ML20236N637
| ML20236N637 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 08/05/1987 |
| From: | Gardner R, Gautam A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236N617 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 50-483-87-15, NUDOCS 8708120042 | |
| Download: ML20236N637 (19) | |
See also: IR 05000483/1987015
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III'
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'eport No. 50-483/87015(DRS)
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' Docket No. 50-483-
License No. NP P-30
Licensee:
Union Electric Company.
Post' Office Box 149
St. Lcais, M0' 63166
s.
Facility Name:
Callaway Nuclear Power Plant
Inspection At:
Callaway Site, Ca11away' County, Missouri
Inspection Conducted:
May 4 through July 22, 1987
M
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Insp'ector:
Anil S. Gautam
Reactor. Inspector, Region III
Date
Also participating in the inspection and contributing to the report were:
R. Smeenge,-'RIII
S. Alexander, NRR
R. Stoffel, INEL
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A. Udy, INEL
V._Nicolette, Sandia
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Approved By:
Ronald N. Gardner, Section Chief
,815- b
Plant Systems Section, Region III
Date
~ Inspection Summary
~ Inspection on May 4 through July 22,-1987 (Report No. 50-483/87015(DRS))
Areas Inspected:
Special announced safety inspection of the environmental
qualification (EQ) of electric equipment within the scope of 10 CFR 50.49 and
Regulatory Guide 1.97, Revision 2.
The inspection included licensee action on
'SER/TER commitments; EQ program compliance to 10 CFR 50.49 and requirements of
Regulatory Guide 1.97; adequacy of EQ documentation, and a plant physical
inspection of.EQ equipment (Module Nos. 30703, 25576, and 25587).
Results:
The licensee has implemented a program to meet the requirements
of 10 CFR 50.49 and Regulatory Guide 1.97.
Deficiencies in the areas
inspected are' summarized below:
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8708120042 870006
ADOCK0500g3
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V_IOLATION
Item No.-
Description
Report Section
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50-483/87015-05(DRS)
Lack of qualified moisture
2a
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(No response required.) seals on Target Rock
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solenoid valves' electrical
connections
POTENTIALLY ENFORCEABLE / UNRESOLVED ITEMS
Item No.
Description
Report Section
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50-483/87015-02(DRS)
Marathon 1600 terminal blocks
6a
found unqualified for 120 VAC
10 CFR 50.49 designated circuits.
50-483/87015-03(DRS)
KULKA terminal blocks found
6b
unqualified for 480 VAC and
120 VAC 10 CFR 50.49 designated
circuits.
OPEN ITEMS
50-483/87015-01(DRS)
NRR review of the licensee's
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evaluation of superheated steam
releases outside the containment and
its effect on 10 CFR 50.49 equipment
(IE 84-90)
50-483/87015-04(DRS)
EQ file on Struthers-Dunn relays to be
6d
clarified and supplemented by additional
qualification infonnation.
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DETAILS
1.
Persons Contacted
Union' Electric Company (UE)
- D. F. Schnell, Vice-President, Nuclear
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- J. D. Blosser, Manager, Callaway Plant
- G. L. Randolph, General- Manager,' Nuclear Operations
R.-J. Schukai,-General,Manoger, Engineering
C. D. Nashind, Manager-0perations Support
- A. P. Neuhalfen, Manager-QA
D. W. Capone, Manager,'. Engineering
A. Peevy, Assistant Manager, Technical Services
- K. W. Kuechenneister, Assistant Manager-Materials
- J. C. Gearhart, Superintendent-0peration Support
- W. H. Sheppard, Superintendent-Outages
- J. Lanx, Superintendent-Technical Support
- M. S. Evans, Superintendent-Training
- W. R. Robinson, Superintendent-I&C
- J. E. McLaughlin, Superintendent-Administration
- M. E. Taylor, Superintendent-0perations
T. McFarland, Superintendent-Design Control
- G. J. Czesdin, Superintendent-P&S
- A. C. Passwater, Superintendent-Licensing
- D. E. Young, Superintendent-Maintenance
- B. L. Walton, -Assistant Superintendent-Admin.
- G. N. Belchik, Assistant Superintendent-Materials
T. Sharkey, Supervisor-Compliance
- R. E. Carlson, Supervisor Engg-Materials
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K. R. Evans, Supervisor Engg-I&C
- R..A. Hamblen, Supervising Engineer
T. L. Shaw, QC Supervisor
+*D. E. Shafer, Supervising Engineer
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- C. E. Slizewski, Supervising Engineer-Technical Support
- R. S. Nelson, Maintenance Planning
D. R. Miller, I&C Engineer
J. W. Ostroot, Engineer
- J. J. Cassmeyer, Assistant Engineer
- G. G. Yates, Engineer
- W. A. Norton, QA Engineer
- R. J. Davis, PM Engineer
- R. G. Hainer, Engineer
- E. Kahl, Assistant Engineer
Consultants
- B. J. Metro, Westinghouse-Engineering Aide
- D. R. Prichard, WCN0C - EQ Engineer
- J. G. Utt, Bechtel - Senior Engineer
- M. H. Fletch r, CFA, INC
A. Hason, Bechtel
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Nuclear Regulatory Commission (NRC)
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- C. Brown, Senior Resident Inspector
- Denotes'those attending the interim site exit meeting on May 8, 1987.
+0enotes those participating in the exit meeting conducted by telephone
on' July 22, 1987, at the conclusion of the inspection.
2.
Licensee Action on Previous EQ Concerns
a.
(Closed) Unresolved Item (50-483/86007-01A(DRP)).
fhis item addressed
NRC concerns regarding the lack of adequate sealing of the electrical
connections to Target Rock solenoid valves BB-HV-8001 A and B, and
BB-HV-8002 A and B.
These valves are used for venting the reactor
vessel head during an accident.
On February 20, 1986, the licensee
declared these valves inoperable.
Subsequent to this finding, the
licensee took immediate corrective action and installed qualified
CONAX ;3als on the conduit connections of these valves as required
by the EQ documents.
The licensee also provided an operability
analysis, additional EQ documentation, and a similarity analysis to
indicate that the valves in question would perform their safety
function.
The operability analysis indicated that the accident
environment would be less severe than specified; that the valves had
a Thomas Betts " Tite Bite" connector with silicon rubber RTV sealant;
and that the conduit entered a gasketed enclosure from the bottom.
Additional EQ Documentation included tests of solenoid valves
qualified without seals and a similarity analysis (UL NRC-1291,
April 11, 1986) of the tested and installed valves.
During this current review, the inspectors verified the licensee's
corrective action and reviewed the applicable EQ files for sealing
requirements.
The licensee stated that the corrective action taken
would avoid further violations in regard to the sealing of electrical
connections beause the correct sealing requirements are currently
stated on the installation drawings, in the EQ files, and are part of
the EQ maintenance program.
10 CFR 50.49 Paragraph (f) requires equipment important to safety to
be qualified by testing and/or analysis.
Contrary to the above, four
Target Rock solenoid valves were found in an unqualified configuration
past the November 30, 1985 EQ deadline.
The licensee's corrective
action has been reviewed and found adequate.
This is considered a
Severity Level IV violation, Supplement ID, (50-483/87015-05(DRS)) as
discussed in the Appendix.
No response is required.
b.
(Closed) Unresolved Item (50-483/86007-01B(DRP)).
This item addressed
the lack of qualified seals in Target Rock solenoid valves used in the
CVCS line.
The licensee performed a review which indicated that in
the event of any failure astciated with the electrical connections
of these valves, the valves would fail to an accident safe position,
thereby performing their safety function.
No further concerns were
identified.
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c.
(Closed) Unresolved Item (50-483/86010-02(DRP)). This item addressed
NRC concerns regarding a lack of records to evidence sealing of
flexible conduits connected to various Limitorque Valve Motor
Actuators. The licensee took immediate corrective action and applied
sealant to the appropriate conduit connections as required by design
drawing M-24007
After review of their EQ files the licensee determined that a lack
of sealing on the conduit connections did not affect the
environmental qualification of the Limitorque Actuators. This was
confirmed from review of Limitorque Reports B0058, 600456, and B0119
where flex conduit used was not sealed, and where the configuration
allowed drainage through the motor. No further concerns were
identified.
3.
(Closed - TI 2515/87) Review of Licensee Implementation of Regulatory
Guide 1.97
The inspectors reviewed the licensee's effort in qualifying Regulatory
Guide 1.97 equipment within the scope of 10 CFR 50.49 Paragraph (b)(3).
During this review, the inspectors observed that the licensee had
submitted their Regulatory Guide 1.97 responses on April 15, 1983, and
August 16, 1984, to NRR, and the NRC had issued a final SER in regard to
the Callaway Plant (Unit 1) conformance to Regulatory Guide 1.97 on
April 10,1985.
This inspection was conducted in accordance with the requirements of NRC
Temporary Instruction 2515/87. Areas reviewed included verifying the
adequacy of the licensee's Regulatory Guide 1.97 list and inclusion of the
Regulatory Guide 1.97 Category I and II items in the licensee's
10 CFR 50.49 list; verification that selected Regulatory Guide 1.97 items
had redundancy, physical separation, isolation, and uncompromised power
supplies; examination of selected Regulatory Guide 1.97 items in the plant;
detailed review of appropriate EQ files for Regulatory Guide 1.97 items; and
verification of maintenance and surveillance activities performed on
Regulatory Guide 1.97 items installed in the plant. The following drawings
were reviewed:
DRAWING LIST
Drawigg
Drawing No.
Type
Title
M-22BN01Q)
Borated Refueling Water Storage System
E-23BN07 Q)
S
Miscellaneous Instruments
8756D37 sh38
PC
Refueling Water Storage Tank Level
Safety Class 1E Differential Pressure
Electronic Transmitter Qualification
Group B
M-22GN01(Q)
Containment Cooling System
OP-E-3GN05(Q)
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Miscellaneous Instrumentation
8756D37 sh40
PC
Containment Pressure Wide Range
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M-22AB01(Q)
Main Steam. System
E-23AB21(Q)
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Miscellaneous. Circuits
E-23AB28(Q)~
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Control-Active Side
8756D37 sh13
PC-
Steam. Generator Loops'1,.2, 3,
and 4, Pressure Protection Set'I
7250064 sh7
F
Steam Generator Trip Signals
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7246D92 sh2-
EC
Process Cabinet 1 Protection Set I
OP-E-03SE07(Q)~
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Ex-Core Neutron' Flux Monitoring System
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Detector Units NE 60 and NE 61
900030
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Neutron Flux Monitoring System
(Gammametrics)
M-22LF09(Q)
'P&ID
. Reactor Building and Hot Machine Shop
Floor _and Equipment Drain System
OPE 23LF12(Q)
S.
Instrumentation and Annunciation-
ILD
RHR System Containment Recirculation
Sump B Level
57765
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Delaval Installation Specifications,
Transmitter Wiring and Mounting
M-22BB01(Q)
E-23BB15(Q)
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Instrumentation
8756037 sh5
PC
Wide Range Reactor Coolant Temperature
(Hot and Cold Legs)
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7250064 sh17
F
Pressurizer Pressure. Relief System
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E-03NN01(Q)
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Class IE Instrumentation AC Schematic
Diagram
E-21010(Q)
SLD
DC Main Single Line Diagram
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P&ID Process & Instrumentation Diagrams
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Schematic Diagrams
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Process Control Diagrams
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Functional Diagrams
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External Connection Diagrams
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. Assembly Diagrams
ILD
Instrument Loop Diagrams
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Installation Diagrams
SLD
Single Line Diagrams (Electrical Schematic)
The inspectors selected the following Regulatory Guide 1.97, Category I
variables and corresponding equipment for review:
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Variable
Equipment Tag No.
Range
RCS Hot Leg Temperature
TI-413Y.
0-700 F.
TI-413A
TR-413
Normal Sump Water Level (Containment) LI-9
0-156 in.
LI-10~
EJ-LR-6
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Refueling Water Storage Tank L'evel
LI-930
0'to 100%
LR-930
Steam Generator Pressure
P1514A
0 to 1300 psig
PR514
Containment Pressure-High Range
PI-938
-5 to +180 psig
PI-939
PR-938
-Neutron Flux-
SE NE-60A/B
10 8 to 200%
SE NE-61X/Y
Full Power
The inspectors reviewed the environmental qualification files, drawings,
. portions of the Callaway Technical Specifications, Final Safety Analysis
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Report, and Safety Evaluation Report to verify the qualification of the
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above equipment.
The equipment.was confirmed to be listed on the Master
Equipment List;Lthe' Category 1 equipment and variables were verified to
be in redundant channels and separated'in accordance with thel
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requirements of Regulatory Guide 1.75; and the power sources for the-
above equipment were from redundant sources.
The licensee also stated
that Regulatory Guide 1.97 instrumentation was not shed from their power
sources during an accident.
The use of signal isolation was verified on
process and. instrumentation diagrams (this prevents failure in
non-Class 1E instrumentation from degrading the indication for the
. audited variable).
During this review, the inspectors noted that surveillance and maintenance
is scheduled by the plant computer. The inspectors verified that the
plant. computer had the next surveillance scheduled, noted the frequency
of the surveillance, and identified the procedure number for the
appropriate work.
The inspectors also verified that the indicators and
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. recorders were tagged in the control room and had the ranges reported in
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During review of the environmental qualification files for the dentified
equipment, the inspectors noted that in June 1984, Westinghouse Electric
Company informed the SNUPPS utilities of a potential safety concern
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related to the analysis of equipment qualification following a postulated
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main steamline rupture with resultant superheated steam. releases outside
of containment (IE Information Notice 84-90).
Westinghouse concluded
that the peak temperature expected in the steam tunnel exceeded the
qualification requirements previously used for equipment in the steam
tunnel.
In regard to such an event the Regulatory Guide 1.97 main
steamline pressure transmitters would be required to operate outside of
their environmental qualification envelope.
A Justification for Interim
Operation was submitted to the NRC on October 2, 1984, to support the
licensing review of the SNUPPS units--Callaway and Wolf Creek.
On
April 4,1986, SNUPPS provided an additional submittal to the NRC
addressing main steamline break superheated steam effects on equipment
qualification.
This issue is being reviewed by NRR.
The Barton pressure
transmitters, Model No. 763, in the auxiliary building steam tunnel are
used to monitor main steam loop pressure and RCS wide range pressure.
The component numbers of the pressure transmitters affected by super
heated steam are as follows:
Main Steam Pressure
RCS Pressure
AB-PT-0514, AB-PT-0525, NJ-PT-0536
BB-PT-0403 and BB-PT-0406
AB-PT-0515, AB-PT-0526, AB-PT-0544
AB-PT-0516, AB-PT-0534, AB-PT-0545
AB-PT-0524, AB-PT-0535, AB-PT-0546
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The licensee stated that these steamline pressure transmitters perform
their safety function of providing a Steamline Isolation Signal well
before their qualified temperature of 420 F is exceeded.
The licensee
also stated that for long-term, post-accident monitoring these transmitters
may be unreliable; however, alternate steamline pressure transmitters
AB-PT-01, 02, 03, 04 could then be used to monitor this RG 1.97 variable.
The licensee stated that these alternate transmitters are Class IE devices
and that they are located outside the steam tunnel such that they are
unaffected by any superheated steam environment.
Union Electric committed to include a cautionary statement in their
operating procedures which would direct operators to rely on the
identified alternate instrumentation (transmitters AB-PT-1, 2, 3, and 4
for loops 1, 2, 3, and 4 respectively) should superheated conditions occur.
Pending further review of this issue with NRR this is an Open Item
(50-486/87015-01(DRS)).
4.
Licensee Action on SER/TER Commitments
The NRC inspection team evaluated the implementation of the licensee's EQ
corrective action commitments made as a result of EQ deficiencies
identified by the NRC in their May 84 NUREG-0830, Supplement 3, Safety
Evaluation Report (SER).
The majority of the deficiencies identified in
the SER addressed documentation, similarity, aging, qualified life, and
replacement schedules.
All open items identified in the SER were
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discussed with the NRC staff, and the licensee's proposed resolutions to
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these items were found acceptable by the NRC, as stated in their-
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October 1984, NUREG-0830, Supplement 4 SER.
The primary objective of the
Region III' EQ Audit'in this area was to verify that appropriate analyses
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and necessary' documentation to support the licensee's proposed and
accepted resolutions to NRR were contained in the licensee's EQ files,
and that appropriate modifications or replacements of equipment had been
implemented.
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During this review, the NRC inspection team selectively reviewed EQ
documentation and examined equipment in the plant relevant to prior
discrepancies. identified in the SERs.
For example,.the inspectors
reviewed the licensee's commitment to complete the qualification of
Rockbestos Firewall III Cables, Amphenol penetrations, Limitorque
operators, Westinghouse Motors, and other. equipment identified in
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Table 3.2 of.the May 1984 NUREG-0830, Supplement No. 3 SER.
During this review deficiencies were identified in the qualification of
Marathon 1600 terminal blocks used in 264V applications and in the
qualification of KULKA terminal blocks.
Details of these deficiencies
are further discussed in Sections 6a and 6b of this report.
The licensee also stated that they had qualified all installed TMI Action
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Items and that they were committed to maintain this equipment qualified
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within the scope of 10 CFR 50.49 on their Master Equipment List.
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5.
EQ Program Compliance to 10 CFR 50.49
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The inspectors reviewed selected areas of the licensee's EQ program to
. verify compliance to 10 CFR 50.49. The licensee's EQ program was found
to identify methods of equipment qualification; provide for evaluation
and maintenance of EQ documentation in an auditable form including
maintenance records; provide for upgrading of replacement equipment; and
provide for.the' control of plant modifications.
Based on their review,
the inspectors determined that the licensee had established an adequate
EQ program in compliance with the requirements of 10 CFR 50.49.
The
licensee's methods for establishing and maintaining the environmental
qualification of electrical equipment were reviewed in the following areas:
a.
EQ Program Procedures
The inspectors examined the adequacy of the licensee's policies and
procedures for establishing and maintaining the environmental
qualification of equipment within the scope of 10 CFR 50.49. .The
licensee's EQ program was reviewed for procurement of qualified
equipment; maintenance of qualified equipment; modifications to the
plant that could affect qualified equipment; updating of the EQ
manter list; and review and approval of EQ documentation.
Procedures reviewed included the following documents:
APA-ZZ-00101, " Preparation, Review, Approval, and Control of
Procedures, " Revision 1, dated March 16, 1987.
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APA-ZZ-00110, " Qualification of Qualified Reviewers,"
Revision 4, dated February 10, 1986
(Superseded by APA-ZZ-00910).
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APA-ZZ-00140, " Safety, Environmental and Other Licensing
Evaluations," Revision 6, dated March 11, 1987.
APA-ZZ-00200;'" Document Control," Revision 6, dated April 20,
1987..
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APA-ZZ-00320, " Initiating and Processing Work Request,"
Revision 10, dated November 13, 1986.
-APA-ZZ-00330,." Preventive Maintenance Program," Revision 6,
dated March 6, 1987.
APA-ZZ-00390, " Review of Electrical Equipment Qualification,"
Revision 4,-dated April 8, 1987.
APA-ZZ-00400, " Procurement of Parts, Supplies, Materials and
Service," Revision 9, dated December 18, 1986.
~APA-ZZ-00600, " Design Change Control," Revision 9, dated
-July 17, 1986.
.APA-ZZ-00601, " Category 1 Modifications (CMP's), Revision 0,
dated July 17, 1986.
APA-ZZ-00910, " Qualification of Qualified Reviewers,
Revision 0, dated March 13, 1987.
EDP-ZZ-04043, " Distribution, Review and Processing of Generic
Letters, IE Bulletins and Notices," Draft
(To replace QE-346).
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MDP-ZZ-QN008, " Inspection of Safety Related Instrumentation and
Control Cables Inside Containment," Revision 2,
dated January 9, 1987.
NEDP-18,
" Quarterly Status Report of NRC I&E Bulletins,
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Circulars, and Information Notices," Revision 1,
dated July 10, 1985.
QE-346,
" Distribution, Review and Processing of Generic
Letters, IE Bulletins, and Notices," Revision 0,
dated April 25, 1986 (To be replaced by
Specific areas reviewed in these procedures included definitions of
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harsh and mild environments, equipment qualified life, service
conditions, periodic testing, maintenance and surveillance, and
upgrading of replacement equipment purchased after February 22,
1983.
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No violation of NRC requirements were identified.
b.
10 CFR 50.49 Master Equipment List (MEL)'of EQ Equipment
IE Bulletin ~No.79-018 required licensees cf all power reactor
facilities stith an operating license to provide.a MEL that
identified each Class IE electrical equipment item relied upon to
perform a safety function during a design basis event.
10.CFR 50.49,. Paragraph (d), requires licensees to prepare a list of
electrical equipment important to safety and within the scope of the
rule.
The NRC inspectors reviewed the licensee's MEL for compliance
to 10 CFR 50.49. . Areas reviewed included adequacy of the MEL,
technical justifications for removal of items from the MEL, and
licensee reviews of the MEL for changes due to field modifications.
The inspectors' verified the completeness / adequacy of the list in
terms of equipment needed during accident conditions through review
of piping'and. instrumentation drawings (P& IDS), emergency
procedures,. technical specifications, and FSARs.
Plant Emergency
operating procedures reviewed included, " Reactor Trip or Safety
Injection" E-0, Revision 3; " Loss of Reactor or Secondary Coolant,"
E-1, Revision 4; and "SG Tube Rupture," E-3.
No violations. of NRC requirements were identified.
c.
EQ Maintenance and Surveillance Program
The inspector reviewed specific maintenance, replacement,
surveillance tests, and inspections necessary to preserve the
environmental qualification of EQ equipment identified on-the MEL.
EQ requirements in the licensee's maintenance procedures and EQ
binders were reviewed against maintenance records of selected
equipment to verify performance of maintenance and surveillance
activities at prescribed intervals, as scheduled by and recorded in
the Computerized History and Maintenance Planning System (CHAMPS).
Included in this review were the maintenance and surveillance
records for Tobar, Barton, and Rosemount transmitters; Target Rock
isolation valves; Limitorque valve operators; and ASCO solenoid
valves.
Check sheets attached to work packages were observed to be
very satisfactory.
The NRC inspectors found the licensee's methods for scheduling
maintenance and surveillance adequate.
During this review the
inspectors found no discrepancies in the implementation of EQ
maintenance / surveillance activities in the field.
The inspectors
considered this a good reflection of the licensee's ongoing EQ
efforts.
No violations of NRC requirements were identified.
d.
Plant Procurement and Upgrading of Replacement Equipment
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Licensee procurement procedures and docunents were found to
adequately address appropriate _ quality and regulatory requirements
regarding the environmental qualification of equipment within the-
scope of 10-CFR 50.49. . Checklists were observed to have been used
to provide evidence of reviews and approvals.
For. example,
procurement packages for Masoneilan transducers and pressure
regulators, Brand Rex cables, and Raychem splices were found to
properly address procurement of replacement equipment to-
requirements of IEEE.323-1974.
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The NRC inspectors reviewed plant equipment modification packages
CMPs 86-0069A,.86-0076A, and 86-0079A.
An EQ Checklist in each
package provided evidence that the licensee had reviewed and
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evaluated the effects of.the modification on the qualification
status of the EQ equipment prior to implementing the modification
of plant equipment.
No violations of NRC requirements were identified.
e.
Quality Assurance (QA) and Training Program
During this review the inspectors determined that the licensee had
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implemented a program to monitor the quality of EQ activities
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.through surveillance, audits, and the reviews of the records and
files for_ plant modifications and equipment. procurement.
NRC
inspectors reviewed four QA audits, which included the audit of
specific.EQ activities, and six QA surveillance inspections of EQ
activities performed between October 1985 and April 1987. -Various
QA/QC witness points had been included in surveillance activities.
The inspectors found the methodology, results, and followup
corrective action relative to the audits and surveillance very-
acceptable.
The NRC inspectors also reviewed the licensee's staff training
program and associated records' relative to the performance of EQ
activities.
The training records indicated that the licensee
had implemented a training program for key personnel, including
management, operations, and maintenance personnel responsible for EQ
activities.
The training program was found to adequately address
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key aspects of 10 CFR 50.49 requirements.
No violations of NRC requirements were identified.
6.
Detailed Review of Qualification Files
The licensee qualified their EQ equipment to the requirements of
IEEE 323-1974, NUREG-0588, Category I.
The inspectors reviewed
over 40 equipment qualification files for evidence of the environmental
qualification of equipment within the scope of 10 CFR 50.49 and evidence
of equipment qualification to NUREG-0588, Category I.
Files were found
'to include a full description of the equipment; similarity analysis of
tested equipment to that installed in the plant; allowed mounting methods
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and orientation; qualification of interfaces (conduit housing, seal,
etc.); evaluation.of aging effects'on equipment; description of test
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sequence and methodology; environmental conditions for the equipment
during an accident; qualification for submergence of applicable
equipment;. resolution of. test anomalies; and maintenance / surveillance
criteria for the preservation of the qualified status of equipment.
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The inspectors selectively reviewed the above areas, as applicable,
including special reviews for the required duration of operability of
equipment; licensee evaluation of tested materials and configurations
relative to actual plant installations; adequacy of test conditions;
aging calculations for qualified life and replacement intervals; effects
of' decreases in insulation resistance on equipment performance; adequacy
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of demonstrated accuracy of equipment and interfaces during an accident;
and licensee evaluations of discrepancies identified in IE Notices and
Bulletins.
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EQ files were reviewed for electrical cables, cable splices,
terminations, terminal blocks, electric motors, solenoid valves,
electrical penetrations, seals, lubricants, transmitters, temperature
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elements, radiation monitors, control and position switches, switch gear,
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control panels, and miscellaneous electrical devices.
In some' instances
the inspectors questioned the performance acceptance criteria, adequacy
of test results, and the required safety functions of the equipment;
however, the. licensee provided relevant test data and references to
mitigate.any concerns. The inspectors found that in most cases the files
allowed verification of equipment qualification to a specified
performance for accident conditions.
Exceptions are noted below:
a.
Marathon 1600 Terminal Blocks (TB) File E-028
The inspectors reviewed Wyle Laboratories Report 45603-1, dated
February 18, 1982, for evidence of the qualification of the
Marathon 1600. terminal blocks to IEEE 323/74.
The report indicated
that samples of Marathon 1600 terminal blocks were tested at
528 VAC, 264 VAC, and 132 VAC.
In each test the circuit current was
33 Amps.
In the case of the 528V (25 Amp fuse) and 264V (18 Amp
fuse) circuits, fuses failed during the LOCA test, indicating very
low IR values (the 25 Amp fuse could not be replaced as it continued
to fail).
The inspectors concluded that these samples had failed the
LOCA test at 528V and 264V.
'The licensee stated that the Marathon 1600 terminal blocks were only
used in 120V applications at Callaway.
The inspectors noted that
during the test a 12 Amp fuse had been connected in the 132V circuit;
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however, no irs were measured.
The licensee indicated that the irs
were not Eelow 11 ohms since the fuse did not fail.
Since no irs
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were measured during the LOCA test for the 120V circuit, and since
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significant fuse failures had occurred at higher voltages, the
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inspectors co-luded that there was no evidence to determine if the
irs at 132 VAC were any higher than 11 ohms.
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The inspectors also observed that all fuses including the 12 Amp
fuse in the:132V circuit had failed late into the LOCA test. This
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failure occurred subsequent to a temporary loss of power. The
licensee stated that the-fuse failures were due to the abrupt _
restoration of power; however, the NRC inspectors concluded that
since such power excursions simulated plant conditions, the fuse
failures were considered a reflection of _ the performance of the
terminal blocks in circuits energized intermittently.
Based on the above concerns, the inspectors concluded that the Wyle
test.#45603-1 did not support the qualification of the Marathon 1600
terminal blocks for 120V applications. The licensee was informed
that their Marathon 1600 terminal blocks were unqualified and that
they were required to submit a justification'for continued operation
or perform adequate currective action prior to startup from their-
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current outage. The licensee took immediate corrective action by
replacing terminations to Marathon 1600 terminal blocks with Raychem
splices.
Subsequent to their corrective actiot, the licensee submitted
additional responses to demonstrate that the Marathon 1600 terminal
blocks were qualifiable as originally installed in the plant. These
responses included a Limitorque Corporation Report B0119, dated
July 1,1982, which documented LOCA testing of Marathon 1600 terminal
blocks in Limitorque actuators. The inspectors noted that during
this Liuitorque test the terminal blocks were not continuously
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energized; the peak temperature of the test was 312 F while the
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Callaway MSLB peak temperature is 389 F; and that this Limitorque
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report had incomplete records and data. Further details and records
of the Limitorque test were requested from the licensee.
In addition,
NRR was consulted for information regarding the Limitorque Report
B0119. Further inspection revealed that IR readings taken during
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the test were apparently taken at six to ten volts rather than at
250V originally considered by the licensee. The inspectors were
concerned that the terminal block insulation resistance of 1000 ohms
measured at'6V would not be maintained at 120 VAC. The inspectors
concluded that the Wyle test and the Limitorque test did not provide
sufficient evidence that the Marathon 1600 terminal blocks were
qualifiable for 120 Volt applications during accident conditions at
Callaway.
The licensee followed up these NRC concerns by submitting a third
report for the qualification of the Marathon 1600 terminal blocks.
This test, done at Wyle, Report #17657, dated December 21, 1983,
indicated measurements of leakage currents for terminals in junction
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boxes with top entry conduits, as installed at Callaway. Leakage
currents for the terminal blocks directly below conduit openings
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(with. chemical spray on) were a maximum of 360 mA, except for two
instances of fuse failures (one at 4A in a 135 VDC circuit and one
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at 3A in a different 135 VDC circuit). The NRC inspectors are
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currently reviewing the justification for such leakage currents at
Unresolved Item (g further review, this is a Potentially Enforceable
Callaway.
Pendin
50-483/87015-02(DRS)).
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b.
- KULKA Terminal Blocks, File E-035, Volumes 1&2
These terminal blocks are used with Amphenol Bunker-Ramo
penetrations and are connected to power and control circuitry inside
the containment. The inspectors reviewed the Amphenol (Bunker-Ramo)
LOCA test report number 123-2222, " Design Qualification Report for
Electrical Penetration Assemblies," dated March 2,1979, contained in
the EQ file. The Kulka terminal blocks had been tested along with
the Amphenol penetrations. During the LOCA testing, IR values were
measured periodically and five 10 mA fuses were used to monitor
leakage currents. The 10 mA fuse failed during the second LOCA peak
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(340 F). The circuit voltage then had to be reduced from 600V
to 370V to keep the fuse from failing again. The rated voltage
(600V) could not be reapplied until the seventh day- of the test.
There was no discussion of this test anomaly by the licensee in the
file. Since the licensee had applied these terminal blocks in
480V power circuits, the inspectors were concerned that based on
the test report, the KULKA terminal blocks were not qualified for
480V applications. The licensee, however, stated that the lack of
failure of the 10 nA fuse in the 370V circuit demonstrated a
leakage current of 13 mA in the 480V circuit, and that based on
their evaluation this leakage current did not compromise any 480V
EQ application.
The inspectors also noted that the Bunker-Ramo report indicated very
low IR values of 100 ohms for the terminal block at the lowest applied
voltage of 370V. This indicated a high leakage current; however, the
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10 mA fuse monitoring leakage currents in the test did not fail once
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the voltage was reduced to 370V.
It was not clear to the inspectors
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gwh the 10 mA fuse did not fail when such low irs were indicated. The
file and test report did not contain sufficient information to allow
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resolution of this discrepancy. The licensee maintained that the
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fuses were a more reliable indicator of leakage current and that the
measured low irs were in error and should not be considered. The
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inspectors were concerned that the reason for such a significant
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discrepancy (low irs of 100 ohms not causing a failure of a 10 mA
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fuse) could be that the fuses were not connected properly in the
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test. The inspectors also noted that the licensee's EQ
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documentation did not include appropriate details of the Bunker-Ramo
test plan, circuits energized, method of the measurement of IR values,
and a justification for the low IR values. The licensee performed
further evaluations and discovered that the KULKA terminal blocks had
in fact been disconnected from their voltage source after seven hours
into the test due to failures.
Subsequent to this finding, the licensee submitted a CONAX Test
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Report No. IPS 675, dated April 20, 1971, for further evidence of the
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qualification of the KULKA terminal blocks. This report documented
an environmental test of un-aged Models of KULKA terminal blocks,
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as well as a separate aging test done on different models of KULKA
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terminal blocks. The aging test included irradiation of the blocks
and indicated no failures. This methodology, however, is contrary
to the requirements of NUREG-0588, Category I (IEEE 323/74), where
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the same specimen must be tested for aging and an accident
environment. The licensee stated that testing of.un-aged KULKA
terminal blocks in a' harsh environment was appropriate as the terminal
block material, Glass filled Diallyl Phthalate, was extremely durable
and not affected by aging. The inspectors noted, however, that one
KULKA Model 604-3.. had failed during the test, and that numerous cracks
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that could cause excessive leakage currents had been identified.
The inspectors were also concerned that the CONAX test report
indicated a zero leakage current throughout the test for the
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terminal block 604-3 even though the fuse in the circuit had
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failed. The licensee stated that since this failure occurred late
into the test', sufficient qualified life could be extrapolated for
the terminal blocks in EQ applications based on the period they did
not fail during the test. The licensee also stated that other KULKA
Models, having. lesser center to center terminal spacings in the same
test, had not failed.
Based on their review of the Wyle and Limitorque reports, the
inspectors determined that sufficient evidence had not been provided
to demonstrate that the KULKA terminal blocks were qualified for
480V and 120 VAC applications during accident conditions at the
Callaway plant. The licensee was informed that based on the guidance
in Generic Letter 86-03 the licensee was required to submit an
operability analysis to justify the continued operation of the plant.
The inspectors are also considering-data from testing done at Sandia
on unaged terminal . blocks. Pending further review of the licensee's
responses, this is a Potentially Enforceable Unresolved Item
(50-483/87015-03(DRS)).
c.
General Atomics Radiation Monitor, EQ File J-361A
This file addresses the qualification of General Atomics,
Model RD-23, containment radiation high range detectors to the
requirements of 10 CFR 50.49. The tested environmental conditions
envelope the postulated worst-case post-accident conditions. The EQ
documentation in this file supported-the environmental qualification
of these radiation detectors in their application.
On February 23, 1987, a 10 CFR 21 report was submitted to the NRC by
the manufacturer General Atomic Technologies, Inc. (Sorrento
Electronics). This report identified the Rockbestos extension
coaxial cable assembly, used in conjunction with the monitor inside
the containment, to be subject to insulation resistance changes due
to elevated temperatures. The report stated that the insulation
resistance of the Rockbestos coaxial cables is lowered due to the
increased temperature, such that these cables could not operate
within a factor of 2 accuracy as required by Regulatory Guide 1.97.
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The insulttion resistance is required to be greater than 5 x 10
ohms to maintain the accuracy requirements (within a factor of 2)
for this instrumentation. The cable manufacturer responded to the
Union Electric request for information on March 6,1937, referring
to the cable qualification report, QR 6802.
This test report does
not show an insulation reristance greater than 5 x 10B ohms.
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Subsequent to this finding on June 30, 1987, the licensee presented
additional'information regarding their prior reviews with Sorrento
Electronics. Sorrento Electronics had identified a cotential
deficiency in measured cable irs for extended periods above 240 F.
The licensee stated that their preliminary evaluation noted that
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their LOCA temperature peaks at 308.6 F for less than 50 seconds,
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and that their individual LOCA transients on FSAR Figure 3.11(B)-1
show that the temperature exceeds 240 F for approximately 17 minutes
with a peak temperature of 243.2 F.
The licensee further stated that
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surface temperature effects would also lower this profile somewhat,
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taking credit for conduit and an air gap, and that since a LOCA is an
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ANS Condition IV event (20 minutes assumed before operator action)
the GT-RE-59 and -60 readings do not, of themselves, direct operator
action in the E0P's. Rockbestos test data on the temperature /IR
relationship for the cables in question was reviewed by the
inspectors and is also available at the Rockbestos Office in New
Haven, Connecticut. Based on this data the IR values for a 1000 ft
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of cable at a Callaway peak temperature of 243.2 F will be
4.13 x 108 ohms.
Since Callaway uses only approximately 110 ft of
this cable, the licensee determined that on the monitors the irs
will be over 8.26 x 108 ohms for this cable.
This IR value is
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within the accuracy requirements of R.G 1.97.
No further concerns
were identified.
d.
Comsip Auxiliary Relay Cabinet RP-209, File E-093
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This file addressed the qualification of Struthers-Dunn Model 219
Relays and Model 236 Time Delay Relays mounted in the relay cabinet
to the requirements of NUREG 0588, Category I.
Qualification for the HELB was based on thermal and radiation aging
and DBE irradiation. The thermal aging was done in near 100%
Relative Humidity (RH), time and temperature significantly exceeded
the relatively mild HELB environment, but the radiation exposure was
only to 1.210E03 Rads.
Functional tests of the relays in the
original test sequence resulted in the failure of fifteen of the over
twenty samples.
The failure mode was not described in detail, but
failures were attributed to aging of the PVC armature hinge bearing
pads at too high a temperature. A second test sequence was conducted
with adequate aging of the PVC pads but the samples were not irradiated.
The qualification was ultimately based on a combination of partial
type testing with analysis. The licensee provided information which
clarified what materials were in the relays installed at Callaway
(which was not explicit in the file) and also produced the EPRI
references cited in the file for their radiation tolerances which
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indicated considerable margin over plant requirements. The large
amount of margin was also used to justify lack of sequential testing
to address synergistic effects. The licensee planned to incorporate
all clarifications and additional amplifying information in the
file.
Pending NRC review of this corrective action this is an Open
Item (50-483/87015-04(DRS)).
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7.
Plsjt Physical. Inspection
The.NRC inspectors selected over 40 items on the MEL for examination in
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the plant.
The EQ file of each item had been reviewed and information
regarding the. location, manufacturer, model/ serial number, mounting,
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orientation, environment,.and interfaces had been noted. The inspectors
examined the selected. items in the field, as accessible, and verified
that the method of installation of each item was not in conflict with its
environmental qualification.
Specific areas reviewed included
traceability of installed items of EQ files, ambient environmental
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conditions, qualification of interfaces (connectors, wires, seals,
. insulation, lubricants, etc.), evidence of'significant temperature rise
from process,. drainage, mounting methods, physical conditions and
housekeeping.
In almost all cases, items examined in the field during
this walkdown were found to meet their appropriate EQ requirements with
the following exceptions:
a.
During review of the power leads for Motor SG LISB in the Auxiliary
Building Penetration Room the inspectors noted that the leads were
indented and.apparently damaged.
The licensee took immediate
corrective action and repaired the leads under Work Request
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WR-W101158.
No further concerns were identified.
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b.
Limitorque Actuator EG-HV-0060 Heater Wires
During examination of Limitorque Actuator EG-HV-0060 the inspectors
noted that contrary to EQ requirements, the heater wires for this
actuator were energized.
The licensee provided an adequate
evaluation to indicate that the energization of the heater wires did
not affect the EQ of the Limitorque Actuator.
The licensee also
agreed to clarify their files through plant document RFR 3748-A, to
state that the heater wires were still energized and that the wires
would be deenergized during the next plant outage.
No further
concerns were identified,
c.
Valve EC-V-0120 Boron Leak
During review of EQ equipment at containment elevation 2000 ft., the.
inspectors noted excessive leakage in valve EC-V-0120.
This valve is
in the skimmer line for the refueling pool inside the containment.
The inspectors verified that the leakage did not affect any EQ
equipment.
During plant operation this line is not used and does not
contain fluids under pressure.
The licensee took corrective action
and prepared work request No. 062142 to repair this valve during the
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scheduled refueling outage in September 1987.
No further concerns
were identified.
No violations to NRC requirements were identified.
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8.
Open Items
Open items are matters which have been discussed with the licensee, which
will be reviewed further by the inspector, and which involve some action
on the part of the NRC or licensee or both.
Open Items disclosed during
this inspection are discussed in Paragraphs 3 and 6d.
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9.
Potentially Enforceable / Unresolved Item
An unresolved item is a matter about which more information is required
in order to ascertain whether it is an acceptable item, an open item, a
deviation, or a violation.
Potentially Enforceable / Unresolved Items are
unresolved items, which if ascertained to be a violation, may be followed
up with enforcement action in accordance with NRC enforcement guidance on
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Potentially Enforceable / Unresolved Items are discussed in
Paragraphs 6a and 6b.
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10.
Exit Interview
The Region III inspectors met with the licensee's representatives
(denoted under Paragraph 1) during an interim exit on May 8, 1987, and
discussed their findings by phone at the conclusion of the inspection on
July 22, 1987.
The inspectors summarized the purpose and findings of
the inspection and the licensee acknowledged this information.
The
licensee did not identify any documents / processes reviewed during the
inspection as proprietary.
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