ML20148B423

Forwards Response to NRC Re Reactor Vessel Matl Properties & Noncompliance w/10CFR50 Toughness & Surveillance Requirements

ML20148B423
Person / Time
Site:	Diablo Canyon
Issue date:	10/02/1978
From:	Crane P PACIFIC GAS & ELECTRIC CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
NUDOCS 7810310226
Download: ML20148B423 (8)

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K NRC PUBLIC DOMENT R00E t

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' PACIFIC GAS AND' E LE C T R,IC C O M PANY MOL'41ti}.

1 77 BEALE STREET,31ST FLOOR SAN FR AN CISCO, C ALIFOR NI A 94106 (415) 781 4211

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Mr. John F. Stolz, Chief cu Light Water' Reactors Branch NO. 1 S

Division of Project Management c}

U.'S.-Nuclear. Reg"latory Commission T A

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Re:

Docket No. 50-275-OL Docket No. 50-323 -OL W

Diablo Canyon Units 1 & 2 6

Dear Mr. Stolz:

03 Attached.are 40 copies of responses to your 7

' letter. dated July 28,'1978 regarding reactor vessel material properties,non-compliance with 10 CFR 50,

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-Appendix G-fracture. toughness requirements,'and 10 CFR 50^, Appendix H-reactor vessel material sur-i veillance program requirements.

Five. copies.of these reports have been cent directly to Mr. Dennis Allison.

. Kindly acknowledge receipt of the above i

material'on the enclosed copy of this letter and re-turn it to me' in the enclosed addressed envelope.

very truly yours,

'i Philip A. Crane,.Jr.

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. Enclosures'

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Mr.. Dennis'Allison

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Areas of Non-Compliance with 10CFR50 - Appendix G

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The ferritic materials in the Diablo Canyon Units 1 and 2 reactor ves-sels were not all tested to meet the later ASME Section III Boiler and-Pressure Vessel Code required by 10CFR50 Appendix G since these vessels j

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were fabricated to the 1966 and 19.68 Editions of the Code, respec-I tively.

Therefore, Charpy impact test orientation was parallel to the working or rolling direction of the base materials rather than normal to the working or rolling direction as required by Appendix G.

Westing-house did, however, perform additional impact tests on the intermediate and lower shell course plates of both vessels which surrcund the effec-tive height of the fuel assemblies.

Full Cha py test curves were obtained on these plates from specimens oriented normal to the principal i

l rolling direction.

A summary of these results is presented in Tables 5.2-21A & B.

Full Charpy curves for all the base material on the ves-l-

sels have been obtained by the fabricator on impact specimens oriented i

parallel to the principal working or rolling direction.

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materials not in the beltline region were estimated using methods iden-1 tified in the NRC Star.dare. Review Plan 5.3.2.

l Weld metal test specimens were removed from a separate weldment using excess material from the beltline region plates.

Reactor vessel bolting material toughness tests were not performed to i

meet a' minimum requirement of 25 mils lateral expansion and 45 ft-lb ener'gy at the preload temperature'or at the lowest service temperature, i

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Tests were performed to meet 35 f t-lbs at 10 F.

The results of the tests are shown in the attached tables.,

These results show that some of the bolting material did not meet the 45 f t-lb requirement of Appendix G at 10 F.

It is expected that these _ materials would exhibit at least 45 f t-lbs if tested at 50 F which is considered to be the lowest pre-load or service temperature. Although lateral expansion measurements were not performed in most cases, it is also expected that these mate-rials would exhibit at least 25 mils lateral expansion if tested at 50 F based on test results on other bolting materials where both impact energy and lateral expansion were obtained.

Stress intensity f actors for various vessel locations were not calcu-lated to determine if they are lower than the reference stress intensity ~

f actors specified in' ASME Section III Code Appendix G.

Westinghouse has performed these calculctions for many vessels and has shown that the

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calculated stress ' intensity f actors are always lower than the reference 1

stress intensity. Westinghouse is confident that if calculated stress intensity factors for the Units 1 and 2 vessels were obtained, the values would fall below the reference stress intensity f actors.

Pressure-temperature limits for the reactor vessels have been estab-lished using the methods identified in NRC Standard Review Plan 5.3.2.

Th( assumptions and sample calculation identified in the Standard Review Plan are similar to those used by Westinghouse.

Drop weight specimens were machined parallel.to the rolling direction. Charpy impact test specimens were oriented as. identified earlier.

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' Areas of Non-Compliance With 10CFR50 - Appendix H 1.

The surveillance program for Unit No. I does not comply with ASTM E-185-73.

This program was designed to conform with ASTM E-185-70; therefore the n' umber of test specimens and their. orientation is not in agreement with ASTM E-185-73.

The surveillance weld material is repre-sentative but not identical to that used in the beltline region of the vessels.-

Four of the six capsules in Unit No. 2 will receive a neutron' flux 3.6 l

times as high as that received by the vessel inner surface and therefore are not in agreement with the requirement of Appendix H which requires that the neutron flux received by.the specimens is not more than threel times as high as that received by the vessel inner surface.

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