ML20034B725
| ML20034B725 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/23/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034B724 | List: |
| References | |
| GL-88-11, NUDOCS 9004300254 | |
| Download: ML20034B725 (5) | |
Text
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ENCLOSURE
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.r SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION, UNIT ?
p DOCKET NO. 50-353 1.0 1HTRODUCT10N in response to Generic letter 88-11. "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Effect on Plant Operations," Philadelphia-Electric Company (the licensee) presented the: current pressure / temperature (P/T) limits in the Limerick Generating Station Unit 2 Technical Specification.-
Section 3.4 The response was documented in. letters from the licensee dated November 23, 1988 and March 31, 1989. The purpose of the response is to show that the current P/T limits of 2.5 and 8 effective full power years (EFPY)-
satisfy Regulatory Guide (RG) 1.99, Revision 2..
The P/T limits are used 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: 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 comercial 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.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing-l 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 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 the licensees and permittees use the g
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. methods in Regulatory Guide 1.99, Revision 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 (HA7) materials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the 1.imerick 2 reactor vessel. The amount of neutron irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART (most embrittled) at 2.5 and 8 EFPY was plate 14-2 (heat number B3416-1) with 0.14%
j copper (Cu), 0.65% nickel (Ni), and an initial RT f 40*F.
ndt Presently, no surveillance capsules have been withdrawn from the reactor pressure vessel. All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HA7 metal.
l Since the current P/T limits consist of two sets of curves that are valid for 2.5 EFPY and 8 EFPY, the staff has to calculate the highest ART for both EFPY cases.
For plate 14-2 at 2.5 EFPY, the staff calculated the ARTS to be 61'F at 1/4T (T = reactor vessel beltline thickness), and 52.3'F at 3/4T. The staff used a fluence of 9.0E16 n/cm2 at 1/4T and 4.0E16 n/cmr at 3/4T. At 8 EFPY, the ARTS were calculated to be 83.9'r and 67.7'F using fluences of 2.9E17 n/cmr and 1.3E18 n/cmr at 1/4T and 3/4T, respectively. The ARTS were determined using Section 1 of RG 1.99.~Rev. 2, because no surveillance capsules have been withdrawn from the reactor pressure vessel.
The licensee calculated that the RT would shift upwards by 44'F from an d
of40*Fasaresultoffhefluenceat1/4 Tat 8EFPY._There-initial RT fore, the b would be 84'F.
This agrees well with the staff's calculated i
value of 83.9'F.
Substituting the ART of 84'F-into equations in SRP 5.3.2, the staff verified that the current P/T limits for 8 EFPY for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR t
Part 50. The staff reviewed the licensee's P/T limits for 2.5 EFPY for heatup.
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cooldown, and hydrotest and found that they also meet the requirements of Appendix G of 10 CFR Part 50.
l In addition to the beltline materials, Appendix G of 10 CFR Part 50 also i
imposes P/T limits based on the reference temperature for the reactor closure vessel flange materials.
Section IV.2 of Appendix G states that when the i
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' pressure exceeds 20% of the preservice system hydrost:itic 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% 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 preservice 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 s
flange reference temperature of 10'F, the staff has determined that the current 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 (EOL) be above 50 ft-lb. The Limerick 2 Final Safety Analysis Report indicates that no USE values for the beltline materials were determined. Using figure 2 of RG 1.99 Rev. 2, the licensee has calculated the minimum unirradiated Charpy impact upper shelf energy.(USE) required to meet the Appendix G requirement of 50 ft-lb or greater at EOL. The licensee has also presented data showing that plates from the same manufacturer of similar composition meet that minimum unirradiated USE value. A similar approach has been used to show that the welds will also meet the Appendix G requirement.
The staff has accepted this approach in the timerick I and 2 SER, NUREG-0991, and considers that the Appendix G requirement of 50 ft-lb or greater has been met.
3.0 CONCLUSION
The staff concludes that the current P/T limits on the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 2.5 and 8 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.
Hence, the current P/T limits may be maintained into the Limerick 2 Technical Specifications.
Dated: April 23,1990 Principal Contributor:
J. Tsao
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4.0 REFERENCES
1.
Regulatory Gu!ie 1.99, Radiation Embrittlement of Reactor Vessel Materials.
Revision 2, Mai'1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2'P,ressure-Temperature Limits t
3.
Limerick Generating Station, Unit 2, FSAR i
4 Limerick Generating Station, Unit 2 Technical Specifications 5.
November 23, 1988, Letter from J. W. Gallagher (PECo) to T. Murley (USNRC),
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3, Limerick Generating Station, Units 1 and 2: Response to Generic letter 88-11 6.
March 31, 1989, LetterfromJ.S.Kemper(PECo)toUSNRCDocumentCdntrol Desk,
Subject:
. Limerick Generating Station, Unit 2, Response to Generic Letter 88-11
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NUREG-0991 Safety Evaluation Report Related to the Operation of Limerick Generating Station, Units 1 and 2, August 1989 l
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Mr. George A. Hunger, Jr.
Limerick' Generating Station-Philadelphie Electric Company Units 1 & 2:
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cC' Troy B. Conner, Jr., Esquire Mr. Thomas Gerusky, ' Director
'4 Conne" and Vetterhahn Bureau of Radiation Protection 1747. Pennsylvania Ave., N.W.
PA Dept. of Environmental Resources Washington, D. C.
20006 P. O. Box 2063
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Herrisburg, Pennsylvania 17120 Mr. Rod Krich 52A-5
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Philadelphia Electric Company' Single Point of Contact.
955 Chesterbrook Boulevard P. 0. Box 11880 Wayne, Pennsylvania 19087-5691 Harrisburg, Pennsylvania 17108-1880-Mr. Graham M. Leitch,-Vice President Mr. Philip J. Duca Limerick Generating Station Support Manager-Post Office Box A Limerick Generating Station Sanatoga, Pennsylvania 19464 P. O. Box A Sanatoga, Pennsylvania 19464 Mr. Marty J. McCormick, Jr.
J Plant Manager Mr. Gary Edwards Limerick Generating Station Superintendent-Technical P.O. Box A Limerick Generating Station Sanatoga, Pennsylvania 19464 P. O. Box A Sanatoga, Pennsylvania. 19464 Mr. Larry Doerflein U.S. Nuclear Regulatory Comission Mr. Gil'J. Madsen Region 1 Regulatory Engineer 475 Allendale Road Limerick Generating Station King of Prussia, PA 19406 P. O. Box A-Sanatoga, Pennsylvania 1946_4 Mr. Thomas Kenny Senior Resident inspector US Nuclear Regulatory Comission P. O. Box 596 Pottstown, Pennsylvania 19464 Mr. John Doering Project Manager Limerick Generating Station P. O. Box A 1
Sanatoga Pennsylvania 19464 Mr. Larry Hopkins Superintendent-Operations Limerick Generating Station P. O. Box A Sanatoga, Pennsylvania 19464 r