ML20214J829

Safety Evaluation of Matl Properties & Fast Neutron Fluence for Fracture Toughness Requirements for Protection Against PTS (10CFR50.61).Pressure Vessel Meets Requirements Until End of Current OL on 201021

ML20214J829
Person / Time
Site:	Maine Yankee
Issue date:	11/20/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20214J827	List: ML20214J825 ML20214J829
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8612010496
Download: ML20214J829 (3)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FELAltD TO VATERIAL PROPERTIES AFC FAST NEUTRON FLUENCE FOR FRACTURE TOUGFNE55 REQUIREMENTS FOR PPOTECTION AGAINST PRESSURIZED THERVAL SHOCK EVENT 5 (10 CFR 50.61)

MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC F0WER STATION DOCKET NO. 50-309

1.0 INTRODUCTION

By letters dated April 22, 1983 and January 21, 1986, the Maine Yankee Atomic Power Company, licensee for the Maine Yankee Atomic Power Station, submitted information on the material properties and the fast neutron fluence (E > 1.0 MeV) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61.

2.0 EVALUATION 2.1 The following evaluation concerns the estimation of the fluence to the pressure vessel to the end of the current operating license on October 21, 2008 and the corresponding value of the RT PTS

• The calculation on which the Maine Yankee Atomic Power Company based its January 21, 1986 submittal was performed by Westinohouse in connection with an earlier pressurized thermal shock (PTS) submittal which was submitted by letter dated April 22, 1983. The calculation was performed with the DOT code using pinwise source distribution but it utilized a P scattering approximation 3

, and a cross section set which was not based on ENDF/B. Mcwever, Maine Yankee surveillance capsule 263 which was included in that calculation showed a measurement to calculation ratio of 1.51. The measurement of the capsule was performed by Battelle Columbus Laboratories (Reference 1). Because of the wide discrepancy between measurement and calculation, the capsule analysis was repeated by Hanford Engineering Development Laboratory (HEDL) and Brookhaven National Laboratory (BNL). These analyses established that the original discrepancy was due to the fact that the in-capsule water was omitted, affecting the calculated flux at the dosimeters and the neutron spectrum (References 2 and 3). The estimated measurement to calculation flux uncertaintywasabout12%(Reference 3). The licensee normalized the fluence estimate on the original capsule measurement which is conservative. Therefore, with calculations independent of the original licensee sponsored calculation, i it was established that the capsule 263 results were within the calculational uncertainty limits and normalized toward the ':pper limit.

In view of:

(a) the Pressure-Temperature updating requirements for the fracture toughness of the beltline material in 10 CFR 50 Appendix G, and (b) the fact that the RT value is readily available from the calculations ofthePressureTempbturelimits,and l

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(c) the staff desire to be informed on the current value of the RTPTS for all PWRs the licensee should submit a re-evaluation of the RT and a comparison to thepredictionofthatsubmittedintheJanuary21,Nh6letteralongwiththe future Pressure-Temperature operating limits which are required by 10 CFR 50 Appendix G*. It should be noted that this re-evaluation is a requirement by 10 CFR 50.61, whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.

The licensee projections were then based on acceptable results. The low leakage loading for the Cycle 9 fuel load and subsequent cycle fuel loads were used to estimate the fluence to the end of the current operating license.

The equation specified in 10 CFR 50.61, as applicable for the Maine Yankee plant is:

0 RT PTS = I+M+(-10+470xCu+350xCuxN1)xf .27 where: I = Initial RT = -56*F NDT M = Uncertainty Margin = 59"F Cu = w/o Cooper in circumferential weld 9-203 = 0.31 Ni = w/o Nickel in circumferential weld 9-203 = 0.74 f = peak azimuthal fluence to the end of

, license (E > 1.0 MeV) lower shellcircugerentialweld9-203in units of 10 n/ cme = 1.47 Therefore:

0 RT -56+59+(-10+470x0.31+350x0.31x0.74)x1.47 27 PTS= = 3+216x1.109 = 242.7*F which is lower than 300*F, the applicable PTS rule screening criteria and is acceptable.

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2.2 The following evaluation concerns material properties for fracture toughness requirements for protection against thermal shock events.

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• This request for information is covered under 0MB Clearance No. 3150-0011.

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i The controlling beltline material from the standpoint of PTS susceptibility was identified to be the intermediate to lower shell circumferential weld 9-203, weld wire heat number IP3571.

The material properties of the controlling material and the associated margin and chemistry factor were reported to be:

Utility Submittal Staff Evaluation Cu (copper content, %) 0.31 0.31 Ni (nickel conteit, %) 0.74 0.74 I (initial RT NDT, F) -56 -56 M (Margin, *F) 59 59 CF (Chemistry Factor, 'F) 216.0

3.0 CONCLUSION

S The licensee projections concerning the estimation of the fluence to the pressure vessel to the end of the current operating license on October 21, 2008 were based on acceptable results and the corresponding value of RT (242.7F)islowerthan300F,theapplicablePTSrulescreeningcriter$I3 ,

and is acceptable.

The controlling material has been properly identified. The justifications are acceptable.

givenforthecopperandnickelcontentsandtheinitialRT"Nrthesevalues, The margin has been derived from consideration of the bases following the PTS Rule, Section 50.61 of 10 CFR Part 50. Assuming that the reported values of fluence are correct, Equation 1 of the PTS rule governs, and the chemistry factor is as shown above.

References

1. Perrin, J. S. et. al " Maine Yankee Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule 263" BCL-585-21, dated December 12, 1980.

2. McElroy W. N. et. al " LWR Pressure Vessel Dosimetry Program; LWR Power Reactor Surveillance Physics Dosimetry Compedium" Hanford Engineering Development Laboratory, NUREG/CR-3319 dated August 1985.

3. Carew J. F., et. al " Pressure Vessel Fluence Benchmark Calculations" BNL-NUREG-34715, dated February 1984.

Date:

Principal Contributors: L. Lois and P. Randall