ML19319A908

Forwards Request for Addl Info Necessary to Complete Review of Inservice Insp Program

ML19319A908
Person / Time
Site:	Oconee
Issue date:	02/26/1977
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	Goller K Office of Nuclear Reactor Regulation
References
NUDOCS 7912300136
Download: ML19319A908 (6)

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t MEMORANDUM FOR:

K. Coller, Assistant Director, Operating Reactors, DOR D. Eisenhut, Assistant Director, Operational Technology, DOR i

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e SUPJECT:

DUKE POWER CO. - INCORPORATION OF PROVISIONS OF I

-10 CFR 50.55a (g) " INSERVICE INSPECTION REQUIRFlIENTS"

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INTO TECHNICAL SPECIFICATIONS (TAC 6314) t Plant Name: Oconee - Unit 1 Licensing Stage: Operating plant Docleet Number:

50-269 Document Reviewed: Letter and attachments 1 thru 4 dated October 1, 1976 from William O. Parker, Jr., Duke Power Company to f

B. C. Rusche, Director of NRR responding to the February 16, 1976 Federal Register Notice concerning l

10 CFR 50.55a Branch and Project Manager: 02B #1, D. Neighbors Requested Completion Date: May 1, 1977 - r F

Operating Reactor Branches Involved:. Engineering Branch, Reactor Safety Branch, Plant Systems Branch Description of Request: Review of the proposed inservice inspectionLl--

proeram Review Status: Request for additional information

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The Engineering Branch, Division of Operating Reactors has coordinated and

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reviewed the docunents in the above reference for conformance with the re-quirenents of 10 CFR 30.55a, paragraph (g) "Inse..vice Inspection Require-cents." We find that the information submitted is inadequate and satisfactory renponse to our enclosed request for additional information is required t

before we can complete our evaluation. To provide the requested additional information we reco==end that the licensee use the two letters, dated April 26, 1976 and November 30, 1976 from the NRC to Duke Power Company as guidance.

We also suggest that separate Technical Specifications be established for i

cach unit because of the difference in the initial startup date for the

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facility commercial operation. The Reactor Safety Branch and Plant Systems Branch have collectively reviewed the inservice testing requirements for j

pumps and valves and submitted detailed comments which are incorporated in this evaluation. The Code of Federal Regulation 10 CFR 50.55a(g) requires By 1

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that inservice testing of pumps and valves be performed in accordance I

with the requirements of the ASME Code Section II.

This code specifies j.

that inservice testing of pumps and valves shall be conducted on all

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Code class 1, 2, 3 pumps and valves. This review considered only safety related pungs and valves whose operation is relied upon to shut down' the plant or mitigate the consequences of an accident, since in our judgment, this is the intent of 10 CFR 50.55a(s).

D. G. Eisenhut, Assistant Director for Operational Technology Division of Operating Reactors Enclosure As stated i

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DlVlSION OF OPERATING REACTORS

' Request forfAdditional Information Inservice Inspection' Requirements

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' Attachment'l Comments-l1/. Pr'ov'ide additional 16 formation using the letter-from A. Schwencer,

Cliief,. ORB #1 to W. ~ 0. Parker, Jr., Vice-President, Steam Production

, Duke Power Co.', dated November 30.-1976, as guidance, to justify.the request for relief from. the inservice examination requirements sp :ci-

fied in the 1974 edition of-the Section XI Code through.the summer

1975 addenda for Code. Class 1~, 2 and 3. components.

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12.. Clarify whether tha required inservice examination will be aitnessed or verified by a-third party. inspector even though South Carolina is

.not an ASME. Code' state.

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, Comments 4

~1. - In order to evaluate. the.0conee 1 pump testing. program the following

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information is requested:

. a) A list' identifying each pump - to be tested by system and application.

b) ; -The test parameters that will.be measured for each. pump.

c)' The : test intervals, i.e.. monthly during operation or only during cold shutdown.

2.

When certain parameters are not going to be tested and relief is re-

quested, provide the following :btf ormation:

. a). Specifically identify the ASME Code requirement that-has been determined to be impractical for the pump.

Lb) 7 Provide information to support the determination that the require-

> ment in' (a) is impractical.

'c) -Specify;the. inservice testing that will be performed in lieu of the i

'ASME Code Section XI requirements' that have been determined to be l impractical or provide-the basis for operation of this pump without

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this ISI.

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t. 1 Provide-tne. schedule'for impicmentation of the procedure (s) in d) i (c) above..

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' Comments

. Provide code class' designations for all valves tested.

1.

2. 'On your tab 1'e identify each valve in ASME Section XI Cat. A that will be. leak tested during refueling outages.

For check

3. -Provide the test intervals for all valves to be tested.

valves,1 identify those that will be exercised only during cold

- s' hutdown.

4. 'Where relief has been requested from certain requirements of the code, specify the inservice testing that will be performed in lieu of the ASME Code Section XI requirements that are impractical or provide the bases for operation of this valve vithout this ISI. Also provide

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the schedule for implementation of this test,ing.

Provide simplified piping diagrams of systems which must function to 5.

i safely shutdown the plant or mitigate the consequences of ca accident.

Active components on the above systems which' must change position Also, provide a narrative description of the should be' identified.

- valve line-ups required of the systems identified above each of. their safety functions.

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'6.-

In addition'to the above comments we have found that in some cases, valves important to plant safety were omitted.

Comments on these valves will be.made by drawing number.

a) Dkg PO-100-A Only 2 of the three pressurizer relief valves are

1'isted. " Valve # 1RC-66 should be included.

~ b). Dwg' PO-101A-1. Valves listed on this. drawing are containment iso-

' Check valve #HP-194 should be included if possible

-lation valves.

to the' test program.

.c) Dwg_PO-102A Valves BS-3 and BS-4, Power opeiated valves on the

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Lsuction lines to the reactor building spray pumps are omitted. They

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should be included or. justification provided for not including,them.

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• d) Dwg PO-103A-1,2,3 - Check valves BS-14 and BS-19 are proposed to be tested every 5 years. A source of instrument air exists

.(according to the drawing) for s

.f nozzle testing. The licensee should consider using the instrument air to test the check valves on a more frequent schedule, e) Dwg PO-122A Only one of sixteen main steam safety reliefs is listed. Provide justification for not including the others, f) Dwg PO-127-B - We cannot locate N2 isolation valves 1N-91 thru 1N-94.

These may be mis-numbered on submittal. They should be N-105, 106, and 107.

Confirm and correct technical specification as required.

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_. - Comments 1.

Because of difference in the date of start of facility commercial opera-tion for Oconec Units 1, 2 and 3, we recommend that separate Technical Specifications be established for each unit.

2.

The' language in the Technical Specifications 4.04 and 4.2.1 is not acceptable. The sample technical specification language recommended in the letter from R. A. Purple, Chief, Operating Reactors Branch #1, NRC to Duke Power Company, dated April 26, 1976 should be used, i.e.

4.2.1 - Inservice inspection of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section X1 of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

Inservice testing of ASME Code Class 1, 2 and 3 pumps and 4.0.4 valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR~50, Section 50.55a(g), except where specific written relief has-been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

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- 3. ' Technical Specifications 4.2.6 regarding the pump flywheel inservice inspection program is not acceptable. We require that a surface

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examination of all, exposed. surfaces and a' complete volumetric exami-nation, during the plant shutdown coinciding with the inservice in-spection schsdule as required by -the Section XI Code be performed at 1

approximately ten-year. intervals, in addition to the in-place volu-

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metric examination of the bore and keyway of each reactor coolant pump' flywhel at approximately three-year intervals as specif'ied' in f

Technical. Specifications 4.2.6.. Removal of the flywheel is not required.-

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