ML20126H954
| ML20126H954 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/24/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126H945 | List: |
| References | |
| NUDOCS 9301050373 | |
| Download: ML20126H954 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 76TO FACILITY OPERATING LICENSE'NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 DOCKET NO. 50-483
1.0 INTRODUCTION
By letter dated December 18, 1991, Union Electric Company (the licensee) requested changes to the pressure / temperature (P/T) limits in the Callaway Plant, Unit 1 Technical Specifications, Section 3.4.
The proposed.P/T limits
'are valid for 17 effective full power years (EFPY)-and were developed usi_ng Regulatory Guide (RG) 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2.
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. The P/T limits. provide for the operation of the reactor coolant system during heatup, cooldown, criticality, and hydrotest.
To evaluate the P/T limits, the staff used 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-ll.
Backaround Each licensee authorized to operate _a nuclear power reactor is required:by 10_ CFR 50.36 to provide ' Technical Specifications (TS) for the operation of the plant.
In particular,10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the TSs. -The P/T_ limits are among the limiting conditions of operation in the TSs' 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 PP limits.. An ' acceptable method for constructing the P/T limits-is described in SRP Section 5.3.2.
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 surveillance 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.
9301050373 921224 DR. ADOCK 0500 3
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 licensees 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.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Callaway I reactor vessel. 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 17 EFPY at 1/4T (T - reactor vessel beltline thickness) was the lower shell plate R2708-3 with 0.07% Cu, 0.59% Ni, and an initial RT of 20*F.
The material with the highest ART at 17 EFPY at 3/4T was the" lower shell plate R2708-1 with 0.07%
copper (Cu), 0.59% nickel (Ni), and an initial RT,3 of 50*F.
The licensee has removed two surveillance capsules from Callaway 1.
The results from capsules V and Y in Unit I were published in Westinghouse reports WCAP-11374 and WCAP-12946, 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 at 1/4T, lower shell plate R2708-3, the staff calculated the ART to be 94.6*F at 1/4T at 17 EFPY.
The ART for 1/4T was determined by Section 1 of RG 1.99, Rev. 2, because plate R2708-3 is not in the surveillance capsules.
For the limiting beltline material at 3/4T, lower shell plate R2708-1, the staffcalculatedtheARJtobe83.8' Fat 17EFPY.,Thestaffusedaneutron fluence of 7.62E18 n/cm at 1/4T and 2.74E18 n/cm' at 3/4T.
The ART for 3/4T was determined by the least squares extrapolation method using the Callaway 1 surveillance data. The'1 east squares method is described in Section 2.1 of RG 1.99, Rev. 2.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 95'F at 17 EFPY at 1/4T for the same limiting plate material.
The staff determines that the licensee's ART of 95'F is more conservative than the staff's ART of 94.6*F, and it is acceptable. Substituting the ART of 95'F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, i
cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
l
. 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 cf Appendix G states that when. tte 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 flange reference temperature of 40*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.
The material with the lowest predicted E0L USE is intermediate shell plate R2707-1 with a unirradiated USE of 78 ft-lb. Using Figure 2 of RG 1.99, Rev. 2, the staff calculated that the EOL USE at 1/4T for this material is 61.9 ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Missouri State official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the instal-lation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or a change to a surveillance requirement. The staff has determined that the amendment invol',es 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 this amendment involves no significant hazards consideration and there has been no public comment on such finding (57 FR 28206). Accordingly, this amendment meets 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 this amendment.
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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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6.0 REFERENCES
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.
December 18, 1991, letter from D. F. Schnell (UECo) to USNRC Document -
Control Desk, subject:
Callaway Plant Revision to Technical Specification 3/4.4.9 Pressure / Temperature Limits 4.
R. G. Lott, (t al., Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-11374, Westinghouse Electric Corporation, June 1987 5.
E. Terek, et al., Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-12946, Westinghouse Electric Corporation, June 1991 Principal Contributor:
S. Sheng Date: December 24, 1992 u
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