ML20214V950

Insp Repts 50-369/87-11 & 50-370/87-11 on 870316-20. Violations Noted:Failure to Perform Trending of Reactor Trip Breaker Response Time Test Data

ML20214V950
Person / Time
Site:	Mcguire, McGuire
Issue date:	04/22/1987
From:	Bruske S, Conlon T, Foster L, Merriweather N, Reeves W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20214V919	List: ML20214V915 ML20214V928 ML20214V950
References
50-369-87-11, 50-370-87-11, GL-83-28, IEB-83-04, IEB-83-08, IEB-83-4, IEB-83-8, IEB-85-002, IEB-85-2, IEIN-85-093, IEIN-85-098, IEIN-85-93, IEIN-85-98, IEIN-86-062, IEIN-86-62, IEIN-87-005, IEIN-87-5, NUDOCS 8706150020
Download: ML20214V950 (20)

See also: IR 05000369/1987011

Text

UNITED STATES

ff L9H%gg NUCLEAR REIULATORY COMMISSION

t v( q REGION It

3 g 101 MARIETTA STREET, N.W , SulTE 2900

o * ATLANTA, GEORGIA 30323

s .... /

Report Nos.: 50-369/87-11 and 50-370/87-11

Licensee: Duke Power Company

422 South Church Street

Charlotte, NC 28242

Docket Nos.: 50-369 and 50-370 License Nos.: NPF-9 and NPF-17

Facility Name: McGuire 1 and 2

Inspection Conducted: March 16-20,1987

Inspectors: N

N. Merriweather, Team Leader' b cla - 77

Date Signed

byt -Y

L. E. Foster, Lawrence Livermore National Lab (LLL)

bN- Y]

Date Signed

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S. Bruske, LLL

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Date Signed

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W. Reeves, LLL

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Date Signed

Approved by W4 //- ZZ 4

T. E. Conlon, Section Chief Date Signed

Engineering Branch

Division of Reactor Safety

( SUMMARY

i Scope: This special announced inspection was performed to assess the

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licensee's compliance with their response to Generic Letter 83-28, " Required

Actions Based on Generic Implications of Salem Anticipated Transient without

l Scram (ATWS) Events." Areas inspected included post-trip review, equipment

classification, vendor interface (including followup on NRC notices and

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bulletins), post maintenance testing, and reactor trip system (RTS)

reliability.

Results: One violation was identified - Failure to Perform Trending of Reactor

Trip Breaker Response Time Test Data.

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REPORT DETAILS

1.- Persons Contacted

Licensee Employees

• N. Atherton, Associate Chemist

• R. G. Bledsoe, Sub-Station Equipment Supervisor

• D. Boies, Associate Engineer .

• T. Cline, Instrumentation and Electrical (IAE) Supervisor

• R. Cole, Nuclear Safety Assurance-Production Specialist III

• G. A. Copp, Maintenance Planning Representative

• J. Day, Associate Engineer

• J. Effinger, Quality Assurance Audit Supervisor, General Office (GO)

• E. Faggart, Shift Engineer (Acting Station Manager)

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• J. Ferguson, Document Control Supervisor-G. O.

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• J. L. Freeze, IAE Supervisor

B. Jones, Preventive Maintenance Coordinator

• L. Kunka, Nuclear Production Department Engineer

• N. McCraw, Compliance Enginecr

B. Johnson, Test Engineer

• R. Pierce, IAE Support Engineer

• F. Pope, Administration Supervisor

• F. Siurua, IAE Engineer, G0

• J. E. Snyder, Performance Engineer

• K. Wilkinson, Transmission Support Engineer

M. D. Blackwell, Quality Assurance Supervisor

D. E. Colson, Design Engineer

R. Hardin, Supervisor, Design Engineering

D. M. Jenkins, Design Engineer

D. F. Markle, Quality Assurance

T. E. Mooney, Maintenance Manager

A. W. Roy, Quality Assurance Supervisor

P. T. Vu, Nuclear Production Engineer

P. T. Ward, Vendor Document Supervisor

Other licensee employees contacted included engineers, technicians,

operators, mechanics, security force members, and office personnel.

NRC Resident Inspectors

W. Orders, Senior Resident Inspector

S. Guenther, Resident Inspector

• Attended exit interview

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2. Exit Interview

The inspection scope and findings were summarized on March 20, 1987, with

those persons indicated in paragraph 1 above. The inspector described the

areas- inspected and discussed in detail the inspection findings. No

dissenting comments were received from the licensee. Subsequent to the

- exit meeting, a telephone conversation was held on March 23, 1987, between

N. Merriwether and R. L. Gill (DPC) to discuss upgrading the unresolved

item identified in the exit meeting to the violation identified below.

Violation 50-370/87-11-01, Failure to Perform Trending of Reactor Trip

Breaker Response Time Test Data, Paragraph 9,

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

to or reviewed by the inspectors during this inspection.

3. Licensee Action on Previous Enforcement Matters

This subject was not addressed in the inspection.

4. Unresolved Items

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Unresolved items were not identified during this inspection.

5. Background

On February 25, 1983, both of the scram circuit breakers at Unit 1 of the

Salem Nuclear Power Plant failed to open upon an automatic reactor trip

signal from the reactor protection system. This incident was terminated

manually by the operator about 30 seconds after the initiation of the

automatic trip signal. The failure of the circuit breakers 'was later-

, determined to be related to the sticking of the undervoltage trip

attachment. Prior to this incident on February 22, 1983, at Unit 1 of the

Salem Nuclear Power Plant, an automatic trip signal was generated based on

steam generator low-low level during plant startup. In this case, the

reactor was tripped manually by the operator almost coincidentally with

the automatic trip. Consequently, the failure of the undervoltage trip

attachment was not immediately detected.

As a result of the problems identified with circuit breakers at Salem and

at other plants, NRC issued Generic Letter (GL) 83-28, " Required Actions

. Based on Generic Implications of Salem ATWS Events," dated July 8, 1983.

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This letter required licensees of operating plants to respond to

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intermediate-term actions to ensure reliability of the Reactor Trip System

(RTS). Actions required to be performed by the licensees included

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development of programs to provide for post-trip review, classification of

( safety-related equipment, vendor interface and vendor manual control,

post-maintenance testing, and RTS reliability improvements.

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LThe licensee, Duke Power Company, responded to GL 83-28 in letters, dated

November 4, 1983, May 24, 1985, August 23,1985, April 14, 1986 and

October 1, 1986. In these responses, Duke Power described their

compliance to the NRC positions described in the generic letter.

This inspection was performed to verify compliance to the licensees

responses ' associated with post trip review, equipment classification,

. vendor interface and post maintenance testing for the McGuire Nuclear

Plant. The results of the inspection are discussed in the paragraphs that

follow.

6. Post Trip Review

The applicant was requested, in GL 83-28, to describe their program,

procedures 'and data collection capabilities in order to assure that the

causes for unscheduled reactor shutdowns, as well as the responses of

safety-related equipment, are fully understood prior to plant restart.

The applicant's response to GL 83-28 provided a description of the program

and procedures pertinent to performing post-trip reviews. The inspector

reviewed their response, appropriate plant procedures, and interviewed key

licensee personnel to assess the adequacy of the applicant's program for

post-trip reviews.

The inspection was formatted to verify that a post-trip review program has

been implemented and meets the following attributes:

Procedures and equipment exist to cover post-trip review.

Safety assessments of the reactor trip are clearly delineated as part

of the post-trip review.

Post-trip review procedures are reviewed periodically by an onsite

safety review committee such as the Plaut Operations Review Committee

(PORC) and upgraded in any areas that have. been identified as

deficient.

Plant personnel preparing and/or reviewing post-trip documentation

receive initial training and refresher training in post-trip review

procedures.

Responsibilities and authorities of plant personnel who will perform

the review and analysis of these events are clearly defined.

Criteria for determining the acceptability of authorized restart have

been established.

Criteria for comparing plant information with known or essential

plant behavior have been established.

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Guidelines 'are established for preservation of evidence of reactor

trips.

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The licensee's program, procedures and equipment provided to address each

of these items is discussed as follows:

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a. Procedures and equipment exist to cover post trip reviews.

l McGuire procedure PT/0/A/4700/45, " Transient / Reactor Trip

Investigation" is the primary procedure which addresses post-trip

. reviews. This procedure provides a detailed method to evaluate .

l overall plant response and -important system response following a

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reactor trip. . As mandated by Station Directive (SD) SD 3.1.10,

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" Investigation of Reactor Trips" the reactor trip investigation -

. procedure must be completed following all' reactor trips with the

j following exceptions:

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(1) Control rod drop test at power levels below 5%.

(2) Manual trips at approximately zero power as part of a normal

controlled shutdown.

(3) Planned trips for testing and mai ;enance from approximately

zero power.

SD 3.1.10 does state, however, that if any unexpected protective  ;

function is actuated or other related equipment malfunctions occur

during any of the above situations, then, the post trip investigation

4 .will be performed.

• In addition to these procedures, two-procedures, which govern reactor

restart, require signoffs by the " Reactor Group" prior.to return to

power. These signoffs document that'the trip has been evaluated and

i a restart is recommended. These procedures are OP/1/A/6100/01

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" Controlling Procedure for Unit Startup" and OP/1/A/6100/05, " Unit

Fast Recovery." (NOTE

The first number after OP designates the

unit. Therefore, these are Unit 1 procedures. Unit 2 procedures

i have the same signoff requirements). The inspector reviewed all

procedures related to post trip reviews and verified that all

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licensee commitments contained in their responses to the GL 83-28

relating to procedures are being carried out.

In addition to the review of procedures, a detailed review was

i performed 'of a completed trip investigation report. The post-trip

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review performed following the Unit 2 trip of November 28, 1986, was

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randomly selected for review.

This detailed review verified that the procedure provided a

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systematic evaluation of plant performance following a plant trip and

any abnormal system response was documented and recommended

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corrective action was specified. This review also verified that the

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installed plant equipment utilized to capture data for future

evaluation includes all important data points and the data is

available in clear and usable format. This equipment includes an

alarm typer, events recorder and the plant sequence monitor.

b. Safety assessments of the-reactor trip are clearly delineated as part

of the post-trip review. The safety assessments of the reactor trip

are specified as part of the SD 3.1.10 and PT/0/A/4700/45.

c. Post-Trip review procedures are reviewed periodically by an onsite

safety review committee and upgraded in any areas that have been

identified as deficient.

SD 4.2.1 covers periodic . review and upgrade of all station

procedures. SD 4.2.1 specifies a two year review and upgrade of

safety-related procedures such as the transient / reactor trip

procedyre.

d. Plant personnel preparing and/or reviewing post-trip documentation

receive initial training and refresher training in post-trip review

procedures . ~

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Personneljaalification requirements and requalification requirements

are speci fied in SD 3.1.10. This SD also requires that a current

list of qualified reviewers be maintained and only those on the list

perform reviews.

e. Responsibilities and authorities of plant personnel who will perform

the review and analysis of these events are clearly defined.

Personnel responsibilities and authority are clearly delineated in

SD 3.1.10.

f. Criteria for determining the acceptability of authorized restart have

been established.

This criteria is specified in SD 3.1.10.

g. Criteria for comparing plant information with known or essential

plant behavior have been established.

Guidelines for evaluation of post-trip data are contained in both

SD 3.1.10 and PT/0/A/4700/45.

h. Guidelines are established for preservation of evidence of reactor

trips.

SD 3.1.10 specifies that completed transient / reactor trip reports are

to be retained for the life of the plant.

Within the area examined, no violations or deviations were identified.

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7. Equipment Classification

GL 83-28 Items 2.1 and 2.2 dealt with equipment classification for reactor

trip system . components - and all other safety-related components,

respectively. GL 83-28 Item 2.1 required that licensees confirm that all

components whose ' functioning is required to trip the reactor are

identified as safety-related on documents, procedures, and information

handling systems used in the' plant. Item 2.2 required licensees to

describe their program for ensuring that all other components of

safety-related systems necessary for accomplishing required safety -

functions are also 1dentified as safety-related on documents, procedures,

and infonnation handling' systems used in the plant. In a letter, dated

November 4,1983. Duke Power Company (DPC) responded to GL 83-28 Item 2.1

for the McGuire Nuclear Station. -In this response, DPC stated that "all

components of the Reactor Trip System whose functioning is required to

trip the reactor are identified as safety-related on documents,

procedures, and information handling systems used in the plant to control

safety-related ' activities." In the same response, the licensee also

described the controls used for all other safety-related components. The

licensee indicated that the McGuire Nuclear Station Quality Standards

Manual for Structures, Systems, and Components provides the mechanism for

the determination of whether or not a given station structure, system, or

component is safety-related.

The inspector confirmed that the McGuire Nuclear Station Quality Standards

Manual for. Structures, Systems, and Components (QSMSSC) provides the

mechanism for -determining whether or not a given station structure,

system, or component is nuclear safety-related or requires that certain

quality standards be maintained. This manual identifies four QA condition

codes with ~ QA condition 1 identifying safety-related items. It provides

the criteria for determining the QA conditions of structure, systems, and

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components, ' and it- identifies tables of plant structure, systems and

L components 'with there defined QA classifications. It also provides

evaluation checklists for determining the safety classification of those

i items where conflicts exists between QA conditions shown in the manual

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list and design documents or for items not identified in the QSMSSC or on

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design drawings.The QSMSSC Revision 0 was issued on April 1, 1984. This-

l manual has not been revised or updated since that date. Discussions with

l personnel in the General Office (Licensing) revealed that two requests,

l dated January 8, 1985 and March 26, 1986 have been received in Licensing

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proposing revisions to the manual. As of the date of this inspection, no

, revisions have been initiated. The inspector questioned the licensee

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about the controls for revising and updating the QSMSSC. It appears that

the' manual includes instructions on how the manual can be revised but it

j does not specify a time frame for when the manual will be reviewed. -In

addition, the inspector was informed that station modifications do not

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specify the QSMSSC as a document that must be revised for closeout of

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modification packages. The inspector discussed the above concern

regarding updating the QSMSSC with the licensee in the exit meeting and

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they committed to have the manual revised. In the interim, the manual is

considered adequate for classifying. items because it only provides -

general criteria. In most cases, classifications are provided by design.

drawings. As: stated above in cases of conflicts between the manual and

drawing, the evaluations checklists can be used to classify items using

the criteria provided in the procedure.

SD 3.3.0, " Determination of QA Condition 1, 2, 3 or 4 Structures, Systems

and Components," further defines station personnel responsibilities on the

use of the QSMSSC for determining the QA condition of given Nuclear-

Station Structures, Systems, and Components. The SD states that an

evaluation checklist can be prepared by any member of the station

organization. However, -it must be reviewed by a qualified reviewer,

(SRC member). Qualified reviewers are addressed in SD 4.2.1, " Handling of

Station Procedures." This Directive defines the qualification and

certification of those plant staff authorized to review station

procedures.

The inspector discussed with the licensee the training, qualifications and

certification of qualified reviewers at McGuire Nuclear Station.

Discussions . revealed that all training classes of qualified reviewers

received the same certification exams. (This also included personnel-who

required a retest). The inspector questioned the adequacy of the

licensee's methods for safeguarding tests considering that several classes

have attended the certification program. The inspector reviewed results of

candidates that had failed the exam the first time and also, the results

on the retest. The records did not indicate a significant improvement

from the initial test results and also, there were examples of candidates

that also failed the retest. However, even though the exams are kept

under lock and key this will not ensure that future tests will be

safeguarded. The licensee was informed of this concern in the exit. They

indicated that a pool of questions will be developed to safeguard future

qualified reviewer exams.

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In order to verify implementation of the QSMSSC, the inspector interviewed

! responsible licensee personnel and reviewed appropriate documents to

i determine if randomly selected reactor protection system items were

properly being classified as safety-related on documents, procedures, and

, information handling systems used at the plant. Personnel in both

Mechanical, and Instrumentation and Electrical Planning Sections were very

knowledgeable of the procedures and demonstrated their abilities to

determine QA conditions by use of the QSMSSC, drawings and other

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information documents (i.e., flow diagrams, equipment lists, valve lists

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and bill of materials). Licensee personnel were able to demonstrate use

of the QSMSSC in classifying the following safety-related components

! a. Reactor Trip Breakers

i b. Main Control Board Reactor Trip Switches

i c. Latches for the Control Rod Drive Mechanism

i d. Manual Valve 1 MWG MV 0082

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' Item a is identified in the QSMSSC as QA Conditicn 1. Items b thru d ,

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required the .use of drawings to determine the quality classification.

. Item c required extensive research of the technical manual and drawings-to

conclusively determine the quality classification.

Previously issued purchase orders for Reactor Trip System Components (such

as. Reactor Trip Breakers, Shunt Trip Attachments, Undervoltage Trip

Attachments and Manual Reactor Trip Switches) were found to be properly

classified as QA-Condition 1 (safety-related).

l: SD 4.2.1 requires all QA Condition 1 procedures to be identified with an A

in the procedure number. All station procedures are required to meet-the

intent of.SD 4.2.1. This ensures that work on safety-related equipment is

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performed using safety-related (QA Condition 1) procedures. Drawings for

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safety-related structures, systems and components are stamped QA

. Condition 1. Mechanical equipment and piping systems identified as QA

Condition 1 are given classification codes A, B and C. These mechanical

boundaries are identified on drawings.

The inspector verified by review of records and interviews that selected

plant personnel have received training on the use of SD 3.3.0,

" Determination' of QA' Condition 1, 2, 3 or 4 Structures, Systems and

Components.

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Based on the above, the inspector concluded that the licensee's methods

for classifying items appear to be acceptable, however, enhancements could ,

be made which would minimize the number of evaluations that must be

performed. Similar programs and initiatives are on-going by other

utilities to develop detailed lists of safety-related structures, systems

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and components. This is considered very important because the licensee's

procedures only require those maintenance work orders and requisitions

which are classified as QA Condition 1 to be reviewed by QA.

4 The licensee has also implemented the QSMSSC at Oconee and Catawba.

Weaknesses identified during inspection at these sites are discussed in

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Inspection Reports 50-269, 270 and 287/85-21 and 50-413/86-29 and

i 414/86-29, respectively.

Within the areas examined, no violations or deviations were identified.

8. Vendor Interface and Manual Control

The inspector reviewed the licensee's response, dated November 4,1983,

which described in general their program for vendor interface and control

of vendor technical information. The licensee's implementation of his

responses were examined to verify that the program had been developed in

detail and if the program was being used at the corporate office and at

McGuire Nuclear Station. Results of this inspection revealed that the

licensee had developed a detailed program and that the program was being ,

implemented. l

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The licensee's response stated that the majority of the NSSS components

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were supplied

in place that by Westinghouse

ensures that al (T)technicaland that theyconcerning

information have a continuing

the program

Reactor Trip System and other safety-related equipment is complete,

current, and is controlled throughout the plant life. The program ensures

that the technical information is reviewed and distributed to cognizant

plant personnel for incorporation into applicable plant maintenance and

operating procedures. The program also acknowledges receipt of the

information by written confirmation.

The licensee is an active member of the Nuclear Utility Task Action

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Committee (NUTAC) which has developed the Vendor Equipment Technical

Information Program (VETIP). The licensee had developed procedures to

ensure that the VETIP program is being implemented. This program utilizes

present vendor contacts, the Institute of Nuclear Power Operations (INP0)

Programs, the Nuclear Plant Reliability Data System (NPRDS) which is

managed by INP0, and other industry and government publications such as

Vendor Information Letters and IE Bulletins.

Responsibility for the receipt, distribution, review, update, issuance,

maintenance, and overall control of technical information has been

assigned to individual coordinators within the Nuclear Safety Assurance

Section and the General Office Licensing Section. (Directives have been

developed to specifically assign responsibility). Vendor instruction

manuals and drawings are controlled by the Design Engineering Department

and are -maintained and distributed by the General Services Division,

Drawing Control Group. Technical information is distributed to McGuire

Nuclear Station by the Design Engineering Department, General Services

Division. Control of vendor manuals at the McGuire Station is the

responsibility of the Document Control and Distribution Section. The

licensee stated that they review and evaluate vendor recomendations and

that these items are tracked from receipt to final disposition. Response

to Item 3.1.2 of GL 83-28 listed nineteen W Technical Bulletins and eight

}{ Data Letters that the licensee had compfited action on. The licensee

receives an annual letter from W which lists all of the Technical

Bulletins and Data Letters transmTtted by W. The licensee stated that all

documents relating to the Reactor Trip Bre'akers have been processed. The

inspector reviewed the latest W NSID Technical Bulletin Index, dated

June 18, 1986, and reviewed act'lon taken by the licensee on NSID 84-08,

84-02-R1, 83-11, 84-01, 85-08, 85-11, 85-18, 86-07, 83-03 and NSID 85-17.

The above described activities are detailed in licensee procedures,

directives, and manuals which were reviewed. Quality Assurance activities

associated with the auditing of these vendor manuals and technical

document controls were also examined and found satisfactory.

Procedures and directives associated with the receipt, review,

distribution, tracking, documentation, usage, and overall control of

vendor manuals, vendor information letters, vendor technical bulletins,

operating experiences, INP0 documents, NRC generated documents, and other

industry data were reviewed and are listed below:

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a. Operating Experience Program (OEP) - Generic Keys for OEP Event Data

Form and NSIDB, (0EP Data Base) - Draft Copy, provides instructions

for filling out the Event Data Form by the Station's Safety Review

Group.

b. Directive No. 4.8.1(s) .0perating Experience Program Description

specifies how operating experience information including vendor

information, INP0 information, NRC documents, and other technical

data will be handled to ensure that resolution has been accomplished.

c. Procedure IMF-1, Receipt and Distribution of Operating Experience

Information provides instructions for initial screening, proper

receipt and distribution of information. Licensee stated that this

procedure may be revised due to development of new directive and

computer capabilities.

d. Station Directive 2.0.7, Review of Operating Experience, Revision 4,

provided a method for reviewing NSAC/INP0 SEE-IN and industry

reports. This directive will probably be revised as result of new

directives.

e. Administrative Policy Manual Section 2.1, Document Control specifies

how drawings and vendor documents will be received, identified,

distributed, approved, issued and filed.

f. Station Directive 2.1.1, Control of Vendor Manuals, Revision 3,

Attachment No. 9, depicts how controlled copies of vendor manuals

will be maintained in the master file, operations, I&C and the

mechanical maintenance departments. Document Control Group has an

index of all manuals, makes revisions to manuals, and dispose of

superseded manuals. An audit is performed every 12 months on Working

Copy Controlled Manuals that have been issued outside the Controlled

Satellite / Master File Areas.

g. S. D. 2.1.1, Control of Master File Documents, Revision 27, provides

a program for control of storage, retention, disposal, monitoring,

routing and distribution of Master File Documents. This directive

controls access to the Master File area. Vendor Manual Inserts and

approved by Design Engineering are incorporated into the manuals by

the individuals they are issued to; however, an acknowledgement

receipt is required.

h. Procedure DCP-151, Vendor Manual Check-Out, Revision 0, provides a

method to properly control the check-out and return of vendor

manuals.

1. Procedure DCP-152, Vendor Manual Transmittal Processing, Revision 2,

assures that all vendor annuals are properly approved and completed

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before distribution.

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j. Procedure DCP-153, Audit of Stations's Vendor Manuals, Revision 0,

- provides instructions for auditing vendor manuals which are

maintained at the station.

k. Procedure DCP-154, Processing Audit Discrepancies, Revision 1, gives

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instructions for properly processing and completing discrepancies

. identified within the Vendor Manual Audit.

1. Procedure DCP-156, Retransmitting Insertion Control Forms, Revision

l 1, provides instructions for completing and retransmitting all

incomplete Insertion Control Fonns after the manual has been

incomplete: Insertion Control Forms after the manual has been audited

and did not contain any outstanding discrepancies.

m. Design Engineering Department Manual,Section I.4.9, Vendor

Documents, revised 01-31-87 outlines the handling and checkout

procedures for all vendor documents maintained by the General

Services Division, Drawing Control Group.

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The review of the above procedures, inspection of document control areas,

examination of vendor manuals, review of insert sheets, audits,

transmittal forms, acknowledgement letters and interviews with personnel

confirmed that the licensee has developed an adequate vendor interface and-

4 vendor manual control program and that the program is being implemented.

4 The inspector was advised that all vendor manuals in use at McGuire -

n Nuclear Station have undergone a page by page audit to ensure that the

manuals are correct, current and have all documentation.

I The inspector reviewed the licensee's handling of various IE Bulletins and

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Notices which relates to equipment in the reactor trip system. The

i notices and bulletins are being handling thru the Operating Experience

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Program as part of the vendor interface program. The documents reviewed

, were IEBs 83-04, 83-08 and 85-02 and IENs 85-98, 85-93, 86-62 and 87-05.

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The licensee's records provide evidence that the bulletins were received,

reviewed, acted upon, and a response issued to NRC, where required. The

IE Notices were received and evaluated by appropriate corporate and site

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personnel for applicability. The inspector now considers the above

[ bulletins and notices to be closed for McGuire Units 1 and 2.

To further confirm that the licensee's programs for vendor manuals were

adequate and being implemented, the inspector selected ten safety-related

components from the Master Index List and then physically inspected the

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manuals in the Document Control Areas at McGuire Station and at the

! Corporate Office Design Engineering. The identification numbers assigned

by the licensee were checked against the vendor identification numbers in

the computer and on the manual hard copy. Vendor manuals examined are

listed below:

i

i

!

i

+- ,, _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ . _ . , _ . _ _ _ _ . _ _ _ . . _ . _ . _ _

. _ __ .- . . - - _ _ . _ . _ _ . __ ._ _ _ _ - _ _ _ - .

.

. .

l

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L

12

l

Drawing No.

'

Title . Vendor

~12623 Main Steam valve Inst. Atwood &

c Morrill

DAP-ACAPRH-01/02 RHR Pump I/B Westinghouse

! 79F55799 Aux. Feedwater Pump Motor Westinghouse

l 804-3 Safety-Related Elec. Transm. ITT Barton

l Start-Up Manual Nuclear Steam Supply System Westinghouse

'

Start-Up Manual .

Incore Instrumentation Tech Manual Westinghouse

i MCM 2203.03-0006 Feed Wtr. Pump / Seal Inj. Pump Pacific Pump

SMBI-180C Limitorque Maintenance Manual. Limitorque

i MPM-WOGRTSDS416-01 Reactor Trip Breaker Switch- Westinghouse

Gear Maintenance Manual

4 .I. B. 33-790-1E Reactor Trip Switch-Gear Westinghouse

Breaker Maintenance Manual

Vendor manuals were also examined to determine if the Vendor Document

'

Instruction Insert Control Forms had been attached to the nanual as

! required by procedure. Manuals .were also examined to determine if the

insert material had been inserted as specified on the insert forms.. The

t following manuals identified by the Duke File Number were examined and

l found satisfactory:

4

MCM 1201.00-0006, Vol. 1 Transmittal Nos. 1265,1098 and 4784

-

"

MCM 1205.00-0938 Transmittal Nos. 2304, 0262 and 0704

MCM 1399.06-0067 Transmittal Nos. 2982, 1376 and 2329

!' The inspector selected several vendor technical information letters, and

Westinghouse Technical Bulletins and requested that the licensee use the

computer program to determine the status of corrective action. The

licensee produced three computer readouts which showed the status of

Reactor. Trip Breaker related items, vendor information letters that were-

breaker related, a list of vendor information letters which were still

open and a list of the vendor information letters which had been closed,

plus'the related dates. Based on the easy retrieval of information, the

licensee's Operating Experience Management and Analysis Operating

Experience Program appears to be a viable system for tracking technical

data.

The licensee has assigned an engineer as their Nuclear Plant Reliability

Data System (NPRDS) Coordinator. This coordinator works with INPO and

other Duke plant personnel to help facilitate its effectiveness. McGuire

personnel stated that they use the NPRDS data from the plant to determine

what areas have the most problems. They use this information as a tool to

determine if action is necessary. One problem is that personnel think

they have to be a computer expert to effectively utilize the data. The

licensee's Corporate Office is trying to orient the plant personnel on how

to use NPRDS and other data base information. The licensee demonstrated

. . - . - - . _ - . . . .--- _- . . . - .

.

.

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13

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i

the use of NPRDS to show the inspector how many Reactor Trip Breaker

, records were on the data base. The use of NPRDS program by the licensee

'

could be an asset during component evaluations and trouble shooting plant

problems, provided that plant personnel utilize the data.

F To further enhance vendor manual control and the program to ensure that

vendor technical information .and recommendations are incorporated into

plant procedures, the licensee's computer program is designed to cross -

i

. reference every plant procedure that is affected by each vendor manual or

L technical bulletin. Also when a plant procedure is changed. - the

applicable vendor manuals can easily be identified. The licensee

demoristrated the effectiveness of this system by retrieving plant-

'

4

"

procedures associated with ten referenced vendor manuals. The licensee

also issues a Vendor Manual Status Report every month. The inspector

reviewed the latest status report which shows the number of vendor manuals

! verified, discrepancies noted, inserts re-issued, manuals proofed af ter

checkout, new manuals received and hours worked. The licensee has a

special group to handle vendor manuals to ensure that the manuals are kept

current and to ensure adequate control. The licensee demonstrated how l

they could retrieve vendor manual information by vendor name, vendor

1

drawing or manual number, component name, and by their own unique

j numbering system.

, Audits performed by two licensee's Quality Assurance Groups were examined

to determine if audits were being performed in the document control areas

and at vendors facilities. The audits of the document control activities

were conducted by the General Office Quality Assurance and the audits of

,

the vendors and suppliers of saftety-related equipment and services were

performed by the Quality Assurance Vendors Division. The audits examined

{ are listed below:

k~ QA Audit MC-86-1 through MC-8 was performed in January and February

i 1986. The report was completed and all action items had been signed '

,_

off.

! QA Audit MC-85-66 through MC-85-77 was performed between November 1,

through December 31, 1985. This audit resulted in nine findings in

'

,

! the Station Document Control Areas, including the Satellite Files.

! Four of these findings required corrective actions which were

corrected and closed out in February 1986.

!

! Power Conversion Products, the supplier of station battery chargers,

i was audited on May 23,1984, June 20,1985, and on April 9, 1986, to

! determine if they were to remain on the Approved Suppliers List.

,

Results of audits revealed that this supplier was approved.

'

Pacific Pump Division of Dresser Industries, supplier of charging

i pumps, had been continuously audited since 1973. Latest response was

i on January 20, 1986 when the licensee closed out their findings.

l Present acceptance is based on ASME Certification.

!'

!

!

!

_ _ _ _ _ .. _ _ _ _ _ _ _ _ _ . _ . - _ _ . _ __. _ _ _ _ . _ . _

_ . ~ _. __ _ _ _ _ - _ _ _ . _ .

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14

'

Limitorque Corporation had been audited in 1984, 1985, and on May 27,

and June 26, 1986. The file folder also contained other

correspondence between the licensee and the- vendor concerning

LER 414/86-35 and NRC Information Notice 86-03. Duke had based their

-1987 approval on a History Review.

~

Raychem Corporation was last audited on September 28-29, 1986. This.

audit addressed the safety-related application of the WCSF-N line of

heat shrink tubing which had been qualified for IE service.

i

Westinghouse Electric (NTD) was approved based on results of a CASE

l

Report by Florida Power and Light, NRC Report 99900404/85-1, AIA

Lumbermans Mutual and the history of nuclear work.

Westinghouse Electric (NSID) was audited by the licensee on

August 25-26, 1986, per Duke QAM Procedure QA-601. The licensee's

audit showed that W NSID had a QA program which was being implemented

-

by qualified perso_nnel, therefore, retained W NSID as an approved _

vendor.

Based on the review of audits discussed above, the inspector concludes

that the licensee is performing audits and surveys of their own document

control areas and are continuing to interface with their vendors and

'

suppliers.

3

Audits in'other areas of GL 83-28 were also reviewed with corporate audit

4 staff. The schedule for planned audits and the training and qualification

of auditors were also discussed. The inspector had no concerns in this

, area.

1

Within the areas inspected, the inspector did not identify any violations

,

or deviations.

l 9. Post Maintenance Testing: Reactor Trip Circuit Breaker

! The inspector ascertained that the licensee is committed to a full program

of maintenance and post-maintenance testing of reactor trip circuit

breakers by the following action plan

!

Field observation of the maintenance of a reactor -trip circuit

'

breaker.

!

Review of licensee's procedures on reactor trip circuit breakers.

Review of licensee's procedures versus vendor technical manuals.

Review of completed maintenance work requests on reactor trip circuit

breakers. l

4

.

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- .- . .-_.- - . -- - - . - . .- . - . _ _ _ _ . .

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15

.

a. Field Observations

The inspector witnessed the performance of a maintenance procedure of

a spare Westinghouse DS-416 Reactor Trip Circuit Breaker pulled from

the licensee's warehouse for this purpose. The procedure used was

-

. Maintenance Procedure MP/0/A/2001/06, approved March 10, 1987, and

" Westinghouse Program Manual MPM-WOGRTSDS416-01," Revision 0, dated

November 30, 1986.

, The mai~ntenance of the licensee's circuit breakers for all of their

plants are performed by the same group from the Transmission

'

Department located in Charlotte. This work is performed under the

i direction of Mr. R. Bledsoe, Sub-Station Equipment Supervisor,

Transmission Department. This group performs reactor trip breaker

maintenance for both Catawba and McGuire and was inspected in the

Catawba ATWS ' inspection in July 1986. As a result they were totally

prepared for -the McGuire ATWS inspection. This new procedure,

4

MP/0/A/2001/06, was thought out and dry runned prior to its final

approval one week before this inspection period. It includes all of

f the changes critiqued on the previous inspection and utilizes and

i refers to all specifications in the new Westinghouse Maintenance

Program Manual on DS-416 Reactor Trip Breakers, which also was dated

as being received one week prior to the inspection.

At the start of-the maintenance procedure all of the special tools,

meters, and power supplies listed in the procedure were' logged with

, their serial numbers and calibration dates with verification by the

Quality Assurance QC/QA representative. The procedure went smoothly

and in sequence. The QA/QC representative took an active part in the

.

procedure.

2

Several editorial changes were suggested by the inspector which he

felt would enhance the procedure. These will be reviewed by the

licensee for their possible inclusion in the next revision of the

procedure. When drawings were referred to in the procedure, and

original was used when the copies in the manual were difficult to

read.

The inspector feels that this new procedure MP/0/A/2001/06, in

conjunction with the Westinghouse manual, provide an adequate

, inspection and maintenance of the Reactor Trip Circuit Breakers.

b. Review of Procedures

i The inspector reviewed the following procedures with the appropriate

personnel involved with each.

'

(1) MP/0/A/2001/06 - " Westinghouse DS-416 Air Circuit Breaker

i Inspection and Maintenance." This review is covered in the

previous section.

.

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. 16

(2) PT/0/A/4601/07A and PT/0/A/4601/07B " Response Time Testing of

Reactor Trip Breakers (A and B trains)."

The inspector walked through the complete procedures step by l

-step with the responsible individuals. They were observed to be

complete and accurate. The only inspector comment was on the

possibility of locking the back of the reactor trip rack cabinet

since the licen;ee stressed the importance of contacting control

room personnel and receiving approval before opening the

cabinet. The trip test buttons are located inside the cabinet.

_

The room where the cabinet is located is next to the control

room. The inspector noted that the cabinet room is temperature

and humidity controlled with filtered air and was-immaculate.

(3) PT/0/A/4601/08A and PT/0/A/4601/08B " Solid - State Protection-

System (SSPS) Trains A&B Periodic Test above NC System Pressure

of 1955 PSIG Change #4," dated June 18, 1985. These procedures

.were observed to be complete and accurate.

(4) PT/0/A/4601/09 and PT/0/4601/09A " Solid State Protection System

(SSPS) Trains (A&B) with NC pressure 1955 PSIG " Change #8,"

- dated October 6,1986. These procedures were observed to be

completed and accurate.

(5) PT/1/A/4600/56 and PT/2/A/4600/56 " Manual Reactor Trip Function

Test," dated January 27, 1987. These procadures were observed

to be complete and accurate.

c. Vendor Interface

The vendor manuals that were observed were complete and up-to-date

with the latest revisions. The correspondence observed showed close

cooperation between the vendors and the licensee,

d. Automatic Shunt Trip Modification and Undervoltage Modification

The inspector verified the installation of the automatic shunt trip

and undervoltage vendor modifications.

The documentation is as follows:

Nuclear Station Modification MG-1-1376 performed on Work

Request 92950 NSM, dated April 7, 1983.

Nuclear Station Modification MG-1-1161 performed on Work

Request 92340, dated April 7, 1983.

Nuclear Station Modification MG-2-0285 performed oa Work Request 92972.

. -

17

Nuclear Station Modification MG-2-0119 performed on Work Request 92341.

e. Generic Letter 83-28

The inspector reviewed the licensee's compliance with Items 3 and 4

of GL-83-28. The licensee was .found to be in compliance with the

exception of Item 4.2.2.

Item 4.2.2

Reactor Trip System Reliability (Preventative Maintenance and

Surveillance Program for Reactor Trip Breakers).

Position

Licensees and applicants shall describe their preventative

maintenance and surveillance program td' ensure reliable reactor trip

breaker operation.

(1) A planned program of periodic maintenance, including

lubrication, housekeeping, and other items recomended by the

equipment supplier.

(2) Trending of parameters affecting operation and measured during

testing to forecast degradation of operability.

(3) Life testing of the breakers (including the trip attachments) on

an acceptable sample size.

(4) Periodic replacement of breakers or components consistent with

demonstrated life cycles.

Items 4.2.2 is not being implemented by the licensee at the McGuire

, Nuclear Station. The data is recorded but not compiled and trended.

The following communications refers to this item.

(1) GL-83-28, Item 4.2, NRC Position and Requirements.

(2) Duke Response to Item 4.2.2, dated November 4, 1983.

(3) NRC Request for Additional Information, dated February 22, 1985.

(4) Duke Letter to H. Denton with Information, dated May 24, 1985.

(5) NRC Interim Report, dated February 17, 1986.

(6) Duke Response Letter to H. Denton, dated April 14, 1986.

This item is being evaluated by NRR.

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f. Licensee's Compliance With Amendment Number 2 to Facility Operating

Licensee NPF-17, Dated May 27, 1983

License Amendment Number 2 to McGuire Unit 2 Facility Operating

License dated May 27, 1983, paragraph 2.C.12.c requires the licensee

to implement the reactor trip breaker and reactor trip bypass breaker

testing and reporting as described in Table 1, " Periodic

Surveillance / Maintenance of Reactor Trip Breakers and Reactor Trip

Bypass Breakers." Table 1, Column 3 delineates the six months

surveillance and maintenance requirements for reactor trip and bypass

breakers. Item 1 of Column 3 requires that the data obtained from

response time testing of the UV/ breaker shall be trended.

"**To be performed before and after preventive maintenance."

The inspector found this item not being performed by the licensee.

'

The licensee has not been trending UV/ Breaker response time test

data. After being informed of this concern, the licensee reviewed

all previous test data and provided the inspectors a alot of Reactor

Trip Breaker Time Response Trend Curves. This data substantiated

that a negative trend had not occurred which would have caused a need

to take some corrective action. However, this is considered a violation

of the above license condition and is identified for tracking as

Violation 50-370/87-11-01, Failure to perform trending of reactor

trip breaker response time test data. The licensee submitted proposed

license amendments for McGuire Units 1 and 2 on December 7,1985 which

would eliminate the requirement for trending reactor trip breaker

response time. However, as of the date of this inspection this

amendment had not been approved. Considering the facts that the

licensee was able to perform trending of all previous test data and

that no adverse trends were observed, the staff concludes that the

safety significance of this item is minor and warrants only a Severity

Level V violation.

Conclusion

The inspector found the responsible personnel in each area to be

cooperative and knowledgeable. The personnel performing the

maintenance appeared well trained in their area of expertise and

conscientious in the performance of their duties. The work areas and

the area housing the Reactor Trip Breakers were immaculate.

The procedure MP/0/A/2001/06 was newly developed and incorporated the

critiques from previous inspections. It provides a step by step

sequential approach, provides QA/QC involvement and incorporates the

latest specifications from the vendor manual.

L

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The trending of response time testing of UV/ Breaker on UV signal data

as identified in the Amendment #2 to Facility Operating License

NPF-17 Table 1, Column 3, Item 1, was not being performed by the

licensee. This item is considered a violation.

The inspector concluded that the post-maintenance testing and all

areas audited by him provide a high level of safety in the operation

of the licensee's station.

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