IR 05000298/1990002
| ML20012B880 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/07/1990 |
| From: | Constable G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20012B879 | List: |
| References | |
| 50-298-90-02, 50-298-90-2, NUDOCS 9003190093 | |
| Download: ML20012B880 (15) | |
Text
.
,..
,-
__
_
,
h i
'} },
h 1.t [
-
'
'
o *
- q[.y
- q j
.
-
,
,
'
- .:
.;
,
.,
,
ic. '
"*
"I n
APPENDIX
'
'
[s
,
q
,
b U.S. NUCLEAR REGULATORY COPHISSION
,
REGION IV
,
..
r>
'
,
1%
,
a-
...,
.
50-298/90-02 Operating License: 'DPR-46.
i
'
-.NRC. Inspection Report:
b'
Docket: -50-298
,
- b
' Licensee:; Nebraska Public Power District (NPPD)
=l
.P.O.-Box 499
'
,
Columbus Nebraska 68602e0499
'
Facility Name:. Cooper Nuclear Station (CNS)
.?
Inspection At:
,
I f
Inspection Conducted:. January 16~ through February 15, 1990 Inspectors:' G. A. Pick, Resident Inspector, Project Section C
-
,
,
~
Division of Reactor Projects
"
W. R. Bennett, Senior Resident Inspector, Project Section C
,
Division of Reactor Projects
>
<
%
do Approved:
_
,. -
L, 6'.-t.7:enstable, chief. Project :Section c Date f h
. Division of Reacto'r Projects L-Inspection Summary I
o L
Inspection Conducted January 16 through February 15,1990'(Report 50-298/90-02)
.
m i
Areas Inspected:. Routine, unannounced inspection of operational safety verification, monthly surveillance observations, monthly maintenance w
observations,' scram discharge volume = capability verification, preparation for refueling,;1icensee event report followup, followup of previous inspection m
findings, and fitness-for-duty training programs.
O
_Results: One noncited violation for an-inadequate procedure concerning test F
lead connections was identified. The licensee committed to review all surveillance procedures to clarify instructions for test lead connections and
-'
- make-any necessary revisions by' August 31, 1990.
"
The licensee's response to several minor events demonstrated proper management l
-
attention and conservatism regarding plant safety. Comunication and
cooperation among various plant departments was excellent in plant surveillance, maintenance, and response to plant events. The licensee properly
9003190093 900308 PDR ADOCK 05000298 Q
PDC u
,
a
+'
,
<
,
,
e,
.:
,
.
.
-2-
)
implemented their commitments in response to Multiplant' ActioncItem B-58,
" Scram Discharge Volume Capability." Licensee preparation for the upcoming ~
l
'
refueling outage was good concerning contractor orientation and training;-._
- however, approval of design change packages was not timely.. The licensee had--
-
made improvement in this area prior to the last refueling outage but was not
.
as' successful in preparation for this outage.
j f,
m
.
}
,
-
k
,
i f
>
\\
r i
- .'
't'
-
~-
-
.
_
r#
t;
'
+ 1' m
.c
'
,
,
,
-3-
<
.
Om
,
'
DETAILS
-
1.
Persons-Contacted-Principal Licensee Employees
- G. R. Horn, Division Manager of Nuclear Operations
- J. M. Meacham, Senior Manager of Operations
- S. M.~ Peterson, Senior Manager of Technical Support Services
- E. M.. Mace, Engineering Manager
- R. Brungardt, Operations Manager
. R. L. Gardner, Maintenance Manager l
- J. V. Sayer, Radiological Manager H. T. Hitch, Plant Services Manager
~
- R. L. Gibson Audit and Procurement Quality Assurance Supervisor
- Y. Annstrong, Administrative Secretary I
- G. A.- Schmielau, Instrument.- and Control (I&C) Foreman
]
' * Denotes those present during the exit' interview conducted on February 21, j
1990.
The NRC inspectors also interviewed other licensee employees and contractors during-the inspection period.
2.
Plant States
Plant power decreased from 100 percent at the beginning of this period to 92-percent by February 15, 1990, in a continuing coastdown to the Cycle 14
!
refueling outage.
!
- 1 3.
Operational Safety Verification (71707)
'
The inspectors observed operational activities.throughout the %spection
period. Control room activities were observed to be well controlled.
Proper control room staffing was maintained and professional conduct was continuously observed.
Following discussions with operators, it was
.
'
determined that they were cognizant of plant status and understood the importance of, and reason for, each lit annunciator. The-inspectors observed selected shift turnover meetings and noted that information concerning plant status and planned evolutions was-communicated to the
<
or. coming operators.
Facility operations were performed in accordance with
,
L the requirements established in the CNS Operating. License and Technical
[
Specifications-(TS).
L On January 22, 1990, during conduct of the residual heat removal (RHR)
)
valve operability and valve inservice testing surveillance procedure, the
!
disc for air-operated testable Check Valve RHR-A0V-68A indicated an
,
l internediate position (both the open and closed lights illuminated) after h
stroking; however, the indicating light for the valve operator indicated
'
full-closed after stroking. The testable check valve is capable of being tested.to verify that the valve will open and reseal.
'
m
_._.
__
_ _ _ _ _ _
_.
_
-
I 2,
,
,
.
.
-4-
,
If the indicated dhe position was correct, the RHR system would be inoperable. The licensee developed a Special Test Procedure (STP) to verify that the testable check valve disc was closed. The Station
Operations Review Committee (SORC) approved STP 90-159 on January 22. The
'
test procedure utilized a flow path in a 3/4-inch drain line upstream of RHR-A0V-68A, but downstream of the normally closed inboard injection j
+
valve, to measure pressure to detemine if the check valve was leaking, i
i If pressure increased to primary plant pressure as observed on the temporary gauge installed for the test, RHR-A0V-68A would be detemined to
,
I be open and the RHR system declared inoperable. The initial drain line
'
pressure was 46 psi due to the keep-fill pump. After opening the drain i
valve, the pressure in the line dropped to O psi with minimal leakage from i
the drain pipe..This verified that the check valve was seated and the RHR
.
line was isolated from the reactor coolant system.
i On January 24 the 50RC approved a justification for continued operation to allow the plant to operate with the valve indication in this degraded
>
condition until the refueling outage. As specified in the justification for
continued operation, the operators will perfom STP 90-159 after each
,
'
operation of RHR-A0V-68A to verify its position.
On January 24, 1990, thehighpressurecoolantinjection(HPCI) Outboard Steam Supply Valve HPCI-MO-16 inadvertently closed which caused the HPCI pump to be. inoperable. The steam supply valve closed while I&C technicians
'
t were testina a pressure switch for low steam line pressure in accordance with Surveillance Procedure 6.2.2.3.3, "HPCI Steam Line Low Pressure Calibration and Functional / Functional Test," Revision 16, dated May 4, t
1969. The pressure switch sends a close signal to the valve. After responding to the control panel alarms, operators directed the I&C
,
technicians to stop the test. The operators then opened the steam supply valve to the HPCI turbine. The HPCI pump was inoperable for less than 1 minute.
The licensee's 16vestigation into the cause of the 1n' advertent valve
~
closure determined that an I&C technician had inserted test leads into the wrong test jacks. When the technician valved out the instrument to be tested and applied the test signal, both instruments indicated low steam line pressure, making up the valve isolation logic and closing the valve.
The licensee attributed the root.cause of the actuation to poor procedural
~
t guidance since it failed to list the terminals for connecting the leads and to human factors deficiency / human error since the technician had i
trouble identifying the correct test jacks.
1he licensee plans to issue a licensee' event report describing the event.
,
'
The surveillance procedure will be revised to identify which test jacks should be used for their respective instrument.
The licensee committed to conduct a review of all surveillance procedures and to make any necessary revisions-by August 31, 1990. This procedure inadequacy is an apparent viciation of the requirements of 10 CFR 50, Appendix B, Criterion V.
However, due to the licensee's efforts and prompt corrective actions, a n
l<otice of. Violation will not be issued for the above incident in
?
.
.
__
.
. _ _
_ _. _.
_
__
,m i
-;
.-
.
u
.
'
-5-
,
accordance with Appendix 0, paragraph Y.G.1, of the NRC's Rules of Practice," Part 2 Title 10, of the Code of Federal Regulations.
.
..
{
On February 12, 1990, during conduct of the Emergency Diesel Generator (EDG) No. I monthly operability test, the EDG failed to start on the first attempt and took 21.2 seconds to start on the second attempt. There is no
'
TS start time requirement for the monthly operability test; however, there
,
.
is a 16-second starting requirement in the load sequencing procedure i
performed each fuel cycle. When the EDG start time was greater then
'
16 seconds, the operations staff notified management who decided to
declare EDG No.1. inoperable af ter completion of the surveillance. The
,
operators subsequently completed all TS required surveillances of the other emergency core cooling system components.
'!
Troubleshooting by 18C technicians and electrical engineers determined j
that the input filter capacitor for the relay tachometer, a device that controls the starting' sequence, had degraded, and that the control
.I circuitry zener diode s output voltage had drif ted.
Engineering personnel
.
completed commercial grade dedications for each replacement component and replaced each component on a like-for-like basis.
Postmaintenance testing j
was properly completed for the maintenance perfonned. The licensee is
!
implementing a change to the EDG monthly operability test to direct operations personnel to contact the system engineer and management whenever the EDG start time exceeds 16 seconds.
Control panel walkdowns were conducted to verify that emergency core cooling systems were in a standby condition. Tours of accessible areas at
>
the facility were conducted to confirm operability of plant equipment.
During a plant tour on January.16,1990, the inspector observed licensee personnel conducting an inspection of a control building fire door as part
'
of the annual fire protection inspection. The personnel verified that the
-
door closed smoothly and remained shut, and that the latch did not drag or
,
bind.
.
.
The inspectors verified that selected activities of the licensee's radiological protection program were implemented in conformance with facility policies, procedures, and regulatory requirements.
Radiation
,
and/or contaminated areas were properly posted and controlled. Radiation
.
work, permits contained appropriate infomation to ensure that work could be perfonned in a safe and controlled manner. Radiation monitors were
properly utilized to check for contamination. On January 17, 1990, the Inspector observed a health physics technician conduct a daily operability check of the personnel contamhnation monitors using a Cobalt-60 source.
On January 23 the inspector observed licensee activities for verifying and i
removing contaminants from an employee's clothing.
The inspectors observed security personnel serform their duties of vehicle, personnel, and package search. Velicles were properly authorized and escorted or controlled within the protected area (PA). The PA barrier had adequate illumination and the isolation zones were free of transient material. Site tours were conducted by the inspectors to ensure that
__
_
_
.
__
_
.
__
.
_.
.
_
.
.
-
.
__
f
.
,
.
.
s m
t, i
,.
-6-
"
i compensatory measures were properly implenented as required. The inspectors observed guards perform vehicle searches on several occasions during the inspection period.
In addition, the inspectors verified that i'
all food and drink brought onto the site, other than in personal lunch boxes, is unloaded at the warehouse. The trucks carrying food are not allowed entry to the site.
The inspectors observed the licensee's preparation for the Cycle.34 refueling outage scheduled to commence on March 5, 1990 Contractors were hired early to allow time for site and job familiarization, and contractor craft training was perforned well before outage commencenent. The inspectors, however, noted that numerous design change packages have not yet been approved by the SORC. This has been expressed previously as'an arN of concern by the inspectors. Prior to the last refueling o6: age an improvenent in this area was noted, however, this improvement does not
,
appear to have continued in preparation for the upcoming outage.
l One noncited violation for an inadequate procedure was identified.
Engineering and management involvement in the RHR testable check valve problems, the HPCI valve closure, and the EDG starttine problem resulted in thorough reviews, tinely resolutions, and proper corrective actions.
4.
Monthly Surveillance Observations (61726)
The inspector observed the following surveillance procedure (SP)
perfonnances and/or reviewed the test results:
SP 6.1.4.1, " Main Steam Line Process Radiation Monitor Calibration and Functional Test," Revision 6, dated December 21, 1989.
This. calibration check was performed on January 22, 1990, to verify TS operability. This test verified proper calibration of the radiation.
monitors using a built-in precision current source. The procedure was
'
easy to follow and well written. A senior technician provided instruction to three less experienced technicians during performance of the test.
,
The 18C department is qualifying more technicians on the plant procedures to insure that a sufficient number of technicians will be available to support both the design changes and the normal plant surveillance activities during the upcoming refueling outage.
SP 6.3.12.1, " Diesel Generator Operability Test," Revision 28, dated June 8, 1989.
The licensee operated EDG No. 2 to verify operability as required by TS on January 16, 1990. The test requires the operators to start the diesel generator and, after a warmup period, load the EDG to 80 percent of rated capacity for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This testing of the diesel engine exceeds the TS requirements of loading the EDG to at least 50 percent of rated capacity for a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The inspector observed a station operator log the required surveillance data. During performance of the SP, mechanics
-_..._____._,__-_._..-__J
.-
.-
-
--
--
.-
. - -....
.
- - -
i
' O e
!
-7-l
.
performed a diagnostic analysis on the EDG. The diagnostic analysis is done during each planned run of the EDG, The station operator and
mechanics demonstrated excellent communications and cooperation.
!
.
SP 5.3.12.8, " Diesel Generator Fuel Oil Transfer Pump Flow Test,"
i Revision 7, dated March 22, 1989.
The licensee perfomed this test on January 16, 1990, to verify compliance I
with the inservice testing (IST) program. Since the last performance of i
the test, a temporary change was issued to change acceptance criteria to
'
be consistent with the equipment utilized in test performance and improve
data repeatability. The engineer was familiar with the inservice test equipment operation and all equipment was in calibration. The portable
,
instruments will be permanently installed during the refueling outage by l
i Design Change 89-107. After the licensee removed the test equipment, an
!
inservice leak test was satisfactorily performed on the fuel transfer line.
SP 6.3.4.2
"CS Motor Operated Valve Operability Test," Revision 18. dated May 4, 1989.
The inspector observed this operability test of the core spray (CS) motor o>erated valves on January 22, 1990. Control room operators performed
'
t11s test to verify o>erability of the CS valves because RHR Loop A had been declared inopera >1e.
The station operators were given explicit, easy to understand instructions. The SP requires that the injection valves be
- partially opened to equalize the pressure, then closed prior to testing of the motor operator. This prevents overloading the motor operator due to the 950 psi pressure differential the valve would experience without the pressure equalization. The control room operator infonned the station operators whenever he manipulated the injection valves in their vicinity to prevent personnel injury. All valves cycled as required.
.
SP 14.15.1, " Jet Pump Flow Instrument Calibration," Revision 1, dated February 23, 1989.
The inspector observed the >erformance of this instrument calibration check / calibration for Flow.oop B on January 22, 1990. The jet pump flow
,
t instrumentation provides inputs used for indication and computation, Qualified IAC technicians demonstrated clear and concise coninunications
- '
.
among themselves. Proper radiological practices were followed and the test equipment was properly connected. Only one of ten meters in Flow
,
Loop B required adjustment.
SP 6.3.3.1, "HPCI Test Mode Surveillance Operation," Revision 33, dated July 21, 1989.
The operators perfomed the HPCI quarterly inservice and operability test
<
on January 24, 1990. The quarterly IST required that vibration measurements be recorded. The inspector observed the start of Service Water Booster Pump D and RHR Pump B.
The operators operated the RHR loop
,
$
t
- -
-
-
--
., - -
- - - - - -
- -.
. -,, - <,,
-
-
-
--
.
- _ -
..
.
/
-8-in torus, anticipating heatup of the torus cooling prior to the SP after they started the HPCI pump. The inspector observed a station operator take required vibration measurements. The station operator followed proper.
radiological controls and utilized calibrated test instruments.
SP 6.2.2.3.6, "HPCI Pump Low Suction Pressure Calibration and Functional / Functional Test," Revision 16. dated June 22, 1989.
The inspector observed this calibration check of a pressure switch on January 24, 1990. The pressure switch initiates an HPCI turbine trip in j
the event of water loss to the pum) suction. The technician utilized calibrated test equipment and was (nowledgeable of the purpose, precautions, and limitations of the test.
'
SP 6.2.2.5.5, "RHR Reactor Vessel Shroud Level Indication Calibration and
-
Functional Test," Revision 5, dated June 2, 1988.
'
The inspector observed the perfomance of this calibration test of a
,
nuclear boiler instrumentation level transmitter, NBI-LT-LIC, on January 29, 1990. These instruments provide postaccident level monitoring with reactor pressure at 0 psi and with drywell temperature at 212"F.
The technicians established and maintained proper conwnications. They made minor adjustments to the zero adjustment of the control room level gauge and used calibrated test equipment.
j l
i l
SP 6.1.9, " Reactor Vessel Low-High Water Level Calibration and
]
'
Functional / Functional Test," Revision 24, dated June 8, 1989.
'
l-The inspector observed the performance of this calibration test on level indicating Switch NBI-LIS-101D. This level instrument provides reactor scram signals on low reactor vessel level and provides HPCI/ reactor core 1 solation cooling turbine trips on a high reactor vessel level. The I&C J
technician'used proper caution when removing and returning the instrument
to service since the sensitive nature of these instruments has the i
potential to cause a scram. Both the high and low setpoints required j
small adjustments by the technician to bring the instrument into calibration..The test instruments were properly connected and radiological controls were adhered to.
No violations or deviations were identified in this area. The IAC department is in the process of qualifying technicians to support outage j
activities. Good comunications among licensee departments and adherence L
to radiological controls were evident. Operators and IAC technicians were
,
l knowledgeable and adhered to procedures.
-5.
Monthly Maintenance Observation (62703)
i l
.
The inspector observed in-progress maintenance on January 16 and 17,1990, for Service Air (SA) Compressor B.
After a light dusting of desiccant was
,
i identified on two unloader valves during prior maintenance, work planning issued a maintenance work request (MWR) for rebuilding and cleaning of the
'
'
L
- - - - -
-
-
-
.
-.--
-. - - _ -. -
_. -
- -
-
'
,
,
.
..
air compressor and the air compressor valves. As part of the valve
rvbuild, the mechanics replaced internal springs and gaskets and lubricated moving parts. The mechanics Ifghtly sanded the comprtssor cylinders to remove any roughness, set the clearances for movement of the piston within the cylinder, and assured system cleanliness prior to i
reassembly of the congressor. The mechanics conducted this compressor i
overhaul and valve rebuild in accordance with Maintenance Procedure 7.3.26,
" Air Compressor Maintenance," Revision 3, dated February 6,1986.
'
On January 16, 1990, the inspector observed mechanics conducting a diagnostic test on EDG No. 2.
The instruments monitored the peak cylinder firing pressure, cylinder exhaust pressurt, and fuel injector / head bolt
vibrations. All data met specifications. The mechanics conducted the l
testing skillfully and were knowledgeable about the use of the equipment.
l
On January 16, 1990, the inspector observed the reconnecting of the
-
control and power leads to Circulating Water Pump A.
The leads and terininations were color coded to prevent mislanding. The pump had undergone a preventative maintenance overhaul that checked for impeller and bowl degradation and for bearing wear. Because the mechanics had identified minimal wear, the work planners changed the preventive maintenance frequency from 2 to 6 years.
Since the mechanics had observed a light coating of dust on SA Compressor 8, they issued a preventative maintenance MWR to inspect SA Compressor A.
The inspector observed maintenance activities conducted by the mechanics on January 29 and 31,1990.
The mechanics wiped down all interior surfaces of the air compressor to remove accunulated dust and
,
used SA to blow dust from the intercooler air fins. The inspector noted that little dust had accumulated in SA Compressor A.
No violations or deviations were identified in this area.
The mechanics i
conducted a thorough overhaul of SA Compressors A and B.
!
t
Scram Discharge Volume (SDV) Capability Verification (2515/90)
'
l The inspector verified licensee actions that implement their consnitments to ensure adequate SDV capability in response to Multiplant Action
'
Item B-58.
The documents reviewed and utilized to verify the licensee's implementation are listed in the Attachment.
The inspector verified that the licensee's alarm response procedures specified licensed operator actions when water is present in the scram l
dischargeinstrumentvolume(SDIV). Visual inspection verified that
!
alarms exist which identify the presence of water in the SDIV.
Operations surveillance procedurts implemented the TS requirements to periodically cycle and time the SDV vent and drain valves.
I&C performed calibration and functional tests verifying that the SDV level alarms and trips l
actuated at the setpoints specified in the TS.
i i>
- - - - - - -..
,.. -
-
--
---
7,
.
. - - -. -.
- - -
-
. - - -
- -
..
.
!
.-
.
-
,
..
-10-j From interviews with licensee personnel, review of as-built drawings, and j
review of the design package, the inspector verified that the i
safety-related SDIV level instrument taps are located on the instrument i
volume and not on connected piping. From review of current plant drawings, the inspector determined that the instrument volune vent and drain valves close on loss of air and that a single active failure will
not defeat isolation of the vent and drain valves.
,
From discussions with licensed operators and walkdown of the control room
'
panels, the inspector verified that vent valve and drain valve positions
!
t are indicated in the control room, The d'spector's review of the design package detennined that the system
conft
.3 tion has redundancy designed into it: two instrument volunes, two
!
separate sets of instrument taps for each instrument volume, and two trip level instruments off of each instrument tap. For diversity the licensee utilized two different instrument designs to sense level.
Both instrument
,
designs provide a scram signal to the reactor protection system on high SDIY level. The design package calculations referenced documents which
indicated that the SDVs were designed to accept greater than the 3.34 gallons per drive as reconsnended by the nuclear steam supply vendor.
-
Review of the design package further demonstrated that the instrunent volunes were adequately sized and directly coupled to the SDVs.
.,
The instrument volumes drain into the equipment drain sumps and interface with no other systems..The sumps are located three stories below the SDIV drains.
'
No violations or deviations were identified in this area. Consnitments in response to Multiplant Action Item B-58, " Scram Discharge Volume
'
Capability," were properly implemented.
,
7.
Preparation For Refueling (60705)
i The inspectors observed the receipt inspection of the new fuel transfer casks over the period January 16-26, 1990. The reactor engineers maintained proper control of the Special Nuclear Material. Health physics technicians conducted thorough radiological surveys, including i
contamination snears, of the new fuel casks.
Prior to receipt of the new
!
fuel, the nechanics and electricians had verified proper operation of the refueling floor crane in accordance with Procedure 7.2.32 " Crane Hoist
'
and Cable Inspection," Revision 8 dated August 24, 1989.
The inspect 0rs observed the fuel supplier's technical' representative provide on-the-job training to the licensed operators who were to conduct the fuel receipt inspection. The technical representative explained the
proper use of and purpose for each fuel assembly inspection tool.
!
i j-The operators inspected each fuel assembly. All of the fuel channels were
cleaned with a solvent. The refueling floor supervisor documented the fuel bundle number, the fuel channel number, and the refueling pool l
b
<
y
. - -
- -.
.
.
-
.
,
. -..
.
.
...
. -..
-
-
..
-
.
!
l
.
.
l
.
.-
-11-i location for each new fuel bundle. This information would be needed by reactor engineering in order to map out the spiral reloading pattern steps. The operators identified a fuel bundle (LYU-407) which had a
)
distorted edge on the No. 4 spacer. The licensee returned tne bundle to the vendor and they expect to have the bundle repaired and returned prior to the start of the refueling outage. No violations or deviations were
.
'
identified.
i 8.
Followup on Corrective Actions for Violations and Deviations (92702)
j (Closed) Violation 298/8909-01: This violation concerned the failure to
take adequate corrective action to preclude repetition of a failure in the diesel generator control air system.
Corrective actions included modification of the diesel generators to I
reduce the effects of vibration, revisions to improve the root cause analysis program, and enhancement of the computerized nonconfonnance report (NCR) database.
-
The inspectors verified that immediately after the diesel generator failure, the licensee replaced a large portion of the control air, lube oil, and fuel oil-tubing on both diesels.
In addition, during the
'1989 refueling outage, the licensee implemented design changes to both
diesel generators to reduce the effects of vibration. The inspectors verified the implementation of these design changes. There have been no t
vibration related diesel generator failures since the perfonnance of these design changes.
The licensee has instituted root cause training for all engineering personnel and has revised the procedure for nonconformance and corrective
,
'
action to provide better guidelines for root cause detennination.
In
. addition, the' licensee issued an Engineering Department Instruction
,
delineating the requirements for document research, and completed
'!
_
enhancement of the NCR database to improve the capability to perform
-
document research. The inspectors reviewed CNS Procedure 0.5.1,
"Nonconformance and Corrective Action," Revision 5, dated December 28, 1989; and ED 89-02, " Document Research for NCR Evaluations " Revision 2,
.
dated September 26, 1989. The inspectors have noted improvement in the
!
root cause analyses for NCRs in recent months. This violation is
,'-
considered closed.
No violations or deviations were identified in this area.
9.
Onsite Followup of Written Reports (92700)
F (Closed)LicenseeEventReport(LEPs)89-003: This LER documented a diesel
,
generator control air system leak due to inadequate corrective action.
,
The events and corrective actions associated with this LER are documented in paragraph 8 of this report. This LER is considered closed, b
-]
i
\\V'
.
.-
!
-
.
l
'.
..
-12-i l
l
,
'(Closed)LER88-014: This LER docunented an unplanned Group 6 Isolation with the reactor shutdown.
The ESF actuation occurred because of a failed 24V DC fuse. The root
'
cause of the event was determined to be random equipment failure. The
,
,
' licensee determined that no previous similar fuse failures had occurred.
!
This LER is considered closed.
.
'
(Closed)LERs88-013and89-013:
These LERs documented unplanned engineered safety feature group isolations during maintenance activities
with the reactor shutdown.
,
l The root cause of these events was determined to be personnel error and
.
human factors deficiencies, corrective actions including performing
,
l training on the events, and reviewing and revising, as necessary.
-i applicable maintenance procedures.
'
-The inspectors verified that engineering. I&C, and electrical personnel had received training on these events. The inspectors reviewed Maintenance Procedure 7.3.16. " Low Voltage Relay Removal and Installation," Revision 5, dated September 21, 1989. Among the changes made to the procedure were
-
the addition of a requirement for the individual specifying the lifting of leads or installation of jumpers to verify the location for accessibility.
These LERs are considered closed.
No violations or deviations were identified in this area.
10.
Fitness-for-Duty Training (FFD) Programs (25]S/104)-
The inspector atte'd:d FFD supervisor training on January 26, 1990. The instructcrs described the regulations, the program, the policies, and the procedures used to implement the FFD program. The instructor discussed the role and responsibilities of supervisors and managers in implementing
!
the FFD policy. Additionally, the instructor reviewed the CNS Employee
'
Assistance Program (EAP) and the medical review officer's function in the FFD program.. The instructor discussed specific drugs, their conson names, and human behavior patterns associated with use of each drug. The
.
instructor provided a pamphlet which supplemented this lecture with more detail. Statistics and examples were presented that described how drug and alcohol abuse impacts the work place. A series of video vignettes was used to illustrate behavior patterns conson to each type of drug. The vignettes stressed how the use of drugs would degrade job perform'ince, impair employee abilities, and alter employee behavior.
The instructor presented a method for getting an employee needed help, including enrollment in the EAP. The method for aiding a troubled employee had five distinct steps:
recognition, documentation, discussion, referral, and reintegration. Because the action step is the most difficult, the instructor presented a technique that provided the basics V
.
.
..
.-.
.
-..
-
..
.
__
-
g;? j r-
.i i <j l
r-
.:
.
'
e',
-13-
)
'
l
l for conducting a discussion between a supervisor and an employee. Case
'
studies provided scenarios which the students used in analyzing situations and determining appropriate corrective action.
l No violations or deviations were identified in this area.
>
11.
Exit Interviews (30703)
l
An exit interview was conducted on February 21, 1990, with licensee
!
representatives identified in paragraph 1.
During the interview, the NRC inspectors reviewed the scope and findings of the inspection. Other meetings between the NRC inspectors and licensee management were held
,
periodically during the inspection period to discuss identified concerns.
.
~
The licensee did not identify as proprietary any information provided to.
l or reviewed by, the NRC inspectors.
,
.
.
N
!
..-
,-
-
-.
-
._
0 j>
.
.-
t ATTACHMENT The inspector utilized the following documents during conduct of the scram discharge volume capability inspection.
Procedures Sarveillance Procedure (SP) 6.3.10.13. " North And South SDV Vent And Drain Valves Cycling, Open Verification, And Timing Test," Revision 7 dated November 3,1988 SP 6.1.14(N), " North SDV High Water Level Switches And Transmitters Calibration And Functional Test," Revision 21, dated October 10, 1989 SP 6.1.14(S), " South SDV High Water Level Switches And Transmitters Calibration And Functional Test," Revision 19, dated November 2,1989 L
SP 6.1.20(N), " North SDV High Water Level Switches And Transmitters Calibration And Functional Test," Revision 17 dated January 21, 1988 SP6.1.20(S),"SouthSDVHighWaterLevelSwitchesAndTransmitters Calibration And Functional Test," Revision 17, dated February 18, 1988
,
l l
Nuclear Performance ?rocedure 10.9, " Control Rod Scram Time Evaluation,"
!
Revision 17, dated April 27, 1989
_
l Alarm Procedure (AP) 2.3.2.27, " Panel 9-5 - Annunciator 9-5-1."
,
L Revision 17,' dated July 6, 1989 l
AP 2.3.2.28 " Panel 9-5 - Annunciator 9-S-2," Revision 20 dated t
l July 6, 1989-
!
. Drawings Stone & Webster Engineering Corporation (SWEC) Drawing No. (DWG)
13095.19-EP-1B-2 Scram Discharge Volume Drain Modifications Sections" l
SWEC DWG 13095.19-EP.2A-3, " Scram Discharge Instrument Volume: North
Header Instrument Piping" l
SWEC DWG 13095.19-E?-2B-3, "Scren Discharge Instrument Volune: South
'
<
Header Instrument Piping" SWEC DWG 13095.19-EP-2C-2, " Scram Discharge Volume Isometric Piping" SWEC DWG 13095.19-FSK-1-3, " Scram Discharge Instrument Volume Flow Diagram" Other Docunents Cooper Nuclear Station Technical Specifications i
I s
_
__
-
_
_
-.
-
, - _
[t i
.,
.,
l-l~ :_
,
,.
-
!
-2-t NRC Generic Safety Evaluation Report BWR SCRAM DISCHARGE SYSTEM,
,
Appendix B, dated December 1,1980 i
Minor Design Change 81-010-1, " Scram Discharge Volume Modification " dated March 24,1982
General Electric Memtrandum to Burns & Roe Inc., "CRD Scram Discharge Volume," dated April 6,1971
Anendment No. 5 Contract No. E69-11, dated November 5,1971
!
!
I l
,
P
'h t
1
.*
I a
i lj'
,
,
)
&
-
,-