ML20205C395

Safety Evaluation Re Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61).Util 860123 Submittal Re Matl Properties & Fast Neutron Fluence of Reactor Vessel Acceptable

ML20205C395
Person / Time
Site:	Rancho Seco
Issue date:	03/13/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205C307	List: ML20205C302 ML20205C395
References
TAC-59976, NUDOCS 8703300173
Download: ML20205C395 (3)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST' PRESSURIZE 0 THERMAL SHOCK EVENTS (10 CFR 50.61)

SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 By letter dated January 23, 1986, the Sacramento Municipal Utility District, the licensee for the Rancho Seco plant, 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 (References 1, 2 and 3). Our evaluations of the pressure vessel material properties, and fast neutron fluence for fracture toughness requirements for protection against pressurized thermal shock events (10 CFR 50.61) are as follows:

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Material Properties The controlling tsettline material from the standpoint of PTS susceptibility was identified to be a longitudinal weld in the lower shell course, weld WF70 (Weld Wire Heat No. 72105).

This submittal states that there is a small probability that the controlling weld con,tains " atypical weld metal,",

an off-specification material that B&W found in 1978 in a test piece welded with wire heat number 72105.

This was the subject of Topical Report BAW 10144-A, which was reviewed by the staff on December 12, 1979. Based on a limited amount of test data developed by B&W, the submittal states that the screenino criterion will not be reached before end of life even if the atypical weld metal is present. The staff does not disagree with the B&W evaluation of this material, although their evaluation does not given in the PTS rule.

(The follow the procedure for calculating RT wasdevelopedwithoukThonsiderationoftheatypicalweld formula for RT(Nf believes that the probability of occurrence of atypical metal.) The 5 weld metal in vessel welds is low enough and its properties are such that the results of a plant-specific probabilistic risk analysis would not be af fected significantly if the atypical weld metal was considered to be present.

Thus, the materials input to the calcJlation of Rip will be evaluated without further consideration of the atypical weld Ntal.

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

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The justifications Ihc conten!Iing mJLeelaI has been proper 1y idcoLified.

Jrc.?Cceptabli' given for the Coppcr and niCI:Cl Corttent',,!nd the isti tia l RT,,

r thcsc values.

The margin hJs been dertved f ront can',ideratina of the bases',

followin9 the PTS Rule. Section 50.61 or 10 CFR Part 50.

fas t Neutron fluence Detailed calculations havq been performed with the assistance of our contracter.

Brookhaven National l'4bdratory (8tIL).for the. cycle depender't orcssure v'es-sel inner-wall (E > h0 HeV) flu'xes for the Rancho Seco pl' ant.

Th'e calculated fluxes were used to' determine the vessel (inner diameter) accumulated fluence.

for specific welds in the reactor which was then used to predict the RTPTS The analysis was based on a 00T calculation for the Rancho Seco plant.

vessel.

An 80% load f actor. ' low leakage loading and plant specific data were assumed for the extrapolation.

The following table summarizes the comparison of the Joplicant and the BNL results.

The BNL estimates are in substantial agreement with the applicant values, therefore, they are acceptabic.

The RT values pyg At the expiration of the current license are below the applicable criteria specified in 10 CFR 50.61 and are acceptabic.

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-0.05 a) Ref. I PTS =I+M[-10+470Cu+350CuNi]f.270 0

b) RT 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 f.om the calculation PTS of the Pressure Temperature limits. and (c) the staf f desire to be informed on the current value of the RT for all

PWRs, PTS we request that the licensee submit a reev,aluation of the RT and a compari-PTS son to the prediction of Reference 1 along with the Pressure-Temperature operat-ing limits which are required by 10 CFR 50 Appendix G."

It should be noted that this reevaluation is a requirement by 10 CFR 50.61, whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.

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