ML20055D336
| ML20055D336 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/27/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20055D334 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 9007060191 | |
| Download: ML20055D336 (3) | |
Text
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0,, UNITED STATES i E
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20665
- ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION EVALUATION OF NRC GENERIC LETTER 88-11 RESPONSE l L PUBLIC SERVICE ELECTRIC AND GAS COMPANY q
, HOPE CREEK GENERATIF i' TIE DOCKET NO. (
1.0 INTRODUCTION
In response to Generic Letter 88-11. "NRC Position on Radiation Embrittlement '
of Peactor Vessel Materials and Its Effect on Plant Operations," the Public Service Electric and Gas Compan
.the pressure / temperature (P/T) limitsy (the in thelicensee) requested Hope Creek permission Generating Station to revise Technical Specifications, Section 3.4. The request was documented in a letter from the licensee dated November 28, 1988; March 8, 1990; and April 17, 1990.
This revision also changes the effectiveness of the P/T limits of 32 effective full power ~ years (EFPY) The proposed P/T limits were developed based on
- Regulatory Guide 1.99,-Revision 2. ' The proposed revision provides up-to-date '
P/T limits for the operation of the reactor coolant system d; ring heatup, cooldown, criticality, and hydrotest. '
To evaluate the P/T. limits, the staff uses the following NRC regulations and guidance: Append':es G and H of 10 CFR Part 50; the ASTM Standards and the ASMECode,whicharereferencedinAppendicesGandH;10CFR50.36(c)(2);
RG1.99,Rev.2;StandardReviewPlant(SRP)Section5.3.P;andGeneric
-Letter 88-11.
Each licensee authorized'to operate a nuclear power reactor :s 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 iimits 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..
T 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 '
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embrittlement at the time of capsule withdrawai 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'referencetemperature(ART)andCharpyuppershelfenergy-(USE).
Generic Letter 88-11 requested that 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 defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irrad1ation, 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 the .
test specimens made from plate, weld, and heat-affected-zone (HAZ) y contain materials of the reactor beltline.
2.0 EVALUATION The staff evaluhted the effect of neutron irradiation embrittlement on each >
beltline material in the Hope Creek 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 was intermediate shell-course plate 5K3025-1 with 0.15% copper (Cu), 0.71% nickel (Ni), and an initial RTndt of 19 F.
The licensee has removed one surveillance capsule from the Hope Creek' reactor l vessel. All surveillance capsules contain Charpy impact specimens and tensile specimens made from-base metal, weld metal, and HAZ metal.
For the limiting beltline material,-intermediate shell course plate 5K3025-1, the staff calculated the ART to be 64.5 F at 1/4T- (T = reactor vessel beltline thickness) and2 47.7 F for 3/4T at 32 EFgY. The staff used a. neutron fluence of 5.6E17 n/cm at 1/4T and 2.7E17 n/cm at 3/4T. The ART was determined by using Section 1 of RG 1.99, Rev. 2, because the licensee has removed only one l surveillance capsule from^the reactor vessel.
The. licensee used the method in RG 1.99, Rev. 2, to cat.ulate an ART of 65.8'F at 32 EFPY at 1/4T for the same limiting weld metal. The staff judges that a difference of'1.3*F between the licensee's ART of Sb.8'F and the staff's ART of.64.5'F is' acceptable. Substituting the ART of 65.8'F into equations in SRP L 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, L and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
In addition to belM ine materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materia'Is.Section IV.2 uf Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the i
. 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 ;
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3 least 120'F for~ normal operation and by 90'F for hydrostatic pressure tests
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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-L , 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 stress 9d 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 IY.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 initial USE is the girth weld between-the intermediate and loter intermediate shell courses with 68 ft-lb. Using the method in RG 1.39, Rev. 2, the predicted Charpy USE of the weld metal at the end of life will be 55.8 ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable. !
3.0 CONCLUSION
.The staff concludes that the proposed P/T limits for the 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. ~i Hence, the proposed P/T limits may be incorporated into the Hope Creek
. Technical Specifications.
4.0 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 32 Hope Creek Generating Station Final Safety Analysis Report 1
- 4. November 28, 1988, LetterfromS.E.Miltenberger(PSE&G)toUSNRC Document Control Desk,
Subject:
Response to Generic Letter 88-11
- 5. March 8,1990, Letter from T. M. Crimmins, Jr. (PSE&G) to USNRC Document- !
Control Desk,
Subject:
Request for Additional Information, Generic Letter 88-11
- 6. April 17,1990,. Letter from T. M. Crimmins, Jr. (PSE&C) to USNRC Document Control Desk,
Subject:
Additional Information Regarding PSE&G's Response to Generic Letter 88-11, Hope Creek Generating Station Dated: