ML20235U443
| ML20235U443 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/27/1989 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20235U424 | List: |
| References | |
| 50-285-89-03, 50-285-89-3, GL-87-12, IEB-85-003, IEB-85-3, NUDOCS 8903090199 | |
| Download: ML20235U443 (51) | |
See also: IR 05000285/1989003
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APPENDIX B-
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV p
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NRC Inspection Report: .50-285/89-03 License: OPR-40
- Docket: .50-285'
Licensee: Omaha Public Power District.(0 PPD)
1623 Harney Street ~
Omaha, Nebraska .68102
Facility Name: Fort.Calhoun' Station (FCS).
Inspection At: FCS, Blair, Nebraska
Inspection Conducted: January 1-31,,1989
Inspector: P. H. Harrell,' Senior Resident Inspector'
T. Reis, Resident Inspector. .
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R. P. Mullikin, Project Engineer, Project Section B,.
Division of Reactor. Project's
Approyed: N- ! 9/2 7/hi'T
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T. F. Westerman, Chief, Project Section B Uate
Division of Reactor Projects-
Inspection Summary
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Inspection Conducted January 1-31, 1989^ (Report 50-285/89-03)
Areas Inspected: Routine, unannounced inspectio'n including followup on
previously , identified. items,.. licensee-event. report followup, operational safety
verification,' plant tours',' monthly maintenance observations, monthly .
surveillance observations, security observations, radiological protection.
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observations, in-office. review of periodic and special. reports, installation-
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and; testing of modifications, verification of. containment integrity, followup ;
of onsite' events, observation of activ.ities during plant startup; followup on.
status of operations. memoranda, and review of the. licensee's response to NRC
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Results:- -
Duringthisinspectionperiod,theplankwasstartedupfroma' refueling:
outage. The NRC inspectors made observations of the activities related to
plant startup in the control room and in the field. . The observations included-
the approach to criticality and synchronizing the main turbine to the grid.
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8903090199 890301
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PDR ADOCK 05000285
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l During the observations made by the NRC inspectors, it was noted that the
operations staff performed their duties in a highly professional manner. All
evolutions were well controlled by the use of the appropriate' procedures. .The.
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. shift s'upervisors provided an oversight function to ensure that all evo'lutions
were coordinated between operations personnel, and provided timely direction
for each licensed and nonlicensed operator. Plant startup was performed-
without any problems.
In preparation for plant startup, the licensee was required to establish
containment integrity. The licensee ~has experienced difficulty in the past '
with ensuring that'all piping, components, and penetrations were appropriately
secured to establish containment integrity. Prior to establishing integrity
during plant startup, the licensee performed an extensive plant walkdown and-
revised the appropriate procedure to. include all the necessary hardware that
required checking. The NRC inspectors used the. revised procedure to verify -
that containment integrity had beenLestablished.'. No problems were noted during
the plant walkdowns by the NRC inspectors. In this. case, it appeared that the
licensee had taken positive, proactive: steps to ensure that the problems
experienced in the past did not recur.
-In October and Novembe'r 1988"an operational safety t'eam inspection (OSTI) was
performed at the FCS. 'During the inspection, the OSTI team noted problems with
the control of drawings for' temporary modifications and notified the licensee
of the problems _during the exit meeting.- As a'part of the review performed i,
prior' to startup, the NRC inspectors reviewed drawing control for temporary.: '
i modifications and noted that the licensee had not updated'the control room
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drawings to indicate which temporary modifications were still installed. A ,
discussion was held.with licensee management and_the licensee' updated all
appropriate drawings prior to plant ~startup. Although the licensee took action-
to correct a deficiency identified by the OSTI team, it appeared the action was ,
not initiated until the NRC inspectors identified the need to perform the I
actions prior to plant startup. j
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DETAILS
1. Persons Contacted
- J. MacKinnon, Acting Division Manager, Nuclear Operations
- W. Gates, Manager, Fort Calhoun Station
J. Adams,L Reactor Engineer
- J. Bobba, Supervisor, Radiation Protection
C. Brunnert, Supervisor, Operations Quality Assurance
- J. Fisicaro, Manager, Nuclear Licensing and Industry Affairs
l *S. Gambhir, Division Manager, Production Engineering
- J. Gasper, Manager, Training
- K. Henry, Lead System Engineer, Primary System
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- R.'Jaworski, Manager, Station Engineering
. J. Kecy, Supervisor, System Engineering
D. Lieber Supervisor, Security
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- D. Matthews, Supervisor, Station Licensing
K. Miller, Supervisor, Maintenance
T. Patterson, Assistant Manager, Fort Calhoun Station
- G. Peterson, Assistant Manager, Fort Calhoun Station
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- A. Richard, Manager, Quality Assurance and Quality Control
l *C. Simmons. Station Licensing Engineer-
l *M. Tesar, Supervisor, Technician and General Employee Training
- D. Trausch, Supervisor, Operations
- S. W111 rett, Supervisor, Administrative Services
- Denotes attendance at the monthly exit interview.
The NRC inspectors also contacted other plant personnel, including .
operators, technicians, and administ'rative personnel.
2. Plant Status
During this inspection period, the licensee . completed its eleventh
refueling outage of the Fort Calhoun Station'. The'FCS achieved- '
criticality and: entered. Mode 2.at:9:27 a.m. (CST) on January 29,1989.
Following routine low-power physics testing, the plant entered Mode 1 at ,
4:46 p.m. (CST) on ' January 31, 1989.
In addition to' refueling, the major evolutions performed during this ,
outage were as follows.
- Overhaul and postoverhaul testing of both' emergency diesel generators l
were'. completed. {
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- Plant computer address' points were transferred to the emergency. ,
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response facility computer and the old plant computer was' removed.-
- Weld repair and radiography of the emergency feedwater storage tank 1
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were completed,
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The motors for Reactor Coolant Pumps RC-3C and RC-3D were removed
from containment, sent to the vendor's facilities for rebuilding, and
reinstalled.
Both steam generators underwent eddy current testing. No tubes
required plugging.
The safety injection and refueling water storage tank was inspected
by divers and potential leaks were repaired.
The main turbine was disassembled and inspected. Failed blading was
identified in the first stage of the high pressure turbine. During
the upcoming cycle, the turbine will be run without the first stage.
The removal of the first stage represents approximately a 6 percent
loss of electrical generation capability.
Major modifications were completed on the control room ventilation
system, diesel generators, and instrument air system used to operate
several safety-related valves.
3. Followup on Previously Identified Items (92701 and 92702)
a. (Closed) Deficiency 285/8522-2.1-1: Lack of design analysis to
support. sizing of air accumulators for Valves YCV-1045A and
YCV-1045B.
This deficiency involved the lack of an adequate design analysis
developed for the sizing of the air accumulators for Valves YCV-1045A
and YCV-1045B. The inadequate analysis failed to demonstrate that
the accumulators could provide a sufficient air supply to hold
Valves YCV-1045A and YCV-10458 shut for 30 minutes.
The NRC inspector reviewed the functional test performed by the
licensee to verify the adequacy of the accumulators. The review
performed by the NRC inspector is provided in paragraph 3.e of this
inspection report.
At the time this deficiency was identified, the design basis for the
accumulators stated that the valves would be held shut for
60 minutes. Subsequent to the inspection, the licensee revised the
design basis to state that the valves would be held shut for
30 minutes. The licensee revised the design basis document (DBD) for
the instrument air (IA) system to reflect the change in the basis for
the accumulators for Valves YCV-1045A and YCV-1045B.
The NRC inspector reviewed the DBD for the IA system and verified
that the basis reflected the change made by the licensee. No
problems were noted.
b. (Closed) Deficiency 285/8522-5.1-1: Battery sizing calculation.
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This deficiency concerned the use of the battery current discharge
profile that was the same 1979 sizing calculation used to purchase {
replacement batteries. However, de loading had increased since 1979.
The licensee, aided by a contractor, revised the dc-load profile.
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During an NRC inspection conducted during March and April 1988, NRC
personnel reviewed the licensee's action on this deficiency. The
review was documented in NRC Inspection Report 50-285/88-200 and
noted that the battery sizing calculation contained an assumption
that the NRC inspectors wanted confirmed. This was the lack of a
correction factor for battery capacity degradation. This item was
left open pending the battery manufacturer's (Exide) analysis of the
test results to establish the degree of capacity remaining in the
batteries.
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Subsequently, Exide reviewed the batt'ery test data and in a letter
dated June 24, 1988, stated that Batteries 1 and 2 are in very good
condition with the exception of Cell 45 in Battery 2. The licensee
replaced Cell 45 during the 1988 refueling outage.
c. (Closed) Deficiency 285/8522-5.2-1: Fire wrap protection for cable
raceway.
This deficiency concerned the licensee's use of unverified derating
factors for the power cables for the pressurizer heaters that had
been wrapped with a fire protection. material.
During an NRC inspection conducted during March and April 1988, NRC
personnel reviewed the licensee's' action on this deficiency. The
review was documented in NRC Inspection Report 50-285/88-200 and
noted that the licensee's calculation was based on noncable specific
data and nonconservative thermal conductivity values for the fire
wrap.
The licensee contracted with a consultant to perform a detailed
review of the cable conductor temperatures due to the fire wrap. The - I
NRC inspector reviewed the results of the. study documented in a
letter, dated December 29, 1988. The calculations determined that
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the addition of the fire wrap; therefore, the cables were adequate to
perform their intended safety function. No problems were noted
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during the review performed by.the NRC inspector.
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d. (C1osed) Severity Level III Violation 285/8727-02 (Violation A):
Inadequate preparation of a 10 CFR Part 50.59 evaluation. (
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This violation involved the licensee's inadequate preparation of a j
Part 50.59 evaluation for connection of the IA system to the fire 'l
l water system. During preparation of the Part 50.59 evaluation, the j
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licensee failed to consider the effect on the IA system if water from ?
the fire water system entered the IA system. Water did enter the IA
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system causing. Emergency Diesel Generator (EDG) No. 2 to fail due to
inoperability of the radiator exhaust damper, an air-operated device.
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To address this violation, the licensee revised Procedures GEI-27,
"10 CFR 50.5S Safety Evaluation"; B-11. " Independent Design
Verification"; and GEG-3, " Preparation of Design Packages." These
- procedures provide instructions for the preparation, review, and
approval of safety evaluations. . The revisions provided by the
licensee were intended to strengthen the quality of the safety
evaluations perfonned by the licensee. The procedures. implemented a
detailed checklist to be completed for each evaluation that included
consideration of common-mode failures..
The NRC inspector reviewed the procedure changes made by the
licensee. It appeared, based on.this.. review, that.the licensee had
adequately established a program for generation of safety 4
evaluations.
During the 1988 refueling outage, the NRC inspectors reviewed'the
evaluations issued for eight modifications. In each case, it
appeared that the licensee's evaluation completely addressed all the
appropriate areas.
e. (Closed) Severity Level III Violation 285/8727-03(ViolationB.1):
Inadequate testing program for IA system check valves.
This violation was related to the. licensee's failure to test the
check valves that were installed to prevent water from the fire water
system' from entering the IA system. As 'a result, the check valves
failed. to properly seat'and. thus allowed water to enter the IA
system. Entry of water into the IA system caused the failure of the
exhaust damper for EDG No. 2.-
Subsequent to the water intrusion event, the licensee disconnected
the IA system connection to the fire water system. An air compressor -
unit was installed as a dedicated air supply for'the dry-pipe section
of the fire water system. This modification eliminated the potential
for water to enter the IA system via the fire water' system.
To address the generic? issue of testing for all IA safety-related
check valves, the licensee. revised and submitted the inservice
testing program for inclusion of the IA check valves. - During the
1988 outage, the licensee tested the check valves to verify
operability. The testing was performed using'the maintenance
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orders (MO) listed below:
M0 Number Valve (s). Tested
883630 YCV-1045A and -1045B: Steam
supply. valves for~the
auxiliary feedwater pump-
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883632 HCV-400 A,B,C,D through
-403 A,B,C,D:- Component
cooling water (CCW) to ;
containment coolers !
containment isolation valves j
883631 HCV-712A: Spent fuel pool
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charcoal filter bypass valve
1 883629 HCV-385 and -386: Safety
I injection-(SI) pumps.
recirculation isolation valves
HCV-304, -305, -306, and -307:
High-pressure SI isolation
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HCV-2987: High-pressure SI header
isolation valve
883634 HCV-238, -239, and -240: Loop
injection valves
The NRC inspector reviewed the testing performed by the licensee. It
appeared that the licensee had performed testing to verify that the
sizing of the accumulator assembly.was adequate to meet the
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established design basis.
In addition to the valves listed above, the licensee also committed.
to test additional check valves in the-IA system.- Testing of the-
valves will be reviewed during followup of Violation 285/8815-06 and
Open Item 285/8823-01.
f. (Closed) Severity Level IV Violation 285/8807-02: Failure to.
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maintain the fire brigade complement.
i This violation involved the failure of the licensee to maintain five
individuals for fire brigade duties that did not have other duties
- assigned. Specifically, a security guard was assigned to the fire j
brigade, but was also given the responsibility of operating security ,
doors in the event that a fire occurred in the control room. I
To resolve this apparent discrepancy, the licensee issued a revision
to Procedure AOP-6, " Emergency Fire Procedure." The revision
reassigned the duties of operating the security doors to an
operations individual, thus allowing the security guard to be free of -
all duties except those associated with the fire brigade. l
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On' July 1, 1988, the physical security plan was revised =to require
one additional ~ security officer to be present on site for each shift.
This additional officer will act as a' fire brigade member. 1
The NRC inspector reviewed the actions taken by th e licensee,.as-
described above. Based on this review, it appeared. that the . licensee
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has instituted appropriate actions lto ensure proper fire brigade
staffing.
g. (0 pen) Severity Level IV Violation 285/8810-01i Failure to promptly
resolve test deficiencies.
This violation was related.to the. licenseei s failure to promptly
i. , resolve test deficiencies, identified dusing the performance of
ST-NZ-1, " Verification of'Open Containment Spray Nozzles." When,the;
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test:was performed .the individual ~ performing the test failed to test. V
11 spray nozzles due to the inaccessibility of the nozzles. ~The test
results indicated that the nozzles were not inspected; however, the
test was accepted as-completed without'anfanalysis.being performed _to
indicate the results .were acceptable.
In response to this violation, the licensee inspected.the 11 nozzles..
during the 1988 refueling. outage. .The inspection'was performed in
accordance with Procedure SP-NZ-1. Completion of the nozzle
inspection verified that all nozzles were operable in accordance with
Technical Specification (TS) 3.6. After completion of the testing..- '
the licensee declared the 11 nozzles inoperable since the' nozzles -
were located behind a-ventilation duct. To verify oper' ability of the j
containment spray. system with the nozzles-inoperable, theilicen'see
initiated'a review in accordance with the requirements of-
l 10 CFR Part 50.59. The results of.the evaluation concluded..that the
l containment spray system remained operable.
The NRC's Office of Nuclear Reactor Regulation (NRR) and the NRC. 1
inspector reviewed the Part 50.59 evaluation' generated by the !
licensee. Based on the review, it appeared that the' licensee had
appropriately evaluated system operability. The NRC inspector'also
reviewed Procedure SP-NZ-1. No problems were noted during the- !
reviews. .
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The licensee issued a request for amendment of TS 3.6 ~on January 6,
1989. The amendment'was submitted to change the TS so that the 11
inaccessible nozzles would not have to.be inspected. NRR is.
currently reviewing the submission made by the' licensee.
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This violation remains open pending the appr' oval and issuance of a TS.-
amendment by NRR for a change in the inspection criteria for the j
spray nozzles. '
h. (Closed) Open Item 285/8811-08: Modification of the IA containment
-penetration assembly.
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This item was related to a commitment made-by the licensee'to upgrade
the containment penetration assembly for the IA piping. The. licensee 1 <
1 made the. commitment because the valve used for containment isolation- .'
failed open on a loss of-IA pressure. This type"of. failure" mode
would not ensure that containment isolation could be achiev'ed,-when:
required, in all cases.
To address this problem, the licensee modified the penetration
assembly during the 1988 refueling'. outage. The modification was
performed in accordance with the instructions provided in
Modification Request (MR)'FC-88-11. The NRC inspector: performed a
~ detailed review of the modification. The results of the review are
documented in paragraph 13 of this inspection report.
i . (Closed) Open Item 285/8815-09: Completion of' seismic calculations
for. Valves HCV-238 and HCV-239. .
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Inspection of.the IA water intrusion event of July 1987 led to .
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questions concerning the seismic qu'alification of the IA accumulator
assemblies serving 39 CQE (safety-related) valv'es. In-a letter to
NRR, dated November 20, 1988, the' licensee judged the seismic design
of 32 of the CQE assemblies to be adequate based on field
verification and bounding calculations. Four other assemblies were a
documented as'having calculations which confirm they were designed.in
accordance with" accepted criteria. Three accumulator assemblies were-
not available'for field installation verification due to their
location inside containment.
NRR reviewed the submission made by the licensee. Based on'the
submission and subsequent discussions with licensee personnel, NRR
arrived at the conclusions listed below. The' conclusions weree
documented in a letter dated July 28, 1988.
.At the' time of the. plant design, all safety-related valves were ,
required to be_ seismically qualified. -
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Although Appendix =B to 10 CFR Part 50 was'in effect at the time
of plant construction, specific guidelines on the maintenance of_
l~ equipment seismic qualification did not exist. l
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The licensee is presently undertaking an extensive program} to- ,
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reconstitute the design basis and will, continue;to attempt to '/
locate any information to substantiate the: statements made in M y
the Updated Safety Analysis' Report (USAR). l
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Most..of the equipment installed in the plant:is in service in l
other nuclear, as well as fossil,Lunits which'either ha~vei 'j
adequate documentation or have survived ground shaking of actual j
strong earthquakes in the past. '
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The site is located in a.relatively low seismicity zone as:
compared to some other nuclear facilities. .
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Based on these: conclusions, NRR approved a licensee proposal to delay:
the resolution of the seismic qualification of equipment until.
Unresolved Safety Issue (USI) A-46 has been resolved. .The proposal, h
was approved based on the~ conditions stated below:
Should an earthquake of design acceleration level or higher ,
occur, all' safety-related equipment in the plant shall undergo
an intensive. inspection and evaluation for structural integrity.
For active components, the licensee shall ensure the functional-
operability of the components prior to restart of the plant.
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Once the generic resolution of USI-A-46 is complete, the
licensee will pursue an aggressive schedule for. implementation
of the resolution commitments.
During the 1988 refueling outage,-the licensee performed field
inspections for the remainir.g three assemblies: HCV-238, HCV-239,
and HCV-240. Following field verifications and subsequent structural
anaiysis, the licensee determined that for'each of the three
assemblies, the tubing having the lumped valve' masses (isolation and
check valves) could become structurally unstable under the combined j
effects of seismic inertia and gravity. Based on this conclusion,
the licensee designed and installed additional supports for these
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assemblies per MR-FC-88-177, " Seismic Support of Instruments Air
Assemblies." The licensee reperformed the' calculations, based on the
additional supports installed by the modification, and found.the
configurations to be seismically acceptable.
The NRC inspector has received these calculations, performed a-
preliminary review, and found them acceptable. The calculations will .
be forwarded to NRR for final approval. Based on;the review and the !
conclusions presented by NRR for the previous submittal, this item is i
considered closed. The subject of: seismic qualification of the
accumulator assemblies will be reviewed.during the resolution of
USI A-46.
j. (Closed):0 pen Item.285/8815-10: . Completion of seismic calculations
for. Valve HCV-240.
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This item is considered closed based on the discussion presented.in
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paragraph 3.i (above) of this inspection report.
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k. (0 pen). Unresolved Item'285/8815-11: JReview of:the" licensee's method -
used for hot-leg injection. -j
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This item is related to the licensee's' method used for hot-leg i
injection. Due to the potential for a loss of .IA pressure and the .;
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affect on the safety-injection' header valves, the licensee developed
an alternate means for establishing hot-leg injection.
This item was previously reviewed by the NRC inspector and the review
documented in.NRC Inspection Report 50-285/88-46. In that inspection
report, it was noted that the licensee was required to install a
remote-manual operator for Valve SI-186. The operator was required
to provide operations personnel with a means to reposition the valve
during postaccident conditions.
The licensee installed a remote-manual operator for Valve SI-186
during the 1988 refueling outage. The operator was installed in
accordance with MR-FC-88-35, " Hot-leg Injection During Long-term Core
Cooling."
The NRC inspector reviewed MR-FC-88-35 and the physical installation
of the remote-manual operator for Valve SI-186. In addition, the NRC
inspector reviewed Procedure E0P-3, " Loss of Coolant Accident," to
verify that the appropriate changes had been made to provide
instructions to operatiens personnel. No problems were noted during
the reviews.
The licensee generated an evaluation, Operations Support Analysis
Report (0SAR) 88-36, to verify that the alternate hot-leg injection
flow path was adequate. OSAR 88-36 was forwarded to NRR for review.
NRR reviewed OSAR 88-36 and noted several concerns related to the
evaluation. The concerns identified by NRR were transmitted to the
licensee in a letter dated January 10, 1989. The letter requested
that the licensee respond to the concerns within 120 days and also
noted that NRR considered the plant safe for continued operation.
This item remains open pending the receipt of the licensee's response
to the NRR request for additional information and a review of the
response by NRC personnel.
1. (Closed) Severity Level III Violation 285/8815-14: Failure to
maintain containment integrity.
This violation involved a failure of the licensee to maintain
containment integrity. During surveillance testing of a pressure
switch used for sensing containment pressure, a tubing cap was not
reinstalled. The missing cap represented a loss of containment
integrity.
The NRC inspectors performed a detailed review of this violation.
The details of the review are provided in paragraph 15 of this
inspection report.
m. (Closed) Severity Level IV Violation 285/8822-02: Reporting of
inaccurate information to the NRC.
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This violation cited the licensee's reporting of inaccurate
information to the NRC in a letter dated July 1,1988. The letter ,
was generated by the licensee to identify..to the NRC, errors found i
in the setpoints of the reactor protection system (RPS). The letter
explained the cause of the errors, the significance of the errors,
and the corrective actions taken by the licensee to ensure continued
safe operation for the remainder of the operating cycle.
The RPS setpoint errors resulted in the NRC issuing a Notice of
Violation to the licensee for inadequate design control. A-
discussion of the violation, the corrective actions taken, and final j
disposition are documented in NRC Inspection Reports 50-285/88-22 and j
50-285/88-36. j
In the letter dated July 1,1988, the licensee stated that the shift
supervisor had placed instructions in his log to reduce reactor power
to 80 percent if the excore detectors were used to measure linear
heat rate. The NRC inspector's followup of the event found that the
shift supervisor's log did not contain such instructions.
The licensee admitted to the violation and summarized its root cause
as being haste in issuing the letter on July 1,1988, for a condition
discovered on June 28, 1988. Further. followup of the event by the
NRC inspector found that the instructions reported to have been
placed in the shift supervisor's log had been verbally given to the
shift supervisor by the reactor engineer. However, they were not
recorded in the log as stated.
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These same instructions were formally issued as an operations
memorandum on July 8, 1988. This laps ~e in time with no correct,
formal,
applicablewritten instructions
action was
statement of TSof no safety)would
2.10.4(1)(b significance
not havesince the
required a power reduction within the constraints of TS Figure 2-6,
which was in error because of the setpoint errors discovered at that
time.
On July 25, 1988, the licensee issued an additional letter to the NRC
to provide an update and to correct inaccurate information reported
in the letter of July 1,1988. The corrected version stated that the
reactor engineer contacted the shift supervisor and left him the
instructions that were previously reported to have been recorded in
the log. The NRC inspector was able to confirm that these
instructions were left with the shift supervisor.
In tems of corrective action that will be taken to avoid further
violations of this nature, the licensee stated that its existing
procedures and policies which govern submittal of information to the
NRC are adequate to preclude recurrence of this concern. The
licensee further stated that its current employee education program
stressing attention to detail, accountability for actions, and
procedural compliance will also reduce the likelihood of recurrence.
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The NRC inspector reviewed licensee Procedure N00-QP-9, " Processing H
of NRC, INPO, NUMARC, DEC,_00E, and' EPA Correspondence Documents," l
and concurs with the licensee assertion that the process governing l
submittal of information to the NRC should preclude the' reporting of ;
inaccurate information. In the assessment by the NRC inspector, the ' i
.
inaccurate submittal was an isolated incident and not a programmatic
breakdown. No further licensee corrective action is warranted.
n. (0 pen) Open Item 285/8823-02: Resolve discrepancy for TS limiting
condition for operation (LCO) on raw water pumps. .
This open item was related to an evaluation of the raw water (RW)_ l
system performed by the licensee. .The evaluation included i
measurement of the flow rate in the RW headers for each RW pump. The- !
results of the evaluation -indicated. that two RW pumps were' required i
.to meet the flow rate requirements established by,the USAR. 1
After completion of the evaluation, the licensee noted that TS 2.4
did not establish an LC0 when one RW pump was inoperable. To address- J
this apparent discrepancy, the licensee issued an operations . 1
memorandum to voluntarily establish an LCO for'the inoperability of
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one pump. The operations memorandum is discussed in paragraph 12 of
this inspection report.
In addition, the licensee' submitted, on December 31, 1988, a request
for amendment of th'e TS_to address the,inoperability of one RW pump.
During the preliminary review of the amendment by NRR, apparent
problems were noted with the licensee's . submittal. NRR noted that.
~
the licensee failed to address the flow rate evaluation that had been
performed and failed to appropriately address the actions to be taken
when two RW pumps were inoperable.
The licensee stated that a revised amendment will be submitted in the
near future. This item remains open pending resubmittal of the
amendment and approval of the' amendment by NRR.
o. (Closed) Open Item 285/8827-03: Completion of inspection ~and
modifications for the IA system.
This item involved a commitment made'by the licensee to inspect and-
modify the IA system during the 1988 refueling' outage.- This' item
addressed seven activities to be. completed. Each activity completed
~
by the licensee is discussed below.
Three valves were identified that could not be cycled during
plant operation. Therefore, the licensee stated that the three
valves that were potentially wetted'during the IA event would be
cycled during plant shutdown. The valve cycling was performed
to verify valve operability.
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The licensee cycled the valves and determined that the three
, valves were operable in that the' stroke times were within the
established criteria.' The three valves tested-are listed below.
'
Valve Number Valve Function
HCV-206, Reactor coolant pump (RCP)
controlled-bleed-off is'olation- v
.HCV-4388' CCW inlet to.RCP coolers
HCV-438D :CCW outlet to RCP coolers
I .
The NRC inspector reviewed the results'of the valve testing
performed by the licensee. 'The tests were performed'in
accordance with Procedures ST-ISI-CC-1, " Component Cooling .
Valves Inservice Testing,"- and ST-ISI-CVCS-1, " Chemical and
~
Volume Control Valves ' Inservice Testins . The NRC inspector
_
verified that the cycling; times were within the established
acceptance criteria. No problems were noted during the reviews. "
Replacement of.the damper operators for.the EDGs was performed
during the 1988 outage. The previous operators required air
'
pressure to open and shut the-dampers. Due to problems with-
operation of the dampers, the licensee installed damper'
operators that utilize a spring to open the damper in the' event'
IA pressure.is_. lost.
'
The NRC inspector performed an indepth review of the
in's tallation.of the damper operators. The review-is_ discussed
in' paragraph 13 of this inspection report.
To protect individual components from potential damage due to
! particulate in the IA system,..the licensee committed to instal 1
l filters upstream ofiall air-operated valves'without filters.
Based on a walkdown performed by a licensee contractor, the.
licensee identified 17 valves (HCV-400 A,B,C,D through
HCV-403'A,B,C,0.and.HCV-2861) that did not have filters
.
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installed in the air supply line. :The_ licensee issued.
MR-FC-88-089, ." Filters for HCV-400 Series Actuators'," to' provide
instructions _for" installation of the filters. During the 1988 . -
refueling outage,<the licensee installed the filters.'TWith the
installation'of the filters,'all' safety-related valvesTand
~
components' currently have filters installed.~ -
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The NRC' inspector reviewed MR-FC-88-089 and ' performed a walkdown
of selected components and valves in the plant to verify that
"
filters were installed. No problems'were noted.
.
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To ensure that the IA and service air systems are not
susceptible to . water intrusion, the licensee committed to
install an air compressor system to supply air to the; fire water
- dry pipe deluge system.
The licensee installed an' air compre'ssor system in,accordance
with MR-FC-87-32, " Air Maintenance Compressor for Fire
- Protection Deluge System." The modification was completed in
September 1988. The air compressor system supplies air only for-
'
the deluge. system.
The NRC inspector reviewed MR-FC-87-32 to verify t$at the system
was installed in accordance with.the' instructions. No problems
were noted.
.The licensee committed to install smaller micron' filters on the
outlet .of the dryer in accordance~ with a recommendation made by. !
their consultant. The new filters are' rated at'3 microns.and-
replacedthe20-micron'filtersthatwere'previouslyl installed.
'
- The licensee installed the 3-micron filters in accordance with '
MR-FC-87-46, " Instrument- Air Dryer Upgrade (Tie-In)." The
modification was completed during the 1988 refueling outage.
The NRC inspector reviewed MR-FC-87-46 and noted n'o problems. ;
- The licensee's consultant recommended that the licensee' install
a new dryer to enhance the performance ' capabilities' of.the'IA
system. The installation of a new dryer will be performed as an.
on-line modification during'.1989.
In' anticipation of performing the on-line modification, the
licene.9 instclled system connection! points'..during:the 1988
refue . ug outage. The modification wasLinstalled in accordance ,
with is-FC-87-46. The modification will allow theLlicensee'to d
instal', a newLdryer during operation of the IA system without
disturbing: system air flow; The'NRC inspector reviewed
MR-FC-87-46 and noted no problems., *
The installation of an additional dryer for._the IA' system is
considered a system upgrade beyond what.is required by the >
established design basis. The licensee operated.the_IA system l .
with the dryer'that is currentlyiinstalled and has been able to
- maintain the~ dew point at less than -20 Fr.
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For this reason, it-
appears that the IA system, as currently installed, is capablef
l .. ; of. performing.its; intended' safety functio'n. -Therefore, this;
item is considered closed even though the additional air dryer.
has not yet'been completely installed. ,
iThe licensee performed a valve teardown program to inspect the
internals of air-operated valves. The program was established
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to determine what affect water intrusion into the IA system had
on the operation of the valves.
The licensee selected a sample of valves to be inspected based
on the location of the valves in the IA supply header and the
various models of actuators installed on the valves. Based on
the selection criteria, a total of 72 actuators were identified
for disassembly and inspection.
After completion of the actuator inspections', the licensee
performed an evaluation. The evaluation was developed based on
the results of the data collected during inspection of the
actuator internals. The data included determination of the
presence of water, rust, or particulate matter. Based on the
results of the evaluation, the licensee stated that the need for
future inspections of actuators to specifically address the
water intrusion event was not required. Although small amounts
of water and evidence of rust and particulate matter were
identified, the licensee stated that the conditions did not
affect the operability of the valve actuators. The evaluation
also stated that the routine maintenance program would ensure
valve operability.
During reviews of maintenance activities in past inspection
periods, the NRC inspectors observed the inspection of valve
actuators on numerous occasions. During these observations, no
problems were noted where the as-found conditions would affect
valve operability. The NRC inspector reviewed the evaluation
performed by the licensee. It appeared that the conclusions
reached by the licensee were appropriate as no problems were
noted during the review.
p. (Closed) Severity Level IV Violation 285/8832-02: Fire watch patrols
not performed for all nonfunctional fire barriers.
This violation involved a failure of the licensee's fire watches to
patrol all nonfunctional fire barriers hourly. TS 2.19(7)
establishes the requirement for hourly fire watch patrols.
In accordance with 10 CFR Part 50.73, the licensee issued Licensee
Event Report (LER)88-030 to address this problem. In NRC Inspection
Report 50-285/88-46, the NRC inspector reviewed tne details provided
in LER 88-030. This violation is considered closed based on the
review of LER 88-030.
q. (Closed) Unresolved Item 285/8846-01: Control of drawings related to
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This item concerned the failure of the' licensee to translate
information concerning temporary modifications into drawings of a
form pertinent to the plant staff. Followup of this item found it to
be an apparent violation of NRC requirements. 10 CFR 50, Appendix B,
Criterion III requires that measures be established for control of
design changes, including field changes. Appendix A to the USAR,
Attachment 1, paragraph L, states, in part, that the licensee's
Quality Assurance (QA) Program and QA Plan comply with the
requirement of ANSI N45.2.11-1974, " Quality Assurance Requirements
for the Design of Nuclear Power Plants." ANSI N45.2.11-1974 states,
in part, that documented procedures shall be provided for design
changes, including field changes, which assure that information
concerning the change is transmitted to all affected persons and
organizations.
Paragraph 7.3.5 of Procedure 50-0-25, " Temporary Modification
Control," states, in part, that for temporary modifications of
greater than 14 days duration, the cognizant engineer is to send
marked-up copies of design drawings to the document control
department for incorporation into drawing revisions.
Procedure 50-0-25 does not address the existence of nonrepresentative
drawings used by the plant operations staff in the interim time from
the submittal of drawing revision requests to the actual issuance and
incorporation of approved revisions.
On January 15, 1989, the NRC inspector found that on
modification 88-M-011, made to the Safety Injection Refueling Water
Tank Outlet Valves LCV-383-1 and LCV-383-2 on April 1988 had not been
incorporated into control room drawings. A drawing revision request
had not been submitted until October 1988. This modification
involved the addition of a nitrogen bottle supply to supplement the
sizing of the instrument air accumulators. This is an apparent
violation. (285/8903-01)
The licensee provided immediate response to this apparent violation
by updating all control room drawings, as appropriate, to reflect all
temporary modifications installed in the plant. The control room
drawings were updated pr'or to the commencement of plant startup from
the 1988 refueling outage. The NRC inspector reviewed selected
drawings in the control room and verified that the drawings had been
updated.
r. (0 pen) Open Item 285/8846-04: Review of the safety analysis for
operability (SA0) for the wide-range nuclear instrumentation.
This item is related to a review performed by the licensee of a
10 CFR Part 21 report issued by Gamma-Metrics. The report stated
that the assembly used for connection of the cables for wide-range
nuclear instrumentation may contain defects. The Part 21 report
stated that the solder connections may leak.
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The licensee tested the connectors and determined that'the' solder: -
connections leaked. Since the manufacturer has not. designed and'
tested.a repair method for.the connectors, the licensee. issued-
SA0 88-01 to address continued plant operation'if thelinstrumentation-
failed. <
,
NRRand-theNRCinspectorreviewe'dRehision1ofSA0'88-01, dated
January 13,<1989.' During the review, it was'noted that the SAO
appeared'to contain all the appropriate elemtnts to. address' continued.-
plant' operability.
This item remains open pending completion of repairs.t'o the
' instrumentation connectors by.the licensee. . t
4. Licensee Event Report Followup (92700)
Through direct observation, discussions'with licensee personnel,'and
review of records, the following event reports were reviewed to dete mine
that deportability requirements were fulfilled, immediate- corrective 1
action wa's. accomplished, and corrective action to prevent' recurrence.had.
been accomplished in accordance with~the TS.
The LERs listed below are closed.87-038 Excessive Leakage Through Containment Isolation.
Valve PCV-742B'88-002 Inoperability of High-Pressure Safety-Injection
Valve HCV-2987
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88-004 Containment Penetration for the IA System Outside Design
- Basis ,
'88-008 Surveillance Test for Spray Nozzles Not Fully Completed
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l 88-019- Design Error with EDG 2 Damper CircuitryJ '
'88-027 Contaminated Individual;Transpo ted to Hospit'al.88-028 IA Accumulator Assembly Check Valve. Failed the . - -
. Surveillance Test
i
A discussion of the. review performed by.the NRC inspectors forseach LER is
provided below. ,
'
a. LER 87-038 reported an event'where cont'ainment' isolation.. U
'
Valve PCV-7428, located outside'of containment,' leaked excessively
during the performance of the local leak rate test. The~ leak rate
test identified that leakage through the valve was greater than thet e
acceptance criteria provided in TS 3.5(4)c.
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The licensee attempted to adjust the rubber seats of the valve, but
could not access the' seat adjustments from outside the cor.tainment.
The licensee adjusted the mechanical stops on the valve and the valve
subsequently passed its local leak rate test. .
To ensure that the valve seats would be accessible in the event that
the licensee experienced problems with the valve seats, the licensee
committed to reverse Valve PCV-7428. Reversal. of the valve would
make the valve seat adjustments accessible from outside containment.
'
.The licensee reversed Valve PCV-742B in accordance with the l
instructions provided by M0 875536. After reversing'the valve, the
licensee performed a local leak rate test in accordance with
Procedure ST-CONT-3, " Local Leak Rate Testing-Type C." The results
of the test were satisfactory.
The NRC inspector reviewed the actions taken by_the licensee. No
problems were noted during the reviews. J
'
b. LER 88-002 reported an event where Valve HCV-2987 was declared
, inoperable due to the failure of the valve operator. Valve HCV-2987 '
l is used as a high pressure safety injection header isolation valve. ,
I
To restore operability of the valve, the licensee immediately l
installed a temporary nitrogen system to act as an air accumulator
for the valve. The pressure in the nitrogen bottle was routinely' (
checked by operations personnel to verify that' adequate pressure was ]
maintained in the bottle.
l
The temporary nitrogen system was maintained by the licensee entil
the 1988 refueling outage. During the 1988 refueling outage, the
operator for Valve HCV-2987 was disassembled, repaired, reinstalled,
and tested in accordance with the instructions provided on MO 874581.
The valve was not repaired until the refueling outage due to a lack
of spare parts. During postmaintenance testing of the valve, the i
' licensee noted that the accumulator assembly check valve leaked
excessively. The check valve was repaired in accordance with the j
instructions provided on MO 885665 and the accumulator assembly i
successfully passed the testing performed after repair of the check
valve.
The NRC inspector reviewed the actions taken by'the licensee.to
repair the accumulator assembly for Valve HCV-2987. The review
l; included verification that the valve was stroke tested in accordance
with Procedure ST-ISI-SI-1, " Safety Injection Valves - Inservice
Testing," and that the valve met the acceptance criteria. No
problems were noted during the_ reviews.
l
c. LER 88-004 reported an event where the licensee identified that the
IA penetration piping was outside the design basis established by the
'
USAR. The design basis problem was related to the design of the IA
,
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containment penetration where the isolation valve failed open on,the~
loss of IA pressure.~ If the valve failed open, containment integrity. l
would not be established.
'
During'the 1988 refueling outage, the licensee mo'dified the' s
containment penetration piping in accordance with MR-FC-88-11. The
modification changed the valve from a fail-open to a-fail-closed-
'
mode. 1
l
The NRC inspector performed an indepth review of the installation of
~
the modification. . 'The ~results of the review are ~ documented in 1
paragraph 13 of this inspection report.- +
d. LER 88-008 was issued by:the licensee to report an event where 1
'
Procedure ST-NZ-1, " Verification of Open Containment Spray Nozzles,"
"was not completed in accordance'with the procedural 4 instructions.
When'the test was performed, the individualtperforming the test
failed to test 11 spray nozzles' due to the. inaccessibility of the - 1
nozzles. The test results were subsequently accepted without ;j
- evaluating the; significance of not testing.the 11 nozzles.
The NRC inspector reviewe'd the actions taken by the licensee ~in'
response to this event. The results of the review are provided.in
paragraph 3.g of this inspection report.
.
e. LER 88-019 reported a condition which.left uncorrected, potentially
could have prevented proper' operation ,of a. component needed to
maintain the reactor in'a safe shutdown condition. It was discovered
during the reconstitution of the design bases for the EDGs that the:
circuitry for the inlet and. outlet fresh air dampers for.EDG No. 2
could not be isolated from the control. room by a-local means.
Therefore, if'a fire forced evacuation of the control room there
would be no control over this circuitry. - A smart fire could
potentially damage the control room circuitry and cause a; signal '
which would close the dampers, thus rendering the EDG inoperable.
The licensee reported this condition' pursuant'to 10 CFR Part 50.72 on
August 16, 1988.
.As immediate and temporary corrective action, the-licensee-opened the- 1
breaker that powered the circuitry for the dampers. .This action ,
allowed the dampers-to fail open. The licensee then' issued a 1
revision, on August 25, 1988, to Procedure A0P-6, " Emergency' Fire '
Procedure," to instruct operators to isolate EDG'No. 2. controls from
the control room by turning Switch 183/MES to emergency. This
breaker is located in the same room as EDG No. 2'
1
The NRC inspector reviewed the procedure revision and found that the 'l
l
actions stated,. if performed, would result. in the isolation of' i
EDG No. 2 dampers from the control room. The actions would. render l
- the dampers inoperable in an open position. ~;
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As permanent corrective action, the licensee committed to modify the
control circuitry to the dampers so that once control was isolated
from the control room, the dampers could be operated locally. The
licensee scheduled this modification for the 1988 refueling outage.
The NRC inspector reviewed MR-FC-87-63, " Diesel Generator Radiator
Exhaust Damper Valves," and found documentation certifying the
addition of Relays 183MES/ DIX and 183MED/02X to EDG Nos. 1 and 2
damper control circuits. The relays serve to allow local operation
of the diesel dampers.
The NRC inspector then reviewed completed Procedures MR-FC-87-63-M3
and M4, " Diesel No. I and No. 2 Damper Operation Test." The
procedures provided certified records of the proper' operation of the
dampers in an emergency situation. The procedures provided
verification that,-once the 183MES switches were placed in the
emergency mode, the diesel dampers could be operated locally and that
further manipulation of the switches remaining in the control room
had no affect on the dampers.
The NRC inspector then investigated how operations personnel would be
prompted to isolate damper controls from the control room. The NRC
inspector reviewed Procedure A0P-6 and found it had been revised on
December 31, 1988. The procedure clearly provided instructions for
operation personnel to turn the 183/MES switch to the emergency
position. These instructions are included for EDG Nos. 1 and 2.
These actions will isolate the diesels from the control room.
Based on review of the modification package,. test results, and
revision to Procedure A0P-6, it appeared that the licensee had taken
the necessary actions to correct the design deficiency identified in
this LER. The redesigned, installed, and tested circuitry will allow
for local control of the diesel dampers in the event the control room
must be evacuated due to a fire.
f. LER 88-027 reported an event where a contaminated individual was
transported to a hospital in Omaha, Nebraska. The individual was
sent to the hospital in an ambulance due to shortness of breath and
chest pains. This event resulted in the licensee declaring a Notice
of Unusual Event (NOVE).
The NRC inspector performed a followup of this event. The results of
the followup indicated that the licensee's actions were appropriate
in response to the NOUE.
The details of the NRC inspector's review are provided in
paragraph 14.d of NRC Inspection Report 50-285/88-36. Based on the
review previously performed by the NRC inspector, this LER is
considered closed.
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g. LER 88-028 reported an event where the check valve-for the IA
accumulator assembly for Valve-YCV-1045A failed to pass its
surveillance test. The licensee stated in the LER that .the failure
was due to a sliver of material becoming lodged in the check valve
and causing the valve to fail to seat properly.
The licensee routinely' disassembled the tubing fittings. upstream of
the check valve in preparation.for testing activities. The routine
assembly and disassembly of the fitting appeare'd to.cause.a sliver of
thread to break loose and enter the check valve. To ensure that no
other foreign material enters the IA system, the licensee has
committed to install a test tee in the tubing line that supplies air.
to the accumulator. This modification will allow testing of the
^
check valve without disassembly of tubing. fitting.
At the time this problem was .' identified by the licensee, the NRC
inspector was observing testing of the check valve. The review
performed by the NRC inspector is documented'in NRC Inspection
Report 50-285/88-36. As a result of,the review, Open
Item 285/8836-01 was issued to track the licensee's efforts of
installing _ test tees in the IA tubing'for Valve YCV-1045A, as well as
the other accumulator assemblies where the check valves are routinely j
' tested. 4
1
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Based on the issuance of the-open item,-this LER is considered
closed. The licensee's' actions related to the installation of test'
tees will be reviewed during close out of-the open item.
l
Based on the reviews performed by-the NRC inspectors, as described above,
it appears that the licensee took appropriate actions in response"to the
'
identified events to provide timely corrective actions and implementation
of. controls to prevent recurrence of the event.'
No violations or deviations were identified. ,
5. . Operational Safety Verification (71707)
The NRC' inspectors conducted reviews and observations of selected
activities to verify that facility operations were performed in 1
conformance with the' requirements established under 10 CFR, the;11censee's.
administrative procedures, and the TS for the refueling shutdown status of 1
the' plant. The NRC inspectors made several control room ~ observations..to
verify the following:
Proper shift. staffing was maintained and conduct of control room
personnel was appropriate.
Operator adherence to approved procedures and TS requirements was l
evident. q
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' Operability of reactor protective system, engineered safeguards
. equipment, and the safety parameter display system'was maintained.
If not, the appropriate TS. limiting condition for operation was met.
Logs, records, recorder traces, annunciators, panel indications,'and-
switch positions complied with the appropriate requirements.
Proper return to service of components was performed.
M0s were initiated for equipment in need'of. maintenance.
Management personnel toured the control room on a regularcbasis.
Control room access was properly controlled.
Control room annunciator status was reviewed to verify operator,
awareness of plant conditions.
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Mechanical and electrical temporary modification logs were properly ~'
maintained.
Engineered safeguards systems were properly aligned for-the specific.
plant condition.
During observation of the above items,tthe'NRC inspectors noted no
I problems with,the performance'of the licenseA's operations staff. It
l
appeared that the licensee's operations staff was~ adequately performing
their duties to ensure safe operation of the plant.
In the latter part of this' inspection period, the operations staff' started
up the plant from refueling shutdown to_ power operation. A detailed
review of plant startup was performed by the NRC inspectors. . Details of
the review are provided in paragraph 16 of.this inspection report.
,
During the'1988 refueling outage, items were identified,thht were related
to the operational safety of the plant. The'.itemsfwere identified by.the 1
-licensee and require licensee action'to resolve:the items. : The items are
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discussed below:
During an inspection performed in August 1988,-an NRC inspector noted
~
a.
that the lower" temperature. limit specified for the safety. injection
and refueling water tank (SIRWT) may be too low. The lower limit of~
40 F specified by TS 2.-3 appeared to be low because1the licensee
could not provide documentation to indicate that the safety injection.
and containment spray pumps had.been tested with. water at this
l temperature. The concern was that the water may be'too. cold and the.
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clearances within the pumps would be too small due~to' thermal
contraction and may affect pump operability; The licensee evaluated
the lower. limit on the SIRWT and determined that it could be verified
that the pumps would operate satisfactorily at 50 F.
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0'n August'13, 1988, the licensee issued Operations Memorandum 88-07
to' operations personnel to instruct personnel to maintain the SIRWT .s
temperature?above 50 F. This operations memorandum is still~in: ^
effect. ' See paragraph 12 of this inspection. report for a discussion
of the operations memorandum. '
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In' addition to issuing an Operations. Memorandum'88-07, the licensee' ,
also submitted a' request:for amendment:of;the TS to.NRR on
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December 16,.1988. The submittal requested that-the minimum SIRWT
temperature,.as specified in TS 2.3, be' raised from 40 F to 50 F. ,
,
NRR is currently reviewing the; submittal made by.the licensee; This q
item remains open pending-review ~and approval-of the licensee's -
request of amendment of TS 2.3 by_NRR. (285/8903-02)
During outage activities, the licensee.completslyfoffloaded all fuel
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assemblies from the vesselcinto the~ spent fuel pool. The' fuel was -
offloaded-so that the, licensee could shut down' systems for-the 3
performance of maintenance and surveillance on sys.tems that are
required to provide cooling for the" fuel when located in the reactor
vessel.
The TS do not currently address 'which requirements are appropriate
when the reactor is!defueled. For. example, TS'2.1 requires that the
primary system be sampled for boron and chlorides, but does not.
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specify the plant conditions'when'the sampling'mustrbe performed.
When the licensee offloads' all fuel assemblies, primary coolant . , ,
1
i appropriate for sampling is contained only'in'the reactor vessel. To
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sample' the coolant, the : licensee must ' manually ' lower a container into
l the vessel to obtain the' coolant for sampling.'sDuring the 1987
l refueling outage, the licensee requested and obtained permission to
suspend sampling while'the vessel was defueled. .At that time, the
licensee stated that an amendment would be submitted to address which
TS requirements would be appropriate during the defueled condition.
During the 1988 refueling outage, the licensee defueled'the vessel'
and again requested relief.from'the TS requirement for sampling the
primary system. Relief was requested because the licensee did not-
,
submit an amendment subsequent to the 1987 refueling outage. NRR
I approved the relief-request based on an understanding that the~
licensee would submit a request for amendment'of the TS to. address
the defueled plant condition. The licensee stated thatLa submission-
would be made.
- This item remains open pending submission of a request by the
' licensee and approval of the request by NRR. -(285/8903-03)
c. During the 1988 refueling outage, the' licensee modified the RW system
to connect the newly. installed control' room air conditioning (AC)
system to.the RW system. The connections were made by' welding the
supply and return piping for the AC system to.the RW. system. .The RW-
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system has been designated by the licensee as a Class I system in
accordance with USAS B31.7. This code requires that all
modifications to Class I systems be hydrostatically tested after the
completion of modifications.
t
Due to the configuration of the RW system, the licensee determined
that performance of a hydrostatic test was impractical without
ext) 4 ordinary measures. In lieu of performing a hydrostatic test,
the licensee. performed a 100 percent radiograph of.the welds. The
welds were determined to be satisfactory.
~
To address the performance of radiography versus hydrostatic testing,
the licensee issued SA0 89-005. The SA0 evaluation noted that the RW
system is a low-energy and low-temperature system that is consistent.
with designation as a Class III-system as defined by USAS B31.7. The
code allows radiography of modifications in Class III systems. Based
on reevaluation of the system, the licensee stated that radiography
was appropriate.
NRR and the NRC inspector reviewed SA0 89-005. No problems were
noted during the review.
On January 18, 1989, the licensee submitted a relief request to NRR.
The submittal requested a one-time relief of the performance of a
' hydrostatic test.of the RW system. ' NRR is currently reviewing the
licensee's submittal.
This item remains open pending approval of the licensee's submittal j
by NRR. (285/8903-04) .
No violations or deviations were identified.
6. Plant Tours (71707)
The NRC inspectors conducted plant tours at various times to assess plant
and equipment conditions. The following items were observed during the
tours.
General plant conditions, including operability of standby equipment,
I were satisfactory.
!
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Equipment was being maintained in proper condition, without fluid
_
leaks and excessive vibration.
Valves and/or switches for safety-related systems were in the proper-
position.
Plant housekeeping _and cleanliness practices _were observed, including
no fire hazards and the control of combustible material.
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-Performance of work activities was in accordance with approved
procedures.
Portable gas cylinders were properly stored to prevent possible
missile hazards.
Tag out of equipment was performed properly.
Management personnel toured the operating spaces on a regular basis.
During tours of the plant, the NRC inspector noted the items listed below:
I
a. On January 6,1989, the NRC inspector noted that five smoke detectors
for control room cabinets had been removed and placed on a desk in
the control room. The detectors had been removed by operations
personnel because the detectors were alarming to indicate. weak
,
batteries. The NRC inspector. determined that some of the detectors
had been removed for up to a week. In discussions with operations
personnel, it was established that the batteries had not been
replaced and the detectors reinstalled because operations did not
- have a supply of batteries and an I&C. technician was the only craft
l
allowed to replace the batteries.
The NRC inspector discussed this concern with licensee management and
l noted to management that the plant was not within the. established
l design basis with the detectors removed. Licensee management
immediately provided a supply of batteries to operations personnel
and authorized them to change batteries and reinstall the detectors
on an as-needed basic. The NRC inspector verified that the smoke
- detectors had been reinstalled.
b. On January 11, 1989, the NRC inspector noted that temporary Critical
Quality Equipment (CQE) Storage Area 31 was not being maintained in
accordance with established procedures. The area contained items I'
that were not designated as CQE material. No specific problems were
noted since the non-CQE material did not appear to affect the quality
of the CQE material. Upon notification by the NRC inspector, the
licensee removed the non-CQE material from the storage area.
c. On January 16, 1989, the NRC inspector noted that the conduit
connected to the position indication for Valve PCV-1749 had been
pulled loose. Valve PCV-1749 is the containment isolation valve .for
service air. The position indication has been qualified and
installed under the licensee's equipment qualification program. At'
the time of discovery, the plant was in the refueling shutdown mode
and containment isolation was not required.
Upon notification of the problem to the licensee, the conduit was
immediately repaired. The NRC inspector verified that the repairs
had been made during a subsequent tour of the plant.
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c. On January 24, 1989, the NRC inspector noted that contaminated carbon
filters were stored in Room 59 in the auxiliary building. The l
licensee's fire hazards analysis states that contaminated carbon
filters should be stored in metal barrels.
l
The NRC inspector discussed the storage of the filters with the
Supervisor, Radiological Waste who was touring the auxiliary building 4
at the time of discovery. The supervirar stated that maintenance l
personnel were in the process of placing the filters in barrels. '
The NRC inspector toured the auxiliary building a few hours later and i
noted that the filters had been properly disposed of in barrels. 1
No violations or deviations were identified.
i 7. Monthly' Maintenance Observations (62703) .)
1
The NRC inspectors reviewed and/or observed selected station maintenance j
activities on safety related systems and components to verify the
maintenance was~ conducted in accordance with approved procedures,
regulatory requirements, and the TS. The following items were considered
during the reviews and/or observations. l
The TS limiting conditions for operation were met while systems or
components were removed from service.
Approvals were obtained prior-to initiating the. work. 'j
Activities were accomplished using approved M0s and were inspected,
'
as applicable.
Functional testing and/or calibrations'were performed prior to
returning components or systems to service.
Quality control records were maintained.
'
j. Activities were accomplished by qualified personnel. i
l
Parts and materials used were. properly. certified.
Radiological and fire prevention controls were implemented.
The NRC inspectors reviewed and/or observed the-following maintenance ,
activities: !
Replace. steam generator manway covers and studs per MP-RC-2-2-B
(M0 883012)
Replace the stem on Valve SI-214 (M0 886870)
Repair packing leak on Valve HCV-2934 (M0 884744) i
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Repair fitting leaks on the mechanical seal for Boric Acid Pump CH-4A'?
,4
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A discussion of each item.is provided below: .# ]
- a. MO 888012 was issued to replace steam generator. secondary manway
covers and studs. . The NRC inspector' reviewed thel completed work
package. The work was'found to be: accomplished per ~
1 Procedure MP-RC-2-2-B, " Steam Generator-Secondary Manway.;
,
Replacement.'! - With respect to the inspection objectives"citeddin- the
.
i
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body of this inspection paragraph, no problems were noted.
Y b. MO 886870 was issued to replace a bent stem on; Valve SI-214. This
component is a local sample valve for Safety' Injection: Tank SI-68. -l
The NRC' inspector reviewed.the completed work package. After further
inspection of the deficiency, the licensee decided-.it would.be more-
prudent to replace the 1/2-inch socket weld globe valve rather than
~
perform valve disassembly to solely replace.the stem.- Based on. i
review of certified records, it was apparent the valve was replaced,' ,
l
. inspected, and tested in accordance.with approved procedures and l
applicable industry Code USAS'B31.7. *
~
I
c. MO 884744 was initiated to. repair a packing leak evident on j
Valve HCV-2934. This component is a 12-inch motor operated gate
valve which serves as the discharge isolation for Safety Injection
i
Tank SI-6B. Based on review of the completed M0, work was completed-
'
" while the plant was in Mode 3. The documentation indicated that the
l packing gland area was cleaned of boric acid buildup and lubricated.
The packing gland nuts were .then tightened to . secure weepage. All'
work was performed with the valve in the'open position. No
postmaintenance testing was performed on this valve because ,
operations stated it could not be cycled and the valve remains in an-
~
open position during operations,
d. M0 890292 was initiated to repair fitting leakage on the tubing to
t
the mechanical seal for Boric Acid Pump CH-4A. :The NRC inspector
reviewed the~ completed work _ document and, within the'. scope of the
objectives cited in this report paragraph, no. problems were noted.
Based.'on~ observation of maintenance activities by the NRC inspe'ctors, it !
appeared that maintenance personnel satisfactorily performed maintenance l
duties in accordance with the appropriate requirements. '
l
No violations or deviations were identified. '
l
.
8. Monthly Surveillance Observations (61726)'
The NRC inspectors observed selected portions of the performance of,-
and/or reviewed, completed documentation.for the TS-required surveillance. '
> testing on safety-related systems and components. -The NRC inspectors
,
, . verified the followi.ng items during the testing,
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Testing was performed by qualified personnel using approved
procedures. ;
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Test instrumentation was' calibrated. l
!
The TS limiting conditions for operation were. met.<
Removal and restoration of the affected system and/or component were
accomplished. !
Test results conformed with TS and procedure requirements.
Test results were reviewed by personnel other than'the individual
directing the test.
1
Deficiencies identified during the testing'were properly reviewed and -
resolved by appropriate management personnel.
A test was performed on schedule and complied with the TS required. ;
frequency. l
l
The NRC inspector reviewed'the documentation for the following q
surveillance test activities. The surveillance tests were completed ,
during the 1988 refueling outage. 'The procedures used for the test i
activities are noted in parenthesis.
System flow determination for containment air cooling and filtering
.
system (ST-VA-1-F.4)
Containment local leak detection test of the equipment hatch 0-ring
(ST-CONT-1-F.3)
. Inspection of the station batteries (ST-DC-1-F.5)
Containment emergency lighting operability verification (ST-DC-4-F.3)
Calibration of the pressurizer pressure low signal (ST-ESF-1-F.2)
Inservice testing of ' valves in the auxiliary feedwater system i
'
,
1
(ST-ISI-AFW-1-F.1)
Inservice testing of the main steam isolation valves '
(ST-ISI-MS-1-F.2)
1
Inservice. testing of valves in the safety-injection system
'
(ST-ISI-SI-1-F.4)
-
Reactor coolant vent system operability verification (ST-RCGV-1-F.1) i
Inservice inspection of the auxiliary feedwater system- i
'
-(ST-ISI-AFW-4-F.1)
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During review of the completed documentation for surveillance activities,
the NRC inspector noted that it appeared that the testing was perfomed in
accordance with the appropriate requirements. In addition, it appeared
that the licensee adequately evaluated all discrepancies identified during
testing activities.
~
No violations or deviations were identified. .
9. Security Observations (71707)
The NRC inspectors verified that the physical security plan was being.
implemented by selected observation of the following items:
The security organization was properly manned.
,
Personnel within the protected area (PA) displayed their
identification badges.
Vehicles were properly authorized,' searched,'and escorted or
controlled within the PA.
Persons and packages were properly cleared and checked before entry
into the PA was permitted.
I *
The effectiveness of the security program was maintained when
security equipment failure or. impairment required compensatory
^
measures to be employed.
- The PA barrier was maintained and the isolation zone kept free of
transient material.
The vital area barriers were maintained and not compromised by
breaches or weaknesses..
Illumination in the PA was adequate to observe the appropriate areas
at night. .j
Security monitors at the secondary and central alam stations were
functioning properly for assessment of possible intrusions.
During observations of the performance of the ' security, force, the NRC
resident-inspectors noted that'it appeared that- security personnel-
performed their. duties in a professional manner. j
i
A violation was identified during review of. security activities.. The j
violation involved security management's failure.to establish-appropriate
compensatory measures. The violation is discussed in NRC Inspection
Report 50-285/89-06.
No violations or deviations were identified in this inspection report.
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10. Radiological Protection Observations (71707)
The NRC inspectors verified that selected activities of the licensee's
radiological protection program were implemented in conformance with the
facility policies and procedures and in compliance with regulatory
requirements. The activities listed below were observed and/or reviewed.
Health physics (HP) supervisory personnel conducted plant tours to
check on activities in progress.
HP technicians were using calibrated instrumentation, q
Radiation work permits contained the appropriate information to ,
ensure work was performed in a safe and controlled manner.
Persor.nel in radiation controlled areas (RCA) were wearing the
requirad personnel monitoring equipment and protective clothing and
properly frisked prior to exiting an RCA.
Radiation and/or contaminated areas were properly posted and
controlled based on the activity levels within the area.
During this inspection period, the licensee identified an event where an <
individual entered a high radiation area without the proper dosimetry.
The event is discussed in NRC Inspection Report 50-285/89-04.
No violations or deviations were identified.
11. In-Office Review of Periodic and Special Repo'rts (90713)
In-office review of periodic and special reports was performed by the NRC
inspectors to verify the.following, as appropriate:
Correspondence included the information required by' appropriate NRC
requirements.
Test results and supporting.information were consistent with design
predictions and specifications.
Determination that planned corrective actions were adequate for
l
resolution of identified problems. 4
- Determination as to whether any information contained in the
correspondence report shouid be classified as an abnormal. occurrence.
Correspondence did not contain incorrect, inadequate, or incomplete
information.
The NRC inspectors reviewed the following correspondence: ;
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L____________.______________._______._.___._____ _ _ _ __ _ _ _ _ _ _ _ _ _ _
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Application for Amendment.of Operating License for Surveillance. Test
' Interval, dated January 6,' 1989
'
Application;for_. Amendment of Operating License for the Raw Water
System, dated December 31,-1988-
Application for Amendment of Operating License for t'he Contai_nment.
Spray Headers, dated January 6, 1989 - '
Response to NRC Generic Letter 8841 7, dated-January 4, 1989
'
Planned Overhaul of Containment Cooling Fans VA-3A, VA-3B, VA-7C, and
VA-70, dated January 9, '1988
'
,
,
Responses to Request for. Additional.Information Concerning+
'
NUREG-0737, Item II.D.1,- dated January 10, 1989
Review of Open' 0perathons Memorandums, -dat'idlJan0ary 6k 19'89
Response to NRC Bulletin 88-11,' dated January 18, 1989
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December Monthly Operating Report, dated January 13, 1989 j
,
No violations or deviations were ident'ified.
12. Followup on the Status of Operations Memoranda (92701) .
An operational safety team inspection (OSTI) was performed at the FCS from
October 31 through November 10, 1988, by a team of inspectors from NRC .)
Headquarters. During performance of the OSTI, the team identified a '
problem related to the licensee's.use of operations memoranda (0M). The
problem involved the apparent misuse of oms in that the licensee was using
the memoranda to revise or replace instructions contained'in-
safety-related operating procedures. -It appeared that this practice did- 1
not comply with the requirements established by TS 5.8 for changing the- i
content of safety-related procedures, j
On November 23, 1988, a letter was issued to the licensee by the NRC.
This letter stated thatLthe OSTI team had identified the practice of the'
use of oms as a significant finding. The letter further stated sthat the
licensee shall review all outstanding oms to ensure. instructions for
performing activities affecting quality are prescribed in approved _ ..
procedures prior to startup of the plant from the 1988 refueling outage.
.
On January 6,.1989, the licensee responded to the NRC's letter regarding i
the status of a review of oms performed by the licensee. The licensee's l
letter. stated that 26. oms had been reviewed and had been properly
dispositioned. The licensee also stated that, of the 26 memoranda, 22 <!
were. closed and 4 remained open.
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The NRC inspectors performed an indepth review of the disposition of the
oms that were closed by the licensee. A tabulation of the oms is provided
below. The procedure title and revision number provided in the tabulation
indicates which safety-related procedure was revised to include the OM
requirements.
OM 77-12: Control switch operation for pumps with an. automatic start
feature.
This OM discussed the proper method for operation of pumps ~with an
automatic start feature. Under certain conditions during operation
of the control switch, the breaker may not respond to an automatic or
manual start signal if the switch is not properly operated.
The requirements provided by this OM were included in
Procedures 0I-EE-1A, "4160 Volt Circuit Breaker / Switch Operation,"
Revision 1, and 01-EE-2A, "480 Volt Circuit Breaker / Control Switch
Operation," Revision 1.
OM 77-19: Completion of valve checklists.
This OM provided instructions to the plant staff for the proper
method to be used for completion of procedural checklists.
The licensee included the requirement provided by this OM in
Procedure 50-0-29, " Conduct of Operations," Revision 8.
OM 80-16: Fire system impairment. tagging.
This OM provided instructions to operations personnel on.the proper
j method to install impairment tags whenever the fire system is not in i
l
its normal lineup.
The licensee included the information provided by this OM in
Procedure OI-FP-1, " Fire Protection System-Water System,"
Revision 26.
OM 83-01: Operation of the auxiliary building ventilation system.
This OM provided instructions to operations personnel on how to.
operate the auxiliary building ventilation system to ensure
continuous air flow was maintained through the hood in the chemistry
l lab.
The information provided by th1. sM was included in
Procedures 01-VA-1, " Heating, Cooling and Ventilation Normal
Operation Containment," Revision 47, and OP-11 " Reactor Core
Refueling Procedure," Revision 27.
OM 83-02: Definition of maximum / nominal power levels.
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This OM provided licensed operators with the definition of
maximum / nominal power levels that had been issued by.the NRC.
The'information contained in this OM was included in newly issued
Procedure 50-0-45, " Maximum / Nominal Power Level Definition and
Guidelines," Revision 0.
OM 84-02: Transfer of inputs from the P-250 plant computer to the
emergency response facility (ERF) computer.
This OM identified which inputs had been changed _from the plant to
the ERF computer for reference by operations personnel. l
The OM was cancelled and the information was not included in a
procedure since the plant computer was removed during the 1988
refueling outage and all inputs were transferred to the ERF computer.
OM 85-01: Operation of the plant electrical system.
This OM provided instructions to operations personnel on how to
operate the 4160-volt buses if Buses 1A2 and 1A4 were operated from
the 22-kV bus.
The information provided by this OM was incorporated into newly
issued Procedure A0P-31, "All 4160-V Buses Fed From 22-kV,"
Revision 0.
OM 85-04: Shared computer points on the plant and ERF computers.
This OM provided operations personnel information as to which inputs
were shared by the plant and ERF computers.
The OM was cancelled because the plant computer was removed during
the 1988 refueling outage. The ERF computer now contains all inputs.
. .0M 86-04: Loss of the ERF computer during power operations.
<
This OM provided instructions on what actions should be taken in the
event the ERF computer was lost. {;
The information provided by'the OM was included in
Procedure 01-ERFCS-1, " Emergency Response Facilities Computer
System," Revision 13.
- OM 86-05:
~
Subcooled margin monitoring.
This OM provided information stating'that subcooled margin monitoring l
could be performed by accessing the appropriate readouts on the ERF
computer. The OM provided a listing of the appropriate readouts.
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This OM was cancelled and the information was not included in another-
procedure. The basis for not providing the information.in 'a
procedure was that all operations personnel had been adequately;
trained and were aware of the method to'be used for subcooled margin.
monitoring.
OM 86-06: Safety-injection loop injection valve position indicators.
l ' This OMl alerted licensee personnel that the position indication for.
. the loop injection valves was erratic and should not be used'for.
l
- accurate indication of valve position. The OM. stated thst the
I indicator should only be used to show possible valve movement.
The information contained in this OM was-incorporated into-
Procedures,01-SI-1, " Safety' Injection System-Normal Operation,"
Revision 53,'and OI-SC-1, " Initiation of Shutdown Cooling,"
Revision 29.
OM 86-07: Operation of Breakers PCB-2,.PCB-4, and PCB-5.
This OM identified a potential. problem with Break'ers PCB-2, PCB-4,
and PCB-5 related to the breaker being open and air pressure to. the
breaker being lost. Under certain conditions, the breakers have the
potential of exploding.
The information contained in this OM was included ind
Procedures 0I-EE-1, "4.16-kV System-Normal. Operation,"' Revision 15;
01-EG-1, ." Normal Operation 345-kV OPPD Print E-3335-3, " Revision 11;
OI-ST-1, " Turbine Generator Startup," Revision 20; 01-ST-3,1" Turbine
Generator Shutdown,". Revision 18; OP-3, " Plant Startup from Hot
Standby to Minimum Loan," Revision 25; and OP-5, " Plant Shutdown,"
Revision 13.
OM 87-01: Performance of-valve lineups.
This OM defined which plant operations personnel were authorized to.
perform valve position verifications in the field.
This OM.information was included in: Procedure 50-0-29, " Conduct of
Operations," Revision 8.
- OM 87-04: 011 level in Pump DW-46B.
The OM provided operations personnel with information on how to
verify that the oil level in Pump DW-46B, the pure water pump, was
adequate. -i
This OM was cancelled and the information was not included in a j
procedure. During the 1988 refueling outage, the licensee replaced j
Pump DW-46B with a different type that does not-have'~an oil indicator-
sightglass. i
1 1
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OM 87-05: Nuclear detector well Temperature. Indicator (TI) TE-733A.
This OM provided information regarding the' omission of tis TE-733A,_,
TE-736B, and TE-737A for use in determining compliance'with the TS-
limits for concrete temperatures in the. nuclear detector' wells.
The information contained in this OM was included as a note on
Figure III.24 of the Technical Data Book. 4
'
OM 87-06: Security problem in the intake structure. 4
This OM provided information:to operations personnel on what. actions
to take in'the event Circulating Water. Pump CW-1C was' secured.
.
The OM information was included in Procedure 01-CW-1, Circulating I
l Water-Normal .0peration," Revision 11. .;In. addition,. signs;were posted-. '
,1
on Pump CW-1C and Valve CW-14F to alert operations personnel of the I
security problem.
This OM alerted license'd personnel;to the fact that attention should'
be paid to the position of the dampers on.the EDG. Problems were ,
noted with the dampers opening when the emergency diesel generator l
was started.
~ '
.]
This OM was cancelled and the information not included in a j
procedure. During the 1988 refueling outage,.the operator for the
dampers was replaced with an operator equipped with a spring to i
ensure the damper opened on demand. The problem identified by.this .
OM no longer exists. )
OM 88-01: StartingofRadiationM5nborRM-065. q
!
ThisOMprovidedinstructions.tdoperationspersonnel.tomanually
start RM-065 in the event an automatic ventilation-isolation i
actuation signal was initiated. '
)
l This OM was cancelled and the information was not included in a . .;
procedure. During the 1988 refueling outage, the licenseeLinstalled: !
.an automatic start. feature where the operation of RM-065 is initiated '
whenever a VIASris generated. The problem identified by this OM no; ;
longer exists. *
l
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OM 88-02: Operation of the reactor coolant system during mid-loop
operation.
This OM provided information to licensed personnel in response to~
Generic Letter 87-12. The information provided specific items that
the operators should routinely observe during mid-loop operation.
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The OM information was incorporated,into Procedures OP-6, " Hot
Shutdown to Cold Condition," Revision 4; 01-SC-1, " Initiation of. ,
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Shutdown-Cooling," Revision 29; and 01-RC-5, " Reactor. Coolant' System
Draining," Revision 33.
-
0M 88-03: -Alternate hot-leg injection.
This OM provided instructions to operations personnel on.how to .
establish an' alternate;means of hot-leg injection in.the event that
instrument' air system pressure was lost.
The'information contained in this OM'was included in
Procedure E0P-03, " Loss of Coolant Accident," Revision 5.
OM 88-05: ' Cycle 11 calculation error.
i' This OM provided information to the licensed operators regarding
calculational errors made by the licensee that resulted.in
nonconservative thermal margin / low pressure trip setpoints being
issued for Cycle 11. The information provided the: operators with the.
correct setpoints.
-The information contained in this OM has become obsolete.
Amendment 117 of the TS was issued by the NRC and_this amendment
replaced the Cycle 11 information with Cycle 12 information.
OM 88-08: Control room heating, ventilating, and air
conditioning (HVAC) system.
~
This OM provided information regarding~the breach of:the
environmentally controlled envelope for the control room during
modification activities. The information alerted operations-
,
personnel to'take extra precautions while handling potential sources
,
of toxic gas.
This OM was deleted and the information was not included in a
'
procedure. The modifications of the control room envelope have been
completed; therefore, the condition addressed by this OM no longer
-
exists. '
The NRC inspectors reviewed the procedural changes made by the licensee,
as listed above, to verify that the information contained in the oms had
been accurately and completely included in'the procedure change. All the
procedures discussed above were reviewed by the plant review ;
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committee (PRC), approved, and distributed to the appropriate personnel- -!
. for. inclusion in the appropriate manuals. In addition', the'NRC inspectors.
al.so verified that physical plant changes, as. appropriate,- had' been made . , ,
w .to systems or components in accordance with the actions taken during
c_losecut of the OM. No' problems were identified during this. review. ;
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During review df oms, the licensee decided tooleave four o' pen and in
effect pending further actions. The four 0Ms are discussed below:
"' OM 87-07: Fire protection for.the technical support ' center (TSC) '
diesel.
~
This OM was. issued to alert' operations personnel that the. fire
' protection system for the TSC diesel had been deactivated. As an
interim measure, operations personnel were notified'that additional
fireiextinguishers would be-located outside the TSC diesel room.
The PRC reviewed and~ authorized issuance of this OM on January 6,
1989. The OM has been placed in the OM manual in the control. room.
. This 0M will remain in effect until the fuel tank is removed from the.-
TSC diesel room.-
OM 88-04: Light bulb change out.
This OM provides instructions.for operations. personnel on how to- .
track which breaker indicating lights are burnt out so the lights can
be replaced by the electricians. The OM also states that the.
electricians ~ shall request shift supervisor permission prior to bulb -
change out.
,
The PRC reviewed and approved this OM on January 6, 1989. 'A copy of
l the OM has been-filed in the OM manual in the control room. This OM
l will stay in effect until a procedure has been developed and issued'
to address this subject.
l OM 88-06: TS 2.4 and an inoperable raw water pump.
This OM provides operations personnel'with instructions as to the
actions to be taken when a raw water pump becomes inoperable.
Currently, TS 2.4 does not require entry into an LC0 when one pump is
inoperable. This OM supplements the_TS requirement by stating that
entry into an LCO is required when one pump is. inoperable.
~
The PRC reviewed and approved the issuance of this OM on January 6, -i
1989. The OM has been placed in the-0M manual'in the control. room ;
' :and has also been placed Lin the TS . manual for reference' by. operations l
personnel. 4
On December 31,1 1988, the licensee submitted a request for amendment
of.the TS to NRR. This amendment request is currently urider review
by NRR. When the amendment is approved, ~ OM will be cancelled.
OM 88-07: Thermal shock to the safety-injection pumps. l
This OM was istued to change the minimum temperature allowed.in the
SIRWT from 40 F to.50 F'. The minimum temperature of 40 F in the.
.SIRWT is specified in TS 2.3. The change was made by the licensee to
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ensure that the safety injection pumps would not be subjected to
thermal shock.
The PRC reviewed and approved the issuance of this OM on January 6,
1989. The OM has been placed in the OM manual in the control room.
The OM has also been placed in the control room copy of the TS for
reference by operations personnel.
On December 16, 1988, the licensee submitted a request for amendment
of the TS to NRR. The amendment requested a change in the minimum
SIRWT temperature and is currently under review by NRR. When the
amendment is approved, the OM will be cancelled.
In addition to the above actions, the licensee revised Procedure S0-0-13
" Operations Memorandums," to establish requirements for the control of
issuance of oms. The requirements state that the PRC shall review all oms
and that an OM shall not be issued if the OM changes the intent of any
safety-related procedure. In addition, the procedure established a
requirement that all oms shall be reviewed by each licensed operator and
the operator shall sign a sheet to signify that the operator has read and
understood the content of the OM. For each of the four Otis currently
open, a read-and-sign sheet had been completed.
The NRC inspectors reviewed the actions taken by the licensee, as
discussed above. No problems were noted during the reviews.
No violations or deviations were identified.
13. Installation and Testing of Modifications (37828)
During this inspection period, the NRC inspectors examined the
installation and testing of modifications to plant systems performed
during the 1988 refueling outage. The NRC inspectors examined the
modification packages to ensure the following elements were addressed:
Work was performed in accordance with approved procedures.
The licensee had performed a safety analysis pursuant
10 CFR Part 50.59 for the design, construction, and testing of the
modification.
- Modification design was in accordance with applicable codes and
standards.
Modification material had been appropriately designated as
safety-related.
Impact upon the USAR had been addressed.
Impact on the TS had been addressed.
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By direct observation, the NRC inspectors verified that modification
installations were performed by qualified workers and in accordance with
approved procedures and drawings contained in the work package. The NRC
inspectors independently verified that the installed configuration was per
design and that the system licensing criteria remained valid following the
modification. Where applicable, the NRC inspectors verified that jumpers
and/or lifted leads were appropriately used and controlled during the
installation phase of the modification.
. The NRC inspectors also verified the following with regard to
l preoperational testing:
1
The installations were adequately prepared for preoperational
testing.
Testing was perfomed in accordance with approved procedures.
Test results were reviewed and evaluated against previously
established acceptance criteria.
"
Preoperational training was conducted, where appropriate.
The NRC inspectors reviewed the following modifications:
HR-FC-87-63, " Diesel Generator Radiator Exhaust Damper Valves"
MR-FC-88-11, " Penetration M-73 Upgrade, IA System"
Details of the individual reviews are discussed below:
a. On December 18, 1988, the NRC inspector witnessed the connection of
the IA system to newly installed Valve YCV-871E per MR-FC-87-63.
Valve YCV-871E controls the position of the radiator exhaust damper
for EDG 1. It was replaced with a valve of different design to
ensure that the radiator exhaust damper will fail open independent of
valve operation.
The NRC inspector witnessed an I&C technician plumb new copper wb og
to the new Valve YCV-871E per a sketch contained in the modificac son
package. The tubing was installed with compression fittings. The
technician was witnessed to perform a soap-bubble leak test of the
new fittings after installation. No leaks-were noted. To apply
pressure to the newly installed system, the technician had to open an
IA isolation valve. This type of valve manipulation by nonoperations
personnel is allowed by Procedure S0-0-29, " Conduct of Operations."
The technician was observed to return the IA valve to the isolated
position after the successful leak check.
Modification Package MR-FC-87-63, contained Test Procedure liR-FC-87-63-M3
designed to perform a functional test of the newly installed damper
control system. The NRC inspector missed the window of opportunity
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available to witness this test. However, a review of the certified
test records indicated that the dampers functioned properly and
failed open.
The NRC inspector did witness Surveillance Test ST-ESF-6, " Monthly
Load Test of EDG 1," which did verify operability of the diesel
radiator exhaust damper subsequent to the modification. On
December 28, 1988, the NRC inspector witnessed the applicable portion
of ST-ESF-6 and found the dampers to operate within the specified
During January 1989, the NRC inspector reviewed the completed copy of
MR-FC-87-63. In the assessment by the NRC inspector, the
modification package indicated a thorough analysis had been performed
with regard to the effact the modification would have on the plant as
well as preapproved design standards. Additionally, the package was
found to be comprehensive with regard to design installation and
testing.
Based on the reviews performed by the NRC inspector, it appeared that
the licensee installed the new operators for the EDG exhaust dampers
in accordance with the appropriate instructions. No problems were
identified during the reviews,
b. The NRC inspector reviewed the installation of MR-FC-88-11.
MR-FC-88-11 was issued by the licensee to provide instructions for
modification of the containment penetration for the IA line. The
previouspenetrationwasinstalledwithasinglevalve(PCV-1849),
located outside containment, that failed open on a loss of IA
pressure. With this type of failure mode, the licensee could not
ensure that containment integrity could be established in all cases.
The licensee placed interim administrative controls on the
establishment of containment integrity by issuance of a revision to
Procedure A0P-17, " Loss of Instrument Air." The procedure revision
required that an operator shut the manual valve upstream of
Valve PCV-1849 in the event that IA system pressure was lost.
To upgrade the penetration assembly, the licensee installed an
additional valve in the assembly. The new valve was installed inside
containment. In addition, the licensee changed the failure mode from
fail-open to fail-closed.
During observation of the installation of this modification, the NRC
inspector identified problems related to the control of system
cleanliness during installation. The problems identified by the NRC
inspector were issued as a Notice of Violation in NRC Inspection
Report 50-285/88-46.
During review of the completed official copy of MR-FC-88-11, the NRC
inspector identified a number of concerns. These concerns are listed
below:
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Torque wrench information was not recorded in the installation .
procedure to provide for equipment traceability.
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The boundary between the CQE. (safety-related) and non-CQE
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portions: of. the IA. system was apparently . incorrectly
established.
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The modification did not appear to be installed in accordance
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with the approved instructions.
1
LThe NRC inspector and licensee personnel were in the process of
discussing the concerns' listed above~at the end of this inspection
period. Conclusions as to the acceptability of the modification had
not yet been established. No concerns were identified that affected
the safe operation or operability of the IA system containment.
penetration._ This item remains unresolved pending resolution of the
concerns between the NRC' inspector and licensee personnel.
(285/8903-05)
No violations or deviations were. identified.
14. Followup on Onsite Events (93702)
During this inspection period, the licensee experienced a number of
events. Each event is discussed below:
a. On December 31, 1988, the licensee experienced an event where the~
ventilation isolation actuation signal ~(VIAS) was. initiated during
modification activities. Personnel were working to install a .
modification in the main control board, when wires were inadvertently
,
shorted together. Shorting of the wires caused Inverter D~to switch
from an on-line status to the bypass mode. When Inverter D switched
l modes, power was momentarily lost, resulting in the unblocking of
Channel B of the pressurizer pressure low signal (PPLS). The PPLS,
in turn, caused initiation of a VIAS.
At the time of the event, the plant'was in a_ refueling outage and
work was being performed on the control room, envelope. For.this
I reason, the control room ventilation system was in the recirculation
i mode. Consequently,-the' initiation of. the VIAS did not affect any
plant systems.
The NRC inspector reviewed the actions taken by the licensee and.
noted no problems. Review of this event will be performed during
followup on the LER to be issued by the licensee,
b. On January 8,1989, the licensee experienced an event where shutdown
cooling flow was lost. At the time shutdown cooling flow was lost,
the reactor vessel was completely fueled and the vessel head was
fully installed.
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During~ calibration of thecshutdown margin monitor, an'e'lectronic
signal for' simulation of a pressure increase'was applied to the '
, , monitor. This signal caused plant circuitry to sense _ a high pressure
signal. The pressure switch used to sense the pressure on the
shutdown cooling system inlet piping' sensed the over pressure signal
and caused both shutdown cooling system inlet valves.to go shut.
, Shutting of the valves 1 caused the loss of. flow.
The pressure switch senses pressure.in the shutdown cooling system-
'in'fet_ piping to automatically shut the inlet piping:in the event of a
L pressure increase. 'This automatic feature'is provided-to ensure.that..
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'the low pressure' piping in the shutdown cooling system is not,
overpressurized.
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Flown was lost for approximately.5 minutes before.the valves were ,
reopened. During the loss of flow, no, temperature rise.was noted'in
.the reactor vessel. ,
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The NRC inspector reviewed the actions taken by the licensee in.. a
response to this event. No problems were noted. . Followup.will be~ '
performed during review of the LER to be issued-by the licensee. H
c. On January 12, 1989, the licensee was_in the process of' star'ing t up
p the reactor coolant system per; Procedure 01-RC-3, " Reactor Coolant . -
l System Startup." Step IV.A.4 required the establishment of charging _ .
-and letdown per. Procedure 0I-CH-1, " Chemical and Volume, Control- H
, System-Normal Operation." This procedure required operators to H
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verify valve lineups per Checklist A of Procedure 01-CH-1. The !
operators had a verified checklist dated-January 6,1989.
An operator started Charging Pump CH-1B at 2:15:18 a.m. The operator
realized that loop-injection Valves-HCV-238 and HCV-239 were closed-
and stopped the pump.at 2:52:17 a.m. Pressure peaked at 2920.3 psig.
The operator realigned the system and restarted Pump CH-1B at
2:52.52 a.m. At 2:58:06 a.m., operators received a volume' control
' tank low-level alarm. Charging was stopped. Investigation found.the j
packing of Pump CH-1B had ruptured.
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Design pressure of this portion of-the charging system is 2750 psig.
A relief valve, CH-182, is installed to. relieve at 2735 psig.' The
pressure peak exceeded design pressure by 6.2 percent. .
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The NRC inspector met with the system engineer assigned to followup
on the event. M0s had been written to replace the pump packing, .
verify the relief valve setting, verify the pressure transmitter
calibration had not been exceeded, and verify that the flow ,
, transmitter calibration had not been exceeded. . Additionally, an
analysis would be' performed to verify the peak pressure experienced ,
would not have any adverse effects on the piping. ,!
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This:section of piping was last hydrostatically tested to a pressure
of 3420 psig. As an immediate corrective action, the licensee had 4
the piping and associated fsupports inspected by system engineering; $
No deformation was.noted. The pump was inspected and the packing-
replaced by machinists on January 12, 1989. .l System hydrostatic: test
pressure per Procedure ST-ISI-CVCS-4, " Inservice Inspection CVCS- ,
Pressure Test," was verified to be greater than the peak pre'ssure of-
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2920 psig observed during the event. .The licensee verified'that ,
,l
design pressures were.not exceeded on pressure and flow transmitters '
effected by the event. j
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rocess,ofrevisingapplicableproc$dures.t'o
The
ensurelicensee
that anisoperator
in the p? verifies!he has a flowpath prior.to. ,
energizing a positive displacement charging pump. At the end of'this
inspection period, the licensee had not yet determined the most
effective way to supplement the valve lineup sheets. . The issuance of, '
ethis procedure ch'ange and subsequent' review.by the NRC inspector is
considered an open item (285/8903-06).
d. On January 13,1989,l the licensee identified'a problem during.
reconstitution of the design basis. A licensee contractor. identified-
a problem related to the thickness of the concrete directly below-the
main steam header. 1
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A crack in the expansion bellows, installed.in the containment
penetration for the' main steam header could. impinge on the concrete ] .
and the impingement loads could cause the concrete to: fail. Failure
of the concrete would breach the barrier protecting the switchgear ;
room, located directly below the main steam header room, and result
in damage to the electrical components in the switchgear room due to 1 '
environmental conditions.
In approximately 1981, the licensee removed a portion of'the concrete
under the main steam line to facilitate replacement of a' defective
expansion bellows for the main steam line penetration. ~. Upon'.
completion of bellows replacement, the licensee did not. replace the;
concrete that was removed. ~The licensee, in order to ensure that a.
break would not affect equipment operability in the switchgear room, i
installed a barrier around the expansion bellows. .The barrier was ;
installed to deflect- the steam jet in the event of a' break inside the d
containment penetration assembly. The barrier was installed in ,
accordance with MR-FC-89-08, "MainsSteam Line Penetration M-94! !
and M-95 Bellows Pad Reinforcement."
The NRC inspector evaluated the' actions taken by the licensee. The
evaluation included review of MR-FC-89-08 and an.in plant review of. , i
the installation of-the barriers.-- Based on the areas reviewed, no-
. problems were noted.
e. On January 16, 1989, the licensee identified a. problem with the
electrical supply breaker for Motor Control Center (MCC) 4C1. During ,
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plant startup, MCC 4C1 was loaded wii.h the normal plant electrical -!
loads. The supply breaker tripped due-to the amount;of, loads on the 1
MCC.
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.The-licensee removed the breaker, performed maintenance on thei
breaker, and returned-the breaker to. service. After maintenance had
beenLperformed, the breaker operated satisfactorily'in service. :
Infollowuponthi$-problem,thelicenseeidentifiedthat'achange
had been made to Procedure.CP-MCC-4C1, "MCC-4C1 Breaker,"'to change i
the long-term test'value from 320 to.288 amps. The MCC-401 breaker
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was tested during the outage using the new procedure. After the
.
problem developed, the licensee revised Procedure CP-MCC-4C1 to
I- change the t'est value from 288 back.to 320 amps. After the' breaker
was tested using a test'value of 320' amps, no further problems;were1
experienced.
By the end of,this inspection period,-the licensee had not determined
.the exact cause of the supply breaker.for MCC 4C1 tripping. The '
licensee continued to review the potential problem. This item I
remains open pending a completion of the review by'the licsnsee and a J
review.of the licensee's evaluation by the NRC. (285/8903-07)
f. On January,17,1988, the licensee experienced an inadvertent VIAS..
The VIAS was' caused by a high alarm on' Radiation Monitor _RM-050. ,
RM-050 monitors:the radioactive particulate level in containment. At ]
the time of the event, the plant was in Mode 3.
1
Operations. personnel noted t'at the reading on RM-050'had been.
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trending upward as the temperature ~of the RCS was increasing. .This
has been a normal trend during past plant startups. Operations .
. personnel requested that an I&C technician increase.the setpoint for; ,
'RM-050 from the shutdown value to the operating value. =Before the I
I&C technician could adjust the setpoint, RM-050 went into alarm and'
initiated a VIAS. h
i
All equipment' functioned normally upon receipt of the VIAS. The I&C I
technician reset the setpoint to the operating value and no further
problems have'been encountered. <
a
The NRC inspector reviewed theLactions taken by the licensee and
noted no problems. Review of this event will be performed during. .
followup on the LER to be' issued by the licensee. i
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.g. On January"24, 1989, the licensee identified a problem during plant l
startup related to weeping of a pressurizer' code safety valve. 'The !
problem was noted by operations personnel-when the tailpipe
temperature of.the safety valve increased.
!
The-licensee contacted the valve manufacturer who stated that the
weeping was caused by loss'of thecloop seal in-the pipin upstream of
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the safety valve. The valve manufacturer stated' that loss of the
loop' seal would cause uneven heating of the valve disk and-thus would
cause disk warpage. Warping of-the disk caused steam to leak past.-
Licensee personnel entered the containment' andiinspected the loop-
seal piping. They found the piping to be. insulated andLremoved the
insulation. Without. the insulation' installed, the steam in the .
piping condensed and formed a loop seal. L After the loop seal was
established, the safety valve stopped weeping and the ta11 pipe
temperature returned to normal. The licensee replaced the insulation
on the. loop seal and no further problems were noted.
Initially, the licensee experienced difficulty in deciding-whether or:
not it was appropriate to insulate.the loop seal piping.. .The
licensee reviewed the design basis'for the loop seal piping and-
determined that application of insulation was appropriate. The
licensee determined that the insulation was installed throughout'the:
last operating cycle and experienced no problems. However, the-
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licensee could not determine the exact reason as to'why the loop seal-
was not established during plant startup.
At the end of this inspection period, the: licensee was still.in the
process.of determining the exact cause of the weeping and determining
why a loop seal was not established during: plant heatup.- This item.
remains open pending resolution of these concerns and review of the -
licensee's evaluations by the NRC. (285/8903-08)
15. Verification of Containment Integrity (61715)
l
l The NRC inspectors performed a review to verify. that the licensee had,
l established containment integrity prior-to commencing heatup of the RCS
above 210 F. Verification.of containment integrity was established by
reviewing the items listed below:
Verification that selected electrical'and mechanical barriers had
been properly installed.
Containment isolation valves were properly positioned as required by
the appropriate documentation. -
- Local leak'_ rate tests'were performed on' the personnel airlock,
equipment hatch, and fuel transfer tube. +
Containment 1_ntegrity was . verified by. performance of the . items listed-
'
below:
- Review of the local leak rate test performed on the personnel air
lock, the equipment hatch, and the fuel transfer. tube. The leak rate
tests were performed.in accordance with Procedure ST-CONT-2,
" Containment Local Leak Detection Tests-Type B."
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Walkdown.and verification lof selected valve positions related to l
containment integrity as provided:in Procedure'0I-CO-5, " Containment
Integrity Checklist."
slkdown,and verification of. selected valve. positions as provided in.
Procedure OI-RC-2B, ." Reactor Coolant Vent and Leak Test Instruction." l
- Plant tours were perfomed to verify selected mechanical and
electrical penetrations had been properly reinstalled after use '
during the refueling outage.
During performance'of the activities listed above 'no containment'
isolation valves or mechanical / electrical penetrations were found that
were not in the proper position or properly-installed.-
During the last operating cycle, the. licensee identified an event where a -
tubing cap had been removed from an instrument line'and not replaced. The
instrument line was connected to the pressure switch used to detect
containment pressure. As a result of the failure to reinstall _the tubing
cap, containment integrity had been lost.
In response to this event, the licensee extensively revised
Procedure 01-C0-5. This. procedure had previously provided general-
instructions for the technician on how to verify that containment ,
integrity had been established for each mechanical and electrical-
penetration. The revision issued by the licensee to Procedure 01-C0-5
contained specific, clear, and concise instructions for checking each
mechanical and electrical containment penetration.- For example, the .
instructions provided a listing of all caps, valves, and test connections.
for each penetration. A signoff blank was provided so each individual
component associated with a penetration was individually verified to be
installed.
The NRC inspectors performed a detailed walkdown of 109 of the -
approximately 170 containment penetrations installed in the plant. The
109 penetrations reviewed by the NRC inspectors included all mechanical
penetrations that had piping installed through them. During this review,
the NRC inspectors found no instances where containment integrity had not
been properly established. In addition, the NRC inspectors did not
identify any instances where valves,' caps, and test connections were not
identified by the licensee.
Based on the reviews performed by the NRC inspectors, it appeared that the
licensee had adequately established containment integrity. In addition,
the reviews indicated that the licensee had appropriately addressed the.
necessary corrective actions to ensure containment integrity was -
maintained by the revision issued to Procedure 01-C0-5.
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16. Observation of Activities During Plant Startup (71711 and 71715)
The NRC inspectors reviewec' a variety of activities associated with
startup of the plant from tae refueling outage. The reviews were
performed to verify that plant startup was conducted in accordance with
the TS, operating proceduies, and administrative procedures.
During the performance of this portion of this inspection, the NRC
inspectors observed and/or reviewed selected activities in the following
areas:
Plant systems were returned to service prior to plant startup.
Plant startup, heatup, and. approach to criticality were conducted in
accordance with approved procedures.
Operational activities were evaluated to verify operators were
attentive and responsive to plant parameters and conditions,
procedures were used and followed, equipment changes were
appropriately documented, and plant operating conditions were 4
effectively monitored.
Safety-related system walkdowns were performed to verify valve and
switch positions.
'
Response to control room alarms was timely and appropriate actions
were taken.
The NRC inspectors performed the activities listed below to verify proper
completion of the inspection elements listed above:
Walkdowns of selected portions of the high-pressure and low-pressure
safety injection, auxiliary feedwater, and emergency power systems
were performed to verify correct system lineup and equipment
operability.
- A walkdown verification was performed in the control room for all
safety-related switches to verify the switches were in the correct
position.
- Verification of completed system walkdowns was performed to ascertain
that the systems had been aligned in accordance with the appropriate
procedures. The system alignment checklists were reviewed for the
systems listed below: .,
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Checklist Number
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System l
CA-1-CL-C IA System
CO-5-CL-A Containment Integrity
Verification R
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ES-1-CL-A Eng.aeered Safeguards i
Contral' Switches.in the
Control. Room; }
SI-1-CL-A High-Pressure SI-
' System. 1
FW-4-CL-A Auxiliary Fiedwater
System
' Review of the tag-out log to verify that no equipment 'necessary fore l
safe plant operation remained tagged out efter plant startup:had been
initiated.
During review of. the above items, no problems .were noted.
^
The NRC inspectors _also observed operations personnel ' perform major plant
evolutions. The results of the observations are discussed below:- 1
a. On January 29, 1989, the NRC. inspector observed control room
activities during plant startup from hot shutdown'(Mode 3) to hot
standby _(Mode 2). The observation.of activities-was performed to'
verify that the control room operators performed:the. plant'startup
evolution in accordance with the appropriate requirements.
The startup was performed with all rods out, except for Group .IV,.by
dilution of the RCS to reduce the boron concentration.. The'startup.
was performed in' accordance with Procedures OP-2, " Plant .Startup from
Cold Shutdown to Hot Standby," and OP-7, " Reactor Startup."-
During observation of the approach to reactor criticality,' the NRC
inspector made the following observations:
- Operations personnel used and followed the-procedures for
startup. <
"
A' dedicated operator was stationed at .the steam generator
controls for'feedwater.
A dedicated operator was stationed-at the control board to
continually control. and observe the results of each RCS ' dilution
-
evolution.
~* Shift turnover was performed in a professional manner.: The '
offgoing shift briefed the oncoming shift on all evolutions that'
were in progress .at the time turnover was performed.
The nuclear engineer frequently briefed _ operations personnel onL
the data that was being recorded and the significance.of the
data,
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The onduty shift supervisors maintained cognizance over all
plant evolutions and appropriately directed operations personnel
to perform specific duties based on plant status.
b. On January 31, 1989, the NRC inspector observed control-room
operators make the plant transition from hot shutdown (Mode 2) to
power operations (Mode 1). The mode change was made in accordance
with Procedure OP-3, " Plant Startup from Hot Standby to Minimum
Load."
The NP.C insps.ctor noted the following items:
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A senior operator controlled reactivity addition; two operators
I manipulated feed flow, steam flow, and turbine controls; one
l operator prepared for grid synchronization; and two shift
technical advisors monitored plant parameters via the plant
computer.
The operation was under the supervision of an on-duty shift
supervisor and an off-duty shift supervisor and the Supervisor,
Operations.
The NRC inspector observed grid synchronization and subsequent
rise to 30 percent power.
The NRC inspector noted that the evolutions were well coordinated and
performed by knowledgeable individuals in accordance with approved
procedures.
Based on the reviews and observations made by the NRC inspectors, it was
apparent that operations personnel performed the major plant evolutions
associated with plant startup in a highly professional manner. Operations
personnel demonstrated their abilities to operate the plant during the
normal startup operations and to respond effectively to off-normal alarms
that occurred during the startup. Operations shift supervisors
demonstrated their ability to provide oversight of all plant evolutions
and to direct operations personnel in the performance of their duties.
No violations or deviations were identified.
17. Review of NRC Bulletin 85-03 (92702)
NRC Bulletin 85-03, " Motor-0perated Valve Common-Mode Failures During
Plant Transients Due to Improper Switch Settings," was issued to alert
licensees of the potential of improper switch setting on motor-operated
valves. Bulletin 85-03 required licensees to take actions to verify that
switches were properly adjusted.
In response to Item e of Bulletin 85-03, the licensee identified the
selected safety-related valves, the maximum differential pressure of each
valve, and the program established to ensure valve operability. The
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information was'provided in letters dated May;15 and September 15, 1986.
E .
Based'on'a review'ofJthe licensee submittcis, additional information was
requested, in a letter dated August. 4, 19877 from the'NRC. ,The licensee
provided th'e additional"information in a letter dated September 15, 1987.
The NRC reviewed .the ' licensee's submittal and noted that'the licensee met:
'the.'equirementr,
r of Bulletin.85-03. The NRC'also noted that the program
established by the licensee'to ensure _ valve operability was acceptable.
In NRC Inspection Report 50-285/87-17, a review was performed to verify.
the. adequacy of the implementation of the licensee's. program. Duringlthe
review, problems were notbd with the^ method used by.the licensee to set
the switch for the motor-operated' valves'. .The~ licensee is ~still in the
process of resolving .the problems identified in the inspection report.
.
No.
violations or deviations'were identified.
L 18. Unresolved Item
L ~
An. unresolved item is a' matter about which more information is required'in
~
order'to determine whether it is acceptable, a violation, or'a deviation.
One unresolved item is discussed in this inspection report.
Item Paragraph Subject
285/8903-05 13 Installation of the
modification for
upgrading.the IA-
containment. penetration"
'19. Exit Interview . ,
The NRC inspectors met with 197 W. G. Gates (Plant Manager) and other
members of the licensee staff on February <2,1989. The meeting attendeest
are listed in paragraph 1 of this inspection report. At this meeting, the
NRC inspectors summarized the-scope'of the inspection and the findings.
.
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