IR 05000461/1988004

Insp Rept 50-461/88-04 on 880217-0404.Violations Noted. Major Areas Inspected:Licensee Action on Previous Insp Findings,Nrc Compliance Bulletin Followup,Operational Safety Verification & Training Effectiveness

ML20153C689
Person / Time
Site:	Clinton
Issue date:	04/28/1988
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20153C678	List: IR 05000461/1988004 ML20153C674 ML20153C681
References
50-461-88-04, 50-461-88-4, IEB-85-003, IEB-85-3, IEB-87-002, IEB-87-2, NUDOCS 8805060282
Download: ML20153C689 (24)

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, . U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-461/88004(ORP)

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, IL Inspection Conducted: February 17 through April 4, 1988 Inspectors: P. Hiland S. Ray Approved By:

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R. C. Knop, Chief g

Reactor Projects Branch 3 Date Inspection Summary Inspection on February 17 through April 4,1988 (Report No. 50-461/88004(DRP))

Areas Inspected: Routine, unannounced safety inspection by the resident inspectors of licensee action on previous inspection findings; NRC compliance bulletin followup; onsite followup of written reports of nonroutine events at power reactor facilities; operational safety verification; monthly maintenance observation; monthly surveillance observation; training effectiveness; onsite followup of events at operating reactors; and management meetin Results: Of the eight areas inspected, three violations were identified in the area of Operational Safety Verification. The identified violations included: failure to take prompt corrective action to an identified instrument failure (paragraph 5.a.); failure to properly evaluate the impact of a support system when removed from service (paragraph 5.b); and failure to perform a safetr evaluation required by 10 CFR 50.59 when material was staged in containment directly over the suppression pool (paragraph 5.c.). In addition, one violation in the area of onsite followup of events was identified:

failure to comply with a Technical Specification ACTION statement (paragraph 9.b.(2)). All of the above violations are receiving licensee management attentio P 8805060292 880429 PDR ADOCK 05000461 Q DCD

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, DETAILS Personnel Contacted Illinois Power Company (IP)

1. W. Kelley, President

1. W. Gerstner, Executive Vice President

1. • D. Hall, Vice President - Nuclear

• K. Baker, Supervisor - I&E Interface

1. • R. Campbell, Manager - Quality Assurance

1. • J. Cook, Manager - Nuclear Planning and Support E. Corrigan, Director - Quality Engineering and Verification

1. • R. Freeman, Manager - Nuclear Station Engineering Department

1. K. Graf, Director - Operations Monitoring Program D. Holesinger, Assistant Manager - Clinton Power Station

1. • A. Mcdonald, Director - Nuclear Program Assessment

1. • J. Miller, Manager - Scheduling & Outage Management

1. • J. Perry, Manager - Nuclear Program Coordination

1. • F. Spangenberg, Manager - Licensing & Safety

1. • J. Weaver, Director - Licensing

1. • J. Wilson, Manager - Clinton Power Station

1. • R. Wyatt, Director - Nuclear Training Department Soyland/WIPCO

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1. • J. Greenwood, Manager - Power Supply Nuclear Regulatory Commission

1. aP. Hiland, Senior Resident Inspector, Clinton

1. • S. Ray, Resident Inspector, Clinton

1. R. Knop, Chief, Branch 3, Region III

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1. R. Cooper, Chief, Section 3B, Region III

1. Denotes those attending the management meeting on March 11, 198 * Denotes those attending the monthly exit meeting on April 4, 1988.

The inspectors also contacted and interviewed other licensee and contractor personne . Previously Identified Items (92701)(92702) (Closed) Open Items (461/86045-03, 461/86045-04, 461/86077-01, and 461/86077-03): Motor Operated Valve Concern These open items all dealt with inadequacies found during ,

inspections of the licensee's actions taken in response to IE Bulletin 85-03, "Motor Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings", and other concerns with motor operated valves. To reduce duplication, all the above concerns will be tracked under one open item for the response to the bulletin (461/85003-8B). These items are close . . .

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(Closed) Open Item (461/87030-02):

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. Valve Lineup Discrepancie This item was previously discussed in Inspection Report 50-461/87032, paragraph 2.h. and Inspection Report 50-461/87035, paragraph The inspectors reviewed the latest revisions to valve lineup procedures for all manual containment isolation valves listed on CPS Technical Specifica; ion Table 3.6.4-1 to verify that the procedures had been revised to reflect the desired lockea closed position of the valve No discrepancies were found. The inspectors also reviewed documentation provided by the licensee in the form of "

Engineering Change Notices, completed Maintenance Work Requests and dispositioned Condition Reports to show that all of the inspectors'

other concerns with valve lineup discrepancies as documented in the open item have been resolved. This item is close c. (Closed) Unresolved Item (461/86041-01): Delayed Response of Emergency Core Cooling Systems During a simultaneous LOOP and LOC This item addressed a concern identified by the inspectors during review of Hot Start test data for the Division 1 Diesel Generato The test results identified an additional delay in diesel generator start time if a start signal was received during coastdown of the diesel. As documented in Inspection Report 50-461/87041, paragraph 3.b., this issue was verbally resolved between the inspectors and the Office of Nuclear Reactor Regulations (NRR) at the conclusion of that inspection. However, the item remained open pending a formal response from NR NRR memorandum D. R. Muller to N. J. Chrissotimos, dated June 24, 1987, provided the formal response to the inspectors' question on the acceptability of diesel generator start time. As stated in that response, NRR concluded that the probability of a simultaneous LOOP and LOCA during a diesel generator coastdown from a test was less than 1 x 10 to the minus tenth power. Because of that low probability, NRR concluded that the postulated scenario was an incredible event and was not a safety issue. This item is close d. (Closed) Violation (461/87032-01): Six Examples Of Failure To Meet Short Periodicity Technical Specification Surveillance And LCO ACTION Requirement The licensee responded to this violation via IP letter U-601087 dated November 19, 1987, in a timely manner. In addition, the licensee reported each of the six examples in the violation as a Licensee Event Report (LER). The individual corrective actions on each of the LERs have been inspected and closed in previous inspection report The inspector reviewed the generic corrective action for this violation and found it to be adequate. The licensee ins +.alled a tracking board in the main control room to keep track of short periodicity Technical Specification item The licensee also instituted an LER Reduction Plan which included several elements including awareness training, newsletter articles, a status board,

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, . and personnel interviews. Since this violation was issued, the licensee has reported 19 LERs, only one of which was partially attributed to similar problems as discussec in the violatio The LER (88-005-00) was discussed in paragraph 9.b.(2) of this inspection report. Additional corrective actions have been taken for that LER and will be inspected in the future. Based on the completion of the corrective actions as stated in IP letter U-601087, this item is closea, (Closed) Unresolved Item (461/87031-03): Support Systems Not Being Adequately Reviewed For Plant Impact When Removed From Servic During this repcrt period, the inspectors identified a violation directly related to this item. The details of that violation are discussed below in paragraph This item is closed and correctivd action performed by the licensee will be tracked against the violation (461/88004-02).

No Violations or Deviations were identifie . Bulletins and Circulars (25026)

For the Bulletin discussed below, the inspectors verified that the licensee had received the Bulletin, had distributed the Bullatin to appropriate personnel, and had completed appropriate action (0 pen) NRC Compliance Bulletin No. 87-02 (461/87002-8B): Fastener Testing To Determine Conformance With Applicable Material Specification During this report period, the inspectors participated in the licensee's selection and packaging of five additional samples for retestin As documented in Inspection Report 50-461/87039, paragraph 3., five samples of the original inspection did not meet all test parameters for mechanical properties. Further investigation by the licensee determined thtt the testing laboratory had performed the test improperly. The licensee decided to select an additional five samples from the same stock codes and retest them at a different laboratory. This item will remain open pending the inspectors' review of the completed retest result No violations or deviations were identifie ; Onsite Followup of Written Reports Of Nonroutine Events At Power Reactor Facilities (92700)

For the LERs listed below, the inspectors performed an onsite followup inspection to determine whether response to the events were adequate and met regulatory requirements, license conditions and commitments, and to

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determine whether the licensee had taken corrective actions as stated in the LER (Closed) LER 87-019-00 (461/87019-LL): Automatic Actuation of

Safety ReHof Valve 1821-F041C Due To Failed Load Driver Circuit

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This event and the effects of load driver card failure in other circuits on ,ystem operational and plant safety were previously discussed in Inspection Reports 50-461/87011, caragraph 11.b.(13);

50-461/87013, paragraph 6; and 50-461/87015, paragraph 7.a.(9). At the conclusien of those inspections, this item remained open pending final review by NRR of the licensee's analysi NRR memorandum, D. R. Muller to C. E. Norelius, dated October 9, 1987, detailed the staff's safety evaluation of failed load driver circuit cards at Clinton Power Station. That evaluation concluded that while certain load driver failures can disable a system, all failures were bounded by previous Clinton safety analysis and in all cases, redundant or diverse means of protection were availabl Based on the completion of the staff's review as discussed above, this item is close (Closed) LER 87-025-00(461/87025-LL): Manual Actuation of the Reactor Protection System Due to a Fai'-d Feedwater Control Valv This event was documented in Inspet . Report 50-461/87015, paragraph 13.b.(10) and discussed i Inspection Report 50-461/87019, paragraph 6.c. The cause of the svent was failure of feedwater control valve IFWOO4 due to losa of hydraulic control pressur The loss of hydraulic pressure was caused by a ruptured pressure switch diaphragm which apparently failed due to fatique caused by excessive system cycling. The bspectors reviewed the Post Trip Review Report and completed Maintenance Work Request (KdR) C-48027 to verify that the cause of the velve failure had been properly investigated and repaired. Before this event, the licensee had noted sluggish or erratic response in the IFWOO4 valve and had requested a vendor representative to help them investigate the problem and tune the circuit. The vendor had not arrived before this failure and the plant was also restarted before he arrived. However, as discussed in the next paragraph, this failure was not related to the erratic response. The cause of this particular failure was corrected by tuning the hydraulic system crior to startup. This item is close (Ciosed) LER 87-029-00 (461/87029-LL): Automatic Actuation of the Reactor Protection System Due to a Failed Feedwater Regulation Valv This event was documented in Inspection Report 50-461/87020, paragraph 10.b.(1) and discussed in Inspection Report 50-461/87019, paragraph 6.c. The event was caused by a faulty solenoid valve and control circuit board for feedwater control valve 1FV004. These failures caused IFWOO4 to operate erratically. The operators

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attempted to manually control reactor water level but excessive cold feedwater additions caused reactor power to increase to the high i

flux trip setpoint of 40%. The normal high flux trip setpoint of 118% had been reduced in accordance with power ascension program procedures. The plant had experienced previous problems with the j control system for 1FW004 and a K4R for vendor assistance had been

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. . generated well before this event. The vendor had not arrived on site until af ter this trip. The inspectors reviewed the Post Trip Review Report and completed MWR C-30482 which documented the investigation and correction of the control problems with 1FWOO No discrepancies were noted. The inspectors also reviewed other corrective actions implemented in response to the critique of this event. The only critique problem that was safety significant was that the operators stated that, although they were not under management pressure, there was a certain amount cf self-imposed pressure to keep the plant on line to finish the power ascension test program. As a corrective action, management emphasized through weekly shift supervisor's meetings and night orders that operators were expected to shut the plant down when warranted by a safety concern and that they would have management support in such action This item is close (Closed) LER 87-048-00 (461/87048-LL): Violation Of The Plant's Technical Specifications Due To Utility Personnel Error Resulting From Miscommunications And Improper Planning Of Chemistry Samplin This event was documented in Inspection Report 50-401/87030, paragraph 11.b.(6). The event was very similar in cause and happened just before a series of Technical Specification violations which were documented as a violation (461/87032-01). The violation was closed in paragraph 2.d. of this report and the corrective actions to respond to the violation were adequate to correct the generic concerns of this event. The inspectors verified that the specific corrective actions for this event have been complete This item is close No violations or deviations were identifie . Operational Safety Verification (71707)

The inspectors observed control room operations, attended selected pre-shift briefings, reviewed applicable legs, and conducted discussions with control room operators during the inspection period. The inspectors verified the operability of selected emtrgency systems and verified tracking of LCOs. Routine tours of the auxiliary, fuel, containment, control, diesel generator, and turbine buildings and the screenhouse were conducted to observe plant equipment conditions including 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 implemente Inspections were routinely performed to ensure that the licensee conducts activities at the facility safely and in cunformance with regulatory requirement 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:

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Adequacy of plant staffing and supervisio Control room proinssionalism including procedure adherence, operator attentive.0ss and response to alarms, events, and off normal condition Operability of selected safety related systems including i attendant alarms, instrumentation, and controls, f

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Maintenance of quality records and report As documented in Inspection Report 50-461/88003, paragraph 6, the l inspectors had noted a decline in the professional conduct of plant staf l In response to this observation, licensee management reviewed control l room activities and established additional controls over the conduct of plant operations. Specifically, detailed guidance on the number of trainees and other workers in the control room was promulgated. During '

this report period, the inspectors observed strict adherence to that guidance with a resultant improvement in the professional conduct in the control roo On February 23, 1988, during the performance of surveillance procedure CPS No. 9000.010001, revision 28, "Control Room Surveillance Log", the licensee identified a 9% full scale delta between the Reactor Core Isolation Cooling (RCIC) Steam Line Flow instruments IE31-N683A and IE31-N6838. Upon discovery, control room operators requested Control & Instrumentation (C&I) to perform a calibratio Since the identified delta (9%) between RCIC steam line flow instruments was discovered during performance of a "Channel Check",

the licensee considered the results to be qualitative information and did not declare the instruments inoperable. The purpose of those instruments was to isolate the steam supply line on high steam flow which would be indicative of a RCIC steam line brea Additional instrumentation, such as high RCIC room temperature instruments were available to perform the same isolation signal functio Technical Specification Table 3.3.2-1, item 4.a. action statement 27, for the affected instruments required the system isolation valves to be closed within I hour and declare the affected system iroperabl However, as stated above, the licensee did not consider the qualitative information derived from the channel check to be I sufficient information to make a determination that the instruments were "inoperable". Had the licensee declared the instruments inoperable, the associated action stctement would require isolating and declaring the RCIC system inoperabl _ _

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The request for a calibration of the affected instruments was made on Day-Shift February 23. The requested calibration had not been performed when the Mid-Shift on February 24 removed the High Pressure Core Spray (HPCS) system from service for planned maintenance. At shift turnover from Mid-Shift to Day-Shift on February 24, the oncoming Shift Supervisor recognized and questioned why the requested calibration had not been performed on the RCIC instruments prior to taking HPCS out of service. Technical Specification 3.5.1. Action c required the RCIC system to be operable when HPCS was removed from servic The Day-Shif t Shif t Supervisor on February 24 directed that HPCS be returned to service and the requested calibration on the RCIC steam line flow instruments be performed. HPCS was returned to service at about 11:30 a.m. on February 24 and the calibration on RCIC steam line flow instruments was performed. The C&I calibration found instrument IE31-N6838 to be grossly out of calibration, apparently due to a loose Amphenal connectio The loose Amphenal was tightened and the subsequent as-found data was within the calibration acceptance criteria. However, as discussed below in paragraph 9.b.(5),

the same instrument failed again on February 2 CFR 50, Appendix B, Criteria XVI, requires in part that measures shall be established to assure that conditions adverse to quality, such as failures, are promptly identified and corrected. Failure of the licensee to promptly identify and correct the failure of instrument IE31-N683B prior to removing the HPCS system from service is a violation (461/88004-01).

The inspectors noted that clear direction was not provided to the plant staff to address the type of anomaly discovered during the channel check of the RCIC steam line flow instruments. In response to this observation, the licensee initiated Plant Manager Standing Order (PM50)-050, "Execution of CPS Technical Specifications". That PMSO provided direction to plant staff on what action to take and in what time frame when a channel check is identified invalid or has questionable value The inspectors considered the use of PMS0-050 to be a reasonable interim approach pending the licensee's response to the above violation and the corrective action plan required in that response, On February 25, 1988, the inspectors observed plant operators preparing to take out of service the Division II Essential Switchgear Heat Removal system (VX). Plant operators were in the process of tagging out the Division II VX compressor IVX06CB to perform Maintenance Work Request C-2806 As defined in the Clinton FSAR section 9.4.5, the VX system is an Engineered Safety Feature designed to maintain adequate switchgear heat removal under any normal or abnormal pla conditions. Technical Specification 1.27 provides the definition

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of OPERABLE-OPERABILDY; that definition clearly extends to the VX compresso The inspectors discussed existing plant conditions with the Shift Supervisor and the preparation for maintenance work on Division II VX compressor was stopped. Direction was given not to remove Division II VX from service until Division III ECCS (High Pressure Core Spray) was operabl Technical Specification 6.8.1.a. required that written procedures be established and implemented for performance of maintenanc Administrative Procedure CPS No. 1029.01, revision 21, dated December 21, 1987, required in paragraph 8.1.7. that a maintenance request be evaluated for impact on plant systems or component operability by the Shift Supervisor. Failure of the licensee to properly evaluate the simultaneous removal from service of the Division II VX compressor and the Division III ECCS(HPCS) could have resulted in plant operation outside the Limiting Condition for Operation and is a violation (461/88004-02).

The inspectors noted that the licensee's failure to recognize the impact of "support systems" has been a recurring problem. As documented in Inspection Report 50-461/87031, paragraph 6.a. and Inspection Report No. 50-461/87015, paragraph 10.b., the inspectors previously discussed with the licensee the technical specification definition of Operable-Operability and its meaning when support

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l systems are removed from servic Licensee action in response to those discussions was being tracked as Unresolved Item 461/87031-0 Since the items discussed above are directly related to Unresolved Item 461/87031-03, that item is closed and the resolution of the above violation will include the unresolved ite The inspectors noted that the licensee took corrective action when the inspectors initially identified the violation on February 25, 1988. However, the inspectors considered the prior management involvement to assure quality to be ineffective. The inspectors'

conclusion was based on the fact that this same concern had been discussed with licensee management in May of 1987, (Inspection Report 50-461/87015, paragraph 10.b.) and August of 1987, (Inspection Report 50-461/87031, paragraph 6.a.). On March 16, 1988, while performing a plant tour, the inspectors

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preparation for a plant outage that was to commence on March 1 At the time of the inspectors' observation the reactor plant was operating at 90*. power. The material staged included two 55 gallen drums, radiation protection signs, and two rolls of Herculite (approximately 1000 yards). In addition, Herculite had been taped on the access floor between the containment and drywell personnel access points.

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, , The suppression pool in a Mark III containment is open to the containment atmosphere and, as such, the installation or storage of material over the pool must consider the resultant displacement of that material following a Design Basis Accident (DBA). The Clinton FSAR section A3.8 discussed the effects of pool swell following the DB The inspector's immediate concern was the large amount of Herculite that had been placed directly over the pool and the two rolls of Herculite being stored directly over the poo Since the suppression pool was also the source of water for all ECCS pumps, the potential for suction strainer blockage from the large amounts of Herculite following a DBA existe The inspectors immediately discussed the storage / placement of material over the suppression pool with the Shift Supervisor. The Shift Supervisor stated that he was unaware of the condition and took corrective action by directing the removal of the material from the containment. The inspectors verified by direct field observation that the material staged in the containment had been removed shortly after discussing the issue with the Shift Superviso CFR 50.59 requires a written safety evaluation which provides the basis for a change to the facility as described in the safety analysis report. The staging / placement of Herculite in the containment directly over the open suppression pool was a change in the facility as described in the safety analysis report. This activity was accomplished through verbal instructions from radiation protection supervision without a safety evaluatio Failure to perform a written safety evaluation prior to the staging / placement of material in the containment that constituted a change to the facility as described in the safety analysis report is a violation of 10 CFR 50.59 (461/C8004-03).

The inspectors noted that material control in the containment had been a recurring item of concern. As documented in Inspection Report 50-461/87032, paragraph 5.a., material control in the containment was discussed with licensee management. At the time of that inspection, concerns identified by the inspectors were tracked as Open Item 461/87032-02. That Open Item was subsequently closed by the inspectors as documented in Inspection Report 50-461/87036, paragraph 2.d. based on actions taken by the licensee. However, based on the violation discussed above, the inspectors concluded that the actions taken by licensee management were' ineffective to control material in containment. In particular, it was evident to the inspectors from attendance at the critique concerning this event that plant staff per sonnel, including first and second level supervision, did not understand or appreciate the phenomena of pool swell and the need to control material in containment during plant operatio _ _ _ _ _ _ _ _

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, . Control Room Annunciators, Instruments and Recorders

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During this report period, the inspectors reviewed the licensee's plans and actions to date to repair out-of-service (005)/ disabled annunciators, instruments and recorders in the main control roo At the time of issuance of the Full Power License on April 17, 1987, ,

the licensee had 30 control room annunciators that were 005/ disable )

In addition, 34 control room instruments and recorders were identified '

as having reduced capability or other deficiencies that required some compensatory actio I l

Administrative Procedure CPS No. 1406.01, "Out of Service l Annunciator Tracking" delineated the controls in place to identify the compensatory measures required when an annunciator is no longer functioning properly or when it has been intentionally disable On March 17, 1988, (just prior to the licensee's spring outage) with the plant operating at 90% power, the inspectors noted the following conditions in the main control room:

Total Lighted Annunciators - 53 Total OOS/ Disabled Annunciators - 32 Total 00S Instruments - 3 Total Reduced Service Instruments and Recorders - 17 The inspectors compared the annunciators that were DOS / disabled at the time of issuance of the Full Power License to those that were 005/ disabled during this inspection period. All of the former annunciator problems had been corrected. Two of the former annunciators had become DOS again due to other problems. In addition, 31 of the 34 instrument and recorder problems had been correcte Problems with annunciators, instruments, and recorders had received a significant amount of management attention and the status of the problems had been tracked daily. However, since the issuance of the Full Power License, new problems have emerged about as fast as old ones have been corrected. The licensee intends to make a very significant reduction in the number of these problems in the current outag The inspectors will continue to closely monitor the licensee's performance in this are e. Motor Operated Valve Condition Reports During this report period, the inspectors reviewed two condition reports which were of particular interest because of their potential

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to impact the operability of safety-related motor operated valves.

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. - - Condition Report (CR) No. 1-88-02-057 identified a potential '

over/under torque of actuator-to-valve bolting on safetv-related valves. This CR was initiated on February 23, 1988, to resolve discrepancies identified on the applied torque values of about 110 motor operated valves. The inspectors discussed the actions being taken by the licensee to determine if the stated condition had any impact on the operability of the subject valves. The licensee

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performed calculations to identify any potential operability restraints. Following completion of those calculations, the licensee informed the inspectors that all motor operated valves were operable; however, a number of valves would require some rework during the present maintenance outage. The inspectors noted that a determination of no operational impact was made on March 2, 1988 by the licensee.

d Condition Report No. 1-88-03-048 initiated March 18, 1988,

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identified a potential concern with thrust values calculated during the performance of "MOVATS TESTING". This concern was raised by a

craft electrician and converted into a CR on the advice of his i supervision due to the perceived risk on the part of Engineering in I providing a verbal response to the concern without first researching and documenting their position fully. The inspectors noted that

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although the identified condition had a potential for generic implications, the CR had been identified as a Priority 4 (routine -

no adverse impact on plant operations). Verbal discussions with the licensee's Quality Engineering Director indicated that the prioritization had been based on a preliminary judgement that the stated condition was not significant and was probably an invalid CR.

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) The judgement was based on discussions that the Quality Engineering Director had with the initiator of the CR, his supervisor, the

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Compliance Group that had reviewed the CR for normal processing, i

i and Engineering. At the close of this report period, final disposition of this CR had not been accomplished. The inspectors i

will review the final disposition to assure regulatory compliance.

l Three violations were identified.

i Monthly Maintenance Observation (62703)

Selected portions of the plant maintenance activities on 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 l of the activities conformed to the Technical Specifications. The l inspection included activities associated with preventive or corrective l maintenance of electrical, instrumentation and control, mechanical

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equipment, and system The following items were consMared during these inspections: the limiting conditions for operation were met while components or systems were 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

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. . . service; parts and materials that were used were properly certified; and appropriate fire prevention, radiological, and housekeeping conditions were maintaine The inspectors observed / reviewed the following work activities:

Maintenance Work Procedure N Activity CPS No. 8207.02 Emergency Diesel Maintenance CPS No. 8502.04 Protective Relay Inspection Calibration And Functional Testing PMMDGM015 Check Engine Bolt Torque PEMGDA016 Test Division III Diesel Generator Protective Relays C-51550 Plant Modification AR-23 C-46599 Recirc Flow Control Valve Hydraulic Power Unit Repair C-20970 Inboard Main Steam Isolation Valve Repair C-46459 RPS Inverter 1A Trouble Alarm C-49632 1E12-F024B Repair The inspectors noted that the licensee had taken additional precautions while performing shop work on the Main Steam Isolation Valves (MSIV).

Specifically, the licensee had constructed a containment tent around the lathes used to repair the MSIV. In addition, the Radiological Controlled Area (RCA) boundaries in the mechanical maintenance shop had been improved since the Fall 1987 outage by the addition of a boundary fence to clearly delineate the RC No violations or deviations were identifie . 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 conformance with regulatory requirement Items which were considered during the inspection included whether: 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 within the required time limits. The inspectors determined that the test results were reviewed by persons other than those involved with the performance of the test, and that any deficiencies identified during the testing were reviewed and resolved by appropriate management personne ~

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, , , The inspectors observed / reviewed the following activities:

Surveillance / Test Procedure N Activity CPS No. 9281.01 Emergency Diesel Engine Inspection CPS No. 9952.01 Liquid PRM Sampling and Analysis CPS No. 9080.06' Division III Diesel Generator Operability No violations or deviations were identifie . Training and Qualification C/fectiveness (41400 & 41701)

On March 23, 1988, the inspectors participated in an emergency preparedness dress rehearsal drill in preparation for the annual emergency preparedness exercise that is scheduled for April 26, 198 The dress rehearsal was generally well organized and monitored. Several lessons were learned that could improve the licensee's performance in the upcoming exercise and in actual emergency response situations. The inspectors observed that the performance of the crew was acceptable, procedures were being used effectively, and the appropriate individuals clearly took charge of the situation. The scenario was quite difficult but believable. The inspectors commented to the licensee that the simulation of the intensity and volume of off site communications and information that would be requested by the NRC and other agencies was (nrealistic in that too little of such communication was simulat No violations or deviations were identifie . Onsite Followup of Events at Operating Reactors (93702) General The inspectors performed onsite followup activities for events which occurred during the inspection period. Followup inspections 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 the 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 corrective actions noted during the inspectors' followup are provided in paragraph b. belo . .

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. . . Details (1) Operability of Reactor Core Isolation Cooling (RCIC) System Containment Isolation Valves in Question On February 16, 1988, while in the process of revising CPS No. 9054.03, "RCIC Simulated Auto Actuation Test", an operator noted that four RCIC containment isolation valves had never been included in the procedure. When a search of preoperational test data failed to identify a test. demonstrating the ability of the four valves to close on an isolation signal, the Shift Supervisor declared the valves inoperable and ordered the RCIC exhaust vacuum breaker line isolated in accordance with CPS Technical Specification 3.6.4. On February 17, 1988, test data was located which indicated that the four valves had been tested within the required surveillance period and that they had been operable. Had the data not been found, the plant would have been in an ACTION statement of Technical Specification 3. which would have required that plant to be in HOT SHUTDOWN within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The Shift Supervisor thought he was in the ACTION statement for about a three minute period between the time he declared the valves inoperable and the penetration was isolated. The Shift Supervisor did not make the notification to the NRC Operations Center via the ENS in accordance with 10 CFR 50.73(b)(1)(A) because he believed the notification was not required unless an actual power reduction was starte Due to the short period of time the plant was in the ACTION statement, no actual power reduction took place. After discussions between the inspectors and NRC management and staff, it was determined that any entry into a Technical Specification ACTION requiring a plant shutdown should be reported in accordance with 10 CFR 50.72, even if the condition is rectified before an actual power reduction takes place. The license, agreed to adopt this interpretation and promulgated it to the operators via a night order and to all shift supervisors and assistant shift supervisors by a memo. The inspectors noted that this interpretation was adhered to during the event discussed in paragraphs 9.b.(4) and 9.b.(5) below. The licensee determined that since the four RCIC valves were later found to have been operable, the plant would not have had to enter the ACTION statement requiring a shutdown. Thus, the failure to makt the notification to the NRC was not a violation in this case. The inspectors found that the licensee's determination was reasonabl (2) Failure To Declare Drywell Particulate Radioactivity Monitor Inoperable On February 24, 1988, the inspector noted that the drywell atmosphere particulate radicactivity monitoring system (part of the reactor coolant system leakage detection system) was in an alarm condition. The alarm had been received on the evening of

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service for post maintenance testing. The inspectors noted that the annunciator response procedure for the alarm (CPS N .18, revision 21) directed the operator to CPS No. 4001.01, Reactor Coolant Leakage. Among the immediate operator actions of CPS No. 4001.01, revision 5, was an instruction to evacuate the containment if the particulate monitor alarm was vali The inspectors questioned operating shift personnel on the actions taken in response to the alarm. The inspectors determined that the monitor had been considered inoperable at the time of the alarm and had been inoperable since February 18, 1988. Daily grab samples were being taken as a compensatory measure in accordance with CPS technical specifications. The Shift Supervisors had reviewed the most current grab sample results and determined that the drywell radiation levels were normal and the repairs to the monitor were unsuccessful. The actions of the Shif t Supervisor were correct on this issue but the shift crew expressed a concern that it was difficult to interpret the grab sample results and compare them to the particulate monitor readings. The inspectors reviewed the results of grab samples taken on the drywell r atmosphere since February 16, 1988, and discussed the actions taken in response to the monitor failure with the Assistant Manager - Plant Radiation Protection. When it became apparent that several mistakes had been made, the inspectors requested the licensee to perform additional reviews of this even Among the discrepancies found by the inspectors and the licensee were the followin The monitor became inoperable on February 16, 1988, due ,

to radioactivity buildup on the monitor's filter paper '

because the filter paper was not advancin A grab sample taken at about 2:07 p.m. on February 16 indicated that the alarm was invalid but the monitor was not declared inoperabl February 17, 1988, Operations Department supervisory personnel were informed that the most likely cause of the alarm was that the filter advance mechanism wasn't working but the monitor wasn't declared inoperabl Later on February 17, 1988, the problem with the filter advancing mechanism was confirmed but the monitor still wasn't declared inoperable. As a result, no grab sample ;

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The monitor wasn't declared inoperable and daily grab samples were not ordered until directed by the Manager -

Clinton Power Station on the morning of February 18, 198 The next grab sample was taken at about 10:41 p.m. on February 1 !

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In addition to the missed sample on February 17, the sample taken on February 21, 1938, was not taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the previous one. The actual time was about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 12 minute ;

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The grab sample procedure was a 7000 level procedure (Radiation Protection Procedure) rather than a 9000 level procedure (Surveillance Procedure) and thus did not require the same tracking and review that a Technical Specification surveillance would hav Despite the fact that the grab samples were being taken as a result of an LCO ACTION statement, they were not being tracked on the status board discussed above in paragraph 2.d. of this repor Several data entry and independent verification Iroblems ;

were identified in the reviews of the grab sample result Many of the Radiation Protection personnel were not aware that the grab samples were being taken to detect RCS leakage. They were reviewing the grab sample results only in terms of maximum permissible radioisotope concentrations listed in 10 CFR 20 and not trending the results to identify increased leakag There were no procedures for converting the grab sample results to an RCS leakage rate or the equivalent indication that would have been expected on the drywell particulate monito [The inspectors noted that the licensee had discussed this particular item with the staff previously in 1987. The licensee had submitted a proposed amendment to the Clinton Technical Specifications on October 30, 1987. The Commission filed a "Notice of

'W Consideration of Issuance of Amendments to Facility .

Operating License and Opportunity for Prior Hearing" with the Office of the Federal Register on February 10,1988.]

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Calibration sources for the monitor were not availabl The range of operation of the drywell particulate monitor did not appear to be high enough to provide a useful indication as normal plant fission product inventory increased with core life. The licensee predicted that the normal indication of the monitor will be off scale in the futur '

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The drywell particulate monitor had been trending up toward the alarm setpoint for several days prior to the alarm received on February 16, 1988. The operating shifts had taken no action to determine the cause until the alarm actually occurred. This lack of positive action was similar to the problems which led to the violation discussed above in paragraph 5.a. of this inspection repor Because the problems with the drywell particulate monitor were not identified and repaired in a timely manner, the licensee entered a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hot shutdown ACTION statement discussed below in paragraph 9.b.(4) of this repor The licensee determined corrective actions for the issues discussed above and reported these events in LER 88-005-00 dated March 18, 1988. The inspectors will review the LER and make a determination of the adequacy of licensee corrective actions in a future inspection repor Failure to take and analyze daily grab samples of drywell particulate activity at least each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the drywell particulate radioactive monitoring system inoperable between approximately 2:00 p.m. on February 16, 1988 and 10:40 on February 18, 1988 is a violation of CPS Technical Specification 3.4.3.1 ACTION requirements. (461/88004-04),

(3) Power Transient [ ENS No. Information Only]

On February 22, 1988, the licensee experienced an unexpected transient due to an apparent control circuit failure in the

"flux-auto" control system for the Reactor Recire Flow Contro At the time of occurrence, the reactor plant was at 100% power and recire flow control was in flux-auto. Both the "A" and "B" recire flow control valves ramped open from about 73% to 78%

over a time span of 4 seconds (well below the maximum ramp rate of 11% per second). This increase in core flow resulted in a corresponding increase in core flux with APRM-D reaching 112%

and activating "GETARS" (transient test computer). APRM-A,-B, and -C reached flux values between 110% and 111%. The increased flux was able to demand a close signal to the recire flow control valves and the valves returned to their initial position of 73% ope The flux peak (spike) that was of sufficient

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amplitude to activate GETARS was not of sufficient duration to cause an RPS actuation. The duration of the flux spike was 1.6 seconds and since the flow biased trip /high flow clamped trip included a 6 second time delay, the RPS did not receive a trip signal. Also, the neutron flux high trip is set at 118%

and was not actuated due to the small amplitude of the spik In response to the transient, the licensee placed the rectre flow control system in "loop-manual" and does not plan to use

the "flux-auto" mode until completion of troubleshooting and

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corrective actio The flux transient had no appreciable affect on thermal power and the reactor plant remained at 100% power during the transient. The transient experienced was well below the analysis presented in Chapter 15 of the Clinton FSAR for this transien (4) Entering Action Statement Requiring Plant Shutdown

[ ENS No. 11596]

At 10:30 p.m. CST oa February 24, 1938, the licensee entered a Technical Specification ACTION statement requiring them to be in HOT SHUTDOWN within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The plant was operating at 100% power at the time of the event. The licensee corrected the condition and exited the ACTION statement at 11:05 p.m. CST on February 24, 1988, before an actual power reduction was started. The ACTION statement was inadvertently entered when the drywell sump flow monitoring system was made inoperable for a routine periodic surveillance at the same time that the drywell atmosphere particulate radioactivity monitoring system was inoperable due to equipment failure discussed in paragraph 9.b.(2) above. Technical Specification 3.4. requires a plant shutdown if both these parts of the reactor coolant system leakage detection system are inoperabl When the licensee realized the situation they were in, they immediately stopped the surveillance and returned the sump flow monitor to servic Inadvertent entry into the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown action statement was due te plant operations not recognizing the LCO restrictions for 2 of 3 leak detection systems being inoperabl (5) Entering Action Statement Requiring Plant Shutdown IENS No. 11611]

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At 10:07 a.m. CST on February 26, 1988, the licensee entered i

' a Technical Specification ACTION statement requiring them to be in 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 />. The plant was operating at 100% power at the time of the event. The licensee corrected the condition and exited the ACTION statement at 12:17 p.m. CST on February 26, 1988, before an actual power reduction was started. The ACTION statement was entered when a channel check on the reactor core isolation cooling (RCIC)

steam flow instruments indicated that the "B" detector was inoperable. At the time of the failed channel check, the high pressure core spray (HPCS) system was out of service for planned maintenance. With both RCIC and HPCS systems inoperable, plant Technical Specifications require the plant to be in 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 />. The operators planned an orderly shutdown to begin at 12:00 p.m. CST and at the same time they tried to repair the RCIC instrument and return HPCS to service. The licensee informed the NRC via the ENS of the situation at 10:44 a.m. CST. At 12:10 p.m. CST, the licensee informed the NRC that they had not begun a power reduction

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. . - because they thought that they could return HPCS to service within the next few minutes. At 12:33 p.m. CST, the licensee informed the NRC via the ENS that they had returned HPCS to service and exited the ACTION statement at 12:17 p.m. CST with no actual power reduction. The resident inspectors monitored the licensee's actions from the control room throughout the even As a result of the lessons learned by the licensee in the events discussed in paragraphs 5.a. and 9.b.(1) of this report, the crew's actions during this event appeared to have been smoothly execute (6) ESF Actuation [ ENS No. 11636]

At about 6:45 p.m. CST on February 29, 1988, the licensee experienced an unexpected ESF actuation when both trains of the Standby Gas Trectment (SGT) system auto started. At the time of event occurrence, the plant was in Mode 1 operating at 100% power. The auto initiation of SGT resulted from satisfying the 1-out-of-2 twice logic from the containment exhaust radiation monitors. With one monitor (1RIX-PR0010)

placed in a tripped condition for a required channel functional test, a second monitor (1RIX-PR001B) went into "low fail" thereby satisfying the 1-out-of-2 twice logic. The failure of the second channel was determined to be due to a crimped signal wire. Crimping of the signal wire occurred following routine maintenance activities. Physical location of the equipment resulted in difficult access and poor visibility of the signal wire. The licensee was reviewing a requested plant modification to ease access to this monitor. The licensee verified proper operation of the SGT trains and verified a high radiation condition did not exist prior to restoring the SGT trains to a standby mode. The licensee notified the NRC operations center of this event via the ENS at about 9:00 p m. CST on February 29, 1988. The licensee reported this event as LER 88-006-00 dated March 25, 198 (7) Operation Outside Design Basis As Described in FSAR

[ ENS Nos. 11768 and 11852]

At 12:50 p.m. CST on March 16, 1988, the licensee identified, during an engineering review of a maintenance request, that the station's Emergency Diesel Generator Ventilation (VD) system was apparently not capable of performing as stated in the FSA During the conduct of an annual fire protection surveillance test on March 14, 1988, the Division II VD fan failed to trip on actuation of the associated CO2 fire suppression syste The design intent as stated in the FSAR (Section 9.4.5.1.2.e.)

was to trip the diesel generator ventilation fan upon an actuation signal of the CO2 Fire Protection system. This would assure that the CO2 was not removed from the diesel generator

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[ ENS No. 11812]

On March 20, 1988, a local leak rate test (LLRT) on the "D" main steam isolation valves (MSIV) indicated excessive leakage past at least one of the two MSIV The leakage rate was higher than the LLRT rig could deliver. The NRC Operations Center was informed of the failure at about 8:30 p.m. CST on March 20. The licensee reperformed the LLRT with test pressure on both sides of the outboard MSIV and determined that the leakage across the inboard MSIV alone was higher than specifie The licensee decided to proceed with repairs to both the inboard and outboard MSIVs during the current outag MSIVs on the other three lines were tested and repaired during the fall 1987 outage as discussed in Inspection Report 50-461/87036, paragraph 11.b.(2). By April 1, 1988, the licensee had completed repairs on both the "D" line MSIVs and they passed LLRT _ _ _ _ .

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[ ENS No. N/A]

On March 24, 1988, after replacing the four air driven starter motors on the Division III Emergency Diesel Generator as part of routine preventative maintenance, the diesel failed to start during a post maintenance test when 3 of the 4 starter motors failed. The licensee then replaced all four starter motors again and attempted to run the diesel. That time the engine started but smoke was seen coming out of one of the air start motors. The licensee performed failure analysis on the starter motors and investigated stores records to determine the history of the motors. The failures appeared to be due to jelling of the lubricant in the motors and internal corrosion. All four of the failed motors had identical and unique histories. They had originally been built as left-hand drive motors and had been converted by the licensee to right-hand drive when they discovered that the Division III diesel used right-hand drive starter motors exclusively. The conversion was done in 1985, and the motors had been sitting on the shelf ever sinc During the conversion process, the licensee apparently added a lubricant which was incompatible with the lubricant already in the motor. Over the next 3 years a chemical reaction apparently took place between the lubricants. The licensee had no motors of this type left in stock. All other air start motors were ones that had been used and completely rebuil The rebuilding process included removing all old lubricant with solvent and using only one type of lubricant. The licensee had experienced no unusual problems with the rebuilt starter motor The starter motor vendor was Ingersol Rand. The inspector will review the final results of the licensee's investigation into this event to auure regulatory compliance. This will remain an open item pending further review (461/88004-05).

(10) Rod Scram Time [ ENS No.11794]

On March 18, 1988, during the conduct of scram time surveillance testing required by Technical Specification 3.1.3.2, the licensee identified a potential slow ro Control red 24-37 was initially thought to have not met the surveillance requirements of the plant Technical Specifications; however, upon further investigation, the licensee determined that the rod scram time was acceptable. The inspectors reviewed the control rod scram time charts obtained by the licensee and concluded that the licensee was correct in their evaluatio The licensee reported its entry into the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown ACTION statement via the ENS at about 1:30 p.m. CST on March 18, 1988. A followup notification via the ENS advising the Operations Center that the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown ACTION statement was exited was made by the licensee at about 2:30 p.m. CST on March 18, 198 __ __ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ .

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(11) Significant loss of Offsite Notification System [ ENS No. 11814]

On March 21, 1988, at about 8:04 a.m. CST, the licensee informed the NRC via the ENS that the Illinois Nuclear Accident Reporting System (NARS) telephone system was inoperable. The licensee had been informed by the State of Illinois about the inoperability at about 7:35 a.m. CST on the same day. Alternate

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communications capabilities were verified to be available via the normal telephone system. The NARS telephones were repaired by AT&T at about 10:49 p.m. CST on March 21, 198 One violation was identifie i 1 Special/ Management Meetings (30702)

On March 11, 1988, NRC management met with IP management at the Clinton Power Station to discuss the status of the facility, the licensee's

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Monthly Performance Monitoring Management Report and actions being taken

! to enhance the licensee's performance. Key personnel attending this meeting are identified by (#) in paragraph 1 of this report, i The licensee discussed plant operations to date and summarized significant events. The licensee presented the plant status and briefed NRC management on recent events. In addition, the licensee discussed their plans for the spring outage that commenced on March 1 The licensee discussed immediate corrective actions that had been initiated in response to the violations identified above in paragraphs and 5.b. Specifically, the licensee discussed the issuance of Plant

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Manager Standing Orders No. 047, "Technical Specification Operability

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Requirements" and No. 050, "Execution of CPS Technical Specifications".

NRC (Region III) management acknowledged the licensee's status and plans.

. The meeting concluded with a tentative agreement to meet again at the l Clinton Power Station with a similar agenda in May, 1988.

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11. Open Items i Open items are matters that have been discussed with the licensee, which will be reviewed further, and involve some action on the part of the NRC or licensee or both. An open item identified during the inspection is j discussed in Paragraph 9.b.(9).

) 12. Exit Meetings (30703)

The inspectors met with licensee representatives (denoted in paragraph 1)

) throughout the inspection and at the conclusion of the inspection on April 4, 1988. The inspectors summarized the scope and findings of the inspection activities. The licensee acknowledged the inspection findings, i

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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 an documents / processes as proprietar The inspectors attended exit meetings held between regional based inspectors and the licensee as follows:

Inspector Date W. Slawinski 2/26/88 W. Kropp 3/14/88

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