ML20245C933
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| ML20245C933 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/15/1989 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20245C919 | List: |
| References | |
| 50-285-89-22, NUDOCS 8906260331 | |
| Download: ML20245C933 (29) | |
See also: IR 05000285/1989022
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APPENDIX B
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION IV ;
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NRC Inspection Report: 50-285/89-22 "
Licensee: DPR-40 j
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Docket: 50-285
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Licensee: Omaha Public Power District (OPPD) ..
1623 Harney: Street j
Omaha, Nebraska 68102 l
1 Facility Name: ' Fort Calhoun Station (FCS)
Inspection At: FCS, Blair, Nebraska
1 - Inspection Conducted: May.1-31,'1989
Inspectors: P. H. Harrell, Senior Resident Inspcctor
T. Reis, Resident Inspector
R. ' P. Mullikin, Project Engineer-
Approved: 7'? 4 -hI-8
T. F. Westerman, Chief, Project Section B Date
Division of Reactor Projects
Inspection Summary
Inspection Conducted May 1-31, 1989 (Report 50-285/89-22)
Areas Inspected: Routine, unannounced inspection including review of ]
previously identified items; licensee event report followup; operational safety i
verification; plant tours; monthly maintenance observations; monthly
surveillance observations; security observations;" radiological protection
observations; in-office review of periodic, special, and nonroutine event- )
reports; followup of onsite events; and NRC Bulletin followup.
- Results: During this inspection period, the NRC inspectors reviewed the areas !
discussed below. The discussion provides an overall evaluation of each area. J
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The NRC inspectors reviewed the actions taken by the licensee in response to !
previously identified items and licensee event reports. Based on reviews of
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the actions taken by the licensee, it appeared that the licensee had
appropriately implemented.both the short- and long-term actions to prevent- l
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recurrence of the identified problems, events, and concerns.
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8906260331 890620 7
PDR ADOCK 05000285 '
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During observations of activities and evolutinns performed by the operations
staff, the NRC inspectors noted no problems with the performance of the staff.
It appeared that the licensee's operations staff performed their duties in a 3
professional manner to ensure safe plant operation.
With respect to the management of operations, the NRC inspectors noted that,
with the transfer of responsibility to a new plant manager and the appointment
of two new assistant plant managers, the new plant manager failed to designate,
in writing, the succession of authority during his absence. This is a
violation of Technical Specification (TS) 5.1.1 of minor significance.
The NRC inspectors performed numerous tours of the plant during this inspection
period. The NRC inspectors noted continued progress in the industrial coatings
portion of the facilities upgrade program. The NRC inspector also noted that
coatings were being applied in a controlled, preplanned, and industrially safe
manner.
During the followup of an item discovered during a plant tour in March 1989,
the NRC inspectors found the licensee to have installed potentially !
nonfunctional fire doors in a nonevaluated condition. Contrary to TS 2.19, the !
licensee failed to establish a fire watch patrol for the potentially
nonfunctional fire doors.
The NRC inspector noted both positive and negative attributes with respect to )
evaluation of maintenance activities. It appeared that systems engineering was
taking a more proactive role in maintenance as evidenced by the emergency ;
diesel generator (EDG) system engineer's perseverance in troubleshooting and 1
correcting minor anomalies with the static exciter and governor of EDG 2. On l
the negative side, deficient maintenance procedures continued to exist in the
maintenance program.
In the area of surveillance testing, the licensee demonstrated proficient use
of adequate procedures. A concern was noted in the inadequate communication
between licensee management and the NRC with respect to potentially significant
findings from surveillance testing. See paragraph 12.e.
During observations and reviews of activities performed by security personnel,
the NRC inspectors noted prompt action taken with respect to safety concerns.
However, concern was expressed for the failure to communicate an event of
potential significance to the NRC. See paragraph 9.a.
Several discussions in this report relate to engineering support to operations
and maintenance. It appeared the systems engineering and special services
sections were beginning to benefit the licensee. Both are relatively new
employed concepts. Engineering efforts resulted in efficient correction of
problems related to motor-operated valves (MOVs), a steam leak, and the
emergency diesel generators.
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DETAILS
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- 1. . Persons Contacted .j
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*K. Morris, Division Manager, Nuclear Operations j
*R. Andrews, Division Manager, Quality and Environmental Affairs i
*S. Gambhir, Division Manager, Production Engineering j
G. Peterson, Manager, Fort Calhoun Station i
*A. Richard, Assistant Manager, Fort Calhoun Station j
J. Tills, Assistant Manager, Fort Calhoun Station
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J. Adams, Reactor' Engineer .
J. Bobba, Supervisor, Radiation Protection
*J. Fisicaro, Manager, Nuclear Licensing and Industry Affairs'
*J. Gasper, Manager, Training
*J. MacKinnon, Member, Safety Analysis and Review Committee
*R. Jaworski, Manager, Station Engineering
J. Kecy, Supervisor, System Engineering
*L. Kusek, Manager, Nuclear Safety Review Group
D. Lieber, Supervisor, Security Operations
T. Mathews, Station Licensing Engineer
*D. Matthews, Supervisor, Station Licensing
K. Miller, Supervisor, Maintenance- .
*W. Orr, Manager, Quality Assurance and Quality Control
*H. Sefick, Manager, Security Services
*C. Simmons, Station Licensing Engineer
*D. Trausch, Supervisor, Operations-
*S. Willrett, Supervisor, Administrative Services
* Denotes attendance at the monthly exit interview.
The NRC inspectors also contacted other plant personnel, including
operators, technicians, and administrative personnel. j
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2. Plant Status
On April 28, 1989, the licensee reduced plant power and entered Mode 2 j
(hot shutdown) to perform testing and maintenance related to the
following:
The' licensee tested the mechanical overspeed trip on-the main .
turbine. The test was requested by the turbine manufacturer,. General ,
Electric (GE), to verify operability. The results of the test j
indicated the trip operated satisfactorily. j
The 161-kV offsite power supply was removed from service to replace
insulators and ground wires on the offsite distribution system power
lines. The 161-kV power supply n s removed from service on April 29, ,
1989, and returned to service the same day. No problems occurred
I while the system was out of service. i
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April 29,1989, the licensee removed the 345-kV offsite power supply- 1
for~ maintenance activities. ~The 345-kV power supply was- returned to -
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service'on April 30, 1989. No problems were encountered while the-
' power supply was-out of service.
.During. this outage time, the licensee' inspected the main condenser
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and found two' leaking tubes. The tubes were plugged without' ,
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On April 30, 1989, the licensee performed full flow tests of the . ,
motor-driven and steam-driven auxiliary feedwater~ pumps. Preliminary-
data indicated that the tests were. satisfactory.
On May 2,'1989', the licensee discovered that' safety related feedwater stop
Valve HCV-1386 was inoperable.due to a torque switch set too low to shut
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the_ valve for design conditions. This finding resulted in power ~ ascension
being delayed until appropriate testing could be performed on;other
similar valves. The' Motor Operated Valve Actuator Testing' System (M0 VATS)
was performed on four valves and completed on:May 4, 1989. The plant:
reached Mode 1 (power operation) on May 5,L1989.
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While in power ascension on May 6, 1989, Main Feedwater Check Valve FW-162' ]
was'found to be blowing steam to theLcontainment atmosphere through a. f
retainer plug. On May 7,_1989, the plant was taken to cold shutdown for-
repair of the valve.
L While the plant was down for repair of Valve FW-162, the licensee also
plugged four additional tubes in the main condenser, performed maintenance
on the threaded plugs in Valve FW-161, the check valve for Steam
Generator A, and repaired a leak on Atmospheric _ Dump Valve HCV-1040.
Repairs were complete and the plant was brought critical again on-May.13,.
1989. On May 14, 1989, the threaded plug in Valve FW-162 was found to be ;
leaking again. Plant engineering devised a' procedure to weld a cap over ]
the faulty plug to stop the_ leak, This was accomplished on May:16,_1989, j
and a normal power ascension was performed. The plant reached 100 percent i
power on May 20, 1989, where it operated for the remainder.of the- )
inspection period. )
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3. Review of Previously Identified Items- (92701 and 92702)- .j
a. (Closed) Severity. Level IV Violation 285/8810-01: Failure to , l
promptly resolve test deficiencies.
This violation was related to the licensee's failure to'promptly )
resolve test deficiencies identified during the performance of -
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ST-NZ-1, " Verification of Open Containment Spray Nozzles." When the~
test was performed, the individual performing the test failed to- - .
test 11 spray nozzles due to the: inaccessibility'of the nozzles. . The' j
test results indicated that the nozzles were not inspected; however, '
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the test was accepted as completed without an_ analysis being. I
performed to indicate the results were acceptable. j
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In response-to this violation,.the licensee inspected the-11 nozzles f
during the 1988 refueling outage. The' inspection was performed in g
accordance with Procedure SP-NZ-1. . Completion of the nozzle j
inspection verified.that all nozzles were operable in accordance with $
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 operability of the containment! spray:
system with the n'ozzles inoperable, the licensee initiated.a. review
in accordance with the requirements of 10 CFR.Part 50.59. The-
results of the evaluation concluded that the containment spray system
remained operable.
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The licensee issued a reque'st 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.
The NRC's Office of Nuclear Reactor Regulation (NRR) has reviewed the
licensee's TS amendment request. ' Based on the review, NRR issued
Amendment-121'to the TS to approve the licensee's request. The
amendment changed the TS requirement such that the 11 inaccessible
nozzles are no longer required to be tested.
Based on the issuance of the TS amendment, this. item is considered
closed.
b. (Closed) Deficiency 285/8620-13: Inadequacy of_the dose assessment
model.
This deficiency' involved the failure of the" licensee to use a
more realistic model in dose assessment calculations (i.e., the
segmented plume model).' The model was found: inadequate because the .
model made no use of tabular and graphic options.available through .
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the Emergency Assessment of Gaseous'and Liquid' Effluents idAGLE) 4
computer program. j
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~The licensee has taken actions to incorporate the tabular and graphic -i
l- options'of the EAGLE computer program. Presently, the licensee has l provided.several enh.ancements to the program such as. serial
transformation of daughter radionuclides, dose from ground
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i. deposition,' and liquid releases. The licensee currently plans to c.
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provide validation'and verification of the program in the near. '
future. The validation and verification of the EAGLE program will be
reviewed by the NRC during a future inspection. This remains open 1
pending a completion of the NRC review. (285/8922-01) 1
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% c. (Closed) Open Item 285/8731-01: Level and pressure instrumentation-
for the safety injection tanks (SIT).
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This open item involved the. requirement for the environmental
qualification of the level and pressure instrumentation for the SITS.
During a review of the installed instrumentation, it was noted that ]
it aid not appear that the instrumentation met the provisions of :
Regulatory Guide (RG) 1.97, " Instrumentation for Light-Water-Cooled j
Nuclear power Plants to Assess Plant and Environs Conditions During j
and Following An Accident." .I
To provide an indepth review of the instrumentation required to be
installed, the NRC has issueo Temporary Instruction (TI) 2515/87, i
" Inspection' of Licensee's Implementation of Multiplant Action /
Item A-17," to provide specific instructions for NRC inspector
followup.
This item is considered closed based on the followup to be performed
during completion of TI 2515/87 in the near future,
d. (Closed) Unresolved Item 285/8913-04: Review of licensee's
evaluation of operability of fire doors.
This item involved an observation by the NRC inspector on March 21,
1989, that several fire doors. installed throughout the auxiliary
building appeared to be potentially nonfunctional. The licensee had
replaced, during the recent refueling outage ending January 1989,
several fire doors in the auxiliary building with doors that had a
3/4-inch diameter hole in the door frame. The hole was designed to
accept a conduit so that electrical wiring could be connected to the
door strike mechanism. The configuration found by the NRC inspector j
was a 3/4-inch hole on one . side of the door frame with electrical i
cable through it, but without the conduit installed. The NRC j
inspector notified the licensee of the concern and a fire watch was
initiated on March 22,~1989.
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On March 30, 1989, the licensee performed an evaluation to determine !
the ability of the newly installed fire doors to perform their 1
intended functions in the existing configuration. This evaluation R
concluded that the fire doors may not meet the required-3-hour fire i
barrier rating. However, the evaluation also concluded that, based '
on the low combustible loadings of the areas protected, the fire door .
assemblies with 3/4-inch unprotected penetrations remain adequate'for i
the hazards involved, with a,significant safety margin. The ,
licensee's evaluation was supplied to the NRC inspector on March 31, !
1989, for review. The licensee removed their fire watch patrol based
on the evaluation performed. '
The NRC inspector reviewed the licensee's evaluation and found it
acceptable. The removal of the fire watch patrol was also determined
to be acceptable based on the results of the evaluation. However,
the NRC inspector was concerned that these fire doors were installed
during the refueling outage and a fire watch patrol was not deemed
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necessary by-the licensee. The. licensee informed the'NRC that, at
the time of.insta11ationL a fire watch patrol was considered
unnecessary based on-the Underwriter's Laboratory '(UL) 3-hour! label
. . attached to the doors. .The doors were received from a UL-approved
vendor'with a UL 1abel. 'The 3/4-inch holes were drilled into the
door. frame by.the vendor.'
- The licensee'was asked by:the NRC inspect'or on May 26,c1989' whether
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the UL-label was" contingent on the conduit ~being installed. The.
licensee stated that,they did not. know but would . contact either. UL or-
the' vendor for clarification. The. licensee supplied, to the NRC
inspectorion May 30, andvJune 1, 1989, the results'of'a' telephone
communication between the licensee and the vendor (Meta ~1 Doors and
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' Hat dwa're) on May 26,-1989.' ? The vendor stated that thel holes are not
significant with regard to the integrity of the frame under fire '
conditions and, in. the vendor's opinion, the 3-hour-label of the door'
assembly is valid with'the unprotected hole'in the frame. This
response did not-indicatetthat the. vendor had any documentation to
show that a certified fire test had been made'on the door assembly.
with a 3/4-inch unprotected hole'in the frame.
The NRC inspector contacted UL on June 1,.198'9, to ask a
representative on fire door ratings about the UL label. UL stated
that a fire door assembly may be supplied to the custome'r with'
hardware not installed (locks, latches; reclosers, etc.) but that the
UL rating is based on a totally' finished assembly. Thus, the'NRC
, inspector concluded, and the UL representative concurred, that the
fire doors do not have a 3-hour UL rating until'the conduits are
installed to cover the 3/4-inch holes.
The NRC inspector determined that the licensee did not'take a
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conservative approach to safety when they accepted, without further.
evaluation, the UL rating of the fire doors.with the open 3/4-inch
hole in the frame. Although the subsequent' licensee evaluation
showed that a safety. concern;did not exist, the adequacy of the doors
as a 3-hour fire barrier was not known from the time of installation
of the doors, during the refueling outage, until March 30, 1989, when
the evaluation was performed. The licensee had not complied with the
requirements of TS 2.19 in that a fire watch patrol was noti
instituted. The failure to provide a fire watch, patrol is an.
apparent violation. (285/8922-02).
4. Licensee Event Report (LER) Followup (92700)
Through direct observation, discussions with licensee personnel, and
l review of records, the following event reports were reviewed to determine
that deportability requirements we're fulfilled, immediate. corrective
action was accomplished, and corrective action to prevent recurrence had
been accomplished in accordance with the TS. ,
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The LERS listed below are closed:
88-038 Inadvertent Actuation of Engineered Safeguards Equipment
89-003 Missed Operability Verification for Fire
Damper 1-HVR-DMPF-1
89-007 Inadequate Analysis for Feedwater Regulating Valves
89-009 RM-061 Inversion Switch Outside the Design Basis
A discussion of the review performed by the NRC inspectors for each LER is
provided below:
a. LER 88-038 reported an actuation of emergency safeguards
features (ESF) equipment due to personnel error. On December 31,
1988, while the' plant was in a refueling shutdown, a technician was
working on a modification of the pressurizer level instrumentation
and caused a ground while installing a jumper. The ground caused a
loss of Inverter D and; subsequently, a momentary loss of power to
Instrucent Bus AI-40D which, in turn, caused a loss of-,the Channel B
pressurizer pressure low signal (PPLS) block. Loss of the Channel B
PPLS block caused Channel B PPLS to actuate which initiated a safety
injection actuation signal, a containment isolation actuation signal,
and a ventilation isolation actuation signal. All systems
functioned as required for a plant in refueling conditions. During a
refueling outage, the safety injection pumps are, for example, in a
pull-to-lock condition and did not, therefore, actually run.
Normally, the static switch to the inverter is designed to transfer
the critical loads without int.rruption, from the inverter to the
bypass transformer. In normal operation,' static switch transfers are
effected at zero crossing current and no detectable loss of power to
the load is experienced. However, a fast-acting inverter fault
circuit is:used to shut off the inverter very rapidly when a current
overload con-dition occurs. This generates a fault signal that is
held.for approximately 1/4 second, which is long enough to cause the
relay for the PPLS block to drop out, consequently initiating a PPLS,
The NRC inspector reviewed the investigation performed by the
licensee and concurs with the assessment that the safeguards
actuations were caused by personnel error and not malfunctions in the
hardware.
The licensee issued a new Procedure 50-M-100, " Conduct of
Maintenance," on February 25, 1989. This procedure provides ;
instructions regarding care to be taken when performing maintenance !
in close spaces. The licensee believes this will help prevent ]
further occurrences of this sort. '
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The NRC inspector has reviewed training records verifying maintenance
personnel have recehed instruction on the contents of
Procedure 50-M-100. Based en the licensee's actions and the review
performed by the NRC inspector, this LER is considered closed.
b. LER 89-003 reported a missed operability verification for Fire
Damper 1-HVR-DMPF-1. The licensee determined that the damper should
have been included in the surveillance test program upon
installation. It was determined that surveillance requirements were
not met prior to 1988. The damper was discovered during the 18-month
surveillance test performed in March 1988. The damper is located
behind grating that is not normally removed.
The licensee's corrective actions included the addition of Fire
Damper 1-HVR-DMPF-1 to Procedure ST-FP-9, " Fire Barrier Penetration
Inspection," and changing Drawing 11405-M-85 to designate this damper
as a fire damper.. In addition, the licensee intends to perform a
walkdown of safety-related portions of the plant to verify the
accuracy of fire barrier penetration drawings and surveillance tests.
This walkdown is scheduled to be completed by the end of the 1990
refueling outage.
The NRC inspector reviewed the changes to the surveillance test and
drawings. These changes, along with the proposed walkdown of the I
fire barrier penetrations, adequately resolve this LER.
c. LER 89-007 reported an event where the licensee discovered that an
inadequate analysis had been performed during a modification to the
feedwaterregulatingvalves(FRV). During reconstitution of the
design basis for the main feedwater system, a problem was identified
where the licensee had not completed a nigh-energy line break
analysis when the FRVs were redesigned and the system piping was
modified in 1983.
During reevaluation of the modification, it was discovered that
maximum stresses had been exceeded. The existence of the stresses
required new break points to be postulated.
The licensee took actions to resolve the identified problem by the
installation of additional piping supports. The NRC inspector
reviewed the actions taken by the licensee and documented this review
in NRC Inspection Report 50-285/88-13. Based on the review, this LER '
is considered closed.
d. LER 89-009 reported an event where the licensee installed a
modification in 1980 and a seismic analysis had not been performed.
The modification involved installation of a switch box in a control
room cabinet for changing the setpoints for Radiation Monitor RM-061
- from operating to shutdown setpoints.
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, The licensee corrected the prob 1cm by removing the switch box. The.
NRC inspector reviewed this event'and the actions taken by the
licensee. The results of.the review are provided'in NRC: Inspection ].
Report 50-285/89-13. Based on the review performed by the NRC:
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-inspector, this'LER is considered closed.
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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 controis 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
conformance with the requirements established under-10 CFR, the licensee's .;
administrative procedures, and the.TS. 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. I
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Operator adherence to approved procedures and TS requirements was
evident. ,
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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 (LCO) was
met.
Logs, records, recorder: tr aces, annunciators,-' panel indicatioris, and .
switch positions complied with the appropriate requirements.
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Proper return to service of components was performed. l
' Maintenance orders (MO) were initiated for equipment in need of )
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maintenance. ,
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Management personnel toured the control room on a regular' basis.
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Control room access was proper'ly controlled.
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Control room annunciator status was reviewed to verify operator !
awareness of plant conditions.
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' Mechanical and electrical temporary mo'dification ' logs were properly l
maintained.
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Engineered safeguards systems were properly aligned for the specific
plant condition.
During this inspection period, the following items were reviewed:
a. On May 1, 1989, the incumbent in the position of P.lant Manager,
Mr. W. G. Gates, was reassigned as the Technical Assistant to the
President of OPPD. The_ responsibilities of the plant manager were
assumed by Mr. G. R. Peterson, the former Assistant Plant Manager.
.Mr. A. W. Richard was assigned the position of Assistant Plant.
Manager,
On May 11, 1989, due to the number of. changes made in plant
management positions, the NRC' inspector requested that the p unt
manager provide a copy of the letter of succession of authorir.y that
designates who would assume'the responsibilities of plant man 6ger in
~the event the plant manager was absent. At the time the request was
made, the letter had not been written. There had also been
confusion on t'le pert of plant personnel on the succession of -
authority when the plant manager left the site to visit the Omaha l
corporate office of OPPD.
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TS 5.1.1 states, in part, that the Manager, % rt Calhoun Station
shall delegate, in writing, the succession to this responsibility
during his absence.
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Contrary to the above, the delegation of responsibility, in writing, l
had not been issued to identify the succession of authority.. This is 1
an apparent violation. (285/8922-03) j
On May 23, 1989, the NRC inspector was presented with a letter, dated l
May 11, 1989, that established the delegation'of responsibility for
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the plant manager.
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Since the licensee has corrected this violation by issuance of the
letter of succession of authority and has revised other appropriate
plant procedures, this violation is considered closed. For this
reason, a response to.this violation is not required. The licensee
also identified that Chapter 12, " Conduct of Operations," of the
Updated Safety Analysis Report (USAR) is currently under revision to
reflect the changes in plant management. The revision'will be issued
during the next scheduled update of.the USAR in July 1989.
b. TS 2.7.i addresses the actions to be taken by the licensee in the
event that one EDG is declared inoperable. .The TS states, in part,
that one EDG may be inoperable for up to 7 days provided that the
other EDG is started to verify operability.
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lne licensee has interpreted the TS requirement for starting the-
other EDG to verify operability to mean that the EDG can be started
, and operated unloaded for 5 minutes, then be shut down. The NRC i
inspector was concerned about the licensee's interpretation of
operability. The question regarding the interpretation of the TS was
forwarded to NRR.
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In a memorandum dated May 16, 1989, NRR provided the interpretation l'
regarding EDG operability. The memorandum stated that the NRR staff
has reviewed this concern and the present operation of the engines by :
the licensee. The action-requirement to start the diesel to verify ,
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operability is necessary when taking the other diesel out of service. j
The staff considers that a risk exists when paralleling and loading a !
diesel to the grid in that'a grid fault may cause the loss of the
only operable diesel generator. Thus, the diesel generator should
not be tested loaded in this situation. However, the time period
that these two-cycle diesels are operated in'the. unloaded condition
needs to be minimized. The unit should be operated for only that
time necessary to verify that rated speed and voltage can be
attained and ascertained. Extended operation unloaded and at rated
speed causes sn excess combustion air condition which does not allow ,
proper fuel burning and keeps the cylinders cool. Running the unit
at idle speed is actually less detrimental with a periodic rise in
rpm to clear the engine and exhaust. In either regard, the time that ,
the unit is operated unloaded should be kept to a minimum.
Based on the response by NRR to this' concern, it appeared that the
licensee was properly verifying the operability of' the EDGs.
c. On May 22, 1989, the NRC inspector reviewed the licensee's records 1
maintained to document the completion of physical examinations of
'
licensed operators. The review was performed to verify that the i
licensee's documentation system complied with the requirements stated ;
in 10 CFR Part 55.27. <
!
During the review, the NRC inspector noted apparent inconsistencies
with the licensee's record system for documentation of the completion
of physical examinations. During review of the records system, the
NRC inspector also noted concerns with the records' system for the
completion of pulmonary examinations and the mask-fit testing for
those individuals that are qualified to wear respirators. The NRC
inspector related the concerns to licensee personnel so that the
licensee could address the concerns.
The licensee did not provide any response to the concerns identified
by the NRf inspector prior to the completion of this inspection i
period. The licensee stated that the concerns were still under
review and would be addressed in the near future. The determination
of the adequacy of the licensee's records system is considered
unresolved pending a review of the licensee's response to the NRC
inspector's i'dentified concerns. -(285/8922-04)
6. Plant Tours (71707) )
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The NRC inspectors conducted plant tours at various times to assess plant l
and equipment conditions. The following items were observed during the i
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 j
leaks and excessive vibration. q
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Valves and/or switches for safety-related systems were in the proper-
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a
, Plant housekeeping and cleanliness practices were. observed, including
no fire hazards and the control of combustible material.
Performance.of work activities was in accordance with approved
procedures. j
'
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 examined the licensee's
storage of hydrogen cylinders based on a concern identified at a different )
licensee facility. The concern was related to hydrogen cylinders being '
stored on the roof of the control room. This information was disseminated
to NRC inspectors and licensees by NRC Information Notice 89-14, " Hydrogen
Storage on the Roof of the Control Room." The NRC inspector found that, .
with the exception of one bottle, the licensee appropriately stores its .
hydrogen reserves in an outside area remot'e to safety-related equipment.
One bottle is stored in the auxiliary building for hydrogen addition to
the volume control tank. The bottle is' procedurally controlled and
secured.
No violations or deviations were identified.
:
7. Monthly Maintenance Observations- (62703) I
The'NRC inspectors reviewed and/or observed selected station maintenance
activities on safety-related systems and components to verify the
maintenance was conducted in accordance with approved procedures, i
regulatory requirements, and the TS. The following items were considered
during the reviews and/or observations:
*
l' TS LCOs were met while systems or components were removed from.
service.
Approvals were obtained prior to initiating the work.
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.
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Quality cont M records were maintained.'
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Activities were accomplished by qualified personnel.
Parts and materials use'd were properly certified.
: Radiological and fire prevention controls were' implemented.
The NRC inspectors reviewed and/or obs'erved-thelfol. lowing' maintenance
activities:
*
Remove, inspect, and reinstall the pin retainers for Valve FW-162
-(M0s 892861 and 892866).
Remove, inspect, and reinstall the pin retainers for Valve FW-161
(M0 892860)
Welding of cap on a pin; retainer for Valve:FW-162.(M0 892988)
Raw Water Pump AC-10B motor bearing oil change (PM 8904624)
*
Performance of MOVATS testing per Procedure MP-M0V-3A on
Valve HCV-1386 (M0 892755)'
Performance of MOVATS testing per Procedure MP-MOV-3A'on.
Valve-HCV-1385 (M0 892754)
Testing of the' static exciter on EDG 2 (M0 893034)
1
EDG 2 ' governor instability (MO.892680) "
'
A' discussion of each. item is provided below: '
a. '.On May 6, 1989, the plant was-operating in Mode 1 (power operation)' ;
when a leak was discovered on Valve FW-162:, the main feedwater system -;
ch'eck valve inside containment'for Steam Generator A. The licensee- !
issued M0s 892861 and 892866 to provide. instructions to repair the '.
'i
-leak on Valve FW-162. A discussion of this event is providedcin W
paragraph 12.c of this inspection report. No problems were noted, i
b. In addition to performing corrective maintenance on Valve FW-162, as
discussed above, the licensee also decided to perform' maintenance on
~
Valve FW-161. Valve FW-161 is the feedwater check valve for Steam. j
Generator B. The licensee issued M0 892860 to' provide: instructions
'for inspection of the pin retainers on Valve FW-161. The instructions
provided by MO 892860 were'the same as was provided for M0 892861 for ,
the repair of Valve FW-162. No problems were,noted.
(
c. On May 15, 1989, the licensee noted that the retainer plug on
Valve FW-162 was-leaking-again. The licensee issued MO'892988 to' d
provide instructions to craft personnel to weld a cap over the
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leaking retainer. plug to stop the' leakage. The.dethiledinstructions-
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for accomplishing'the task'were. attached to MO 892988. The- >
. instructions provided:a detailed step-by-step procedure; the. weld.
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des ign data form, and the weld verification formi Nol problems were
noted during' review of M0 892988.,
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9 s d. On:May 3,fl989, the NP.Ctinspector witnessed electrical 1 maintenance-
personnel; performing.a bearing oil change under the direction of.
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. Preventive Maintenance (PM); Procedure 89046'24. The component
undergoing maintenance..was,the motor.on Raw Water Pump AC-108. 'The
. . -NRC; inspector; examined.the procedure in'use by the craftsmen and felt--
it had the following. deficiencies: g
<
(1) The quantity of lubricant Lneeded was not specified.
(2) .There were'no instructionsjas to a torque required for the oil
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' drain plugs;
(3) There were no instructions for removing, cleaning, and
reinstalling the wire mesh cage surrounding the motor, which the
craftsman was observed to have performed.'
(4)' There was'no reference on the'PM task' sheet to a. vendor
technical' manual for more detailed instructions.
*
The above is an example of a weakness in procedures at the FCS. .This i
is not a new finding,. _Oneiof..the concerns of'the FCS from an audit;
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standpoint is that procedures;: in general,' are weak'and .that. 'l
management relies heavily;onithe skill of its seasoned. operators and ;
craft personnel < to get tasks performed properly. !
When this particular.. example was brought to the. attention of-a i
supervisory engineer in the! maintenance department, he. conceded the
deficiencies * existed.in the procedure and stated'they wera.being.
addressed by the procedures: upgrade program, which is part of the
licensee's corporate-improvement program known as' Project-1991. The i
supervisory engineer did request and was pr'ovided the specific j
concerns identified above'so that;they could be.immediately
addressed. The NRC inspector expressed concern.that the oil drain
plugs on the pump motors.may'not be sufficiently; tight due to the
lack of torquing instructions and verification. Craftsmen were
immediately deployed to the intake structure to verify integrity of !
oil drain plugs on the raw water 4 pump motors. A loose plug was. I
discovered on the. upper bearing of the' motor for Raw Water
Pump AC-10D. The' problem was_ corrected.
The Supervisor, Maintenance learned of the def'icient procedure and
the loose oil drain plug through his: subordinates. 'He arranged to
have the PM subsection of the. procedures upgrade program demonstrate
to the'NRC inspector, the improvements' underway with PM procedures.
On May 31, 1989, the NRC inspector met with the personnel cognizant-
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of.the upgrade'of PM' procedures. The methodology of the development q
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" of the new procedures was discussed. Drafts of_new PM procedures
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, . were examined by the'NRC inspector and'were found'to be comprehensive ,
with' respect to.the task required. Once implemented, the new ,
procedures should'significantly reduce licensee management dependence i'
on skill of the craft to have~ tasks performed correctly. The new j.
procedures l illustrate proper preplanning with respect to task
accomplishment.
- e. 'On May.4,'1989, the NRC' inspector witnessed.the performance of M0 VATS
1 testing on.the main'feedwater isolation valve ~HCV-1386 to Steam.
Generator A. The work was initiatbd per MO 892755:and performed in
a
accordance.with' Maintenance Procedure MP-MOV-3A, " Calibration and
Adjustments of Motor and Operated Gate and Globe: Valves." -A-
discussion of the events which led to'the testing of.this valve-is-
provided-in' paragraph 12.a of.this inspection report.- No problems
,
were noted. *
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f. The NRC inspector also reviewed the completed procedure for the :
MOVATS testing of HCV-1385. This testing was initiated by M0 892754.- y 4
No problems were noted. i
g. On May 17, 1989, the NRC inspector witnessed the performance of
M0 893034 when technicians' installed a chart. recorder, oscilloscope,-
and other monitoring equipment'to examine and record EDG 2 static !
exciter voltages during the performance of Procedure ST-ESF-6, I
" Monthly Testing of the Emergency Diese1~ Generator." A discussion of
the problem associated with the static. exciter'can be found -in
paragraph 12.e of this report.
h. In NRC Inspection-Report 50-285/89-17, the NRC inspector noted that,
due to fluctuation in the kilowatt output of EDG 2, the systems
engineer was concerned with proper.. operation of the governor. . ,
Engineering initiated MO 892680 which required consulting with the
Woodward Corporation, the governor. manufacturer. On.May 31, 1989, a
.
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Woodward representative visited the FCS, examined the governors for i
EDGs 1 and 2 and found their droop speeds to be improperly set- The
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governors were reset with a procedure developed by theJsystems. . .
l engineer and the vendor representative and the operability of each j
diesel ~was proved after.the resetting by performance of J
Procedure ST-ESF-6. The adjustment of the' droop setting proved to- ;
l signif_icantly reduce the variance in kilowatt output of the EDGs, " i especially EDG 2, which was found to droop significantly more than l EDG 1. The NRC inspector witnessed-portions of the resetting of the j
governors as well as the subsequent retest of each diesel. 'All work <
and testing were found~to be' conducted in.-accordance with approved '
procedures by cognizant personnel. To preclude recurrence of the-
situation, the systems engineer has. issued a revision to i
Procedures 01-DG-1 and 01-DG-2, " Normal Operation of EOG 1" and '
" Normal Operation of EDG 2," respectively, to add a verification that
the droop speed settings are properly set.
No violations or deviations were identified. , j
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8. Monthly Surveillance Observations (61726) J
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The NRC inspectors observed selected' portions of the. performance'of, j
and/or reviewed,' completed documentation for the TS-required surveillance i
testing on; safety related systems and components. The NRC' inspectors
verified the following items during the testing:
Testing was performed by qualified personnel using approved -)
procedures.- '
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Test instrumentation was calibrated. 1
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TS LCOs were met. )
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. Removal and restoration of the affected system and/or component we're- 'I
accomplished.
'
Test results conformed with TS and procedure requirements.
Test results were reviewed by personnel other than.the individual
directing the test.
Deficiencies identified during the testing were properly reviewed and
resolved by appropriate management personnel.
Test was performed on schedule and complied witti the TS. required -
frequency.
The NRC inspectors observed and/or reviewed the documentation for the
following' surveillance test activities. The procedures used for the test
activities are noted in parenthesis:
Monthly. Testing of the Emergency Diesel Generator (ST-ESF-6): {
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Shutdown Margin Verification (ST-SDM-1)
Monthly Channel Check of the Core Exit Thermocouple (ST-CET-1)_ d
A discussion of each surveillance observed is provided below: j
!
l a. 20n May117 and 21, 1989, the NRC inspector witnessed the performance- !
of.ST-ESF-6 for EDG 2. The purpose of this test.is to' verify.that- l
the emergency diesel generator starts on demand, comes up to rated.
speed-and voltage, synchronizes with the grid, loads, and maintains "
y
load for a period of 1 hour. The NRC inspector had been following
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testing of EDG'2 because of_the problem of excessive voltage on the
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static exciter while running unloaded. The static exciter anomaly is
discussed more thoroughly in paragraph 12.e of this inspection
report.
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b. On May 6,1989, the NRC inspector observed the reactor engineer, the
shift technical advisor, and the shift supervisor perform the
calculation of shutdown margin portion of Procedure ST-SDM-1. The
NRC inspector observed the calculation to be performed in accordance-
with a plant review committee approved algorithm.
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c. On May 23, 1989, the NRC inspector observed a senior reactor
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operator gather the necessary data for the completion of 4
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Procedure ST-CET-1. This' procedure verifies the operability of
postaccident monitoring core exit thermocouple by analyzing their-
output to the quality systems parameter display system (QSPDS). At
the time of the test, the NRC inspector noted that Channel B of QSPDS-
was out of service for maintenance. The NRC inspector,found that the~
,
shift supervisor was well aware of the actions required by TS 2.21-
when the number of ~ computer points available for the monitoring of
core exit thermocouple was diminished.
No violations or deviations were identified.
9. Security Observations (71707J
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The NRC inspectors verified that the physical security' plan was being
implemented by selected observation of the following items:
*
The security organizatice 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. ,
* The effectiveness of the security program was maintained when
security equipment failure or impairment required compensatory
measures to be employed.
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* The PA barrier was mainta1ned 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.
*
Security monitors at the secondary and central alare stations were
functioning properly for assessment of possible intrusions.
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During this ;nspection period, the. following items were reviewed:
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a. On May 17, 1989, the NRC inspector was told'by a licensee employee
i< that an altercation between.a security guard and a.nonlicensed-
operator had occurred on May 12, 1989..'It was' stated;that the
security guard had instructed the operator not to enter a particular '
area without another person. The officer was enforcing a health
physics department two-man rule. The operator insisted the rule was
not pertinent to his' task. Nonprofessional' language was exchanged'
and the operator continued to enter the area contested by the guard.
At this point,'the officer instructed the operator'to stop and ,
~apparently) unlatched the strap.on his sidearm weapon. The officer.
'did not draw the weapon.
In response'to'this event,;1'icensee~ management held meetings.with the
security; force and with operations personnel. . In the meetings, the
plant manager stressed the importance,of professional courtesy.
between the two groups. .In addition, security personnel were
instructed about the'use of deadly force.
~
The NRC inspector considers likensee management-to have taken an
aggressive approach in resolving this altercation and determining its
root cause. . The contract security force'is a relatively 'recent
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concept at the FCS and the actions taken by licensee management as a
result of this altercatio_n have led lto a better understanding of the
roles the security force versus the plant staff and'should preclude
any, future altercations.
The NRC inspector was_ concerned that the incident had'to bel learned
5 days later from an employee in_ lieu of' f' rom appropriate management in
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a prompt manner. This concern w'as' discussed with the plant manager.
b. On May 26, 1989, licensee security personnel noted, during a search-
of all hand-carried packages, that a visitor's lunch box contained-
four cans of beer. Security-personnel confiscated the-beer and
refused to allow entry of the visitor into the PA;:-All' hand-carried
packages were searched by the security force on that day and no other
problems were noted.
All hand-carried packages were being searched by security personnel
because the x-ray machine _was out of service. The licensee does not
routinely, search hand-carried packages other than by the use-of the
x-ray machine. During an interview of the visitor conducted by the
licensee's security personnel, the visitor-stated that he had. brought,
beer into the PA on both days just prior to the discovery made by
security personnel. The visitor stated that he was not aware;that
alcoholic beverages.could not be brought;into'the PA. The visitor's-
escort stated that he had never observed the visitor. consume any
alcoholic beverages while onsite. The visitor' stated that.the
alcoholic beverage's were kept in his lunch-box cooler so the-
beverages would be cool when he consumed them after leaving the site.
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In' follow up on this item,.the NRC inspector determined the following
items:
. (1) No training or briefing is' routinely ^provided to visitors as 'to i
the kinds of; items that are prohibited from being brought into i
the plant. !
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;)
(2) The x-ray machine can not differentiate between afcan of i
alcoholic. beverage an'd any other can of beverage.
In response to this problem, the licensee provided the4following l
information:
. 1
(1).ALverylargesign".hasbeeninstalledlforsome'timeadjacent'to
'the roadway leading to the plant. The sign lists all. items that-
are prohibited from being brought into the'PA.
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(2) 'A handout was printed and all visitors are being provided a" l
copy. The handout lists all prohibited items. y
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(3) Signs *were installed in the security building as a reminder of 4
what items are prohibited. !
(4) The visitor's'name was entered:into a log of persons that are-
not allowed access.to the facility as visitors.
(5) Actions'have not'yet been taken to establish a means of
identifying the contents'of cans that pass through the x-ray
machine. An assessment of the: actions to.be taken:was still in
progress.
Based on the occurrence of this-event,'a technical; review of the
licensee's program will be performed by security' specialists from the
NRC's Region IV office in.the near future. This item remains
unresolved.pending the completion of..the review by the security
specialists. (285/8922-05)
No violations or deviations were' identified.
10. Radiological Protection Observations (71707):
The NRC inspectors verified that selected. activities of the licensee's
u radiological protection program were implemented in conformance with.the- !v 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 toursJto
check on activities in progress,
HP technicians were using calibrated instrumentation.
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Radiation work permits contained.the appropriate information to
ensure that work was performed in a safe and controlled manner.
Personnel in radiation. controlled areas (RCA) were wearing the
required' personnel monitoring equipment and protective clothing and
were 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.
;
.No violations or deviations were identified.
11. In-Office Review of Periodic, Special, and Nonroutine Event Reports
(90712 and 90713)
In-office review of. periodic, special, and nonroutine event 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.
Planned corrective actions were adequate for resolution of identified
problems.
Whether or not any information contained in the correspondence report
should be classified as an abnormal occurrence-or additional reactive
inspection is warranted.
Correspondence did not contain incorrect, inadequate, or incomplete
information.
The NRC inspectors reviewed the following correspondence:
Radiation Monitor RM-061 Inversion Switch Outside Design Basis
(LER 89-009), dated May 1, 1989
Response to NRC Bulletin 88-10;-Nonconforming Molded-Case Circuit
Breakers, dated May '1,1989 '
Nuclear Material Accountability Report, dated April 28, 1989
Radiological. Environmental Operating Report for 1988, dated April 28,
1989
Fort Calhoun Station 1988 Refueling Outage Inservice Inspection
Results and NIS-1 and NIS-2 Forms, dated April 28, 1989
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Special Report on Inoperability of Fire Barrier and Fire Detection
Systems, dated May 1, 1989
Request for Extension of Implementation of Operating License
Amendment, dated May 4, 1989
:
Cycle 11 Fuel Performance Report, dated May 5, 1989
l
- Licensed Operator Requalification Simulator Examinations, May 5, 1989 l
Reappraisal of Independent' Nuclear Appraisal, dated May 12, 1989
Omaha Public Power District Fitness for Duty Program, dated May 11, !
1989 !
!
Response to Request for Further Information Rc arding Reactor Vessel j
Embrittlement for Further Consideration of Application for Life !
Extension, dated May 18, 1989 j
Toxic Gas Monitor Inoperability Due to Inadequacies in Modification !
Process (LER 89-010), dated May 19, 1989 ,
April Monthly Operating Report, dated May 18, 1989
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Monthly Operations Report for April 1989, undated i
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Correction to Semiannual Radioactive Effluent Release Report, dated
May 18, 1989
'
Special Report on Inoperability of Inadequate Core Cooling
Instrumentation Used for Postaccident Monitoring, dated May 19, 1989
Safety Enhancement Program Monthly Status Report, dated May 22, 1989
Special Report on Inoperability of Wide-Range Noble Gas' Stack
Monitors RM-063H, RM-063L, and RM-063M for Postaccident Monitoring,
dated May 23, 1989
Failure to Conduct Hourly Firewatch Patrol due to Procedural
Inadequacies (LER 89-011), dated May 23, 1989 i
Updated Status of Various Regulatory Items, dated May 24, 1989
No violations or deviations were identified.
12. Followup of Onsite Events (92702)
l
The NRC inspectors reviewed the following events:
l a. During a power ascension at 5 percent power, a' condition outside the l design basis was discovered. Main Feedwater Isolation Valve HCV-1386
was found to have an improper torque switch setting. The valve
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functioned under existing'(normal) conditions, but it was determined
that it might not operate =under the accident-condition of a main.
steam line break inside containment due'to high. differential pressure
across the valve. The condition was identified when two upstream
MOVs (HCV-1103'and 1104) failed to shut during a power ascension'
- evol uti'on. ~These, valves, h V-1103 and.1104,!are not part of-the
safety analysis.
"As..immediateaction,theflicenseeorderedpowerheldatlessthan
5 percent.while the torque. switch setting on'HCV-1386 was reset.
M0 VATS testing was performed on Valves HCV-1385, -1386, -1104,= and :
~1105. The, torque' switch settings of all' safety-related valves were
verified by review of documentation.
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b. As discussed above, the licensee experienced some operation'l' a
problems with. torque switch; settings of some' safety-related MOVs '"
which prompted review by the NRC inspector of the requirements of NRC' '-
*
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Bullstin-85-03, " Motor-0perated Valve Common Mode Failure During m
Plant Transients Due to Improper Switch Settings." In summary, the
bulletin required licensees to develop and 1mplement a program to
ensure that specific valve operator torque switches'are selected,
set, and maintained' properly. The valve operators'within the scope
of the bulletin are those.in high pressure safety and emergency.
feedwater systems that are required to be tested for operational'
readiness in accordance with 10 CFR 50.55 a(g);which is.the
licensee's ASME Section XI inservice. inspection.(ISI) program.
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The extent of the program for the. valves encompassed by the bulletin-
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was to review and document the. design basis for the operation of each i
valve. This documentation was'to: include the maximum differential
pressure expected during both opening and closing of the; valve for' a
both normal and abnormal events'to the extent that these valve
operations and events are included in the existing approved design
basis. From this data, the licensee was to ensure that valve -l
operators were selected, set, and maintained properly. .j
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In response, the licensee performed the review and instituted a 1
MOVATS program. M0 VATS is a system which' permits testing, adjusting,' j
and' setting of limit switches, torque switches, and torque switch
bypass'which are part of the controls for an MOV.
The NRC reviewed the licensee's program for. design review and MOVATS' l
r J found it to satisfactorily' meet the' requirements of NRC j
Bulletin 85-03. The results of this inspection are documented in NRC' !
Inspection Report 56-285/87-17. j
Since the requirements of the bulletin were inspected by the NRC, the ;
licensee has added a valve in the high pressure. safety
injection (HPSI) system, HCV-308, to its ISI program., The maximum
-differential. pressure expected during the opening and closing of.the
valve during both norm &l and abnormal events to the extent'that this'
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valve operation is included in the existing design basis has not been
determined by the licensee. Valve HCV-308 has not been M0 VATS
tested. 3
The NRC inspector has reviewed this concern and found that the i
safety-related function for this valve is not for HPSI but for
long-term core cooling. Under long-term core cooling considerations,
Valve HCV-308 would be utilized to provide simultaneous hot-leg and
cold-leg injection from an HPSI pump. This is to prevent boric acid
plateout in the core. The simultaneous hot-leg and cold-leg injection
mode is initiated at 3 hours post-LOCA (loss-of-coolant-accident) if
the reactor coolant system is below 700 psia, l
On May 12, 1989, the NRC inspector was provided with a calculated '
thrust and torque verification data sheet from a licensee special
services engineer. This calculation demonstrated that the thrust ,
value for the valve operator for HCV-308 was designed based on an )
assumed opening and closing differential pressure of 2735 psig. Data i
from Limitorque Corporation, the operator vendor, confirmed that !
Model SMB-000 with a torque switch setting of one, which is the
lowest setting, is adequate to provide the thrust necessary for the
valve at an assumed differentia 1' pressure of 2735 psig. l
As previously stated, the licensee has not yet determined what the
maximum differential pressure across HCV-308 will be during its
design function. However, it is clear that it will be significantly
less than 2735 psig since the design pressure of the HPSI pump is
1735 psig and injection through this valve does not occur until RCS !
pressure is below 700 psia.
The licensee performs stroke testing of Valve HCV-308 during cold
shutdown. This was last satisfactorily performed in October 1988,
At that time, the torque switch setting was 2.25. Again, the minimum ,
torque switch setting based on Limitorque's recommendation is one. . I
Based on the above discussions, the NRC inspector does not consider
the lack of MOVATS testing for Valve HCV-308 to be a safety concern.
The modification and M0 VATS testing of Valve HCV-308 will be tracked
as an open item. (285/8922-06)
c. On May 6, 1989, the licensee commenced a plant shutdown due to the
discovery of a leak in Valve FW-162. Valve FW-162 is the main
feedwater check valve located inside containment for the main
feedwater line to Steam Generator A. The leak was from a pin
retainer into the containment atmosphere. The leak was generating a
steam plume approximately 2 feet long. No estimate was available as
to the actual leak rate value. The licensee opted to shut down the
plant based on the size and location of the leak.
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Valve FW-162 is a dual-disc check valve with two pins installed in i
the valve body. One pin is used as a hinge pin and the other pin is j
used as a stop pin. The stop pin is instelled to limit the travel of ;
the valve discs in the open direction. This function ensures that ]
the discs will shut during a reverse flow condition. The pins are- i
held in place by pin retainers. A retainer is a threaded plug 1
installed on each end of the pin to hold the pin in place. The
retainer is threaded into the valve body. One retainer on the hinge j
pin was the source of the leak. J
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The licensee repaired the retainers on Valve FF-162 and, as a {
precautionary measure, also inspected and repaired the retainers on 1
Valve FW-161. Valve FW-161 is the main feedwater check valve for the J
main feedwater line to Steam Generator B. A discussion of the f
repairs performed on the valves is provided in paragraphs 7.a and 7.b
of this inspection report. -
On May 14,.1989, the plant entered Mode 1. On May 15, 1989,.the
i licensee noted, during a routine tour of containment, that one of the
pin retainers on Valve FW-162 was leaking. The leakage was not from !
the same retainer that had previously leaked. The licensee stopped I
the leakage by welding a cap over the retainer. The details of the
! actions taken by the licensee are provided in paragraph 7.c of this i l inspection report. Since the repairs were completed, no additional I
leakage has been identified with the valves.
The NRC inspectors observed portions of the shutdown and startup of I
the plant by operations personnel. During the. observations, it I
appeared that the evolutions were performed in accordance with I
procedural requirements. No problems were noted during the 1
observations. )
d. On May 19, 1989, the licensee, as part of their design basis
reconstitution effort, identified a problem with some alternate
shutdown panel (ASP) indications. The ASP is needed when a. fire ,
requires the evacuation of the control room. It was discovered that i
the cabling providing power for steam generator pressure indication
l at the ASP is routed through the control room. Thus, this indication
could be unavailable in the event of a fire in the control room or
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cable spreading room. !
In addition, the licensee discovered that two indications on the ASP
l (pressurizer pressure and Steam Generator A narrow-range level)
receive power from Battery 1. However, Battery 1 was not analyzed in
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the original 10 CFR Part 50, Appendix R, evaluation for use as a
power scurce for safe shutdown from outside the control room during a
fire. Thus, credit could not be taken for having these indications
available at the ASP.
The licensee completed, on May 20, 1989, Safety Analysis for
Operability (SAO) 89-008 to address the problem with the ASP
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indications. The SAO, with existing conditions, was based upon the
following-
(1) Steam generator pressure can be determined locally from PT-1046 l
and PT-1048 in Room 81 assuming Main Steam Isolation l
Valves-HCV-1041A ari HCV-1042A are open. j
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(2) Steam generator maximum pressure will be controlled by the main 1
steam safety valves if the main steam isolation valves are !
closed.
(3) A cable spreading room hourly fire watch will be maintained
until a modification can be implemented to permanently correct
the problem. The modification will be completed during the next
refueling outage.
(4) A change to Procedure A0P-6, " Emergency Fire Procedure," was
made to load shed Battery 1 of all loads except for Inverter C
and emergency lighting Panels 1 and 5. This improves the !
integrity of Battery 1. !
(5) Procedure A0P-6 was changed to remove power from the steam i
generator pressure indicators and jumper power from the ASP. '
This is considered a repair and is only used during the cold _
shutdown phase, which is allowed under Appendix R.
(6) Procedure A0P-6 was changed to load shed emergency lighting
Panels 1 and 5 after this lighting is no longer required. This ,
helps to further ensure the integrity of Battery 1. i
The NRC inspector reviewed'the SA0 and the changes to ,
Procedure AOP-6. In addition, the inspector walked down, with j
licensee personnel, the procedure requirements to jumper power from j
the ASP during the cold-shutdown phase. The NRC inspector found that !
the jumper cable was readily available near the ASP and that each
conductor was clearly labeled as to which cabinet and terminal
connection point it went to. The only concern noted was that the ]
labeling at the terminals for spare Fuses D-F15 and 0-F16, in the
back of the ASP (AI-179), was not easily recognizable. The licensee ]
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stated that this would be done.
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Also, during the review'of Procedure A0P-6, the NRC inspector noted
problems with the added step to send an operator to Room 81 to read
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steam generator pressure locally. First, the procedure does not
specify what the operator is to do when he reads the two pressure r ;
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l, transmitters (PT-1046 and PT-1048). It was not clear whether the
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operator should stay and monitor pressure or report back to the-ASP i
operator with initial readings. Also,'if communication equipment is 1
involved, no mention was made of that. Second, the NRC inspector
noted that PT-1046 and PT-1048 were reading between 50-75 psia lower- ,,
than actual steam generator pressure indications in the control room.- l
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With retpect to the adequacy of emergency lighting with the
procedure changes, NRR has allowed the licensee to use portable
hand-held lights in the interim until a permanent fix is made.
The above concerns with Procedure A0P-6 (labeling of terminals at
AI-179, clarification of what the operator making local readings {
should do,- and the discrepancy between actual steam generator j
pressure and local indications) are considered an open item.
(285/8922-07)
e. Several times-during the past months, the licensee noted a problem l
during the performance of ST-ESF-6, " Monthly Loading of the Emergency j
Diesel Generator (EDG)," for EDG 2. The specific problem was that i
upon initial start the voltmeter measuring generator output would peg i
high and indicate voltage in excess of 6000 volts. The voltage was
observed to be high for 10-30 seconds and then decay to the
appropriate 4160 volt range. The problem was first documented'in
August 1988 when an M0 was written to address the problem. This M0 i
was voided in November 1988 because the diesel had passed its ,l
surveillance test three successive times. I
In the February through May 1989 surveillance tests, the problem
recurred each month. The NRC inspectors were notified of the anomaly .l
for the first time on May 11, 1989. In February, March, and April )
several calibrations and mechanical evolutions were performed in
attempts to correct the problem. None succeeded. On May 11, 1989,
an open resistor was found in the static exciter which was replaced.
This was thought to have fixed the problem. However, on May-17,
1989, the voltage pegged high again. EDG 2 was declared inoperable.
This was the first time since the anomaly was noticed that the diesel
was declared inoperable.
At this time GE engineers were summoned to investigate the problem.
GE is the static exciter vendor. The GE representative felt the
problem stemmed from contamination under the wiper brush of the auto
voltage adjustment Potentiometer IP. This resulted in open circuits
in various spots on the potentiometer. This, in. turn, resulted in
loss of feedback to the voltage regulator and. full excitation was
applied to the generator.
The above described problem was resolved and between May 17 and
i May 21, 1989, EDG 2 was started and run at rated voltage ten times I without any overvoltage. EDG 2 was declared operational following
successful completion of ST-ESF-6 on May 21,1989.
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The NRC inspector reviewed the entire chronology of events associated
with excessive voltage of the static exciter. It was found that
engineering was thorough, cautious, and persevering in their analysis
of the anomaly. The evolutions indicated that the newly employed
systems' engineering concept is beginning to benefit the FCS.
No violations or deviations were identified.
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j.: 13. NRC Bulletin Followup (92703)
NRC Bulletin 88-10, " Nonconforming Molded-Case Circuit. Breakers,"
requested-that licensees take' actions to provide reasonable assurance that s-
molded-case circuit breakers (CBs) purchased for use in safety-related-
applications without verifiable traceability to the circuit breaker
manufacturer (CBM) perform their safety functions.' Several NRC
inspections'.and licensee findings in.1988 found that.there existed on the
.
market refurbished circuit breakers which did not conform tolthe original
manufacturers or UL specifications.
The licensee responded to the action items contained in the bulletin in a
letter dated May 1, 1989. The licensee's response will be evaluated in a.
future NRC inspection report._ Of an immediate concern, however, was the
licensee's finding'that four of the suspect breaken were known to be
installed as output breakers in safety-related Inverters A'(EE8H),
B (EE8J), C (EE8K), and D (EE8L).
In compliance with the actions required bY the bulletin, the licensee-
prepared an SA0 for the four installed nonconforming breakers. This
licensee document is' designated SAO-89-007. -In conclusion, the licensee ,
determined that the installed breakers meet the: functional requirements as
.
stated in the bulletin based on documented testing by the supplier and
functional testing by the licensee during)the~ Cycle 10 refuelingfoutage.
In_ compliance with the bulletin, the. licenseeha's committed to replace the
suspect breakers during the. Cycle 12 refueling outage' scheduled for 1990.
The review of. the < replacement of-the. suspect breakers by NRC inspectors
will'be-tracked as an open item. (285/8922-08)
The NRC inspector has v 3 cussed the content of the'SA0'with NRR. Based on
the testing performed, NRR considers the evaluation for continued-
operation to be adequate.
No violations or deviations were identified.
14, Unresolved Items
, 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. 1
Two unresolved items discussed in this inspection report"are listed below:
Item Paragraph Subject j
'285/8922-04- 5.c .
Medical Records System-
. 285/8922-05 9.b Adequacy of Searches-
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15. Exit Interview (30703)
The NRC inspectors met with Mr. K. J. Morris (Division Manager, Nuclear
Operations) and other members of the licensee staff on June 6, 1989. The
meeting attendees are listed in naragraph 1 of this inspection report. At
this meeting, the NRC inspectors summarized the scope of the inspection
and the findings.
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