ML20148Q914

Insp Rept 50-482/87-24 on 871019-23.Violations Noted.Major Areas Inspected:Implementation of Equipment Qualification Commitments Identified in SER (NUREG-0881) & Suppls 4 & 5

ML20148Q914
Person / Time
Site:	Wolf Creek
Issue date:	04/05/1988
From:	Ireland R, Andrea Johnson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20148Q864	List: ML20148Q914
References
RTR-NUREG-0881, RTR-NUREG-881 50-482-87-24, NUDOCS 8804140033
Download: ML20148Q914 (33)

See also: IR 05000482/1987024

Text

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APPENDIX B

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

a

NRC Inspection Report: 50-482/87-24 Operating License: NPF-42

Docket: 50-482

Licensee: Wolf Creek Nuclear Operating Corporation-(WCNOC)

P.O. Box'411

Burlington, Kansas 66839

Facility Name: Wolf Creek Generating Station (WCGS)

Inspection At: Burlington, Kansas

Inspection Conducted: October 19-23 (onsite) to November 20, 1987

-(NRC Region IV Office)

Inspector: M Y Mo M ) M/6 82$

AfR.Johnsn,ReactorInspector,TeamLeader Date '

[Kyision af Reactor Safety

Also participating in the inspection and contributing to the report were:

U. Potapovs, Chief, Special Projects Section, VIB, DRIS, NRR I

S. Alexander, Equipment Qualification & Test Engineer, VIB, DRIS,

NRR

M. Trejovsky, Consultant Engineer, Idaho National

Labcratory (INEL)

4 T. Humphrey, Consultant Engineer, INEL

D. Kirg, Member of Technical Staff, Sandia National Laboratories

R. Ireland, Chief, Plant Systems Section, DRS, Region IV

D. Norman, Reactor Inspector, PSS, DRS, Region IV

R. Mullikin, Project Engineer, DRP, Region IV

R. Evans, Reactor Inspector, PSS, DRS, Region IV

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Approved: _

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E. Ireland, Ac ing Chief, P nt Systems

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_R.

Sec; ion, Division of Reactor S~afety

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8804140033 880400

PDR ADOCK 0500040p <

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Inspection Summary

Inspection Conducted October 19-23, 1987 (onsite) to December 18, 1987

(NRC Region IV Office) (Report 50-482/87-24)

Areas Inspected: Special, announced inspection to review the licensee's

implementation of a program for establishing and maintaining the qualification

of electric equipment within the scope of 10 CFR 50.49. In preparation for this

inspection, the NRC team included a review of WCNOC's implementation of EQ

corrective action commitmer,ts, identified in SER, NUREG 0881, April 1982, and

Supplements 4 and 5 (December 1983 and March 1985). These documents provide the

NRC staff acceptance with regard to equipment for which justification for

interim operation (JI0s) were provided prior to the November 30, 1985, deadline.

The NRC inspection team reviewed a sample of 38 EQ work packages in the

documentation files out of a total of 67, and walked down 47

components / equipment.

Results: The inspection determined that the licensee has implemented a program

to meet the requirements of 10 CFR 50.49, however, four violations of NRC

requirements and three unresolved items about which more information is required,

were identified. The licensee's letter No. WM 37-0309 to NRC Region IV, dated

November 20, 1987, provided additional information subsequent to the onsite

spection and addressed the inspection findings presentea in the exit interview

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by the NRC on October 23, 1987. The licensee's information and proposed methods

of resolution to the inspection findings have been reviewed and were considered

in preparation and issuance of this report.

The deficiencies identified by the NRC inspection team represent documentation

files which could not establish that this equipment was qualified. These

components were identified during the onsite review of the EQ documentation

files and a corresponding plant walkdown inspection. The licensee was urged to

resolve these concerns and place the necessary justifications for continued

operation (JCOs) in place as soon as possible in accordance with the NRC Generic

Letter 86-15.

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DETAILS

1. Persons Contacted

WCNOC

B. Withers, President

F. T. Rhodes, Vice President, Nuclear Operations

J. A. Bailey, Vice President, Engineering and Technical Services

M. L. Johnson, Nuclear Coordinator

A. A. Freitag, Manager, Nuclear Power Engineering (NPE)

C. M. Estes, Supervisor of Operations

K. Peterson, Supervisor, Licensing

E. Peterson, Supervisor, Qualification Evaluations

R. E. Gimple, Technical Staff Engineer, Materials Quality

A. L. Payne, Suoervisor, Quality Plant Support

C. E. Parry, Superintendent, Quality Engineering

J. C. Goode, Licensing

V. D. Luckert, Engineer, NPE

J. F. McMahon. Supervisor, Technical Training

D. R. Richard, EQ Engineer

C. J. Hoch, QA Tech

W. J. Rudolph II, Manager, QA

J. M. Pippin, Manager, NPE

J. Stokes, Material Services Manager

0. L. Maynard, Manager, Licensing

C. A. Snyder, Manager, Purchasing and material Services

D. Rich, Superintendent ci Maintenance ,

G. Boyer, Plant Manager  ;

S. C. Hopkins, Senior Engineering Specialist

B. McKinney, Superintendent, Technical Support

• M. G. Williams, Superintendent, Regulatory, Quality, and Administrative

Services

• L. L. Cook, Supplier Quality l

• K. Harvey, QC Services Supervisor 1

• T. M. Damashek, Quality Engineer

R. D. Flannigan, Supervisor, Compliance Engineer

• M. R. Bove, Senior Engineer

• J. L. Houghton, Operations Coordinator, Operations

• S. Austin, Operations Coordinator

W. C. Wiseman, Maintenance

Others

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B. J. Metro, Engineer, Westinghouse l

M. H. Fletcher, Consultant, CPA, Inc. l

J. G. Utt, Senior Engineer, Bechtel  ;

D. N. Lorfing, Licensing Engineer, GSU l

W. E. Kahl, UE-Callaway

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NRC

B. L. Bartlett, Resident Inspector

J. E. Cummins, Senior Resident Inspector

• Denotes those not present at the exit interview.

2. Purpose

The purpose of this inspection was to review the licensee's implementation

of the requirements of 10 CFR 50.49.

3. Background

NUREG-0588 was issued in December 1979 to promote an orderly and

systematic implementation of EQ programs by industry and to provide

guidance to the NRC for its use in ongoing licensing reviews. The

positions contained in NUREG-0588 provided guidance on (a) how to

establish environmental service conditions, (b) how to select methods that

are considered appropriate for qualifying equipment in areas of nuclear

plants, and (c) other areas such as margin, ging, and documentation.

A final rule on environmental qualification of electrical equipment

important to safety for nuclear power plants became effective on

February 22, 1983. This rule, Section 50.49 of 10 CFR Part 50, specified

the requirements to be met for demonstrating the environmental

qualification of electrical equipment important to safety located in a

harsh environment. In conformance with 10 CFR 50.49, electrical equipment

for the Standardized Nuclear Unit Power Plant System (SNUPPS) plants are

qualified according to the criteria specified in Category I of NUREG-0588.

In order to document the degree to which the environmental qualification

program complies with the NRC's environmental qualification requirements

and criteria, the licensee prnvided equipment qualification information

since the November 30, 1985, deadline, by letters dated November 29, 1985,

January 17, April 4, May 14, July 28, August 1, August 7, August 29, 1986,

June 5 and November 20, 1987, to supplement the information contained in

Section 3.11 and Appendices 3.11(N) and 3.11(B) of the FSAR.

The Wolf Creek SER, NUREG-0881, April 1982, requested the licensee to

submit his environmental qualification program for. safety-related

information as outlined in NUREG-0588, Appendix E, and update Section 3.11

of the FSAR, prior to the NRC job site audit of the licensee's files as i

outlined in the NRC Standard Review Plan.

SER, NUREG-0881, Supplement No. 1, was issued in December 1983, which

evaluated the adequacy of the SNUPPS EQ program and provided the NRC

position relating to open items and unresolved issues. These outstanding

items were either resolved prior to the issuance of the operating license

or JI0s in accordance with 10 CFR 50.49 were provided. Such outstanding

items included (a) information which demonstrated qualification of all

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electrical equipment required by the Three Mile Island (TMI) action plan,

and installed in accordance with Regulatory Guide 1.97; and (b) the

surveillance / maintenance program to implement the program before fuel

loading and a program to detect age-related degradation of electrical

cables inside containment that includes a periodic inspection of selected

cables.

SER, NUREG-0881, Supplement No. 5, was issued in March 1985. The NRC

concluded that, for safety-related items not having complete qualification

documentation, the licensee had provided commitments for corrective action

and schedules for completion. For items identified that did not have full

qualification before an operating license was issued, an analyses had been

performed in accordance with 10 CFR 50.49(i) to ensure that the plant

could be operated safely pendir.g completion of environmental

qualification. These analyses reviewed by the NRC and JI0s concluded that

reasonable assurance had been provided that the SNUPPS plants could be

operated safely pending completion of environmental qualification. The

NRC concluded that a license condition would be incorporated into the

SNUPPS plants licenses requiring all electrical equipment with' the scope

of 10 CFR 50.49 to be qualified by November 30, 1985.

By letter dated February 4,1985, WCNOC in response to Generic

Lettt4 84-24, certified compliance with 10 CFR 50.49 in that all equipment

was either fully qualified or a JIO had been submitted pending full

qualification.

By letter dated January 17, 1986, WCNOC provided to the NRC, Revision 3,

"Report of Independent Review of Environmental Qualification Programs to

NUREG-0588" which reflected the completion of the EQ review by WCNOC and

closure of outstanding JI0s. ]

By letter dated April 4,1986, WCNOC provided to the NRC a report

entitled, "Evaluation of Environmental Qualification of Equipment

Considering Superheat Effects of H;;h Energy Line Breaks for Callaway

Plant and Wolf Creek Generating Station." The report concluded that the

equipment which must function to mitigate a postulated high energy line

break (HELB) with superheat effects, to bring the SNUPPS plants to a safe l

shutdown condition, would perform their safety functions following such a '

postulated event

The documents above identified were reviewed by the NRC inspection team

members and used in preparation for this inspection. The inspection

involved an onsite and subsequent NRC Region IV in-office inspection of

records subsequently furnished by the licensee. 1

4. Findings

a. EQ Progran Compliance with 10 CFR 50.49

The NRC inspectors examined the licensee's program for establishing

the qualification of electric equipment within the scope of

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10 CFR 50.49. The program was evaluated by examination of the

licensee's qualification documentation files, review of procedures

for controlling the licensee's EQ efforts, and verification of

adequacy and accuracy of the licensee's program for maintaining the

qualified status of electrical equipment. Based on the inspection

findings, which are disc.ussed in more detail below, tha inspection

team determined that the licensee has implemented a program to meet

the requirements of 10 CFR 50.49 for WCGS although some deficiencies

were identified (refer to Sections 4.f, 4.g, and 4.h).

b. EQ Program Procedures

The inspection team examined the implementation and adequacy of site

policies and procedures for establishing and maintaining the EQ of

electrical equipment in compliance with the requirements of

10 CFR 50.49. The licensee's methods for establishing and

maintaining the EQ of electrical equipment were reviewed in the

following documents:

Type Procedure No. Title

Directive III.24.0, Revision 2 Equipment Qualification

EQ Procedure EQM IC 0908, 4/11/85 EQ Qualification

Requirement for S-R and

I&C Cables Inside

Contairment

General Procedure KGP-1131, Revision 5 Plant Modification

Process

Nuclear Plant KPN-C-309, Revision 1 Plant Modification

Engineering Design Development

Procedures

KPN-D-319, Revision 1 EQ Review of Electrical

Equipment to 10 CFR

50.49

KPN-D-324, Revision 0 EQ Summary Document

(Draft)

Procurement KP-2140, Revision i Material and Services

Procedure Procurement

Operations ADM 01-057, Revision 12 Work Request

Procedures

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ADM 08-202, Revision 9 Planning and Scheduling

Preventive Maintenance

Tasks

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ADM 08-211, Revision 1 Updating Procedure for

Managed Maintenance Data

Base

ADM 08-813, Revision 1 I&C Group EQ Maintenance

Program

Emergency EMG E-1, Revision 1 Loss of Reactor Secondary

Operating Coolant

Procedure

The NRC inspection team reviewed the above licenree's procedures for

meeting the requirements of 10 CFR 50.49 including (1) qualified

life; (2) service conditions; (3) periodic testing; and

(4) maintenanca and surveillance. The licensee's EQ program was also

reviewed with regard to establishment of an auditable documentation

file, including such documents as EQ audit reports, maintenance and

surveillance records, supporting documents which establish EQ

training of personnel, and supporting documents which control plant

modifications, and installation of replacement equipment to the

requirements of 10 CFR 50.49.

The licensee's E0 program procedures and policies are established and

are being adequately implemented to control and maintain the EQ of

electrical equipment at WCGS for compliance with the requirements of

10 CFR 50.49.

The following programs were effectively in place at WCGS:

(1) EQ Maintenance Program  ;

The WCGS EQ maintenance program consisted of

Procedures ADM 08-202, ADM 08-291, and ADM 08-813 which were in l

place at the time of the NRC inspection.

ADM 08-202 provides instructions for planning and scheduling

routine mechanical and electrical preventive maintenance (PM)

tasks by means of a managed maintenance data base (MMDB) which

r,tipulates scheduled PM activities on a computer printout.

Provisions are made for an evaluation of PM activities performed

after a scheduled late date. Lubrication requirements are

stipulated by reference to a Master Lubrication List (MLL) which l

1s addressed in ADM 08-208. '

ADM 08-211 provides the method for documenting and making i

available new information for updating the MMDB. A review of '

change information was made by the NRC inspector.

ADM 08-813 implements an EQ maintenance program which requires ,

procedures to be written for instrumentation and control (I&C)  ;

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components which have a qualified life less than 40 ye *s and/or

contingencies for maintaining the specified qualified li e. '

In addition to procedural reviews and personnel interviews, the

NRC inspection team reviewed the maintenance program

implementation by selecting a sample of eight components from

the EQ Master List (EQML) and comparing the required PM from

(a) the SNUPPS NUREG-0588 submittal to NRC including changes;

and (b) maintenance requirements from the current MMDB printout.

This review revealed an apparent weakness in the change system

in that required maintenance was not always reflected in the

MMDB. The apparent weakness was also identified in QA Audit

Report TE50140-K170 in which a 100~ percent review was made of

all Categories A and B equipment requiring special maintenance

and/or replacement prior to 40 years. In spite of the

identified weaknesses, correction of deficiencies identified in

the QA audit _ should ensure that all required maintenance is

performed during each outage. Implementation of

Procedure KPN-D-324, "EQ Summary Document." which is now in

draft, should provide for maintenance program improvement.

The WCGS EQ program appears to be well planned and implemented.

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

(2) Surveillance of Safety-Related IAC Cables Inside Containment

As a result of badly deteriorated cable insulation identified at

Savannah River during late 1976, WCGS committed in SSER 4, dated

December 1983, to a surveillance program to identify and prevent

significant age-related degradation of I&C cables inside

containment. The program was presented to and determined to be

acceptable to the NRC, as stated in the SSER. During this

inspection, the NRC team reviewed the implementation of this

program to date.

Procedure EQM IC-0908, dated April 11, 1985, "Environmental

Qualification Requirement for Safety-Related I&C Cables Inside

Containment," was reviewed to determine the details and

frequency of this inspection requirement. The work request (WR)

(currently not numbered nor dcted) which will implement the WCGS

inspection program to completion, was reviewed by the NRC

inspection team. The WR schedules inspections to be completed

each 5 years, commencing on March 25, 1990. The WR and

supplements identify the inspection procedures to be followed

and which components are to be inspected. The baseline

inspection was reportedly performed prior to initial startup of

the plant in accordance with WR 03933-85. This program is

apparently being adequately implemented.

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No violations, deviations, or unresolved items were identified

by the NRC inspection team.

(3) Control of Plant EQ Modifications

WCGS Procedures KGP-1131, Revision 5, and KPN-C-309, Revision 1,

were reviewed by the NRC inspection team to ensure proper

control of equipment qualification is being maintained during

plant modifications. Procedure KGP-1131, "Plant Modification

Process," describes the process of initiating a Plant

Modification Request (PMR), obtaining the required approvals,

and closecut of the PMR. KGP-1131 also establishes departmental

and personnel responsibilities with the PMR. The procedure

ensures that the responsible parties review the PMR for effect

on the EQ program. Procedure KPN-C-309, "Plant Modification

Design Development," describes the engineering activities

associated with P' irs. This pro edure cu.eains instructions and

associated EQ checklist to ensure PMR reviews include

evaluations with regards to the effect of changes and station

modifications involving EQ equipment.

The control of the WCGS plant EQ modif.sation program

implementation, to verify that modifications involving EQ

equipment have been incorporated into the PMR process, will be

accomplished during a subsequent NRC inspection

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

(4) Program for Independent Review of EQWPs

Procedure KPN-D-319, "EQ Review of Electrical Equipment to

10 rFR 50.49," establishes the guidelines on performing an

independent review of EQ programs, including EQWP development by

WCGS personnel and review of EQWPs supplied by outside

organizations. The NRC inspection team reviewed EQWP

documentation for evidence that a positive statement was

included in each EQWP by the licensee to ensure that the

equipment is qualified for its application and that the

documentation has been properly reviewed and approved. The

"Electrical Equipment Qualification Signature Sheet,"

(Form KEF-D-319-1), clearly performs this function; however,

EQWPs supplied by other organizations reviewed prior to the

initiation or KPN-D-319, Revision 0, did not contain the

signature sheets.

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

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c. EQ Training Program /QA Audits of EQ Activities

The NRC inspection team reviewed training records and conducted

interviews with WCGS training personnel responsible for administering -

training programs related to EQ activities. WCGS is in the process

of formalizing their EQ training program by written procedure.  ;

Verification of this program will be accomplished during a subsequent l

NRC inspection. i

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WCGS has in place an audit program which prescribes planned and ,

periodic QA audits of EQ activities. The NRC inspection team i

reviewed WCN0C QA audit reports TE 50130-X045 and TE 50140-K170. The l

NRC inspection team concluded that these EQ audits were effectively J

being accomplished.

No violations, deviations, or unresolved items were identified by the

NRC inspection team. )

d. EQ Replacement Parts / Procurement Program

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. The WCGS EQ Replacement Parts Program is an integral part of the

licensee's overall EQ program and is controlled by  !

Procedures KP-2122, "Material and Services Receipt;" KP-2123, i

"Material Storage and Handling;" KP-2124, "Material Issue;" and i

KP-2125, "Stored Item Maintenance." The WCGS E0 Procurement Program '

is controlled by Procedure KP-2140, "Material and Services <

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Procurement."

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No violations, deviations, or unresolved items were identified by the

NRC inspection team.

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e. EQ Master List (EQML)

The NRC inspection team reviewed the WCGS EQML, Revision 0, and

associated documents to verify its adequacy and verify the 1

implementation of WCNOC's EQML development and maintenance

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procedures.

The WCGS EQML is presented as Table 3.11(B)-3 of the WCGS Updated

Safety Analysis Report (USAR). The WCGS EQML was based on a review

of technical specifications, emergency operating procedures (EOPs),

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off-normal operating procedures (ONOPs), piping and instrumentation

4 diagrams (P&lDs), electrical diagrams, Regulatory Guide (RG) 1.97

4 (Revision 3, Categories 1 and 2), NUREG-0737, and plant equipment

i verification walkdowns.

The original EQML, submitted to NRC on March 10, 1983, was compared

to the most recent EQML fcund in the USAR Table 3.11(B)-3, Numerous

. changes were made between the two revisions of the EQML using FSAR -

change request forms. EQML changes will be controlled by a new

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procedure (KPN-D-3?4, "EQ Summary Document") when the procedure is

approved.

The procedures for developing and maintaining the EQML define the

EQML and assign overall responsibility for maintenance of EQ

engineering files to the nuclear plant engineering / technical support

department. These procedures also control deletions from the EQML,

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as well as additions. The process as described was. adequate to

ensure that the appropriate determination could be made for equipment

added to or deleted from the EQML. The NRC inspection team  :

identified no plant equipment that was required to be on the EQML  !

that was not included.

Additionally, the E0Ps were reviewed with WCGS operations personnel.

The NRC inspection team selected 16 items of equipment required to be

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used with the E0P for loss-of-coolant accident (LOCA)/ main steam line

break (MSLB) and verified that they were all listed in the EQML as

qualified.

The NRC inspection team also reviewed the post-accident monitoring

equipment program status. RG 1.97, "Instrumentation for Light Water

Cooled Nuclear Power Plants to Assess Plant and Environs Conditions

During and Following an Accident," was reviewed against USAR

Tables 3.11(B)-3 and 7A-3, and RG 1.97, Table 2 recommendations. All ,

necessary equipment required by RG 1.97 was included on the EQML or a l

satisfactory explanation was documented as to why it was not  :

included. '

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Based on the NRC inspection team's review, the 10 CFR 50.49 EQML is

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considered satisf actory.

No violations, deviations, or unresolved items were identified by the  !

NRC inspection team,

f. EQ Occumentation Files (EQF)  !

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The licensee's EQF at WCGS has been established and is being

maintained to meet the requirements of 10 CFR 50.49. The

requirements are contained in WCGS engineering Procedures KPN-0-319,

Revision 1, KPN-D-324, Revision 0, and WCGS Nuclear Department

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Directive No. 111.24.0, Revision 2. These procedures and the

directive apply to the activities for environmental qualification of

equipment important to safety as committed in the WCGS USAR,

Revision 0, Appendix 3.11(B) and 3.11(N). Procedure KPN-D-319 is

primarily intended to be used for developing EQWPs by Bechtel,

SNUPPS, or the licensee. The EQWPs in the EQF are the result of

qualification documentation (test reports, test plans, vendor

descriptions and data, correspondence, etc.) to the requirements of

1 NUREG-0588, Revision 1, "Interim Staf f Position on Environmental

Qualification of Safety-Related Electrical Equipment." The EQWPs in

the EQF consist of: (1) EQ signature sheet; (2) EQ evaluation check

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sheet; (3) check sheet supplements (qualification contingencies); i

(4) equipment evaluation work sheets; and (5) references used to

complete the EQWP (i.e., test reports, letters, drawings,

calculations, analyses, etc.).

The NRC inspection team examined files for 38 selected equipment

items (EQWPs) to verify the qualified status of equipment within the ,

scope of 10 7R 50.49. In addition to comparing plant service  ;

conditions with qualification. test conditions and verifying the bases  !

! for these conditions, the NRC inspection team selectively reviewed

areas such as: (1) required post-accident oparating time compared to  !

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the duration of time the equipment has been demonstrated to be  ;

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qualified; (2) similarity of tested equipment to that installed in

the plant; (3) evaluation of adequacy of test conditions; (4) aging -

calculations for qualified life; (5) replacement part schedules;  ;

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(6) the effects of decreases of insulation resistance on equipment

performance; (7) adequacy of demonstrated accuracy; (8) evaluation of

test anomalies; and (9) applicability of EQ problems as reported in  ;

IE Information Noti:.es (ins) and IE Bulletins (IEBs) and their

resolution. The files adequately documented qualification of

equipment and were readily auditable, complete, and accurate.

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During this review of the EQF, the NRC inspection team identified

. violations to 10 CFR 50.49, the unresolved items, and open items

described below:

l (1) EQWP-01013, Raychem Cable Termination Material i

WCNOC committed to incorporating Wyle Test Reports 17859-02B and

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-02P (not in the EQF during the NRC inspection) that demonstrates )

qualification for some nonstandard splice configurations at the

WCGS into the Raychem EQ file EQWP-01013. Contrary to

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paragraph 5(1) of NUREG-0588, Category I, the file as reviewed did i

not support qualification of the installed splice configurations

i or establish simile ity between tested and installed splices, as

discussed in paragraph 4.h(1) of this recort. In addition, some

licensee-identified splice configurations appear to have fallen

outside the boundt, of all available type test documentation which

qualification could be supported.

The licensee letter WM 87-0309 to NRC Region IV dated ,

November 20, 1937 further discussed Wyle Laboratnries

Qualification Test Report No. 17859-02P which documents the

testing of a nur.ber of configurations of splices from various

suppliers. Included in thfs test were Raychem WCSF-N splices ,

with seal length significantly less than 2 inches (as little as l

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1/2-inch seal on some test samples) and bend radii to

approximately 1.2 times the 0.D. of the splice. This report was

obtained through membership in the Nuclear Utility Group on

Equipment Qualification (NUGEQ) and is now available for review

in the WCGS EQF. The Raychem splice bend radii less than 5

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times the 0.0. installed at WCGS were because of the small

physical size of the terminal boxes used for transmitters. A

calculation based on conformance of the Raychem splice

configurations to the insioe dimensions.of.the terminal boxes ,

yielded the smallest bend radius configuration at WCGS to be

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approximately 1.43 times the splice 0.D. EQWP E-01013 wili be

revised to reflect the above test report results and the results

of other ongoing programs addressing Raychem splices as

appropriate. j

, A licensee program to walkdown approximately 150 of the Raychem

l splices at WCGS in harsh environment areas revealed: (a) one j

splice with a seal length less than 1/2-inch which has been '

3

~

replaced; (b) two splices that had adhesive missing from the end

of the shrink tubing; (c) two splices that were identified as not

having the tubing fully shrunk; and (d) one splice that was

missing a shim. These splices have been replaced by-the licensee.

Further, evaluations of installed configurations are being

conducted in accordance with station orocedures.

This item is considered a violation to 10 CFR 50.49 (also refer

i

to paragraphs 4.g(1), 4.h(1)(a), 4.h(1)(b), and 4.h(1)(c) of this

report). (482/8724-01) i

1

(2) EQWP BOP-Limitorque, HE-1, and HE-4, Limitorque Motor Operater,

'

Energized Space Heaters ,

1

The NRC inspection team reviewed the above EQWPs in the licensee's '

EQF and identified the following deficiencies as a result of the i

file review associated with a walkdown of selected equipment

identified in each file.

I

(a) Limitorque Motor Operator Space Heaters (EQWP r:

B0P-Limitorque) Inside Containment I

'

,

These motor operators, reviewed by the NRC inspectors, are

primarily located inside the containment. These operators

a-e single voltage motors with "Radiation Class" or "Class

'

RH" insulation, and all are above the postulated flood

elevation for the areas in which they are located.

i Qualification of the space heaters for use in the Class IE

'

control circuits of these Limitorque motor operators was not'

established in the EQF. The analysis, used in determining  !

i

that no failure mode of the space heaters existed that could

adversely affect the operation of these operators, was not

i established fully in the file. Qualification of the space

heaters was based primarily on Bechtel's conclusion that the

only failure mode of these heaters is an open circuit. The

'

i analysis found in the file did not show how this conclusion l

was reached. Supplementary information present in the file

1

i

i l

'

l

- - . , . - , , . . - - _-. . - - - - - . - - - .-.--c- - - h L- l

. <

\

. . ._ _ -

,

'

, a

.. ..

, ...

'

>

14

,

N

d i

,

that was apparently used as a basis-for this conclusion was ^

Limitorque Laboratory Report 681T.41. In addition, this  !

report describes testing of a Wird Leonard ter,  ;

ID No. 30/25F750WL8005, which was installe, in a simitorque  !

motor operator with the actuator assembly subjected to a i

design basis accident (DBA) test. This test report does not  !

adequately' describe: (1) the test sotup regarding

'

I

orientation and electrical interfaces (such as conduit i

entries and seals, if any), (2) the actual test instruments l

used to monitor the test, and (3) the purpose of the test.  ;

,

Also, the heater was onl/ energized during the last i

'

10 minutes of the 30-day LOCA type test stimulation. During *

this test, resistance measurements were made between the two

heater leads ant. to ground; however, only four measurements

were taken throughout the 30-day LOCA type test. The icwest l

recorded resistance measurement was 15K ohms; however, the *

voltages >t which three of these resistance measurements

were taken, are not recorded in the test data. It was ,

l concluded by the NRC inspection team that the test report

used to qualify the heaters, and provide the basis for

Bechtel's avaluation, did not conform to current industry

standards used in establishing qualificction of electrical

equipment used in Class IE circuits.

, The air temperature rise because of the close proximity of

j degradable material to space heaters had not been properly

l addressed by the licensee. A review of the licensee's ,

) methodology indicated the.t bulk temperature rises inside  !

'

the limit switch compartment were used to establish the i

thermal iging temperatures and thi no censideration was

given to the temperature rises bc ase of the close i

i pecximity of components to the heaters. During the ,

walkdown, the NRC inspector noted the heateri were within  !

0.5 to 1-inch of either: (1) wire insulation; (2) the  :

phenolic material of the terminal strips; or (3) the 9

fibrite/melamina material of the limit switch finger board.

~

In one case (LF-FV-0095), the NRC inspector noted the X1

jumper wire from points 18C to 17Z on tr.a torque switch was

within 0.010-inch of the heater, but showed no evidence of

insulation degradation. The heater lead wires, which were ,

not connected, were found in contact with the space heater j

]

and showed evidence of insulation degradation,

j

~

(b) Limitorque Motor Operator Space Heaters (EQWP HE-1) Insy

,

Contai_nment

Qualification of the space heaters for use in the Class 1E

control circuits of these Limitorque motor operators was

i not established in the EOF. The analysis used in

j determining that no failure mode of the space heaters

1

1

!

. .

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

-

.

. .

15

existed that could adversely affect the operation of these

operators was not fully established in the file. (Refer to

paragraph 4.f(2)(a) above.)

During the NRC walkdown inspection of EP-HV-8808A, two

space heaters were observed to be installed in the limit

switch compartment with the closest component to the heater

being the heater lead wires. There was approximately

0.5 to 1-inch clea-ance between the lead wire insulation

and heater. The licensee's position on qualification of

space heaters and response to IN So-?1 regarding the

reported burned or damaged wires identified in Limitorque

motor operators with space heaters is outlined in

paragraph 4.f(2)(a) above.

The air temperature rise because of the close proximity of

degradable material to 'e heaters also had not been

properly addressed. Retei to paragraph 4.f(2)(a) above.

(c) .Limitorque Motor Operator Space Heaters (E0WP HE-4)

Outside Containment

The NRC inspection team found that some of these Limitorque

motor operator switch compartment space heaters were

connected and energized. The following concerns were

identified:

1) The licensee had p(rformed an analysis of the aging of

the operator motor switch compartment because of the

temperature rise caused by the heater. A worse-case

condition of an SMB-000 operator (smallest switch

compartment) and a 40-watt heater were used in the

calculatior, in which a temperature rise of 32*F

,

resulted inside the compartment. This would cause a

sizable reduction in qualified life of the operator

based on the limiting degradable material it the

switch compartment.

2) The analysis had considered that the air temperature

rise within the limit switch :cmpartment was uniform,

but no effects of localized heating of items in close

proximity to the heater was consfdered. During the

NRC walkdown, including a WCGS walkdown, control and

motor lead wires have been identified near and in

contact with energized heaters. There have also been

reported instancos of burned and degraded wiring by

WCGS. The NRC inspection team concluded that all

possible effects of aging and damage to these motor

operator components and wiring have not been analyzed

by the licensee and t;it wiring damage as a result

from the energized heaters has occurred.

. . . .

. -

's

. .

16

3) The power source for heaters is provided to terminal

blocks inside the operator from the same Class 1E

power supply, and through the same qualified field

cables, that supplies power to the operators. The-

jumper wires from the terminal block to the heater

have also been qualified; however, the heater has not

been qualified (tested), and failure of the heater in

a short-circuit mode would cause a loss of power for

the operator. The licensee stated that the heater

could fail, only in the open mode, which wout? not

ccuse power interruption.

The licensee's position could not be supported by test

data or analysis. (Refer to paragraph 4.f(2)(a)

above.)

The licensee's position on qualification of space heaters and

response to IN 86-71 regarding aging and damaged wiring, and

components is also outlined in the licensee letter WM 87-0309 to

NRC Reg on IV dated November 20, 1987. The licensee stated that

switch compartment space heaters in Limitorque motor operators

located inside containment were to be removed prior to restart

from the second refueling outage. In addition, a portion of

limit switch compartment space heaters located in all EQ

Limitorque operators outside containment would be removed, as

time and schedule permits. The remaining space heaters would be

removed in conjunction with scheduled maintenance on all EQ

operators throughout the third refueling outage. During this

work effort, the compartments would be examineo by the licensee

for signs of aging and degradation, and deficiencies identified

would be corrected. The licensee also committed to revision of

the EQWPs which would be revised to reflect the new

configv'ation without space heaters. A JC0 for me compartment

space heaters installed and energized was provided with the

WM 87-0309 letter.

Ir. conclusion, the post-inspection information provided by WCGS

which cited a worst-case 3?"I rise in bulk switch compartment

internal air temperature for the last 2.5 years, and calculated

4

an effect on qualified life, did not adequately establish

4

qualification in that there was no discussion of the heat rise

effect on components subject to thermal aging degradation. The

reduction in qualified life under these conditions could be as

large a factor of four for some materials. Based on the length

of time installed and original qualified life, some Limitorque

motor operator qualified life would be substantially modified,

based on consideration of these factors.

These items (paragraphs 4.f(2)(a), 4 f(2)(b), and 4.f(2)(c)

above) are considered a violation to 10 CFR 50.49. (482/8724-03)

.

l

- -

--4 . . - .

=

a

, .

17

(3) Limitorque Motor Opera _ tor Undervoltage/ Frequency Operation

(EQWP HE-4 and EQWP BOP-Limitorque)

(a) Limitorque Motor Operator Undervoltage/ Frequency

Conditions (EQWP HE-4) Outside Containment

Limitorque motors for valve actuators are required to

operate at voltages above and below the motor-rated voltage

of 460 volts. The EQWP check sheet supplement stated that

variations of 210 percent in voltage and 25 percent

frequency are part of the required NEMA motor design

standards for starting duty requirements considered in the

design of each individual motor operator. The test report

used to qualify the EQ4P HE-4 motor operators (Westinghouse

Report WCAP-8687, Supplement 2, H04A, Revision 3, dated

August 1986) did not demonstrate qualification of the

operators for either the high or low voltage requirements.

Documentation could not be presented during the NRC

inspection to show that these voltage requirements had been

considered in the Limitorque motor operator valve sizing

particularly with regard to thermal overload protective

device selections.

(b) Limitorque Motor Operator Undervoltage/ Frequency

Conditions (EQWP BOP-Limitorque) Inside Containment

Documentati q could not be presented during the NRC

inspection which addressed testing the valve actuators for

undervoltage/ frequency conditions during accident

environments. IEEE Standard 323-1974 requires that voltage

in power supplies be applied to the tested equipment.

Statements found in the file regarding the NEHA standards

used in designing motors were found to be unacceptable.

Because these standards may not consider the degraded

condition of a motor under the postulated environments of a

DBA, the licensee was required to ensure their actuator

motors were properly oversized for undervoltage/ frequency

applications in harsh environment, to support Limitorque's

position as found in Limitorque Tast Report B0058.

The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, stated that the operation of Limitorque

motors for valve actuators at 10 percent reduced voltage is

talen into consideration by Limitorque when sizing the

operator. During normal sizing for a reduced voltage

app'ication, the motor size is i'm eased to provide the

required motor torque at the reduced voltage. Limitorque,

in Qualification Test 90212, subjected the test actuator to

degraded voltages and frequencies with no effect on

actuator performance. The results of the test are

documented in the Limitorque EQWPs HE-1, HE-4, and

!

-

. _ _ _ _ . _

. -

.

. .

18

B0P Limitorque. WCGS stated that a separate evaluation was

being prepared to determine the operator capability at a

reduced voltage for the Limitorque operators qualified by

EQWPs HE-1, HE-4, and BOP Limitorque. The available thrust

cr torque for each qualified operator at 80 percent voltage

(a value 10 percent beyond the tested reduced voltage), as

compared to the required thrust or torqLe for each operator

based on system parameters, would be evaluated. The

schedule for completion of this task is March 1988.

The above concerns in paragraphs 4.f(3)(a) and 4.f(3)(b) are

considered an Unresolved item (482/8724-07).

(4) EQWP BOP.Limitorque, HE-1, and HE-4 Limitorque Motor Operator

Thomas & 3etts (Models RB-4 and RB-6) Crimp Connector Splices

The NRC inspection team reviewed the above documentation files

as follows:

(a) Limitorque Motor Operator Crimped Connector Splices

(EQWP HE-4) Outside Containment

During the documentation file review, the EQWP check sheet

supplement stated that dual voltage motor leads for Class B

insulated motors are spliced using Thomas & Betts crimp

connectors RB-4 and RB-6 as discussed in WCGS PMR 01787.

During the NRC walkdown inspec tion, motor operators I

BN-LCV-112E, BN-HV-8812B, and EM-HV-88218 were found to l

have these crimp type connectors installed ia the motor

leads of the dual voltage motors. Based on information 4

provided by the licensee, a total of 49 Limitorque 1

operators art quipped with dual voltage motors and have

the RB-4 and RB-6 crimped connections. Thirty-eight motor

operators were provided by Westinghouse (HE-4

specification) and eleven were provided directly from

Limitorque (M-221 and M-223A speci fication). Qualification

test reports provided for the operators contained in the

EQWP HE-4 did not include qualification of the RB-4 and

RB-6 connector splices.

Type testing of motor operators iocumented in Limitorque

reports B0003, 60U376A, and 600198 addressed dual voltage I

motor testing, however, documentation as to the type of

connector splice was not available. Also, no configuration

control was evidenced in documents contained in the file,

to demonstrate the prevention of contact of the crimp

connector splices with each other, or shorting to ground.

__

_-

.

, . .

.

. .

19

(b) Limitorque Motor Operator Crimped Connector Splices

(EQWP HE-1 and SOP-Limitorque) Insic'e Containment

The principal documentation used to establish qualification

of these dual voltage motor operators was Limitorque

Report 600456, contained in Limitorque Report B0058.

Several concerns were identified during the file review

regarding the qualification of the Thoma; & Betts,

Models RB-4 and RC-6, crimp conaector splices used in the

dual voltage motor actuators installed inside containment.

This file referer.:ed WCAP-8687, Supplement 2-H04A, where a

similar valve actuator with a dual voltage motor operator

was tested using similar crimp connector splices under

outside containment conditions. It was established by the

NRC inspection team that all nuclear steam supply

system (NSSS) and balance of plant (BOP) supplied

actuators, equipped with dual voltage motors, are

classified as NUREG-0588, Appendix E, Category C, type

equipment. Category C equipment is equipment that will

experience envitonmental conditions of DBAs through which

it need not function for mitigation of said accidents, and

whose failure (in any mode) is deemed not detrimental to

plant safety or accident mitigation, and need not be

qualified for any accident environment, but will be

qualified for its nonaccident service environment. The NRC

inspection team reviewed the licensee evaluations regarding

the reouired nonaccident service environment (radiation)

and concluded that qualification of EQWP HE-1, and

B0P-Limitorque dual voltage motor operators were

established.

(c) The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, in response to paragraphs 4.f(4)(a) and

4.f(4)(b) concerns, stated that a qualification test for

Thomas & Betts RB-4 and RB-6 nylon crimp connector splices

used in dual voltage Limitorque operators was being l

performed by Wyle Laboratories. An 8 year screen

qualification test was completed successfully in late

October 1987 with a completed test summary. A 40 year

qualification test is being conducted and completion and

issuance of a final test report is scheduled for early

1988. The results of the test program will be incorporated i

in EQWP HE-4 and the BOP Limitorque EQWP to document the

qualification when the final test report is issued. A JC0

was submitted to the NRC with letter WM 87-0309. The JC0 l

indicated that the Thomas & Betts connector splices were

qualitiable for 14.7 years, taking heater effects and WCGS

ambient temperatures into account, on the basis of the

completed Wyle tests,

i

L_______-__

. .

. .

20

NUREG-0568, Revision 1, Category 1, paragraph 5, requires that

qualification documentation shall verify that each type of

electrical equipment is qualified for its application and meets

its specified performance requirements. The basis of qualification

is required to be explained to show the relationship of all

facets of proof needed to support adequacy of the complete

equipment. Data used to demonstrate the qualification of the

equipment is regained to be pertinent to the application and

organized in an auditable form. The documentation is required

to include suffici'nt information to address those items identified

in NUREG-0588, Appendix E, which includes splices (Item 16).

The NRC inspection team's concern in paragraph 4.f(4)(a) above

is considered an Unresolved Item (482/8724-05).

(5) EQWP, ESE-lC, Tobar Model 32PA1212 Pressure Transmitter

The EQF indicated a design modification made to Tobar

Model 32PA1212 transmitters as a result of EQ type testing. The

NRC inspection team could not verify that the modification was

incorporated on the Tobar tran>mitters installed at WCGS.

The licensee letter WM 87-0309 to NRC Region IV dated November 20,

1987, indicated that the modifications which added silicon to

the Veritrak Model 76PH2 transmitters qualified by WCAP-8687

Supplement 2-E01B, Revi:1on 1, was incorporated in all

Westinghouse-supplied Tobar transmitters. In addition, transmitters

provided by Tobar directly to end users are also qualified to

the same test report.

Refer to paragraph 4.h(1) for NRC walkdown concerns. l

No violations, deviations, or unresolved items were identified l

5y the NRC inspection team. j

(6) EQWP ESE-49A, Barton, Model 581-4, Differential Pressure

Indicating Switch

l

(a) The NRC inspection team reviewed EQWP ESE-49A which included '

an accuracy requirement of 10 percent. This 10 percent

accuracy requirement documented in ECWP ESE-49A was found

to be inconsistent with the intent of the test and should

not have been included in the file.

The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, stated that the EQWP would be revised to

delete the 10 cercent accuracy requirement for needed

clarification of the file.

No violations, deviations, or unresolved items were identifieo

by the NRC inspection team.

4D

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- _ _ _ _ _ _ .-

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.

. .

. .

21

(b) The test report in EQWP ESE-49A en RVLIS hydraulic isolator

documents i problem with qualification of the hydraulic

isolator far harsh environment applications. The Barton

differential pressure indicating switch was found qualified

only for hydraulic isolation as documented in EQWP ESE-49A.

The switching mechanism of this switch does not perform a

safety related function and is not qualified for harsh

environment applications at WCGS.

Also, the performance requirements for the RVLIS hydraulic

isolator documented in EQWP ESE-49A does not address

sensing line errors and did not include performance data.

Hov:ever, WCAP-8687 is referenced in EQWP ESE-49A. The

performance requirements, Section 3.2 of WCAP-8687,

Supplement 2-E49A, address the performance of the pressure

switch as a hydraulic isolator. It includes sensing line

errors and consideration for use of this switch in other

applications (where its switching function is utilized).

The performance data taken during the qualification testing

of the pressure switch is available at Westinghouse. This

data was recorded via strip chart recorders, but because of

the voluminous nature of the raw data, and that most data

taken were to be used to support qualification in other

applications, the data was not included in the report. The

summary statements found in Section 7.0 of the WCAP-8687

supplement are based on review of this raw test data.

No violations, deviations, or unresolved items were

identified by the NRC inspection team.

(c) The NRC inspection team had a concern with regard to the

licensee providing an auditable link between the tested and

installed configuration of the RVLIS hydraulic isolator.

Investigation into the file revealed a documentation

inconsistency. The EQ data package (contained in

WCAP 8587, Supplement 1) documents the testing of Barton

pressure switch Model 581-4 whereas the EQ test report and

EQWP ESE-49A documents Model 581.

The licensee letter WM 87-0309 to NRC Region IV dated

November 1987 stated that the hydraulic isolator of both

Models 581 and 581-4 are identical. Improvements made by

Barton in the switching mechanist are incorporated in

Model 581-4. Westinghouse has coritrmed that Barton

pressure switch Model 581-4 was tested. EQWP ESE-49A will

be revised to reflect the co-rect model number. The

improvements of the Mooel 581-4 are to be documented in the

EQF.

No violations, deviations, or unresolved items were

identified by the NRC inspection team.

.. ..

,

. ,

. .

22

(d) The NRC inspection team had a concern with regard to the ,

tested configuration of the RVLIS hydraulic isolator as -

documented in EQWP ESE-49A not matching the installed

configuration. The NRC inspection team found that the test

configuration is only typical of the plant-installed

configuration for the equipment tested. As a hydraulic

isolator, the pressure switch is located in the sensing

line between a transmitter and the process line measured.

For the measurement of the pressurizer level (RVLIS

function), two pressure switches are used, one each in both

of the transmitters process line connections (loops A

and B).

The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, responded to this concern in tnat a'i

error in tne EQ test report, fcund in WCAP-8687,

Supplemer.t 2-E49A, Revision J Section 3.2, page 5 was

identified. EQWP ESE-49A would be revised to reflect

correct information, regarding verification, during the

test, by monitoring the output of the correct transmitter

(loop B) receiving the pressure signal. The purpose of the

loop A test configuration was to gather performance data

relevant to the pressure switches performance in other

applications. Loop A was intended to confirm that the

ptocess line measurement by the nressure switch was

independent of variations in process fluid volume. The

data taken for this configuration was via the indicating

switch electrical output and is considered to be relevant

to the pressure switch's performance of functions other

than as a hydraulic isolator.

This item is considered an Open Item (482/8724-10).

(7) EQWP E-035B, Kulka Terminal Blocks, Used in Control Circuits i

Inside Containment i

The NRC inspection team reviewed the application of Kulka

terminal blocks in control circuits inside containment. Conax

l

Qualification Test Report on Kulka terminal blocks, IPS 675,

! 1981, is used as the basis for qualification. Test Report

IPS 675 documents a maximum leakage current of 0.16 milliamps at

539 volts during LOCA testing.

The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, indicated Kulka terminal blocks located in

the electrical penetrations and terminal boxes, are qualified by i

!

EQWP E-035B. They are used in the control circuits of inside

containment safety-related eouipment qualified by the following

EQWPs and to the effects of the following leakage currents as

-

evaluated by the licensee:

.-

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. - .

. .

23

(a) Limitorque Operators (EQWF HE-1 and B0P-Limitorque)

The motor starters for tha Limitorque operator are either

size 1 or size 2. Size 1, the worst-case, requires

1.26 ampere to pick up its coil and draws 0.185 ampere

during continuous uperation. Leakage currents of

0.16 milliamps will not affect the operation of the starter

coils.

(b) ASCO Series 8316 Control Valves (EQWP HE-2)

ASCO series 8316 control valves have power removed from

them on a containment isolation signal. Therefore, leakage

current during an accident condition does not apply.

(c) Target Rock 79AB Series Control Valves (EQWP HE-10A)

These solenoid valves, which ara used in head vent, excess

letdown, and accumulator vent applirrtions, draw 0.8 amps

at 125VDC. Leakage currents of 0.16 milliamps will have no

, affect on these vilves.

(d) Bettis Actuators (EQWP M-237]

Bettis actuators control the containment minipurge

isolation valves. The isoistion valves are closed by

removing power from ASCO stjenoid valves which are

appurtenances to the Bettis actuators. Since power is

remeved from the ASCO valves during accident conditions,

leakage currents during accident conditions do not apply.

(e) Valcor 52600 Series Control Valves (EQWP J-603A)

These valves are used as process sarnple line valves and as

containment isolation valves. In accordance with

Instruction Manual J603A-00080 for these valves, the

solenoids draw 1.5 ampere at 120VAC. Leakage current of

0.16 milliamps will not affect th( operation of these

valves.

The leakage currents identified by the licensee in l

paragraphs 4.f(7)(a) through 4.f(7)(e) above, regarding Kulka

terminal blocks qualified by EQWP E-035B, are sufficiently small

I

during LOCA conditions so as to not affect the safety-related

control circuits inside containment at WCGS. EQWP E-035B will

be revised to reflect this information in the EQF.

1

No violations, deviations, or unresolved items were identified i

by the NRC inspection team. i

,

_____ _______________________ ______ _ ___ _. _ _ __ ________-________________________!

. " .

. .

24

(8) EQWP E-028, Marathon Terminal Blocks, Model 1600

These terminal blocks are used inside and outside containment in

Limitorque motor operator limit switch circuits including valve

position indicators. They are also used in control circuits for

pressurizer power operated relief valves (PORVs) Applied

voltage is limited to 120V. The licensee stated that operation

of circuits which have Marathon 1600s is not affected by leakage

currents of up to 300 milliamps. The file contained Wyle

Laboratories Test Reports 45603-1 and 17657 as the basis for

qualification.

(a) In test report 17657, leakage currents were weasured during

LOCA testing. Leakage currents did not exceed the

300 milliamps criteria given by the licensee. The circuits

were always energized during testing.

(b) In test repnrt 45603-1, insulation resistance was not

measured during the LOCA testing. Leakage currents were

monitored indirectly by placing a fuse in the test circuit

such that leakage curren*.s greater than the fuse rating

would open the fuse. A 12-amp fuse was used in the 132V

test to monitor the leakage currents. After a power

fa;1ure during testing had been corrected, the facility

power was abruptly reapplied. Upon reapplication of power,

the 12-amp fuse optned. The NRC inspection team's concern

with this file is ti.at the 12-amp fuse was blown when the

testing e.rcuit was reenergized. Also, testing the

terminc1 block circuitry under r.onstant power is not the

worst case as demonstrated by the interruption of facility

power. Terminal block cir::uitry in containment may

experience similar power profiles as experienced in the

test, and this test has demonstrated an unacceptable

leakage current during a power transient. The

Marathon 1600 file indicates that unacceptably high leakage

currents may occur during power transients, and documented

test results do not demonstrate tha' equipment can perform

its required function for all postulated harsh

environment.

The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, responded to the NRC inspection concern

by addressing specifically in detail all aspects of the

test anomaly described above, in Wyle Test Report 45603-1.

,

'

The licensee did not agree with the dispositioning of this

aromaly as contained in the test report as follows:

1) The Wyle test setup, described in the report, applied

voltage and current to a serpentine arrangement and

measured total leakage current from six terminal

, blocks to ground, of which four of the blocks are not

.

, *. .

,

.' .

25

!

used at WCGS. Because of this test setup, it was not

possible to determine positively the source of the

leakage current. <

,

2) The 132 VAC terminal blocks in the Wyle test were

allowed 12 amps leakage current, however, the leakage

current was a summation of the leakage from 54 possible  :

paths to ground. The fuses were not examined during

'

the power outage, but were only discovered to be blown

upon restoration of power. It was highly probable the

fuses blew upon loss of power when an inductive voltage

transformer at the laboratory generated high voltag?

'evels.

The licensee further stated that justification for the

acceptability of Marathon 1600 terminal blocks for WCGS

application is contained in Wyle Test Report 17657.

The licensee will be revising EQWP E-028 to contain an

aralysis which would reflect a better dispositioning of the

Wyle test anomaly.

This item is considered a violation of 10 CFR 50.49

(482/8724-04).

(9) E0WP ESE-43C, ESE-43E, and ESE-43G, Westinghouse Cable Splice

Assemblies

Westinghouse hardline potting adaptor cable splice assemblies are

used for thermocouple splicing above the reactor vessel head and

at the reference junction box. EQWP ESE-43C applies to these

assemblies before the second refueling outage, and EQWP ESE-43E

and ESE-43G will be in ef fect af ter the second refueling outage.

The EQF contained Westinghouse Test Reports WCAP-8687,

Supplement 2-E43C, WCAP-8687, Supplement 2-E43E, and WCAP-8687,

Supplement 2-E43G so support qualification of these cable splice

assemblies.

During type testing, Westinghouse used a gamma dose cf

8.6 x 10' rads to simulate a required lifetime beta dose of ,

4.15 x 10' rads. The equivalent gamma dose was determined by an  ;

analysis which took credit for design shielding at the location

of the assemblies. The report did not describe the analysis or

the shielding used. The utility supplied the NRC inspection team

with additional information and analysis with regard to the

shielding requirements, and this concern was resolved.

WCGS will incorporate the additional documentation in the

appropriate EQWPs of the EQF.

1

No violations, deviations, or unresolved items were identified by

the NRC inspection team.

. *. .

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26

(10) EQWP E-062, Boston Insulated Wire (BIW) 600V Cable

Ethylene propylene rubber insulation /chlorosulfonated

polycthylene jacket (EPR/CSPE) 600V cable is used for instrument

applications inside containment. The file contained BIW test

document No. 10466-E-062. A licensee report wnich addr:ssed

instrumentation accuracy requirements was submitted separately

to the NRC inspection team. The report indicated that the BIW

cable used in instrumentation circuits produced acceptable

accuracies for their applications. The licensee should

incorporate this report and any additional information to

establish acceptable accuracies for their applications, into

EQWP E-062 in the EQF.

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

(11) EQWP E-061, Eaton (Samuel Moore) Thermocouple Extension Cables

This cable consists of ethylene propylene base (EPDM) insulation

and Hypalon jacket. The file contained an Isomedix test report

dated June 1978 to establish the radiation qualification as

applicable to WCGS. The NRC inspection team identified the

following concern. The licensee reduces the beta dost by taking

credit for jacket shielding. The Isomedix test report did not

demonstrate that the jacket survived the radiation test intact.

An addendum to the June 1978 report was supplied to the NRC

inspection team which did tate that the cable jacket was intact

aftte testing. This addendum (September 19. 1978) will be added

to EQWP E-061 in the EQF.

No violations, deviations, or unresolved items were identified l

by the NRC inspection team. ,

g. IE Information Notices and Bulletins

The NRC inspection team evaluated the licensee's activities related

to the review of EQ-related ins and IEBs.

Procedure KGP-1311, Revision 0, controls the licensee's Industry

Technical Information Program (ITIP) which inr.ludes handling:

(1) ins and IEBs; (2) industry technical information; (3) INP0

Significant Operating Experience Reports; and (4) NSSS Technical

Bulletins. The ITIP did not contain guidelines to determine whethe-

an issue has EQ implications when not directly applicable to WCGS.  ;

Personnel who screen this information are not required to have EQ  ;

training (even overview), so that issues with generic aspects that '

might be pertinent to WCGS are not likely to be recognized. The  !

licensee stated that this proceaure is being revised to route all  !

such correspondence to EQ personnel in the Nuclear Plant Engineering

Department, which was being implemented at WCNOC. The licensee

, <. .

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27

stated in their letter WM 87-0309 to NRC Region IV deted November 20,

1987, that personnel who initially screen ITIP information would

receive EQ training by June 30, 1988.

The NRC inspection team reviewed the status of actions on EQ-related

ins and verified implementation of WCGS's program for processing and

tracking NRC IEBs and ins as they relate primarily tr EQ. The

licensee had reviewed and evaluated appropriate EQ-related ins and

IEBs from IEB 79-01 through IN 87-08. t'arrative summaries of the

evaluations of IE8s and ins in the data base printout were reviewed.

Actions pertaining to selected ins were reviewed in detail and

are discussed below:

(1) IN 86-53, Improper Installation of Heat Shrinkable Tubing

(a) WCGS actions regarding this IN were reviewed in accordance

with Temporary Instruction (TI) 2500/17. The file on

IN 86-53 indicated that the licensee had initially

determined that the Raychem installation procedures in use

at WCGS were consistent with the manufacturers

instructions. The file also contained a KG&E internal

memorandum, dated December 5, 1986, which discussed the ,

December 1986 meeting of the Electric Power Research

Institute ('cPRI) EQ Advisory Group in which Raychem issues

were addressed. It stated that utilities had foud

improper splices despite having proper procedures and noted

that WCGS EQ documentation did not address impre.erly

installed splices.

WCGS's failure to maintain equipment in a qualified

configuration or provide acquate documentation to qualify

splices is identified as a violation to 10 CFR 50.49 as

discussed in paragraph 4.f(1) of this report.

(b) It was not clear frcm the IN 86-53 file that WCGS had

performed subsequent walkdowns to resolve the above issue

until improper splices were found by NRC inspectors during

this inspection as discussed in paragraph 4.h(1) of this

report. Subsequent to the inspection, WCNOC provided

additional information on this issue, in response to NRC

concerns, and stated that they had now conducted a walkdown

of 150 splices in harsh accident environment areas. Most

fell within Raychem specification limits or within the

bounds of those qualified in recent testing at Wyle

Laboratories (Project 17859-2). WCNOC had obtained the

test reports through the Nuclear Utility Group on

l EQ (NUGEO). Deficient splices including those identi'ied

by the NRC were replaced. While reworking these splices,

WCNOC also found some bolted splices in which the

heatshrink tubing was determined to be too small for the

application. This condition was reported to the NRC in a

November 18, 1987, Licensee Event Report (LER).

. . a

,

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28

WCGS's failure to maintain equipment in a qualified

configuration or provide adequate documentation te qualify

splices is identified as a violation to 10 CFR 50.49 as

discussed in paragraph 4.f(1) of this report.

(c) The post-inspection information also detailea WCNOC's plans

to inspect 100 percent of remaining Raychem splices in

instrument circuits to verify that they are acceptable per

Raychem recommendations. WCNOC also reported that they are

developing a sample plan based on the total population of

Raychem splices in control circuits. Acceptance criteria

for the control circuit! will be based sn tested

configuration of splices reported in Wyle Report

No. 17859-028. Power circuits were qualified in a

different configuration and are considered by WCNOC to be

outside the scope of this sample plan inspection program.

(d) The NRC inspection team reviewed Procedure INC S-0506,

Revision 2, "Wire (.plicing and Termination with Raychem,"

dated August 11, 1987. Previous revisions, effective when

most splices would have been installed, were unavailable

to the NRC inspection team for review during the

inspection.

No violations, deviations, or unresolved items were

identified by the NRC inspection team.

(e) Raychem Application Guide WCSF-N, Revision 1, was listed as

a reference, but not referred to in votedure INC S-0506.

A precaution was listed to adhere to minimum bend radius

standards for field wire during installation in or on any

terminal; however, minimum bend radius requirements for

splices were not covered. The procedure required

cable / wire surfcces to be cleaned, but there was no

requirement to pill back braided jacketing and ;nstall

splices only on smooth nonwoven surfaces. The procedure

contained a table that specified total required length of

splices based oniliameter. The procedure also requires

shims to be cut ar. least 2 1/4 inches long for accident

condition applica& ions, at least 1 1/8-inch for nonaccident

condition applications, and requires that the entire length

of the shim be applied to a suitable sealing surface to

maintain a minimum seal length. The NRC inspection team

determined (a) what was suitable was not defined;

(b) minimum seal length was not defined; (c) the tern.s

"accident" and "nonaccident conditions" were not defined

including how the determination was made was not specified;

and (d) minimum shrink overlap length for the joint between

_

, . .

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29

any sealing component of the splice (not just shims) was

not specified. The procedure included a section on

installation of heatshrink on a bolted V-splice. It is not

clear that the V-splice would be a qualified configuration

in that it relies on inserting filler material between the

wires and then a heatshrink end cap over them, in lieu of

using a breakout boot, to seal around the adjacent wires as

they exit the splice, as used in the standard Raychem kit

for V-splices. Without specifically requiring use of

Raychem procedures to make up for these deficiencies, this

procedure is considered inadequate for installation of

solices in accordance with Raychem specifications for

Raychem qualified configurations.

WCGS's failure to maintain equipment in a qualified

configuration or provide adequate documentation to qualify

splices is identified as a violation to 10 CFR SC.49 as

discussed in paragraph 4,f(1) of this report.

(2) In 8E-03, Unidentified Internal Wiring in Limitorquq Motor

Ogerators

r

WCGS acticns regarding this IN was reviewed in accordance with

TI 2515/75. Review of the file on this including licensee

walkdown and maintenance records indicated that WCGS hao

conducted plant walkdowns and replaced all unidentified wiring '

with qualified wiring. No discrepancies in this area were noted

in the NRC's plant physical walkdown inspection of selected

Limitorque switch compartment internals.

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

i

(3) IN 84_-90_u_Superheated Steam Release During a Main Steam Line

_

Break (MSLB)

i

'

This issue has seen analyzed in detail at WCGS and several

formal submittals have been made to the NRC. The issue is

,

currently under review by the NRR, and the qualification of <

various components subject to the superheated steam environment I

will remain unresolved pending completion of the review. WCNOC l

has prepared JCOs for affected equipment. The NRC inspettion

team found them to adequately address safe interim operation,

but most of the analyses are not sufficient to demonstrate final

qualification.

This item is considered an Unresolved Item (482/8724-06). .

l

(4) IN 86-71, Energized Space Heaters in Limitorque Valve Operators I

The licensee's resoonse to this concern addressed the issue of

the qualification of heater lead wires and heater failure modes

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .___-_ ______

, ., ..

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and effects. Reduction in qualified life for valve operators,

as found by the NRC inspection team during this inspection, are

discussed in paragraph 4.f(2) of this report, and is identified

as a violation to 10 CFR 50.49.

(5) IN 86-104, Unqualified Crimp Type Connectors

The licensee's response and evaluation of this issue addressed,

in particular, Thomas & Betts "blind-barrel" nylon crimped

connectors used in Limitorque valve operators with dual voltage

motors. Insufficient documentation for qualification, as found

by the NRC inspection team during this inspection, are discussed ,

in paragraph 4.f(4) of this report, and is identified as an  !

Unresolved Item.

h. Plant Physical Inspection

The NRC inspection team walked down and physically inspected  ;

approximately 47 components. The inspection team examined attributes

and characteristics such as mounting configuratica, orientation, j

interfaces, model numbers, ambient environment, and physical condition.

(1) Raychem Splices to Pressure Transmitters

During the plant walkdown, the NRC inspectors identified Raychem

splices that were not installed according to Raychem

specifications, in configurations not qualified under existing

WCGS EQ documentation. WCNOC's failure to maintain this

aquipment in a qualified configuration or provide adequate i

documentation to qualify splices is identified as a violation to

10 CFR 50.49 as discussed in paragraph 4.f(1) of this report.

l

(a) Tobar Model 32PA1212 pressure transmitter (Plant 10

No. BB-PT-0455, pressurizer pressure) is located in the

containment. The pigtail lead Raychem splices were found .

bent with less than the Raychem-specified minimum bend

radius (5 x splice 0.D.) and had less than 2 inches of '

seal. The EQWP ESE-lC file listed only model Nos. 32PA1

and 32PG1; however, the installation test data sheet listed

model No. 32TA1212. Inspection of the installed

transmitter indicated the "TA" on the test data sheet had

been recorded in error and information provided by the

licensee resolved differences in documented and installed

model numbers. The numbers following "pal" simply referred

to different instrument application ranges and should not

affect DBA performance. (Refer to paragraph 4 f(5) for EQF

EQWP ESE-1C review.)

,

(b) ITT Baron Model 764 level (differential pressure)

transmitter (Plant ID No. BB-LT-0459, pressurizer level) is ,

located in the containment. The pigtail lead Raychem l

l

, .. N

f

e' .a

31

i

splices had no adhesive protruding from the heatshrink ,

tubing ends. This appears to be indicative of

under-shrinking. The splices were also bent through 90

angles with much less than 5 x splice 0 D. bend radius, and ;

seal lengths were less than 2 inches.

(c) ITT Baron Model 764 pressures transmitters (Plant ID

Nos. AB-PT-0514 and AB-PT-0544, steam generator steamline l

pressure) are located in the steam tunnel. The pigtail

lead Raychem splices had no adhesive protruding from some

of the sleeve ends.

(2) Minco RTO Installed Configurations

The NRC inspection team walked down the Minco Model S8809 RTDs

(Plant ID Nos. TE-1318 and TE-1327 for the reactor vessel level

indication density compensation system) located in the

containment. The documented configuration of Minco RTDs in

EQWP ESE-42A did not match the installed configuration. EQWP

ESE-42A documents that the terminal boxes used as splice boxes

for Minco RTD cables in the Reactor Vessel Level Instrumentation

System are to be locat'd above maximum postulated containment

flood level. The EQi;P does not document qualification for

submergence of the terminal boxes. The NRC inspection team

determined that several cable termination end seal assemblies of

these RTDs in terminal boxes TB-23208 and TB-23210 (at the

2004 foot level as measured against wall markings and confirmed

by Bechtel Drawing No. E-1R23220(Q)) were located below flood

level. The documented DBE flood level in the area is 2004 foot

6 inches indicating that the potted end of the seal assemL11es

and che external lead wires in the junction boxes would be

subject to submer;ence under accident conditions during which

they would be required to function.

The 'icensee letter WM 87-0309 to the NRC Region IV Office dated

Novemt,e r 20, 1987, indicates that the terminal boxes have been

relocated above flood level consistent with configuration

documented in EQWP ESE-42A and evaluation of the previously

installed configuration by the licensee is beino conducted. ,

The unqualified condition of these RTD configurations is

considered a violation to 10 CFR 50.49. (482/8724-02)

'

(3) Limitorque Motor Operator (EQWP-80P Limitorque, HE-1, and HE_-4j

Compartment Degradation Because of Energized Heaters

Refer to paragraph 4 f(2) of this report.

.

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..,a.

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32 l

l

(4) Limitorquc Motor Operator Greasas (EQWP HE-4) Outside

Containment

Two valves whose qualification was covered under WCAP-8687,

Supplement 2-H04A (EQWP HE-4), EJ-HV-8804A, and EJ-HV-8804B were

identified by WCGS personnel as having questionable greases. A

review of the maintenance records revealed that the greases of

, these two valves had not been changed out and discussions with

licensee personnel disclosed that no schedule had been

established to replace the greases. It should also be notad

that these valves are NUREG-0588, Appendix E, Category C type

equipment. Category C equipment is equipment that will

experience environmental conditions of DBAs through which it

need not function for mitigation of said accidents, and whose

failure (in any mode) is deemed not detrimental to plant safety

or accident mitigation, and need not be qualified for any

accident environment, but will be qualified for its nonaccident

service environment.

The licensee letter WM 87-0309 to NRC Region IV dated

November 20, 1987, assured grease change out on the identified

Limitorque operators to be completed prior to startup from the

second refueling outage.

i

'

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

(5) Limitorque Motor Operator (EQWP HE-4) Outside Containment,

Damaged Motor Leads

1

Two motor leads had been camaged and subsequently repaired with

tape in an SMB-000 series operator with mark number EMHV-88218.

The damage had apparently resulted when the wires were caught

between the limit switch compartment and cover during the cover ,

installation. It could not be determined during the NRC "

inspection if the defect had been reported, analyzed, and

j repaired by an approved method.

, The licensee letter WM 87-0309 to NRC Region IV dated

l November 20, 1987, indicated a WCGS investigation which revealed

an open work request on EM-HV-8821A. Pending completion of the

repair, black electrical tape was wrapped around the damaged

wire for personnel safety. A revision to the work request was

"

issued October 12, 1987, to replace the damaged motor lead.

,

No violations, deviations, or unresolved items were identified  ;

{ by the NRC inspection team. '

I

-. . _ - _ _ _. -.

_

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.' . .'

33

(6) Delaval Containment _ Sump Level Switch LF-LE-0009A

Model XM-54852 (EQWP J481)

Ouring on NRC walkdown, it was observed that level switch

.

LF-LE-0009A junction box had a locse connection at the switch

tube permitting silicon fluid leakage.

The licensee letter WM 87-0309 to NRC, Region IV dated

November 20, 1937, indicated repair on Level Switch LF-LE-0009A

has been completed. Other containment sump level switches would

be inspected and repaired, as required prior to startup from the

second refueling outage.

No violations, deviations, or unresolved items were identified

by the NRC inspection team.

6. Open Items

Open items are matters which have been discussed with the licensee, which

will be reviewed further by the NRC inspector, and which involve some

action on the part of the licensee.

6. Unresolved Items

Unresolved items are matters about which more information is required in

order to ascertain whether they are an acceptable item, an open item, a

deviation, or a violation. These unresolved items, which if ascertained

to be a violation, may be followed up with enforcement action in

accordance with NRC enforcement guidance on 10 CFR 50.49.

7. Exit Interview

An exit interview was conducted with WCGS personnel on October 23, 1987,

at the conclusion of the onsite inspection during which the scope of the

inspection findings were summarized.

.