IR 05000461/1988030
| ML20235N056 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 02/15/1989 |
| From: | Cooper R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235N051 | List: |
| References | |
| 50-461-88-30, IEB-88-076, IEB-88-76, NUDOCS 8903010004 | |
| Download: ML20235N056 (16) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-461/88030(DRP)
Docket No. 50-461 License No. NPF-62 Licensee:
Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name:
Clinton Power Station Inspection At:
Clinton Site, Clinton, Illinois Inspection Conducted:
December 6, 1988, through January 27, 1989 Inspectors:
P. Hiland S. Ray SMf Approved By:
W.
o I
ef
Reactor Projects Section 3B Date
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Inspection Summary Inspection on December 6, 1988, through January 27, 1989 (Report No. 50-461/88030(DRP))
Areas Inspected:
Routine, unannounced safety inspection by the resident
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inspectors of licensee action on previous inspection findings; review of Information Notices; preparation for refueling and refueling activities; operational safety verification; monthly maintenance observation; monthly surveillance observation; onsite followup of events at operating reactors; maintenance programs; and environmental qualifications.
Results: Of the nine areas inspected, one violation was identified in the area of licensee action on previous inspection findings. The violation concerr.ad operation with one diesel generator not meeting its design l
basis.
One unresolved item and one open item were identified in the area of environmental qualification (Paragraphs 10.a. and b.).
The unresolved item concerned inadequate corrective action to a previous violation of unqualified butt splices.
The open item concerned a 10 CFR 21 notification from Limitorque Corporation identifying potential failure of torque switches.
In addition, two violations were identified in the area of operational safety i
verification.
The violations concerned an inadequate surveillance which did not include all the items required to be checked, and a missed surveillance due to personnel error.
However, for these two violations, in accordance with 10 CFR 2, Appendix C, Section V.G.1, a Notice of Violation was not issued. All of the above violations were receiving management attention.
8903010004 890213
PDR ADOCK 05000461
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DETAILS I
1.
Personnel Contacted Illinois Power Company (IP)
W. Kelley, President
- W. Gerstner, Executive Vice President
- D. Hall, Vice President - Nuclear
- J. Perry, Assistant Vice President
- J. Wilson, Manager - Clinton Power Station
- K. Baker, Supervisor - I&E Interface J. Brownell,-Project Engineer / Specialist
- R. Campbell, Manager - Quality Assurance
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- J. Cook, Manager - Nuclear Planning and Support H
- R. Freeman, Manager - Nuclear Station Engineering Department
- D. Holesinger, Assistant Manager - Clinton Power Station
- D. Holtzcher, Acting Manager - Licensing & Safety
- A. Mcdonald, Director - Nuclear Program Assessment
- J. Miller, Manager - Scheduling & Outage Management
- J. Weaver, Director - Licensing
- R. Wyatt, Manager, Nuclear Training Soyland/WIPC0
- J. Greenwood, Manager - Power Supply Nuclear Regulatory Commission
- E. Greenman, Projects Division Director, Region III
- R. Knop, Chief, Projects Branch 3, Region III
- R. Cooper, Section Chief, DRP, Region III
- B. Drouin, Project Inspector, DRP, Region III
- P. Hiland, Senior Resident Inspector, Clinton
- S. Ray, Resident Inspector, Clinton
- Denotes those attending the monthly exit meeting on January 30, 1989.
- Denotes those attending the Management Meeting on January 10, 1989.
The inspectors also contacted and interviewed other licensee and contractor personnel.
2.
Previously Identified Items (92701, 92702)
a.
(Closed) Unresolved Item (461/88023-01):
Design Deficiency in the Division III Diesel Generator Exhaust.
l This item was previously discussed in Inspection Reports No. 50-461/88023, Paragraph 4.a. and No. 50-461/88027, Paragraph 10.b.(4). On October 7, 1988, the inspectors noted that
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a rupture disk had been left installed in the exhaust line for the Division III Diesel Generator (D/G) but similar disks had been removed from Divisions I and II during plant construction, in accordance with ECN 6674 issued in November 1985. The item remained unresolved pending the licensee's inve:tigation to determine the correct configuration and safety significance.
On December 13, 1988, the licensee issued Licensee Event Report l
(LER) No. 88-027-00 which reported that the Division III'D/G had been inoperable from initial plant startup on February 27, 1987, until the rupture disk was removed on November 11, 1988.
During that' period, the. plant was operated in a. condition outside the design basis because missile damage to the unprotected portion of the Division'III D/G exhaust line concurrent with a loss of
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offsite power may have rendered the D/G and thus the High j
Pressure Core Spray (HPCS) system inoperable.
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The LER also reported that the licensee's decision on October 7, 1988, when initially informed by the inspectors of the installed rupture disk, had been incorrect. The Shift Supervisor had determined that the Division III D/G was still operable since it would take a concurrent loss of offsite power and tornado generated missile to render it inoperable.
The Shift Supervisor believed that the D/G still met its design basis because-it met the single failure criteria but he failed to consider 10 CFR 50,
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Appendix A, General Design Criteria 2, which required that i
safety systems be designed to withstand the effects of natural phenomena without the loss of capability to meet their safety functions.
Technical Specification 3.8.1.1 required that three separate and independent diesel generators be OPERABLE in OPERATIONAL CONDITIONS 1,
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2, and 3.
In addition, Technical Specification 3.8.1.1 ACTION d.
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required that, with the Division III D/G inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, HPCS be declared inoperable. Technical Specification 3.5.1.
ACTION c. required that, with HPCS inoperable for more than 14 days, the plant be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and ir.
COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Operating the plant extensively in OPERATIONAL CONDITIONS 1, 2, and 3 oetween February 27, 1987, and November 11, 1988, with the Division III D/G inoperable due to being unable to meet its. design basis is a violation (50-461/88030-01(DRP)).
Unresolved Item 461/88023-01 is closed and elevated to a violation.
b.
(Closed) Unresolved Item (461/88023-03):
Failure to report an Engineered Safety Feature (ESF) Actuation.
This item was discussed in Inspection Report No. 50-461/88023, Paragraph 11.b.(2). The item was unresolved pending the licensee's clarification of which systems described in the Updated Safety Analysis Report (USAR) are considered to be ESFs.
10 CFR 50.72 required that automatic actuations of ESFs be
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reported to the NRC within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> but did not require reports
for automatic actuations of Essential Auxiliary Support (EAS)
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systems.
The licensee determined through review of various regulatory documents and discussions with their architect-engineer (Sargent & Lundy) that the Essential Switchgear Heat Removal (VX) system and Diesel Generator Ventilation (VD) system were improperly classified as ESFs in their USAR.
The systems were actually EASs.
In Attachment B to Illinois Power letter U-601326 dated December 9, 1988, the licensee informed the NRC that they will use the list of ESFs in Chapter 6 of the USAR.
They committed to update the lists in Chapters 7 and 9 during the next normal USAR revision cycle. The inspectors determined that those actions were reasonable. This item is closed.
One violation was identified.
3.
Followup of Information Notice / Regional Request (92701)
For the Information Notice discussed below, the inspectors verified that the licensee had received the Information Notice, had distributed the Notice to appropriate personnel, and had completed appropriate actions.
(Closed) Information Notice No. 88-76 (461/88076-NN):
Phenomenon Not Previously Considered In The Design Of Secondary Containment Pressure Control.
This Information Notice was received by the licensee on September 26, 1988.
Following receipt, the licensee assigned review responsibility in accordance with Licensing and Safety Procedure L.1, " Feedback Program." IP Review Sheet Y-209467 dated October 10, 1988, assigned responsibility for review to the Nuclear Station Engineering Department (NSED).
The subject Information Notice identified a phenomenon whereby instruments measuring the differential pressure (delta P) between the interior of the secondary containment and the atmosphere did not take into account the temperature-induced difference in the pressure gradients inside and outside the secondary containment.
The NSED review of this Information Notice was documented in IP memorandum Y-90331 dated December 31, 1988.
l That review concluded that the phenomenon discussed in Information l
Notice No. 88-76 was not applicable to Clinton Power Station because inside and outside pressures were measured at approximately the same l
temperatures and elevations. As detailed in the licensee's evaluation, the inside pressure taps were located at the base of the containment gas control boundary at elevation 800'.
The licensee's evaluation stated that the temperature within the containment gas
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control boundary at elevation 800' would be about the same as the
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outside temperature since the boundary was made of metal siding, maintained at a-negative pressure, and there were no heat sources.to raise the temperature.
In addition, the licensee's evaluation stated that at elevations below 800' the delta P would be higher than-indicated delta P because the lower areas had higher space temperatures'which result in lower air density.
Based on the above, the inspectors concluded that the licensee had adequately reviewed the subject Information Notice. This item is closed.
No violations or deviations were identified.
4.
Preparation For Refueling and Refueling Activities (60705, 60710)
The inspectors observed several of the activities conducted in preparation for refueling such as fuel handling systems repairs and-surveillance, radiation protection preparations, and reactor disassembly.
The inspectors also observed some new and spent fuel movements. The licensee encountered several problems with the Inclined Fuel Transfer system and other fuel handling systems which caused delays in actual refueling activities.
During the removal of.
the steam dryer from the reactor on January 7, 1989, an unexpected
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problem was encountered when an area radiation alarm activated an interlock on the polar crane and stopped its movement.
Investigations of that event disclosed inadequacies in the training and supervision of the refueling contractors.
On January 7,1989, a violation of Technical Specifications occurred when the single rod out interlock system was defeated for one control rod by installing an improper jumper. The jumper caused an indication that.the rod was fully inserted when it was actually fully withdrawn. At that point, a second control rod could have been withdrawn contrary to the requirements of Technical Specifications.
The error was attributed to an improperly written procedure.
On January 22, 1989, a second violation of Technical Specifications occurred when core alterations were performed for several hours with required source range monitors inoperable. This event was caused by the Shift Supervisor not adequately determining the status of the
"D" source range monitor before declaring it operable. Additional problems with low signal-to-noise ratios on source range monitors caused other delays in refueling activities.
The events above as well as other refueling activities will be discussed in Inspection Report No. 50-461/89002 which will detail the findings of a regional specialist's inspection conducted during the refueling outage.
During the reactor disassembly, with the reactor cavity drained, the i
inspectors observed personnel working around the open pit inside the
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I handrails without using safety harnesses.
In one case, a worker was
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leaning over the pit hauling a bucket of tools up and down.
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inspectors brought the situation to the attention of the supervisor of the refuel floor operations but improvements in safety precautions were temporary and ineffective.
The inspectors then brought the situation to the attention of the site safety organization who promptly verified that the conditions existed and took prompt action to correct the deficiencies.
Violations and deviations in this area will be reported in Inspection No. Report 50-461/89002.
5.
Operational Safety Verification (71707)
The inspectors observed control room operations, attended selected pre-shift briefings, reviewed applicable logs, and conducted discussions with control room operators during the inspectior, period.
The inspectors verified the operability of selected emergency systems and verified tracking of LCOs.
Routine tours of the auxiliary, fuel, containment, control, diesel generator, turbine buildings and the screenhouse were conducted to observe plant equipment conditions including the potential for fire hazards, fluid leaks, and operating conditions (i.e., vibration, process parameters, operating temperatures, etc).
The inspectors verified that maintenance requests had been initiated for discrepant conditions observed.
The inspectors verified by direct observation and discussion with plant personnel that security procedures and radiation protection (RP)
controls were being properly implemented.
Inspections were routinely performed to ensure that the licensee conducted activities at the facility safely and in conformance with regulatory requirements. The inspections focused on the implementation and overall effectiveness of the licensee's control of operating activities, and the performance of licensed and nonlicensed operators and shift technical advisors. The following items were considered during these inspections:
Adequacy of plant staffing and supervision.
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Control room professionalism, including procedure adherence, operator attentiveness and response to alarms, events, and off-normal conditions.
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Operability of selected safety-related systems, including attendant alarms, instrumentation, and controls.
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Maintenance of quality records and reports.
During the first part of the inspection period the plant was operated in coastdown with control rods fully withdrawn.
On December 18, 1988, a degrading seal on the "B" Reactor Recirculation Pump reached
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the point where plant management decided to go into single loop operation. The plant remained in single loop until it was shutdown
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-on January 2, 1989.for the first refueling outage.
During the remaining portion of the inspection period the plant remained L
shutdown for refueling and other outage activities, a.
On December 1, 1988, while performing an independent technical review of a revision to surveillance procedure 9082.01 for electrical distribution system verification, the licensee identified that the Division 4 Regulating Isolation Transformer power switch was not included in the surveillance.
Thus Technical Specification surveillance 4.8.3.1.1 and 4.8.3.2 which required a verification of AC breaker alignment be performed at least once per 7 days had never been fully conducted.
As a result of this event, the licensee reviewed their surveillance procedures for similar missing switches and identified three additional regulating isolation transformer switches that were not being checked. The licensee reported their findings to the NRC in Licensee Event Report (LER)
88-29-00 dated January 3,1989. The inspectors determined that the licensee's evaluation and corrective actions for this event were good and that it had been properly reported and should not have been prevented by corrective actions for any previous violation.
Therefore, in accordance with the discretion allowed by 10 CFR 2, Appendix C, Section V.G.1, a Notice of Violation will not be issued. The inspectors will follow-up the remaining corrective actions in the LER in a later inspection.
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b.
On December 18, 1988, while placing the plant into single loop operation because of degradation of the "B" Reactor Recirculation Pump seal, reactor power was reduced below the High Power Setpoint of the Rod Pattern Control System.
Subsequently rods were withdrawn and power was raised above the setpoint.
Technical Specification Table 4.3.6-1 Note c required that the Control Rod Block, Rod Pattern Control System channel functional tests be conducted within one hour prior to rod movement whenever one of the power setpoints was crossed.
Due to a personnel error on the part of the Line Assistant Shift Supervisor, a licensed Senior Reactor Operator, the surveillance were not performed prior to rod withdrawal.
The error was discovered by the licensee and reported to the NRC in LER 88-031-00 dated January 16, 1989. The inspectors found that the licensee's actions were prompt, the event was properly reported, and the event should not have been prevented by corrective actions for any previous violation.
Therefore in accordance with the discretion allowed by 10 CFR 2, Appendix C, Section V.G.1 a Notice of Violation will not be issued. The inspectors will follow-up the remaining correctiva action of the LER in a later inspection.
Two violations for which Notices of Violation will not be issued were identified.
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6.
Monthly Maintenance Observation (62703)
511ected portions of the plant maintenance activities on
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i safety-related systems and components were observed or reviewed to ascertain that the activities were performed in accordance with approved procedures, regulatory guides, industry codes and standards, and that the performance of the activities conformed to the Technical Specifications.
The inspection included activities associated with i
preventive or corrective maintenance of electrical, instrumentation and control, mechanical equipment, and systems. The following items
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were considered during these inspections:
the lining conditions for operation were met while components or systems i.ere removed from service; approvals were obtained prior to initiating.the work; activities were accomplished.using approved procedures and were inspected as applicable; functional testing and/or calibration was performed prior to returning the components'or systems to service; parts and materials that were used were properly certified; and appropriate fire prevention, radiological, and housekeeping conditions were maintained.
The inspectors observed / reviewed the following work activities:
Maintenance Work Procedure No.
Activity C-45216, Modification DG-38 Division 3 Turbocharger Installation CP.S No. 8121.03 Control Rod Drive Leak Test C-56314 Disassemble Reactor Vessel D-01770 Control Room HVAC "A" Motor Repair PMMDGM031 Division 1 EDG Cylinder Head / Liner Seal Replacement C-45387 Low Pressure Turbine Repairs C-22112 Hydraulic Control Unit Inspections No violations or deviations were identified.
7.
Monthly Surveillance Observation (61726)
An inspection of inservice and testing activities was performed to ascertain that the activities were accomplished in accordance with applicable regulatory guides, industry codes and standards, and in con +;rman..a with regulatory requirements.
Items which were considered during the inspection included whether l
adequate procedures were used to perform the testing, test instrumentation was calibrated, test results conformed with Technical Specifications and procedural requirements, and tests were performed
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within the required time limits. The inspectors determined that the test results were reviewed by someone other than the personnel involved with the performance of the test, and that any deficiencies
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identified during the testing were reviewed and resolved by appropriate management personnel.
The inspectors observed / reviewed the following activities.
Surveillance / Test Procedure No.
Activity CPS No. 9861.02 MSIV Local Leak Rate Test CPS No. 9431.13 Source Range Calibration CPS No. 9432.62 Containment Vent Exhaust Radiation Monitor Calibration No violations or deviations were identified.
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8.
Onsite Followup of Events at Operating Reactors (93702)
a.
General-The inspectors performed onsite followup activities for events which occurred during the inspection period.
Followup inspection included one or more of the following:
reviews of operating logs, procedures, condition reports; direct observation of licensee actions; and interviews of licensee personnel.
For each event, the inspectors reviewed one or more of the following:
the sequence of actions; the functioning of safety systems required by plant conditions; licensee actions to verify consistency with plant procedures and license conditions; and verification of the nature of the event. Additionally, in some cases,.the inspectors verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel errors and were taking or had taken appropriate corrective actions. Details of the events and licensee l
corrective actions noted during the inspectors' followup are l
provided in Paragraph b. below.
b.
Details (1) Entering ACTION Statement Requiring Plant Shutdown due to Surveillance Testing on Drywell Leak Detection System On December 16, 1989, the licensee informed the NRC via the ENS that they had entered the Technical Specification ACTION statement for Specification 3.4.3.1 for the Reactor j
Coolant System Leak Detection system. That specification required the plant to be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following
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24' hours.. 'The Drywell Sump Flow Mbnitoring system had been
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inoperable since. December 4, 1988. A channel functional test became due on the Drywell Atmosphere Particulate Radioactivity Monitoring system and the Drywell Atmosphere
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Gaseous Radioactivity Monitoring system. The surveillance i
would render the systems inoperable for a short. period of
. time which would require entry into the ACTION statement.
The licensee discussed the situation with the inspectors and Region III management in advance.to inform the NRC of their intent to enter the required shutdown ACTION statement. The inspectors determined that the surveillance activity was.well planned and the licensee was able to perform it in less than l' hour. The licensee exited the ACTION statement without performing any actual power.
reduction.
(2) Entering ACTION Statement Requiring ~ Plant Shutdown due to Blown Fuse on Drywell Leak Detection System On December 19, 1988,-the licensee informed the NRC via the ENS that they had entered the ACTION statement of Technical Specification 3.4.3.1 for the Reactor Coolant System Leak Detection system. The event was caused by an instrument technician who inadvertently touched the wrong terminal with a multimeter probe while performing a monthly surveillance on the Reactor Water Cleanup system area temperature detector. The probe caused a short which resulted in a blown fuse on the Drywell Atmosphere Particulate and Gaseous Radioactivity Monitoring systems.
Since the Drywell Sump Flow Monitoring system was still inoperable as described in the paragraph above, Technical Specifications required that the plant be in at least HOT SHUTDOWN in the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The licensee replaced the fuse in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and exited the ACTION statement before any actual power reduction.
In the critique of the event, several problems were discussed which contributed to the event. The most significant was that there was insufficient physical space in many of the instrument panels for taking the measurements required in several routine surveillance.
The same situation contributed to the event discussed in Paragraph 8.b.(6) below.
(3) Notification of Unusual Event Due To Shutdown Required by Plant Technical Specifications [ ENS No. 14388]
On January 2, 1989, the licensee informed the NRC via the ENS of a Notification of Unusual Event when a plant shutdown l
was required by Technical Specification 3.0.3.
At the time of the event, the licensee was in the process of performing
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a scheduled plant shutdown for their 'first refueling'.
The plant had been. operating.in single loop since December 18, 1988, due to a degraded seal.in the "B" Reactor Recirculation Pump. Technical Specification surveillance.4.4.1.1.3.d.
required verification of core flow greater than 39% when thermal power was within the unrestricted zone of Figure 3.4.1.1-1.
With thermal power.in the unrestricted zone, core flow was reduced to less than 39% to facilitate plant shutdown. The licensee noted their inability to meet the surveillance requirement and entered Technical Specification'.3.0.3. The licensee contacted Region III
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management to advise them of their intent to carry out the requirements of Specification 3.0.3 which required that the plant be placed in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and at least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The licensee continued with the plant shutdown and terminated the Unusual Event about four hours later when the plant reached HOT SHUTDOWN.
The event was caused by an error in the Technical Specifications. The word " unrestricted" in Specification 4.4.1.1.3.d. should have been " restricted." The licensee submitted letter U-601356 dated January 26, 1989, to propose an amendment to the Technical Specifications to correct the error.
(4) Main Steam Isolation Valves Fail Local Leak Rate Test
[ ENS No. 14416]
On January 5, 1989, the licensee informed the NRC via the ENS that three of the four main steam lines had failed to meet their local leak rate test criteria of 28 standard cubic feet per hour per line. Only the two "B" Main Steam Isolation Valves (MSIVs) were found to be acceptable.
The licensee determined that five of the other six MSIVs on the other three steam lines would have to be repaired during the current outage.
Excessive MSIV leakage had been previously identified and repaired on three main steam lines during an October 1987 outage and on one main steam line during a March 1988 outage. These events were discussed in Inspection Reports No. 50-461/87036, Paragraph 11.b.(2); No. 50-461/88004, Paragraph 9.b.(8);
No. 50-461/88009, Paragraph 3.e.; and No. 50-461/88027, Paragraph 6.c.
The licensee also described the former problems in Licensee Event Reports'(LERs) 87-062-00 dated
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November 24, 1987, and 88-008-00 dated April 8, 1988. As discussed in Inspection Report No. 50-461/88027, the licensee was planning to incorporate Field Alteration MSF012 in the second refueling outage which should improve the valves'
seating characteristics.
The repairs to the MSIVs will followed by a regional maintenance team inspection scheouled
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to begin on March 13, 1989.
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l (5) ESF Actuation due to Pressure Transient While Valving in a j
' Reactor Water Level Instrument l
On January 9,1989, th'e licensee reported to the NRC via -
i the ENS that they had experienced an unexpected automatic-actuation of an ESF. The event was caused when improper restoration of reactor level instrument IPT-CM265 following excess flow check valve testing resulted in the. generation of isolation signals in instruments sharing a common sensing line. This resulted in isolation of instrument air containment isolation valves IIA 006 and IIA 007 and the generation of a Reactor Core Isolation Cooling (RCIC)
system auto start signal.
RCIC was tagged out of service at the time and did not actuate. This event was similar to previous ESF actuations while restoring pressure and level instruments following maintenance and surveillance reported by LERs 87-004-00, 87-014-00, 87-022-00, and 87-026-00.
Inspection Report No. 50-461/88003, Paragraph 5.c. discussed the corrective actions for those previous events.
Review of the procedure being used, Technical Surveillance CPS No. 9864.01, revealed that the procedure did not identify any automatic trips or actions which could have been generated-as a result of the test.
While that was true of the instrument being tested, it did not recognize the possible effects on other
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instruments sharing the same sensing line.
Review of the corrective actions taken as a result of previous LERs revealed that instrument restoration techniques and precautions required to prevent trips /actuations had been identified and implemented for maintenance actions via maintenance procedure CPS No. 8801.12. The corrective actions focused on maintenance and operations activities and had not been incorporated into technical surveillance procedures.
The inspectors will follow-up corrective actions in conjunction with review of the LER which will report this event.
(6) ESF Actuation Causing Loss of Shutdown Cooling due to Inadvertent Grounding of Terminal While Conducting Surveillance [ ENS No. 14457]
On January 10, 1989, the licensee reported to the NRC via the ENS that they had experienced an unexpected ESF actuation which caused a temporary loss of shutdown cooling. The event was caused by an instrument technician who inadvertently grounded a lead in the Reactor Core Isolation Cooling system
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during the conduct of a Logic System Functional test.
Insufficient physical space in instrument panels contributed to this event similar to the event discussed in Paragraph 8.b.(2) above.
Corrective actions for that event had not been completed at the time of this event. Grounding
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of the lead caused a fuse to blow in a 24VDC power-supply circuit causing a loss of' power and isolation si.gnal to the Residual Heat Removal (RHR) system.
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l The isolation signal caused the outboard containment l
isolation valve for RHR (1E12-F008) to close.
Since this was the suction path for RHR-B, the pump tripped.
Plant
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technicians replaced the blown fuse and RHR-B was restarted
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l in a shutdown cooling mode in about 37 minutes.
During the event, the plant was in Mode 5 (Refueling) with Reactor q
Recirculation Pump "A" running.
In addition, shutdown j
cooling was available from the Reactor Water Cleanup system
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and the Fuel Pool Cooling system.
The inspectors will followup the licensee's corrective actions in conjunction with review of the LER which will report this event.
No violations or deviations were identified.
9.
Maintenance Programs (62700, 62702)
During the inspection period, an NRC contractor conducted a preliminary review of the recently revised procedure CPS No. 1029.01 " Preparation
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and Routing of Maintenance Work Requests." Most of the inspector's findings were minor and should be resolved by a more extensive inspection of the program scheduled for a future team inspection. Among the significant findings, the inspector did not note any formal procedure for maintenance review of amendments to Technical Specifications to see if changes to maintenance surveillance, preventative maintenance, or corrective maintenance procedures were required to incorporate the amendments. The inspector also noted several cases where required signatures were missing on maintenance work request revision sheets CPS No. 1029.01F011.
Other maintenance related events and issues which occurred during this inspection period will also be followed up by the Maintenance Team Inspection scheduled to begin on March 13, 1989. Among the issues were the valid test failure of the Division 3 Diesel Generator on January 8,1989, and questions concerning the qualified lifetime of the air start motors on the Division 1 and 2 Diesel Generators.
No violations or deviations were identified.
10.
Environmental Qualification (93702)
a.
As documented in Inspection Report No. 50-461/87026, Paragraph 6.b.,
a previous NRC inspection identified that unqualified AMP KYNAR butt splices had been installed at Clinton Power Station.
On October 20, 1988, the NRC issued an Order Imposing Civil
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Monetary Penalty for violations of 10 CFR 50.49(f).
One example i
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of the violation was the identification of one hundred and ninety-six AMP KYNAR electrical butt splices installed in valve actuators, solenoid values and electrical junction boxes affecting multiple safety systems.
During this report period, the licensee identified three additional components where AMP KYNAR electrical butt splices were installed. As detailed in IP letter U-601335 dated December 21, 1988, during performance of preventive maintenance, the licensee identified two motor operated valves (1FC037 and IE12-F027A)
that, contrary to previous inspection efforts, had unqualified butt splices installed.
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As detailed in Licensee Event Report (LER) 88-032-00, the licensee's immediate corrective action was to rework the two valves and implement a sample inspection effort. The sample plan and current status was discussed with NRC (RIII) management on January 10, 1989, at a management meeting discussed below in
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Paragraph 12.
However, on January 18, 1989, the licensee informed the inspectors that a third valve (1E51-F013) was found to have an unqualified butt splice installed.
The licensee then stated that all devices originally inspected and reworked in response to the 1987 violation would be reinspected during the current refueling outage.
The adequacy of the licensee's corrective action to assure all unqualified butt splices have been identified and reworked will remain an Unresolved item (50-461/87030-02) to be reviewed by a Region III specialist.
b.
On November 3,1988, Limitorque Corporation. issued a 10 CFR 21 notification concerning common mode failures of white and gray colored Melamine torque switches in Limitorque actuators.
The notification reported evaluation of torque switch failures at Washington Public Power Supply System which determined that the Melamine torque switches were subject to long term post mold-shrinkage accelerated by high temperatures. The shrinkage could potentially affect the valve / actuator's ability to perform its safety function.
Limitorque recommended that all Melamine SMB-000 and SMB-00 torque switches be replaced with environmentally qualified Fiberite (brown) torque switches.
On November 10, 1988, Licensing and Safety Department issued letter Y-209692 assigning investigation of the issue to the Nuclear Station Engineering Department.
On December 8, 1988,
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Licensing and Safety Department issued letter Y-90145 detailing I
the investigation plan for the issue.
The plan called for replacement of all Melamine torque switches
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in safety-related applications by the end of the second refueling outage. The plan also included testing of all applicable Limitorque operators prior to the first refueling
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outage and replacement of those which would be involved in MOVATS signature analysis during the first refueling outage.
The licensee's response to this Part 21 notification will be followed up by a regional er.vironmental qualification specialist and is considered an Open Item (50-461/88030-03).
One unresolved item and one open item were identified.
11. Management Meeting (30702)
On January 10, 1989, NRC management met with IP management at the Clinton Power Station to discuss the status of the facility, the licensee's Monthly Performance Monitoring Management Report and planned activities to be performed during the licensee's first
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refueling outage.
Key personnel attending this meeting are identified by (#) in Paragraph 1 of this report.
The licensee discussed plant operations to date and summarized significant events.
Included in their discussion, the licensee summarized current status of their ongoing review of Feedwater Heating System Transient reported in LER 88-025-00. The licensee discussed current inspection efforts and planned inspections in response to identifying unqualified butt splices (subsequent to this meeting, the licensee issued LER 88-032-00 on this matter).
The licensee then discussed plant response to operating in single loop between December 18, 1988 and the start of their current refuel outage on January 3,1989.
The licensee presented their plans and current progress of their first refueling outage scheduled to last 69 days.
NRC (RIII) management acknowledged the licensee's status and plans.
12.
Violations For Which A " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requirement.
However, because the NRC wants to encourage and support licensees' initiatives for self-identification and correction of a
problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.G.I.
These tests are:
(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and (5) it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
Two violations of regulatory requirements identified during the inspection for which a Notice of Violation was not issued were discussed in Paragraphs 5.a and 5.b.
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13. Open Items
Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspectors, and which will involve some action on the part of the NRC or the licensee or both.
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l One open item disclosed during the inspection was discussed in l
Paragraph 10.b.
14. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations. One unresolved item disclosed during this inspection was discussed in Paragraph 10.a.
15.
Exit Meetings (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1)
throughout the inspection and at the conclusion of the inspection on January 30, 1989.
The inspectors summarized the scope and findings of the inspection activities. The licensee acknowledged the inspection findings. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any documents / processes as proprietary.
The inspectors attended exit meetings held between regional / headquarters based inspectors and the licensee as follows:
Inspector Date G. Pirtle 1/12/89 P. Rescheske 1/26/89 l
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