IR 05000461/1992022

Insp Rept 50-461/92-22 on 921208-930119.Non-cited Violation Noted.No Violations or Deviations Noted.Major Areas Inspected:Plant Operations,Radiological Controls,Maint & Surveillance & Engineering & Technical Support

ML20128B097
Person / Time
Site:	Clinton
Issue date:	01/26/1993
From:	Hague R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20128B078	List: IR 05000461/1992022 ML20128B074
References
50-461-92-22, NUDOCS 9302020397
Download: ML20128B097 (10)

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U.S. NUCLEAR REGULATORY COMMISSION REGION 111 Report No.

50-461/92022(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 8, 1992 - January 19, 1993 Inspectors:

P. G. Brochman F. L. Brus Approved By:

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_ichard L. Itag itef

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Reactor Projec ction 1C Jnsnection Summary

JAspfttion from December 8. 1992 throuah January 19. 1993. (Beport No.

50-461/92022(DRP))

Areas inspntmli Routine, unannounced safety inspection by the resident inspectors of plant operations, radiological controls, maintenance and surveillance, engineering and technical support, and safety assessment and quality verification.

Results: Of the five areas inspected, no violations or deviations were identified in four areas: one non cited violation was identified in the remaining area: (failure to translate design requirements into surveillance procedure acceptance criteria - paragraph 4.b).

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The following is a summary of the licensee's performance during this intpection period:

P1antJaqrjttions

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Position indication for all control rods was lost when a power supply breaker for the rod control and information system opened.

Operator response to this event was very conservative, llowever, operator familiarity with the alternate control rod position indication system was only adequate and indicated the need for additional periodic training.

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Control room operator's' policy on how to use the thermal overload bypass

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switches for motor operated valves was not consistent among the

operating crews. However, the various methods of operation were not in conflict with technical specifications.

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Chemistry personnel did not recognize that an alarm on the post _ accident-

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sampling system (PASS) required that the system be shutdown._ This

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condition existed for over 2 years.

The PA5S panel was not

safety-related and the ventilation system was always able'to exhaust air

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from the panel, even though the panel's rear covers were reinoved.

Maintenance and Surveillance j

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The licensee failed to include instrument inaccuracies when it developed acceptance criteria for the emergency core cooling system-(ECCS) pumps.

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surveillance procedures.. The licensee's saicty evaluation concluded I

that no degradation in pump performance had occurred and that the_ pumps had always been capable of delivering their design flow rates.

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The licensee identified that electrochemical pitting had contributed to the previous failures of the reactor feedwater pumps actuator bushings.

The licensee installed grounding straps to eliminate a voltage potential between the actuator linkages and the station ground.

S a f e t y A s s e s s mRQi_ tad _Ap3.lity_Vedf.ic a t i o n

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The licensee completed an evaluation of an Agastat relay that was

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Installed without a locking clip.

The lack of a locking clip

l invalidated the relay's scismic qualification.

l The licensee's self-assessment efforts remained effective.

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DETAILS 1. Persons Contacted 1111noit Power Company (IPl

• J. Perry, Senior Vice President

• J. Cook, Vite President and Manager of Clinton Power Station (CPS)

• J. Miller, Manager - Nuclear Station Engineering Department (NSED)

• R. Wyatt, Manager - Quality Assurance

• f. Spangenberg, Ill, Manager - Licensing and Safety

• R. Morgenstern, Manager Training

• J. Palchak, Manager - Huclear Planning and Support

• L. Everman, Director - Radiation Protection

• P. Yocum, Director - Plant Operations

• W. Clark, Director - Plant Maintenance R. Phares, Director - Licensing

• K. Moore, Director - Plant Technical

• W. Bousquet Director - Plant Support Services

• C. Elsasser, Director - Planning & Scheduling D. Kerneman, Director - Systems and Reliability, NSED

• R. Kerestes, Director - Nuclear Safety and Analysis

• J. Langley, Director - Design and Analysis, NSED

• J. Sipok, Supervisor - Regulatory Interface The inspectors also contacted and interviewed other licensee and contractor personnel during the course of this inspection.

. Denotes those present during the exit interview on January 19, 1993.

• 2.

Plant Operations

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The unit operated at power for the entire period.

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a.

Operational Safety (71707)

The inspectors observed control room operation, reviewed applicable legs, and conducted discussions with control room operators. During these discussions and observations, the operators were alert, cognizant of plant conditions, attentive to changes in those conditions, and took prompt accion when appropriate. The inspectors verified the operability of selected-emergency systems, reviewed tagout records, and verified the proper return to service of affected components.

Tours of the circulating water screen house and auxilbry, containment, control, diesel, fuel handling, rad-waste, and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, excessive vibrations, and to verify that maintenance requests had been initiated for equipment in need of maintenance.

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The inspectors observed plant housekeeping and cleanliness conditions and verified implementation of radiation protection controls and the physical security plan, b.

Containment Isplation Valve Circuit Brealer F@nd Of f At 8:06 a.m. on October 23, 1992, the main control room received an alarm that the circuit breaker for reactor water clean-up isolation valve 1033-f 001 was open.

The operator responding to the scene found the breaker in the off; vice, tripped position.

An electrician and quality assurance inspector were performing maintenance in the same switchgear, several cubicles away from the F001 breaker.

No other personnel were in the area and the individuals stated they had not heard anything.

The breaker was designed to open to the trip position when a fault occurred.

The licensee attempted to replicate this problem and could not.

There was no electrical logic between the f001 breaker and the one the electrician was working on. The manufacturer was contacted and could not identify any similar events.

There was no previous history of this oroblem at Clinton.

Based on this information, the licensee concluded that the cause of the breaker cpening was indeterminate. The inspectors have reviewed this information and have no further concerns. This issue is considered closed.

c.

Loss of Control Rod Position Indication At 3:30 a.m. on December 27, 1992, circuit breaker CB-1 on the rod control and information system (RCIS) tripped.

This caused a loss of all control rod position indication and the ability to move control rods; however, control rods could still be tripped.

The operating crew was unable to obtain reliable alternate position indication using the rod action control system ident generator.

The operating crew reviewed the technical specifications (TS) and concluded that they were beyond the applicable TS and entered TS 3.0.3.

By 6:52 a.m., the licensee had inspected the RCIS equipment for damage, and upon finding none, CB-1 was reclosed. A surveillance test was successfully completed and TS 3.0.3 was er i.

The licensee initiated efforts to replace the circuit breaker.

Before the documentation for that work was completed, CB-1 tripped open three additional times. Operations personnel did not enter any limiting conditions for operations (LCO) after the breaker tripped the second and third time, it was reclosed the first two times and left open the last time, to replace the breaker.

An LCO was entered after the fourth trip. The licensee stated this was a voluntary entry into the LCO to track the maintenance activity.

The break.er was successfully replaced and tested by 5:35 p.m.

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The licensee reviewed this event and determined that the breaker was original eculpment with no history of problems.

The breaker

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was shop testec and verified to trip at the correct values.

The circuit breaker was a molded case 120 Vac breaker.

The licensee decided further analysis of the breaker was not necessary.

The licensee subsequently concluded that the shift supervisor's

interpretation of TS was overly conservative and that no entry into TS 3.0.3 had occurred. The inspectors have reviewed the

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lictnsee's interpretation of TS and agree with it.

The 4nspectors concluded that the operating crews were not as familiar with the alternate control rod position equipment as they-i should be. Also, this. piece of equipment was not replicated in -

the simulator; thereby minimizing hands-on training.

The licensee

discussed this event with all of the operating crews and will

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provide additional training.

The inspectors have no additional concerns. This issue is considered closed.

d.

tio_ tor Operated Valve (M0_E.Jhprmal Over1nd Bvoass Switches During a routine control room walkdown, the inspectors noted that the MOV thermal overload bypass switches for the Division 1,-II, and 111 shutdown service water (SX) systems were all in the test position.

The test position enables the thermal overloads in the

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protective circuitry of safety related MOVs.

The overloads are normally bypassed so they do not prevent the MOVs from perfortning their function during an emergency.

However, during routinc operation of these valves, the licensee uses the test position to provide the additional protection for.the valve.

Technical Specification 3.8.4.2 required that if the thermal overload protection for an MOV was bypassed for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the valve should be declared inoperabic and the applicable-TS followed.

The operating crews indicated that the switches had not

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been in the test position for more than the TS limit.

However,

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reactor operators had not made log entries indicating that a.short term limiting condition for operation (LCO)-had been s.

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Licensee management stated it was acceptable to 31 ace all'three SX divisions in test at the same time, as'long as tie period was not greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

NRR agreed with the licensee's position.

Licensee management subsequently issued guidance to the. reactor o)erators on minimizing the use of the bypass switches and logging t1e short term entry into LCOs.

The inspectors have no further concerns in this area and considered the issue closed.

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Cold Weather Preparations (71714i The inspectors reviewed the licensee's cold weather preparation-

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performed in November 1992.

This included preventative

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e maintenance. procedures to drain ventilation cooling systems, operation procedures to check heat tracing, and performing a

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walkdown of various systems.

The inspectors have reviewed the documents and have no concerns in this area.

No violations or deviations were identified.

3.

Radioloalcal Contro11 East Accident Sampling System (PASS) Ventilation On December 3, 1992, the licensee identified that covers on the rear of the liquid PASS panel were not installed.

The design of the system required that the rear covers be installed-to allow the ventilation system to maintain a -0.25 inch WG (water gage) pressure inside of the panel. The PASS system did not perform any safety functions and the PASS panel itself was nonsafety-related.

This system implemented the

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licensee's commitment to NUREG-0737, item II.B.3.

The licensee's investigation determined that the rear covers had been removed in August 1990 to allow access to the back of the panel for modification activities.

However, thert ce no specific instructions to reinstall the doors; consequently, they remained removed until December 1992.

This problem was discovered during a licensee initiative to reduce the number of lit control room annunciators.

There was no safety significance to this event.

The ventilation system.

was operable and was taking a suction on the pass panel and no leaks had'

occurred.

However, annunciator 5231-2A, " Negative Cabinet Pressure Low" had boon, illuminated for 2 years. Clinton annunciator response procedure 5231.04, Operator Action 2, required that the PASS panel be shutdown per procedure 3222.12, " Pass Panel Operation" or 6005.01, " Post Accident Sampling" if sampling was in progress when the alarn, was received.

Also, there was no requirement in either of these procedures to ensure ventilation was operable before drawing the samples.

As corrective action the licensee reinstalled the covers on the PASS panel, discussed this_ event with chemistry personnel, and revised three procedures.

Further corrective actions were under development at the end of the report period.

The Clinton Updated Safety Analysis Report (USAR), Table 3.2-1, " Quality Assurance Requirements," Component XXXVII.12(a) stated that'the requirements of 10 CFR Part 50, Appendix b, did not apply to the PASS

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panel. Clinton USAR Section 9.3.7.3, stated that any gases vented within the PASS panel were removed by the auxiliary building or drywell purge system ventilation system; however,.no pressure differential was specified.

Based on this information, the inspectors concluded that:no violations or deviations had occurred.

However, the inspectors did discuss their concerns with chemistry department management on the procedure weaknesses and personnel response to annunciators'.

No violations or deviations were identified.

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4.

Maintenance and Surveillante (61720 & 621_Q1).

a.

Qbservations Of Work Actlyiliti Station maintenance and surveillance activities of both safety-related and nonsafety-related systems and components listed helow were observed or reviewed to ascertain that they were conducted in accoroance with approved procedures, regulatory guides, industry codes or standards, and in conformance with technical specifications.

Document

&ctivity 9431.12 APRM Channel Calibration b.

Instina of Emeroency Core Coolina System (ECCS1 Pump _i in September 1991, the licensee identified that the flow rates specified in Technical Specification (TS) 4.5.1.b for the ECCS pumps did not incorporate the inaccuracies associated with the flow measuring instruments.

General Electric Design Specification 22A3131AL, Revision 12, Section 4.1.3 required that the flow characteristics of the high pressure core spray (HPCS) system, considering flow measuring instrument inaccuracies, shall be equal to or greater than 5010 gpm with reactor vessel pressure 200 psi above suction pressure.

Technical Specification 4.5.1.b.3 required that the HPCS pump develop a flow of at least 5010 gpm with a differential pressure greater than or equal to 363 psid, when tested pursuant to TS 4.0.5.

Furthermore, the licensee's procedure for testing the HPCS pump also did not add any instrument correction factors. The licensee subsequently determined that all of the other ECCS pumps (low pressure core spray and low pressure coolant injection) had the same problem.

The licensee had documented the problem in condition report 1-91-09-028; however, the resident inspectors were not aware of this issue until November 1992. The licensee had evaluated the inservice testing (IST) data for all of the ECCS pumps and determined the actual flow rates - when adjusted for the flow instrument inaccuracy - did not show any degradation from the certified pump performance curves.

Based on this analysis, the licensee concluded that all of the ECCS pumps had been capable of performing their design requirements.

The inspector agreed with this conclusion. As corrective action, the licensee revised the HPCS and other ECCS pump surveillance procedures to increase the acceptance criterion to adjust for the instrument inaccuracy.

The inspectors questioned if this condition was reportable under 10 CFR 50.73(a)(2)(1)(B), since the actual tested pump flow rates were less than the values specified in TS 4.5.1.b.3.

Technical Specification 4.5.1.b specified that the pumps be tested pursuant to TS 4.0.5.

Since the actual instrument inaccuracy was within the allowed tolerance, the licensee concluded that pumps had been

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tested in accordance with the TS; and this condition was not reportable.

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While not disputing the licensee's position, the inspectors believe it is more appropriate to focus on whether design requirements were properly translated into surveillance procedures.

Part 50 of Title 10 of the Code of federal Regulations, Appendix D, Criterion 111 required that measures shall be established to assure that applicable design basis were r

correctly translated into procedures and instructions.

The failure to adequately translate-the ECCS pump design criteria into surveillance test procedures' acceptance criteria was a violation of Criterion III.

However, as the licensee identified this violation, it is not being cited because the criteria specified in Section VII.B.2 of the " General Statement of Policy and Procedures for NRC Enforcement Actions," (Enforcement Policy,10 CFR Part-2, Appendix C) were satisfied.

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c.

Reactor Feedwater Pumo (RFP) Failure Analysis The licensee initiated a failura analysis of the bronze bushings that were used on the RFP throttle linkages (see Inspection Report 461/92020, paragraph 5.b.).

Its investigation determined that electro-chemical pitting had occurred due to a voltage potential between the throttle linkages and station ground. The licensee installed-a grounding strap on both of the RFP linkages which eliminated the potential. The source of the potential could not

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be determined.

The licensee installed double tapered roller bearings as a temporary replacement for the bronze bushings, while-it completes a failure analysir.

The inspectors will review the iicensee's final analysis in a subsequent report.

No deviations were identified. One t.on-cited violation was identified.

5.

Enaineerina and Technical Support i-The_ inspectors received several regaests for information from Region Ill. The inspectors discussed the issues with.the licensee and forwarded the information to Region Ill. The licensee was very prompt-in-responding to these: questions and had done a good evaluation of_the_

issues.

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6.

Safety Assessment and Ouality Verification a.

Missina Lockina Strao Affects Seismic Oualification of Electrical Etl1Y On March 25, 1992, electrical maintenance personnel identified that the locking strap for relay 62S5 in switchgear IE22-S004 was missing. This problem was documented in condition report (CR)

1-92-03-067. The locking strap was required to maintain the seismic qualification of the 62S5 relay. The 62S5 relay's function was to protect the Division III emergency bus from a degraded voltage condition by isolating the bus from the offsite electrical grid, stripping loads-on the bus, and starting the Division 111 emergency diesel generator.

The licensee conducted an investigation and could not determine how long the strap had been missing. A review of records a

determined that no mention was made of the locking strap in either the; original installation instructions, maintenance work done on the switchgear, or surveillance tests on relay 62S5.

No evidence was found thit would indicate that the locking strap had ever been installed on the relay.

As corrective action, the licensee revised surveillance procedure 9333.03 to verify that the locking strap was reinstalled,_ after

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the relay was removed for testing.

The licensee verified that maintenance personnel were familiar with-the procedure.

The licensee inspected the Division III switchgear and no other Agastat relays were found with missing locking straps.

However, in reviewing this condition report, the inspectors did identify two questions.

First, there was no clear statement in the CR on whether this event was' reportable, given that the relay was not seismically mounted.

Second, while the corrective actions specifically addressed the 62S5 relay, they did not address any

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other Agastat relays.

Licensee management reviewed the reportehility of this event, after the safety evaluation was compirited and concluded that it was not reportable. However, that conclusion and its basis-were not documented in-the CR.

The licensee agreed that the corrective act4 as were too narrowly _

focused and has subsequently revised periodic maintenance training to discuss this event.

This issue has been passed on to the Electrical Distribution System Functional Inspection (EDSFI) team for further review, b.

Licensee Self-Assessment Capability (40500)

The inspectors attended a meeting of the Facility Review Group -

(FRG) [onsite review committee) on January 7, 1993. The FRG reviewed various procedure changes and condition reports. There was a free: flowing discussion between members when various issues

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were discussed.

The inspector reviewed the qualifications and expertise of the FRG members. The members were a highly experienced and diverse group.

Previous meeting minutes and a quality assurance surveillance report of FRG activities were reviewed.

The inspector concluded that the FRG continued to remain a highly effective oversight organization, whose focus was on the safe operation of the facility.

The FRG has kept a high degree of independence and was not dominated by licensee management.

The inspectors did not have any concerns.

No violations or deviations were identified.

fgn-Cited Violation

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The NRC uses the Notice of Violation to formally document failure to meet a legally binding requirement. However, because the NRC wants to encourage and support licensee's initiatives for self-identification and correction of problems, the NRC will not issue a Hotice of Violation if the requirements set forth in 10 CFR 2 Appendix C, are met. A violation of regulatory requirements identified during the inspection, for which a Notice of Violation will not be issued, is discussed in paragraph 4.b.

8.

fxit Interview The inspectors met with the licensee representatives denoted in paragraph 1 at the conclusion of the inspection on January 19, 1993.

The inspectors summarized the purpose and scope of the inspection and the 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 such documents or processes as proprietary.

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