ML20148Q914
| 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: |
| References | |
| RTR-NUREG-0881, RTR-NUREG-881 50-482-87-24, NUDOCS 8804140033 | |
| Download: ML20148Q914 (33) | |
See also: IR 05000482/1987024
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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,
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
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
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 :
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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 '
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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
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a
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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
!
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. -
.
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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.
__
_-
.
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.
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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
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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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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.
.. ..
,
. ,
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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
,
_____ _______________________ ______ _ ___ _. _ _ __ ________-________________________!
. " .
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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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(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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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).
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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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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
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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
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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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(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
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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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(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.
.