ML18230A218

Consisting Supplementary Clarifications to CP&L 5/17/1977 Letter Requesting Authorization to Use Certain Components Which Have Been Certified to Earlier Code Addenda of ASME III than Those Prescribed by 10 CFR 50.55a(c),(d)&(e)

ML18230A218
Person / Time
Site:	Harris   (NPF-063)
Issue date:	06/28/1977
From:	Mcduffie M Carolina Power & Light Co
To:	Case E Office of Nuclear Reactor Regulation
References
Download: ML18230A218 (6)

Text

NRC FQRM 195 (2 76I U.S. NUCLEAR REGULATORY COM

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6i NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL FILE NUMBER TO:

Mr. Edson Gi Case FROM:

Carolina Power & Kight Company

Raleigh, North Carolina Mi Ai McDuffie DATE OF DOCUMENT 6/28/77 DATE RECEIVED 7/1/77 BETTE R RRRIGINAL OCOP Y DESCRIPTION ONOTORIZED PQJ NC LASS I F I E 0 PROP INPUT FORM ENCLOSURE NUMEER OF COPIES RECEIVED

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Consists of supplementary clairifications to their 5/17/77,ltr requesting authorization from the NRC to use certain COIIIponents which have been certified to earlier code addenda of ASME III than those prescribed by Seci 50'5a(c),

(d),

and (e) of 10 CFR Part 50, ~ ~ ~ ~

~ inotorized 6/28/77 '

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PLANT NAME:

Shearon Harris Units 1-2-3-4 RJL 7/1/77 (2-P) gpgPT B,EMOTE FOR CTION/INFO RMATION ASSIGNED AD:

BRANCH CHIEF:

ENVIRHNMENTAL V

MOORE LTR CT MANAGER:

CENSING ASSISTANT:

PROJECT MANAGER:

LICENSING ASSISTANT:

INTERNALD TFMS SAFETY HEINE"iiN Bo HARLESS ISTRI BUTION PLANT SYSTEMS TEDESCO BENAROYA SITE SAFETY &

ENVIRON ANALYSIS DENTON & MULLER ENGINEERING IPPOLITO OPERATING REACTORS ENVIRO TECH ERNST BALLARD YO GBLOOD VAK 0

Z OCZY CHEGK AT I LT Zan TBERG EXTERNALDISTRIBUTION BAER BUTLER GRIME GAMMILL 2

SITE ANALYSIS VOLLMER BUNCH Jo COLLINS KREGER CONTROL NUMBER TIC REG IV J

HANCHETT 16 CYS ACRS SENT CAT GO NSIC 771820230

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Mr. Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U.

S. Nuclear Regulatory Commission Washington, D. C.

20555 Carolina Power 8 Light Company June 28 9g Regulatory Docket File'-,~

Igg~'HEARON HARRIS NUCLEAR POWER PLANT, UNITS 1, 2, 3 AND DOCKET NOS. 50-400) 50-401",

50-402 AND 50-403 g

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10CFR50.55a CODES AND STANDARDS

Dear Mr. Case:

Carolina Power

& Light Company (CP&L) hereby submits supplementary clarifications to Mr. M. A. McDuffie's letter of May 17, 1977, to Mr. B. C.

Rusche xequesting authorization from the Nuclear Regulatory Commission to use certain components in t:he Shearon Harris Nuclear Power Plant (SHNPP) which have been certified to earlier code addenda of ASME III than those prescribed by Sections 50.55a(c),

(d), and (e) of 10 CFR Part 50.

Such authorizations are permitted under 10CFR50.55a(a)(2)(i) upon demonstra-tion that compliance with the otherwise applicable requirements of Paragraphs (c), (d), and (e) would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.

These clarifications are related to Items 1 through 3 of the above-referenced letter and are as follows:

1.

Units 1 and 2 Reactor Vessels As indicated in our May 17, 1977, letter, these reactor vessels were originally purchased to the S71 edition of ASME IIIbut were upgraded to the W71 addendum.

Subsequently, upgrading to the S72 addendum was initiated even though fabrication and testing of the vessels had already begun to the W71 requirements.

The'rincipal difference between the S72 and W71 addenda, that is applicable to reactor vessels, is in the requirement to provide additional fracture toughness test-ing to further categorize materials used in fabrication.

These new (S72) requirements were anticipated and were fulfulled for the majority of material utilized in these vessels.

However, because sufficient additional test specimens from a limited number of heat lots of material were no longer available, the additional S72 testing requirements could not be fulfilled for five heat lots of welding electrode material which were used in the noncore regions of the vessels and the base material for one coolant nozzle.

Thus, for these limited heat lots of material, all xequirements of the W71 addendum were met (including fracture toughness testing requirements),

whereas for all other portions of the vessels, the S72 requirements were met as well.

(Fracture toughness properties of the coolant nozzle for which tests could not be conducted were estimated, using the method described in the NRC Standard Review Plan 5.3.2 and Branch Technical Position NTEB 5-2.

These estimated properties met or exceeded the requirements of the S72 addendum.)

Thus, although the majority of heat lots of material 336 Fayetteville Street

~ P. O. Box 1551

~ Raleigh, N. C. 27602 77l820230

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Case June 28, 197 7 comply with the S72 addendum, the W71 addendum was the latest code to which these reactor vessels could be certified.

An additional clarifica-tion to our letter referenced 'bove is that tes ts for the heat affected zones were not mandatory for these reactor vessels, and the welding procedure qualification tests are in compliance with the S 72 addendum.

2.

Units 1 and 2 Reactor Coolant Pumps The reactor coolant pumps were originally ordered to the S72 edition of the ASME code.

During the'anufacturing process, later codes were satis fied when practicable.

The pump casings of the pump were manufactured to the S 73 edition.

(Different pump components are fabricated at different plants at different times with final assembly at the site.

Due to the time frame of manufacture, some components will be normally supplied to a later code. )

The only part which was not upgraded to at least the W72 addendum was the thermal barrier heat exchanger, due to its advanced stage of design and manufacture at the time the W72 edition was published.

Thus, although the majority of components comply with W72 or later addenda, the S72 addendum was the'atest code to which these pumps could be certified.

3.

Units 1,

2,

3, and 4 Class I Control Valves The Safety Class I control valves had progressed too far in fabrication to permit upgrading to the'equirements of the W72 addendum because Paragraph NA-5200 of ASME IIIrequires the authorize'd inspector to review des ign details prior to fabrication.

Even though the control valves were in fact manufactured to the W72 addendum, the certification requirement of NA-5200 could not be retrofitted.

References to the W72 edition in our previous letter of May 1 7, 19 77, i. e., making changes 'n welding materials, NB-2420, inspection requirements,

NB-2510, and hydrotes ting NB6111. 1, were intended only to illustrate paragraphs where differences exis t between the S72 and W72 addenda.

We believe that this information is sufficient to enable the Nuclear Regulatory Commission to grant authorizat ion to us e the'quipment covered by this letter and the 'eferenced letter of May 17, 19 77 in the SHNPP.

Yours very truly, s

M. A. McDuffie Senior Vice President Engineering 6 Construction MAM/gsm Sworn to and subscribed before me this j

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28th day of June, 19)7.

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N ary Public My commission expires July 4, 1980,

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