IR 05000413/1988029

Insp Repts 50-413/88-29 & 50-414/88-29 on 880801-05. Violation Noted.Major Areas Inspected:Areas of Catawba Safety Review Group Activities & 10CFR21 Reportability Determinations

ML20154A624
Person / Time
Site:	Catawba
Issue date:	08/31/1988
From:	Belisle G, Mellen L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154A588	List: IR 05000413/1988029 ML20154A583 ML20154A614
References
REF-PT21-88 50-413-88-29, 50-414-88-29, NUDOCS 8809130034
Download: ML20154A624 (6)

Text

UNITED STATES go.mteo.s "'o,

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NUCLEAR REGULATORY COMMISSION g

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M,e 101 MARIETTA STREET.N.W.

• AT L ANT A. GEORGI A 3o323 v

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Report Nos.:

50-413/85-29 and 50-414/83-29 Licensee:

Duke Power Ccmpany 422 South Church Street Charlotte, NC 2S242 Docket Nos.:

50-413 and 50-414 License Nos.:

NPF-35, NPF-52 Facility Name:

Catawba 1 and 2 Inspection Conducted:

August 1-5, 1983 Inspector:

$. 'Nh Sbl f E n

.'. S. Mellen, Lead Inspector Date Signed L

Accompanying Personnel:

T. Cooper Approved by:

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G. A. BB i sle, Cn ef Oate' Signed Quality Prograns Section Operations Branch Division of Reactor Safety SUVRARY Scope:

This routine, unannouncad inspection was conducted in the areas of the Catawba Safety Revie,< Group (CSRG) activities and 10 CFR 21 reportability determinations.

The inspection was conducted on-site and at the Duke Power Company general offices.

Results:

One violation was identified related to the CSRG not meeting Technical Specification requirements.

Within the areas inspected, the following findings were identified:

Failure to cenply with Technical Specification requirements for

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CSRG functions, paragraph 2.a.

A weakness (IFI) relative to lack of adequate procedural guidance

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in 10 CFR 21 reportability determinations, paragraph 3.

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REPORT DETAILS i

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Persons Contacted

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Licensee Employees l

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L. Burba, Licensing Engineer

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• M. Cote', Complience Engineer

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C. Criminger, CSRG Member

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H. Edwards, Design Engineer

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J. Fraedrir.h, former CSRG Member

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• R. Futrell, Nuclear Safety Review Board (NSRB) Chairman

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j W. Green, former CSRG Member l

• R. Glover, Compliance Engineer

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R. Hall, NSRB Member l

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M. Hone, former CSRG Member

j R. Kirk, CSRG Member i

M. LaForrest, former CSRG Member

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"P. LeRoy, Licensing Engineer l

i J. Lines, CSRG Member t

B. r,cNeill, CSRG Member

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j 0. Murdoc, NSRB Member

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ll S. Rose, Technical System Manager I

N. Rutherford, Technical System Manager l

L. Schlise, former CSRG Member

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H. Smith, former CSRG Member i

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• R. White, CSRG Chairman l

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NRC Resident Inspectors j

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K. Van Doorn, Senior Resident Inspector i

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• M. Lesser Resident Inspector

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j Other licensee employees contacted included engineers, technicians, licensing personnel, compliance personnel, security force members, and office personnel.

• Attended exit Interview 2.

CSRG Assessment

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Technical Specifications Section 6.2.3, implements some requirements of

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NUREG 0737, Clarification of TMI Action Plan Requirements, and outlines j

requirerents for CSRG activities.

The inspectcrs reviewed CSRG actions

taken to fulfill each requirement and discussed them with selected CSRG

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members, former CSRG eemeers, and the NSRB and CSRG chairmen. The various areas covered by the requirements are outlined in the following four j

sections:

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Technical Specification (TS) 6.2.3.1:

FUNCTION The Catawba safety Review Group (CSRG) shall function to examine plant operating characteristics, NRC issuances, industry advisories.

REPORTABLE EVENTS, and other sources which may indicate areas for improving plant safety. The CSRG shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Director, Nuclear Safety Review Board.

The inspector identified that all requirements of TS 6.2.3.1 were not being fulfilled by the CSRG.

The functions that not were being performed by CSRG were being perferned by other Duke Power organi:a-tions.

While NRC issuances, industry advisories, and selected documents from other sources are reviewed by the various members of the CSRG, this review is not performed with the task of developing detailed recommendations for improving pl ar,t safety, as the TS Section requires. Discussions with the C5RG Chairman revealed that this function is coordinated by members of the Safety Assessment group in the General Office.

The members of the CSRG are occasionally requested to verify the effectiveness of the implemented improvements, but this is not done on a routine basis.

The CSRG is required to perform reviews of plant operating characteristics.

The in-plant reviews performed by the CSRG are considered to fulfill this requirement.

The inspectors reviewed the monthly reports from January 1933 through June 1938 and found that during this period only one in plant review was performed. The CSRG semonstrated that there are six in plant reviews presently being performed, in response to

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I findings at another site. The directions for this activity do not explicitly describe it as the medium under which the review of plant operating characteristics will be performed.

l Detailed procedural guidance for the performance of the assigned I

responsibilities is not available to the nembers of the CSRG. This reduces the consistency and continuity of the performance of these functions by the various members, past and present, of the CSRG. The CSRG Chairman agreed with this assessment and stated that improvetents in this area would be developed.

The failure of the CSRG to perform all of it's required duties related to all functions outlined in the site Technical Specifications is considered violation 413,414/88-29-01, b.

TS 6.2,3.2:

CCMPOSITICN The CSRG shall be corposed of a chairman and at least four dedicated, full-time qualified individuals located onsite.

The CSRG is co posed of the chairman and six members of the plant staff who rotate into this position for a twelve month period.

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I The inspectors reviewed the training program for the CSRG members and

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while it appears to be a detailed program, the time that can be l

dedicated to the training program is minimal, caused by the short

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time that the various members are present in the group and the large

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work load that must be ccmpleted by this group, The short tenure

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period for the rotating members of the CSRG alto results in the CSRG l

cnairman spending a large amount of his time in the training of the i

new members, as they rotate into the assignments in the CSRG.

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TS 6.2.3.3:

RESPONSIBILITIES

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The CSRG shall be responsible for maintaining surveillance of plant activities to provide independent verification that these activities

are performed correctly and that human errors are reduced as much as

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practical,

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The inspectors reviewed the list of completed tasks for the last six f

months and noted that only one in plant review was included on the i'

list. The CSRG is aware of this, in light of the findings at another site, and were able to demenstrate an increase in the number of i

in-plant reviews currently being performed.

The CSRG has assigned a member to review the number of events caused l

by human error and develop recommendations to reduce the number of l

future events, This is a recently developed project, however, and no i

results were demonstrable, d.

TS 6.2.3.4:

RECORDS

Records of activities performed by the CSRG shall b prepared, i

maintained, and forwarded each calendar month to the Director, l

Nuclear Safety Review Board,

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The inspectors reviewed approximately 15 of the most recent investiga-

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tion reports generated by the CSRG and noted that they were detailed

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and complete, Interviews were conducted with the majority of the i

present CSRG rerbers and with several past CSRG meebers and reviewed several draf t reports and determined that the plant staff did not exert undue influence on the CSRG on the content of the report, p

Changes to the draft reports generated by plant staff were J

clarifications or corrections of incorrect information and were not f

used to change the content of the report,

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10 CFR 21 Determinations I

The inspectors interviewed members of Cattwba Site Compliance, General Office Nuclear Licensing, and General Office Design Engineering to determine the ef fectiveness, process, and philosophy in reporting defects under 10 CFR 21, e

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An example of the process used in a 10 CFR 21 determination was reviewed, I

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involving the shuttle valves used on the emergency ciesel generators (EDGs).

General Office Design Engineering performed an evaluation on

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December 7, 1987, which determined that the continual problems with the r

EDGs caused by the use of these valves was reportable under 10 CFR 21. On

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d December 23, 1987, General Of fice Nuclear Licensing issued a memorandum l

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stating that the initial determination from Design Engineering was l

incorrect and that Licensing's initial determination was that the item was

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l not reportable under 10 CFR 21 requirements.

On March 9, 1998, the l

Licensing group issued a final determination that the item was not

i reportable, using the reasoning that the misapplication of the shuttle

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valve did not constitute a defect and that it resulted in a reliability

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j concern and not an operability concern, even though the use of the shuttle j

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valve resulted in intermittent trips of the EDGs. This determination was

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I made based on operability determinations performed by the site Compliance

i group.

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j The inspectors reviewed the Design Engineering procedure and the Station I

q Directive whien provides the directions for performing an evaluation for l

l 10 CFR 21 reporting.

Design Engineering Procedure, PR-203, Problem i

Investigation Reports, Revision 4 provides detailed directions for the j

j Design Engineer who is performing 10 CFR 21 evaluations.

Station Directive j

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2.8.1, Problem Investigation Process and Regulatory Reporting, Revision 9

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does not provide detailed directions for performing evaluations.

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j inspectors noted that the procedures do not assign the responsibility of

10 CFR 21 determinations and reporting to any one group other than Nuclear i

Productton.

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This lack of detailed directions for performing 10 CFR 21 evaluations in j

Nuclear Production procedures and the lack of procedural guidance for tha r

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responsibility of performing these evaluations is considered a weakness and is identified as Inspector Followup Item 413,414/SS-29-02.

f Within this area, no violations or deviations were identified.

Exit Interview (30703)

l The inspection scope and findings were sum ari:ed on August 5, 1983, with

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those persons indicated in paragraph I above.

The inspectors described l'

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the areas inspected and discussed the inspection findings listed below.

The licensee did not identify as proprietary any of the materials provided i

to or reviewed by the inspectors during the inspection.

Dissenting I

con +nts were not received f rom the licensee.

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Item Number Description and_ Reference 413,414/88-29-01 Violation - Failure to comply with Technical Specification requirements for CRGR functions, paragraph 2.a.

1.ack of adequate 413,414/88-29-02 IFI

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procedure guidance in 10 CFR 21 reportability detenninations, paragraph 3.

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