IR 05000298/1990018
| ML20055C777 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/15/1990 |
| From: | Constable G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20055C775 | List: |
| References | |
| 50-298-90-18, NUDOCS 9006250103 | |
| Download: ML20055C777 (10) | |
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APPENDIX m
U.S. NUCLEAR REGULATORY COMMISSION
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REGION IV
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.NRC Inspection Report:
50-298/90-18 Operating License: DPR-46'
Docket:' 50-298
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Licensee:
Nebraska Public Power District (NPPD)
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P.O. Box 499 LColumbus, Nebraska 68602-0499
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Facility Name: Cooper Nuc' ear Station (CNS)
Inspection fi.: CNS, Nemaba County, Nebrana Inspection Conducted: April 16 through May 15, 1990 Inspectors:
G. A. Pick, Resident Inspector, Project Sectior; C Division of Reactor Projects-W. R. Bennett, Senior Resident Inspector, Project Section C Division of Reactor Projects
Approved:
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_. CmtWhief, Project-Section C D(te Division of Reactor Projects Inspection Summary i
Inspection Conducted April 16 through May 15, 1990 (Report'50-298/90-18)
Areas Inspected:
Routine,.unannounted inspection of operational safety
. verification, monthly surveillance observation, complex surveillance, monthly maintenance observation, refueling activities, and licensee event report fol.l owup.. Within these areas, the inspection consisted of selective
. examinations of procedures and representative records, interviews with
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personnel, and observations by the inspectors.
Results: Within the areas inspected, no violations or deviations were identified. The plant completed the Cycle 13 refueling outage on May 5, 1990.
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The operators conducted startup activities in-a safe, conservative manner.
After being notified o' a potential material deficiency affecting their 125 Vdc
and 250 Vdc battery racks, the licensee verified that the battery racks would i
-' withstand a design. basis seismic event assuming the worst-case conditions.
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9006250103 900618 PDR ADOCK 05000298 Q
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During performance of both complex and normal surveillance activities, the licensee demonstrated excellent coordination and communication among the plant staff, The conscientiousness of a mechanic prevented a lack of maintenance preplanning
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for a complicated maintenance activity (diesel generator cylinder reassembly)
i from becoming a problem. The mechanic requested that special work instructions be developed, because he felt uncomfortable implementing the number of critical steps without written instructions.
Discussions with the licensee about additional quality control concerns resulted in a self-audit of their program.
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I DETAILS
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Persons Contacted Principal Licensee Employees
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- G. R. Horn, Division Manager of Nuclear Operations
- J. M. Meacham, Senior Manager, Operations
- S. M. Peterson, Senior Manager, Technical Support
- R. L. Gardner, Maintenance Manager
- R. A. Jansky, Outage and-Modifications Manager
- J. R. Flaherty, Engineering Manager
- J. V. Sayer, Radiological Manager
- G. E. Smith, Quality Assurance (QA) Manager
- H. T. Hitch, Plant Services Manager
- L. E. Bray, Regulatory Compliance Specialist
- G. R. Smith.. Licensing Supervisor The inspectors also interviewed other licensee employees during the-inspectios pt, iod.
- Denotes those present during the exit interview on May 17,1990, 2.
' Plant Status The 61-day Cycle 13 refueling outage ended on May 5, 1990, when the licensee oeclared the reactor critical.
The plant was synchronized-to the grid on May 6.
The reactor operated at essentially full power from May 7 through the end of the inspection period.
The licensee completed-construction of a new building for_ storage of large, low turnover components.. It will also be used as a drug testing facility and office space for design engineers from the corporate office.
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Operational Safety Verification (71707)
The inspectors observed operational activities throughout the inspection period. Control room activities were observed to be well controlled.
Proper control room staffing was maintained and professional conduct was continuously observed.
Discuasions with operato': determined that they were cognizant of plant status and understood the importance of, and reason for, each lit annunciator. The inspectors obsercad selected shift turnover meetings and noted that information concerning p unt status and planned evolutions was communicated to the incoming operators.
Control panel walkdowns were conducted to verify that emergency cece cooling systems were in a standby condition. Tours of accessib'e areas at the facility were conducted to confirm operability of plant eqaipment, including the fire suppression systems and other emergency equipment.
Facility operations were performed in accordance with the requirements established in the CNS operating license and Technical Specifications (TS).
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A Nuclear Utility Procurement Issues Council (NUPIC) audit of C&D power systems cor. ducted February 12-16,.1990, identified a problem with the material used in their manufacture of battery racks.
Wyle Laboratories had qualified the C&D battery racks to AISI 1010 or ASTM A36 steel; l
however, C&D, subsequent to the qualification testing, provided AISI C1010-l or C1008 steel instead, without justifying the~ change of material..- The i
licensee's QA department received the NUPIC audit on March 23. After
'l completing their evaluation on April 13, they contacted design engineering about this material deficiency.
The licensee updated their 125 V and 250 V battery rack seismic calculations based on these' worst-case material conditions.. The calculations proved that'the potentially reduced
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-allowable stresses of the battery rack material were acceptable.
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The inspectors attended the startup review meeting which was convened to
- i complete General Operating Procedure 2.1.1.1, " Plant Startup Review and Authorization," Revision 4, dated September 7, 1989.
The procedure-outlines a method to ensure that all outstanding commitments which could
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impact plant startup have been resolved satisfactorily.
Items discussed
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and resolved included:
(1).an update on the wall-thinning justification l
lfor interim operation and (2) the calibration project status.
The licensee declared the reactor critical on May 5.
The plant startup was performed'
in a safe, conservative manner with only minor problems occurring.
The inspector observed no evidence of management pressure to perform the startup quickly.
_j 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 l
and/or contaminated areas were properly posted and controlled.
Radiation-
work permits contained appropriate information to ensure that work could
be performed in a safe and controlled manner.
Radiation monitors were-
properly utilized to check for contamination.
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The inspectors _ observed security personnel performing their duties of a
vehicle, personnel, and package s'earch.
Vehicles were properly authorized
and~ escorted or controlled within the protected area (PA).
The PA barrier
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had adequate illumination and the isolation zones were free of transient i
material.
Site tours were conducted by the inspectors to ensure that
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compensatory measures were oroperly implemented as required.
l No violations or deviations were identified in this area.
The licensee
responded to an issue identified in the NUPIC audit in a timely, conservative manner.
The litansee conducted a comprehensive review of
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issues affecting plant startup to assure that all commitments had been met.
The reactor startup following the Cycle 13 refueling outage was performed in a coriservative manner, with no evidence of management pressure to start' up quickly.
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5-4.
Monthly Surveillance Observations (61726)
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.The inspectors observed the performance'of and/or reviewed the following surveillance procedures (SP):
l The inspectors observed portions of the following SP. and reviewed the test
results.
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NPP 10.4, " Control Rod Drive Friction Test," Revision 0, dated (
May 25, 1989.
On April-20, 1990, the inspectors observed licensee personnel performing this test for 11 control rods.
The friction test measures times for full t~ avel withdrrwal and insert for each control rod to
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determine potential degradation.
The test required coordination by
reactor engineering, instrumentation and controls (I&C), and plant
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operations.
Communications were properly established and maintained
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during the test.
Discussions with personnel performing the test
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incided that they were aware of the test purpose and limitations.
The I&C pe,xonnel followed proper radiological practices and used i
calibrated eqv oment.
After withdrawing a control rod, operators checked tne stall flow to I
assure that a minimum amount of cooling was provided to the control
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rod drives and to verify that the control rods were coupled, i
SP 6.3.10.2, " Instrument Line Flow Check Valve Test," Revision 17,
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dated March 1, 1990.
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The inspectors observed.the performance of this test on April 27, 1990.
This. procedure tests the excess flow check valves once per i
operating _ cycle to ensure that, upon a postulated instrument line-
break, the leakage would be limited to between 0.2 and 0.7' gallons'
i per minute, thereby, not exceeding ASME Section XI requirements. The
licensee divided the excess flow check valves into 11 groups with the
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testing divided between two-crews. The groups reflected the physical i
location of instrument racks in the reactor building.
Each crew l
consisted of a licensed operator, a station operator, an I&C i
l technician, and a health physics technician.
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The 1icensee tested 69 excess flow check valves located throughout the reactor building. Additionally, I&C technicians isolated several instruments on each instrument rack to prevent inadvertent equipment actuations. The I&C technicians took precautions to prevent the spread.of contamination by pouring reactor coolant water collected during the test of each check valve into equipment drains.
The procedure was well-written, easy-to-follow, ano comprehensive. The licensee completed the test satisfactorily.
Four check valves found out,of specification were cycled several times to bring the as-left values within specification. Two other check valves were replaced
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because their flow could not be brought within specification.
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licensee issued a nonconformance report to evaluate the check valve failures.
No violations or deviations were identified in this area. Coordination and cooperation among the licensee departments,was evident.
Personnel-followed good radiological practices during performance of the testing.
5.
Complex surveillance (61701)
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The inspectors reviewed this area to verify that the licensee conducted their more complex surveillances in accordance with regulatory requirements.
The inspectors witnessed the primary coolant system hydrostatic test and the loss of offsite power / emergency core cooling
system auto-initiation test.
SP 6.3.10.28, "ASME Class 1-N System Leakage Test," Revision 0, dated April 23, 1990.
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The inspectors witnessed the test performance on April 27, 1990.
Prior to start of the test, the test prerequisites were met, as well
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as precautions and limitations for heatup and pressurization of the primary system. The procedure required operators to verify the valve lineup, allow the primary system to be pressurized up to the outboard isolation valve, and insert a manual scram which allowed the control rod drive and scram volume piping to be tested.
Licensed operators and engineers performing the visual inspections had received training on performing visual inspections. Mechanics promptly repaired minor
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packing leaks and a body-to-bonnet leak on an inboard main ' steam isolation valve after the test.
A licensed operator maintained the pressure between 1018 and i
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1038 psig by varying the CR0 flow to the vessel while maintaining a
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constant blowdown. The operator-controlling the pressure compensated for fluctuations created by the ongoing excess flow check valveitest.
The' licensed operator and the engineer conducting visual inspections
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of pipe in the drywell did a thorough inspection and took proper L
safety precautions.
SP 6.3.4.3, " Sequential Loading.of Emergency Diesel Generators,"
Revision 28, dated August 10, 1989.
This test verified that the emergency diesel generators (EDGs)
auto-start, the emergency busses shed their loads, and the EDGs load onto the emergency bus on a loss of offsite power and low low low
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reactor vessel level.
In addition, the core spray (CS) and residual heat removal (RHR) pumps were verified to auto-start in their designed timing sequence.
On April 28, 1990, the inspectors witnessed the EDG No. 1 test and system restoration and the EDG No. 2 setup and testing. All l'
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-7-prerequisites were completed prior to start of the test. All test equipment utilized during performance of the test was verified to be in calibration.
A senior reactor operator (SRO) coordinated the setup and performance of each test.
The SRO reviewed how the loss of various noncritical electrical buses during performance of the test would affect the plant. ' Operators transferred power from the affected reactor protection system motor generator set to a power source on the other train ensuring that a one-half scram would not occur.
The procedure.
required realignment of certain components to minimize the impact on the facility.
The test conductor for each test did an excellent job of coordinating the surveillance preparations. Qualified parsonnel were assigned to the various test stations.
Minor errors in the test procedure were identified and resolved by the licensee.
For example, the test requires the operator to reset a Group VI isolation; however, the Group VI-isolation had apparently. reset itself. The licensee determined that no Group VI isolation signal occurred because the simulated signal for the low low low reactor vessel level did nat provide the signal for the containment isolation. An electrical engineer, familiar with the original installation, determined that the indicating lights lose power on a loss of the critical bus, aut-reenergize when the EDG loads onto the bus.
During the test of EDG No. 1, RHR Pump B sequenced on in 5.8 seconds, compared to the 5.0 +/- 0.5 second specification. The licensee determined that the extended time did not affect the system operability and issued a nonconformance report to evaluate the root cause. All the other pump start times met test specifications.
No violations or deviations were identified in this area.
Qualified personnel performed the visual inspections during the primary system hydrostatic test.
The inspections were' thorough and complete. The licensee demonstrated excellent coordination and communication during the EDG. sequencing test. Discrepancies identified during testing were promptly resolved.
6.
Monthly Maintenance Observation (62703)
On April 17, 1990, the inspectors observed.I&C technicians reinstalling a preregulator for Source Range Monitor (SRM) D.
The repaired preregulator had been satisfactorily postmaintenance tested in a spare SRM cabinet.
The technicians verified the setpoints after installation of the preregulator.
On April 20, 1990, the inspectors observed electricians conducting a functional check of the EDG No. 2 field ground relay. The preventative maintenance activity is conducted every 2 years in accordance with Maintenance Procedure (MP) 7.3.8, " Diesel Generator Field Ground Relay
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p-8-Maintenance and Functional Check." ihe electricians visually inspected the relay contacts looking for pitting and tarnishing. Additionally, they.
checked for loose screws and for proper alignment of the relay solenoids.
The test plug was properly installed. During the functional check,- the relay failed to drop out. The electricians used a relt.y burnishing tool to clean the contacts and made a slignt contact adjustment. The subsequent test functioned satisfactorily.
On April 25, 1990,-the inspectors observed a portion of the reassembly of the 4L cylinder on EDG No.1.
The cylinder required repair because plant personnel observed water leaking from the cylinder head during the diesel generator operability test.
During the inspectors' observation, the maintenance work request package consisted of a work addendum sheet, a postmaintenance test sheet, and a quality control (QC) check sheet.
The QC check sheet specified the required values for torquing the head bolts, the rocker arm bolts, and the.
injector. There were no detailed work instructions present.
From discussions with the mechanics and the lead mechanic observing the in progress work, the inspectors determined that special instructions were being developed and that the reassembly had been stopped.
The lead mechanic had stopped the work because he felt uncomfortable implementing the number of critical steps involved for reassembly of the cylinder
.without written instructions.
The mechanics had replaced the jacket water 0-rings and had torqued the cylinder head under direction of the lead mechanic with no written guidance available. The inspectors questioned the lead mechanic on what he looked for before setting the cylinder head in place and how torquing was. accomplished. The mechanic's response agreed with the special written instructions..He had inspected for cleanliness and indications of damage.
Additionally, he stated that the head bolt torquing was accomplished in three'. increments of 500, 1000, and 1200 ft-lbs., respectively.
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The inspectors' review of the completed work package determined that the special instructions used were well written.
The postmaintenance test of-the EDG was satisfactory.
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The' inspectors identified another QC concern while reviewing the completed corrective actions for Nonconformance Report (NCR)89-149. The NCR documented the corrective actions related to cracking of the injector pump
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pedestal on EDG No. 1.
The special Wk instructions written to perform required inspections on EDG No. I and EDG No. 2 were different. The instructions had been written
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by different departments using the same guidance documents; however, the use of QC steps, the terminology for the type of QC needed, and the number of QC steps differed between the work packages. After discussion with the inspectors, the licensee developed an action plan for evaluation and possible upgrade of the QC program.
The action plan included a QA review
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1 of maintenance activities to assure independence of QC and consistency
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among work packages. This QA surveillance should identify the extent of problems. Based on the evaluation, changes to the QC program and subsequent training on the upgraded program will be conducted.
The licensee established estimated completion dates for the program upgrade.
No violations or deviations were identified in this area. The conscientiousness of the lead mechanic prevented lack of maintenance-
preplanning from causing a problem to occur during repair of an EDG.
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Discussions with the licensee on QC issues resulted in the performance of-a QA surveillance of the licensee current work practices.
7.
Onsite Followup of Written Reports (92700)
(Closed) LER 89-009: This licenree event report (LER) documented concerns with the inservice testing (IST) program identified in NRC Inspection Report 50-298/87-10.
During the performance of a safety system functional inspection from May 11' to June 19, 1987, NRC identified that Service Water Pump 1C was operating-in the alert range; however, pump testing at the required
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increased frequency was not being conducted.
Corrective action included revision of IST program procedures and a programmatic review of all IST pump procedures.
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The inspectors verified that the licensee completed the programmatic review of IST pump procedures. This review resulted in changes to the service water pump and standby liquid control system pump surve*;11ances.
The inspectors reviewed Engineering Procedure 3.9, "ASME Code Testing of Pumps and Valves," Revision 1, dated January 18, 1990. The procedure i
specifies the required ASME testing and describes the. trending required to be performed by the IST engineer.
The inspector verified that increased frequency of testing has been performed whenever pumps are'in;the alert range and that action has been taken whenever pumps are in the. action
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required range. This LER is closed.
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Refueling Activities (60710)
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On April 18, 1990, the inspectors observed the video verification of the L
. reactor core in accordance with Nuclear Performance Procedure (NPP) 10.21,
"Special Nuclear Materials Control and accountability Instructions,"
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Revision 4, dated September 7, 1989.
The three-member team reviewing the film represented QA, operations, and the technical support staff. The team members followed the procedure outlined in Step 8.1.2.2.c.
The personnel selected were experienced and knowledgeable and were not directly involved with the core videotaping.
The Cycle 13 reload had four GE-11 lead test assemblies (LTA) loaded which represented the latest design in GE nuclear fuel.
In response to NRC
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questions about additional surveillances performed on the LTAs, the onsite GE representative received training from the licensee for identifying proper orientation of LTAs installed in the core. On April 17, the onsite GE representative verified proper orientation of the GE-11 LTAs from the
refueling bridge using a video camera.
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-Exit Interviews (30703)-
An exit interview was conducted on May 17, 1990, with. licensee i
representatives identified in Paragraph 1.
During the interview, the inspectors reviewed the scope and findings of the inspection. Other-meetings between the. 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,
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or reviewed by, the inspectors.
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