4.4.6, "Pressure/Temperature Limits" and Bases Section 3/4.4.6, "Pressure/Temperature Limits."
The September 27, 1991 and January 3,
1992 letters provided clar'.fying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION 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, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG 1.99, Rev. 2; Standard Review Plan (SRP) Section 5.3.2; and 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 United States.
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.
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S Appendix 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 survei llance capsules be tested in accordance with Appendix H of 10 CFR Part 50.
Appendix H, in turn, refers to ASTM Standards.
These tests define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature.
Appendix G also requires the licensee 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 license'es and permittees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor vessel materials.
This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 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, 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.
The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Susquehanna 1 and 2 reactor vessels.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 32 EFPY for Unit 1 was the lower intermediate shell plate C2433-1 with 0.105 copper (Cu),
0.63% nickel (Ni), and an initial RT d of 18'F; and the material with the highest ART at 32 EFPY for Unit 2 wa3 )he lower shell plate 6C1053-1-1 with 0.10%
Cu and 0.58% Nl, and an initial RTndt of 10'F.
The licensee has not removed any surveillance capsules from Susquehanna 1 and 2.
All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material in Unit 1, lower intermediate shell plate C2433-1, the staff calculated the ART to be 53.0'F at 1/4T (T = reactor vessel beltline thickness) for 32 EFPY.
For the limiting beltline material in Unit 2, lower shell plate 6C1053-1-1, the staff calculated the corresponding ART to be 44.8'F.
2In the above calculation~,
the staff used a neutron fluence of 4.3E17 n/cm at 1/4T and 2.0E17 n/cm at 3/4T.
The ART was determined by Section 1 of RG 1.99, Rev. 2.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 57.6'F at 1/4T for the same limiting lower intermediate shell plate in Unit 1 for 32 EFPY and 49.5'F for the same limiting lower shell plate in Unit 2.
The staff judges that the licensee's ARTs of 57.6'F for Susquehanna 1 and 49.5'F for Susquehanna 2 are more conservative than the staff's ARTs of 53'F for Unit 1 and 44.8F for Unit 2, and they are acceptable.
Substituting the ART of 57.6'F into equations in SRP 5.3.2 for Unit 1 and 49.5X'F for Unit 2, the staff found
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A further investigation into these non-beltline P/T limits (Ref.
5
& 6) conf'.rmed their'soundness, and therefore verified that the proposed P/T limits for heatup,
- cooldown, and hydrotest meet the beltline material requirements in Append'.x G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when the pressure exceeds 20Ã 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.
Paragraph IV.A.3 of Appendix G states "an exception may be made for boiling water reactor vessels when water level is within the normal range for power operation and the pressure is less than 20 percent of the pre-service system hydrostatic test pressure."
In this case the minimum permissible temperature is 60'F (33'C) above the reference temperature of the closure flange regions that are highly stressed by the bolt preload.
Based on the flange reference temperature of 10'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 that the predicted Charpy USE at end of life be above 50 ft-lb.
Using Figure 2 in RG 1.99, Rev. 2, it was estimated that the lowest end-of-life (EOL)
USE for the limiting beltline material
'.n Susquehanna 1 would be 70.8 ft-lb.
The corresponding EOL USE for Susquehanna 2 would be 53.7 ft-lb when the same estimation procedure was used.
Since both numbers satisfy the 50 ft-lb requirement, they are acceptable.
The staff concludes that the proposed P/T limits for reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 32 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.
An application of SRP 5.3.2 to limiting beltline materials for both units confirmed that they are less restrictive than the proposed non-beltline P/T limits for the feedwater nozzle.
- Hence, the proposed P/T limits may be incorporated into the Susquehanna, Units 1 and 2 Technical Specifications.
The licensee also makes an administrative change to page B 3/4 4-6 of the Unit 1 technical specifications as per a recent telecommunication on November 20, 1991.
Additionally, by letter dated January 3, 1992, the 1'.censee informed the staff that they are committed to include changes to the specimen withdrawal schedule in the next annual update to the FSAR, as recommended in Generic Letter 91-01.
This change in no way affects the no significant hazards determination as submitted by the licensee.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments.
The State official had no comments.
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4.0 ENVIRONHENTAL CONSIDERATION The amendments change a requirement with respect to installation or use of a faci lity component located within the restricted area as defined in 10 CFR Part 20.
The NRC staff has determined that the amendments involve 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 Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (56 FR 22472).
Accordingly, the amendments meet 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 amendments.
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 conducted in compliance with the Commission's regulations, and (3) 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:
C. Y. Cheng J. J. Paleigh Date:
January 10, 1992
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