ML20062D700

Safety Evaluation Supporting Util 881123 & 900917 Responses to Generic Ltr 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Matls & Its Effect on Plant Operations. Submittals Acceptable.Beltline Welds Discussed

ML20062D700
Person / Time
Site:	Farley
Issue date:	11/09/1990
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20062D698	List: ML20062D691 ML20062D700
References
GL-88-11, NUDOCS 9011150093
Download: ML20062D700 (4)

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_ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING RESPONSE TO GFNERIC LETTER 88-11 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 DOCKET NO. 50-348

. 1.0 INTRODUC110N AlabamaPowerCompany(APCoorthelicensee)submitteditsresponsetoGeneric Letter 88-11. "NRC Position on Radiation Embrittlement of Reactor Vessel a Materials and Its Effect on Plant Operations," in a letter dated November 23, 1988, as supplemented on September 17, 1990. APCo's response indicated that the required technical analysis had been completed and that for Joseph M.

Farley Nuclear Plant (Farley) Unit 1, the current pressure / temperature (P/T);

limits in the Technical Specifications are bounded by the methodology contained inRegulatoryGuide(RG)1.99,Revisica2. The current P/T limits are valid-for 16 effective full power years (EFPY).

To evaluate the P/T limits, the NRC staff uses the.following NRC regulations and guidance: (1)A)pendicesGandHof10CFRPart50;(2)theASTMStandards and the ASME Code (wiich are referenced in Appendices G and H); (3) 10 CFR 50.36(c)(2);.(4) RG l'99, Revision 2; (5) NUREG-0800 " Standard Review Plan" (SRP), Section 5.3.2; and (6) 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 ccamercial 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 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 i embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature. Appendix G also requires the licensee to Predict the 9011150093 901109

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effects nf 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 contained in RG 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 CFP, 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 Farley, Unit 1, reactor vessel. The amount of irradiation embrittlement was calculated in accordance with RG 1.99 Revision

2. The staff has determined that the material with the highest ART at 16 EFPY was the lower shell plate B6919-2 with 0.14% copper (Cu), 0.56% nickel (N1),

and an initial RT ndt of 5'F.

The licensee has removed three surveillance capsules from Farley, Unit 1. The results from capsules Y, V, and X were published in Westinghouse reports WCAP-9717, WCAP-10474, and WCAP-11563, 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, lower shell plate B6919-2, the staff calculated the ART to be 143.5'F at 1/4T (T = reactor vessel beltline

', thickness)and of 1.26E19 n/cm{17.6"F for 3/4T at 1/4T and 4.89E R n/cmat at 16 3/4T. EFP{. The ART The staff was used by detemined a neutron fluence using Section 1 of RG 1.99, Revision 2 as the limiting material was not contained in the surveillance capsules.

The licensee used the method in RG 1.99, Revision 2, to calculate an ART of 146.4'F at 16 EFPY at 1/4T for the same limiting plate metal. The staff judges that the licensee's ART of 146.4'F is more conservative than the staff's ART of 143.5'F and is acce) table. Substituting the ART of 143.5'F into the equations of SRP 5.3.2, tie staff verified that the current P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.

P 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

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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 60*F the staff has detemined t that the current P/T limits satisfy Section IV.2 of Appendix G.

g Section IV.B of Appendix G requires that the predicted Charpy USE at end of life (EOL) be above 50 f t-lb. The vessel beltline plates have unirradiated USE data which indicate that their EOL USE values will be above 50 ft-lb.

However, the unirradiated USE's of the three beltline welds are either nonexistent or unreliable. The USE value for the lower shell seam weld G1.08 does not exist. The USE's of the intermediate shell seem weld, M1.33, and girth weld, G1.18 are 103 ft-lb and 101 ft-lb respectively, measured at 10 degrees F only. The USE for reactor vessel materials normally is obtained by performing several Charpy V notch impact tests at various temperatures, not 7

just at one temperature. The staff presently does not have unirradiated USE's to accurately predict the USE's of the welds at end-of life.

ThesurveillancewelddatafromtheCagsuleU,X,andYreportsshowedlibft-lb at a neutron fluence of 2.8E19 n/cm . Tgelicenseeestimatedthatthe fluence at end of life will be 3.75E19 n/cm and as of May 1990 (about 7.3 EFPY), the fluence is about 1.1E19. Based on the surveillance data and fluence, the staff determined that the USE will be above 50 ft-lb at least for operation through the currently licensed 16 EFPY, However, the staff will closely monitor the USE during review of future required licensee submittals of capsule surveillance reports and P/T limit amendment requests to ensure that the 50 ft-lb requirement continues to be satisfied.

3.0 CONCLUSION

The staff concludes that the current P/T limits for the reactor coolant system

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for heatup, cooldown, leak test, and criticality are valid through 16 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies the requirements of Generic letter 88-11 as the methodology contained in RG 1.99, Revision 2 was used to calculate an acceptable ART. Hence, the current P/T limits contained in the Farley, Unit 1, Techt.ical Specifications are acceptable.

Although the USE of the beltline welds is above 50 ft-lb at present, the staff cannot predict whether the USE will be above 50 ft-lb at EOL. As stated above, the 50 ft-lb requirement contained in Appendix G to C CFR Part 50, is met for operation of Farley, Unit 1, through 16 EFPY as currently licensed. However, future licensee submittals will need to demonstrate that the USE of the reactor vessel beltline welds continues to meet the 50 ft-lb requirement.

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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."

3. November 23, 1986, Letter from W. G. Hairston,111 (APCo) to USNRC Docusent Control Desk,

Subject:

NRC Position of Radiation Embrittlement of Reactor Vessel Materials and its Inpact on Plant Operations. 1

4. S.E. Yanichko, et al.; " Analysis of Capsule Y From f.he Alabama Power Company, Farley Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-9717, June 1980.

5. P..S. Boggs, et al.; " Analysis of Capsule U From the Alabama Power Company,

' Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program "

WCAP-10474, February 1984.

6. R.P. Shogan, et al.; " Analysis of Capsule X From the Alabama Power Company, Jose)h M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program," WCA)-11563 Revision 1. September 1987.

7. September 17, 1990, Letter from W. G. llairston III (APCo) to USNRC Document Control Desk,

Subject:

Joseph H. Farley Nuclear Plant Fluence at Vessel Inner Radius.

Principal Contributor: J. Tsao Dated:

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