IR 05000461/1999013

Insp Rept 50-461/99-13 on 990611-0728.Non-cited Violations Noted.Major Areas Inspected:Operations,Maintenance, Engineering & Plant Support

ML20211A902
Person / Time
Site:	Clinton
Issue date:	08/19/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20211A899	List: IR 05000461/1999013 ML20211A893
References
50-461-99-13, NUDOCS 9908240164
Download: ML20211A902 (22)

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U.S. NUCLEAR REGULATORY COMMISSION REGION 111

Docket No: 50-461 License No: NPF-62 Report No: 50-461/99013(DRP)

Licensee: lilinois Power Company I

Facility: Clinton Power Station i Location: Route 54 West )

- Clinton, IL 61727 i

Dates: June 11 - July 28,1999

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inspectors: T. W. Pruett, Senior Resident inspector K. K. Stoedter, Resident inspector C. E. Brown, Resident inspector j D. E. Zemel, Illinois Department of Nuclear Safety Approved by: Thomas J. Kozak, Chief Reactor Projects Branch 4 Division of Reactor Projects

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9908240164 990819 PDR ADOCK 05000461 G PDR l

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EXECUTIVE SUMMARY Clinton Power Station NRC Inspection Report 50-461/99013(DRP)

- This inspection included aspects of licensee operations, maintenance, engineering and plant support. The report covers a 7-week period of resident inspectio Ooerations

. The inspectors determined that operators did not provide sound technical justification before exiting four Technical Specification Action Statements associated with the unexpected opening of a turbine bypass valve (Section 01.1).

. Engineering and training personnel provided effective support to plant operators by providing timely evaluations and explanations for a degraded circulating water pump and a reactor recirculation system flow anomaly (Section 01.1).

... The inspectors concluded that operations personnel frequently referenced the Technical Specifications (TS) and independently assessed issues with the potential to impact TS requirements (Section 01.2).

.' The failure of maintenance personnel to deliver surveillance test results in a timely manner to operators resulted in an unnecessary TS Limiting Condition for Operation entry, and poor communications by operators led to the failure to initiate channel checks on a radiation moniter following its calibration. These examples were indicative of the need to further improve the control of work affecting adherence to TS requirements (Section O1.2).

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. The inspectors determined that the initial operability determination / evaluation conducted i after a high temperature alarm was received for the reactor core isolation cooling (RCIC)

system gland seal compressor, lacked detail in that it did not consider the ;mpact of increased steam leakage on standby gas treatment system operability, main control room dose, or the program for rninimizing primary coolant leakage sources outside containment. Following inspector prompting, engineering personnel completed an additional review of the potential operability impacts and revised the operability determination / evaluation (Section 01.3).

.- One Non-Cited Violation was identified for the failure to follow procedural requirements to post and/or barricade access to plant areas containing equipment considered necessary to maintain on-line safety at an acceptable level. Specifically, plant areas containing equipment associated with the high pressure core spray system were not posted to inform workers that work was not allowed on this equipment while the RCIC system was inoperable. A sample would most likely have been obtained from the Division ill emergency diesel generator (EDG) expansion tank while the RCIC system was inoperable had the inspectors not informed plant management that a technician was in the process of collecting samples from all three EDGs (Section O2.1).

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Maintenance

.: Three examples of a Non-Cited Violation were identified for electrical maintenance and controls and instrumentation personnel using the minor maintenance process to conduct work which required entry into a TS Limiting Condition for Operation or operational requirements manual required action and/or resulted in disabling the safety function of the component (Section M1.2).

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The licensee failed to promptly implement recommendations developed by the probabilistic risk assessment group to minimize plant risk during standby liquid control system surveillance testing even though the testing configuration and methodology placed the plant in an unacceptable risk condition and increased the containment failure frequency by a factor of 180 (Section M1.3).

. The inspectors concluded that the failure of work management personnel to recognize that several activities that were not part of the surveillance test or preventive maintenance task programs, such as RCIC system cold start testing, should have been ,

evaluated for potential operability impacts, was a weakness in the licensee's operability determination program (Section M3.1).

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. ' The inspectors identified one weakness in the licensee's program to control primary coolant leakage sources outside containment. Specifically, the licensee had not defined limits for maintaining leakage "as low as practicable" for those portions of systems outside containment that could contain highly radioactive fluids during a transient or accident (a Technical Specification requirement). As a minimum, operations and engineering personnel were not verifying continued compliance with 10 CFR Part 100 and GDC 19 dose limits following testing or the identification of new leakage sources (Section E1.1).

Plant Suocort

. Licensee controllers and evaluators provided an accurate assessment of activities during the June 16,1999, emergency preparedness training drill. The licensee's observations and post-drill critiques were effective in recognizing strengths, weaknesses, and areas for continued improvement (Section P1.1).

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u Report Details Summary of Plant Status The reactor was operated at 100 percent power for most of the inspection period. On

- June 5,1999, reactor power was lowered to 72 percent due to the unexpected opening of the number (No.) 1 turbine bypass valve. On June 24, reactor power was reduced to 97 percent in response to a degrading circulating water pump. The same day a power reduction of 8-10

. thermal megawatts was experienced due to a reactor recirculation system flow anomal = Following brief periods to address issues associated with the power reductions, the reactor was

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' returned to 100 percent powe . Operations O! Conduct of Operations  !

01.1 Review of Reactor Power Reduction Activities

' Insoection Scope (37551. 62707. and 71707)

The 16spectors reviewed the licensee's actions taken in response to three events which required or resulted in reactor power reduction . Observations and Findinas Opening of No.- 1 Turbine Bypass Valve At 4:06 p.m, on June 15,1999, the steam pressure regulating system unexpectedly ,

transferred control from the "A" pressure regulator to the "B" pressure regulator. The l l

transfer caused a pressure fluctuation and, as a result, the No.1 turbine bypass valve opened approximately 75 percent. The bypass valve quickly closed to approximately 8 percent open. Wdh the bypass valve remaining slightly open, steam was being ~

diverted from the feedwater heating system.' Therefore, operators entered Procedure 4005.01, " Loss of Feedwater Heating," which directed the operators to reduce reactor power to 97 percent. The diversion of steam made it difficult to ' measure first stage turbine pressure and this, in tum, made the instrumentation inoperable for the e control rod block signal, the turbine stop valve closure signal, the turbine control valve closure with oil pressure low signal, and the end-of-cycle recirculation pump trip signa Plant operators entered Technical Specification (TS) Action Statements for the

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inoperable instrumentation, the most limiting of which was a requirement for plant power to be reduced to 40 percent within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> The inspectors determined that operations personnel entered the appropriate TS Action -

Statements. Operators began reducing power to comply with the most limiting action statement and at 73 percent power, the bypass valve fully closed. Although the operators did not know the cause for the bypass valve opening and then not fully shutting, the TS Action Statements were exited when the bypass valve went fully shu As a compensatory measure, plant operators determined that they should keep the ;

plant at reduced power until the cause for the event was determine !

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The inspectors questioned the prudence of exiting the action statements prior to determining a cause for the event. The inspectors were informed that it was plant management's expectation that a rigorous technical justification be provided in cases such as this prior to exiting the TS Action Statements. The licensee subsequently determined that this event occurred due to deficiencies in the maintenance procedures used to calibrate the steam pressure regulating system and to adjust the turbine bypass

- valve bias setting. Following repairs to the pressure regulating and turbine bypass valve circuitry, operations personnel restored the reactor to 100 percent powe Degradation of "A" Circulating Water Pump

On June 24, operations personnel identified an abnormal noise coming from the "A" circulating water (CW) pump and that the pump's running amperage increased from 360 amps to 400 amps. Engineering personnel were requested to determine the source of the noise and training department personnel were requested to simulate removing the CW pump from service to determine the potential impact of this activity on plant operation Engineering personnel determined that the abnormal noise and running amperage

- increase were due to the dogradation of the C'N pump motor control circuitry. Based on this information, operations personnel reduced reector power to 97 percent and

~ removed the CW pump from service. The inspectors observed a portion of the downpower evolution and determined that the use of the simulator contributed to the effective planning for this evolution. The pump was repaired and retumed to service the following da Bistable Reactor Recirculation System Flow During the evening of June 24, operations personnel determined that reactor power had unexpectedly automatically decreased from 2894 megawatts thermal (MW) i to 2885 MW,. In response to this power decrease, the shift manager limited reactor power to 2885 MW, and requested an engineering evaluation to determine the cause of the ,

power reduction. Engineering personnel determined that the power reduction was caused by a bistable flow condition at the intersection of the jet pump header and the reactor recirculation (RR) pump discharge piping, a known phenomenon in boiling water reactors.' The inspectors reviewed the engineering evaluation and determined that the evaluation was Gorough, technically accurate, and considered all possibilitie Conclusions The inspectors determined that operators did not provide sound technicaljustification before exiting four TS Action Statements associated with the unexpected opening of a turbine bypass valv Engineering and training personnel provided effective support to plant operators by l providing timely evaluations and explanations for a degraded CW pump and an RR l system flow anomal )

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. 01.2 Use of Technical Soecifications Insoection Scooe (71707)

During a previous inspection period, the inspectors assessed' the use of TS by

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operations personnel._ The inspectors concluded that although TS usage had improved, operations personnel continued having difficulty recognizing conditions that impacted TS. As a result, the inspectors continued to assess the ability of operations personnel to appropriately 'use TSs during this inspection perio Observations and Findinas The inspectors observed that operations pemonnel frequently used the TSs and continued to independently assess activities affecting TS implementation. However, the two instances' described below indicate more rigor is needed when assessing activities

. affecting TS implementatio Calibration of Intake Radiation Monitor PR009D On June 10,1999, at 9:15 a.m., operations personnel removed main control room (MCR) air intake radiation monitor PR009D from service for a scheduled channel calibration.- When radiation monitor PR009D is operable, TSs require a channel check

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to be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. When the monitor was removed from service, radiation protection (RP) personnel, who normally conduct the required channel checks, were

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informed that the channel checks could be suspended. Approximately six hours later, the channel calibration was completed and the centrol room supervisor retumed radiation monitor PR009D to service without informing RP personnel. At approximately 8:00 a.m. the next day, RP personnel identified that they had not completed a channel check on monitor PR009D since the previous moming. Condition report (CR)

1-99-06-103 was written to document this issue. During a subsequent review, the

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licensee determined that a reactor operator had fortuitously completed a channel check on radiation monitor PR009D within the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during a separate, unrelated

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- channel check) was completed during the channel calibration when, in fact, it was not.'

The inspectors concluded that had the reactor operator not completed a channel check as part of a separate activity, the channel check TS requirement would most likely not have been me Reactor Pressure Channel Surveillance Testing

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At 10:15 a.m., on June 14, controls and instrumentation (C&l) personnel began

. conducting surveillance testing on two reactor pressure channels in accordance with Procedure 9534.02, " Anticipated Transient Without SCRAM [ATWS) Reactor Pressure B21-N401 A, B, E, and F Channel Functional." Technical Specification 3.3.4.2, "ATWS -

Recirculation Pump Trip instrumentation," allows operations personnel to delay entry )

into the associated action statement for up to six hours while this surveillance test is conducted. Prior to beginning the testing, the CRS instructed C&l personnel to hand deliver the test results to the MCR so that the results could be evaluated prior to the ,

expiration of the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowed by TSs. Although testing was completed within i 10 minutes, the results were not immediately delivered to the MCR for review. As the

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day progressed, the shift manager began calling C&l personnel hourly to determine

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when the test results would be delivered to the MCR. No definite time was given or

' demanded even though the six-hour surveillance clock was in effec At 4:21 p.m., the six-hour surveillance clock expired and operations personnel entered TS 3.3.4.2 Action Statement A.1 which required the licensee to restore inoperable channels to an operable condition within 14 days. Approximately 50 minutes later, the test results were delivered to the CRS for review, the results were determined to be

. satisfactory, and the equipment was retumed to service. The licensee determined that this event occurred due to the failure to meet expectations regarding the delivery of test results to the MCR.1 Specifically, the test results were placed in the regular plant mail

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instead of being hand delivered to the MCR as directed by the CR Conclusions The inspectors concluded that operations personnel frequently referenced the TSs and independently assessed issues with the potential to impact TS requirements. However, the failure of maintenance personnel to deliver surveillance test results in a timely manner to operators resulted in an unnecessary TS LCO entry, and poor communications by operators led to the failure to initiate channel checks on a radiation monitor following its calibration.' These examples were indicative of the need to further improve the control of work affecting adherence to TS requirement .3 Review of Operability Detenninations and Evaluations (OD/OE)

. Insoection Scope (37551. 71707)

The inspectors reviewed the OD/OEs listed below for technical adequac *. 1-99-05-191-OD/OE conducted to evaluate the design of reactor core isolation cooling (RCIC) system turbine lube oil cooling restricting orifice,

• 1-99-06-005-OD/OE conducted to evaluate erratic indications on average power range monitor 'A',

• 1-99-06-058-OD/OE conducted to evaluate coolant leakage past a low pressure core spray injection shutoff valve, and

• 1-99-06-138-OD/OE conducted to evaluate high temperatures on the RCIC system gland seal compresso I Observations and Findinas ,

The inspectors determined that three of the four OD/OEs reviewed were technically acceptable and documented an adequate basis to support continued operation of the respechve equipment. However, operability evaluation / determination 1-99-06-138-OD/OE, which documented the basis for continued operability of the RCIC system without the gland seal compressor in service, lacked sufficient detail to support continued operability of the RCIC system. Specifically, a high temperature alarm for the RCIC gland seal compressor occurred during RCIC system surveillance testin Operators stopped the surveillance test after receiving the alarm and engineering 7 j i

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personnel developed the OD/OE to evaluate RCIC system operability. With the gland seal compressor shutdown, steam leakage occurs in the RCIC pump room when the pump turbine is running. The OD/OE stated that the RCIC room temperature would increase due to steam leakage from the gland seal compressor; however, information regarding the magnitude of the temperature increase and the potential impact on i

equipment qualifications was not addressed The inspectors also identified that operations and engineering personnel had not considered the impact of increased steam leakage on MCR dose and standby gas treatment (VG) system operability, or the impact on their program to minimize primary coolant leakage sources outside containment.~ The inspectors questioned operations and engineering personnel to determine why these issues were not discussed in the OD/OE and whether the issues

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the inspectors that information regarding the magnitude of the RCIC room temperature

' increase was not included in the OD/OE, but was addressed in a document listed in the references. The inspectors reviewed this document and verified that the information

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was in the documen In response, engineering personnel completed an additional review and detemiined hat steam leakage from the RCIC gland seal air compressor did not affect VG system operability. In addition, the impact of increased steam leakage on MCR dose and on the program for minimizing primary coolant leakage sources outside containment was minimal. The OD/OE was subsequently revised to include this informatio c. ' Conclusions The inspectors determined that the initial OD/OE conducted after a high temperature alarm was received for the RCIC system gland seal compressor, lacked detail in that it did not consider the impact of incrassed steam leakage on VG system operability, MCR

. dose, or the program for minimizing primary coolant leakage sources outside containment. Following inspector prompting, engineering personnel completed an additional review of the potential operability impacts and revised the O . _02 Operational Status of Facilities and Equipment L 02.1 Control of Division lil Eouloment Durina RCIC System Outaae Insoection Scope (62707. 71707)

The inspectors assessed the licensee's implementation of Procedure 1151.12, "On Line Risk Assessment."

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~ Observations and Findings I On June 15,1999, the RCIC system was declared inoperable for scheduled maintenance.'. On June 16, at approximately 10:00 a.m., the inspectors observed a chemistry technician on top of the Division i emergency diesel generator (EDG)

i obtaining a glycol sample from the expansion tank. The inspectors questioned the technician to determine if operations personnel had been informed that sampling was in progress and if the sampling evolution included the Division lli EDG. The technician stated that operations personnel were aware that sampling would be conducted some ,

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time during the day, but that he had not called the work control supervisor or the MCR

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before obtaining the samples. Additionally, the technician stated that samples were to be taken on all three EDG The Division lli EDG supplies emergency power to the high pressure core spray (HPCS)

system. Procedure 1151.12 requires that plant spaces with HPCS system equipment be posted to inform workers that work was prohibited on this equipment while the RCIC system is inoperable. Since the RCIC system was inoperable, the inspectors questioned the CRS and the shift manager (SM) to determine if work on the Division 111 EDG was authorized and why the Division 111 spaces were not posted. In response to the inspectors' questioning, the CRS and SM stated that they were not aware that the chemistry technician was obtaining glycol samples on the EDGs. Additionally, the SM stated that he was not aware that posting of Division lli spaces was required for this conditio The inspectors discussed the issue with the Plant Manager and Director of Operation In response, the licensee initiated CRs 1-99-06-136 and 1-99-06-137. Additionally, the licensee determined that several actions designed to prevent work on Division Ill/HPCS equipment during the RCIC system outage had not occurred. Specifically:

- The EDG samples were not linked on the work schedule to completion of the RCIC system outage. Consequently, when the RCIC system outage was delayed, the EDG glycol samples were not postpone * During the daily review of the June 15 work schedule, the licensee did not identify the need to restrict completion of the EDG glycol samples with the RCIC system inoperable. The daily schedule, issued at 12:20 p.m., on June 16, included a note specifying that the Division lil EDG was not to be sampled with the RCIC system inoperable. However, the technician had already begun sampling the EDG expansion tank * Postings or barricades were not placed in the plant to restrict access to Division ill components and other equipment whose availability was important to minimize the impact of the RCIC system being inoperabl * Chemistry personnel did not notify on-shift operations personnel before taking the glycol samples on the EDG Technical Specification 5.4.1.s requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, dated February 1978. Section 1 of Appendix A to RG 1.33, recommended administrative procedures be implemented for equipment control. Procedure 1151.12, "On Line Risk Assessment," is an administrative procedure used for equipment control.- Section 8.4.12 of Procedure 1151.12 required, in part, that appropriate postings and/or barricades be used in the plant to clearly identify structures, systems, or components that are protected to maintain on-line safety at an acceptable level. The inspectors determined that the failure to post and/or barricade access to Division lli components while the RCIC system was inoperable was a violation of TS 5.4.1.a. However, this Severity Level IV violation is being treated as a Non-Cited ;

Violation, consistent with Appendix C of the NRC Enforcement Policy (NCV 50 461/99013-01). This violation is in the licensee's corrective action program as CR 1-99 06-136 and CR 1-99-06-13 i

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A i Conclusions One Non-Cited Violation was identified for the failure to follow procedural requirements to post and/or barricade access to plant areas containing equipment considered

, necessary to maintain on-line safety at an acceptable level. Specifically, plant areas containing equipment associated with the HPCS system were not posted to inform workers that work was not allowed on this equipment while the RCIC system was inoperable. A sample would most likely have been obtained from the Division lli emergency diesel generator (EDG) expansion tank while the RCIC system was inoperable had the inspectors not informed plant management that a technician was in the process of collecting samples from all three EDG Miscellaneous Operations issues (92700)

08.1 (Closed) Licensee Event Report 50-461/99-003 Manual reactor scram required to -

insert a control rod caused by a failed transponder circuit board. On January 18,1999,

- the plant was in Mode 4, the mode switch was in the refuel position, and operations personnel were conducting control rod venting activities. During the venting, a rod control and information system failure occurred with control rod 24-17 fully withdrawn from the core. This failure resulted in the inability to move control rod 24-17 by normal means. As a result, the SM directed operations personnel to insert a manual scram in order to retum control rod 24-17 to its full in position. Following the manual scram, control rod 24-17 retumed to position 00 and all other rods were verifed to be fully inserte The licensee determined that the inability to move control rod' 24-17 was crused by a failed transponder circuit board in the rod control and information system. The

. transponder circuit board was replaced and the rod control and information system was retumed to service. The inspectors determined that the licensee's actions to address this issue were adequat II. Maintenance M1- Conduct of Maintenance

' M1.1 . General Comments (61726. 62707)

The inspectors reviewed or observed portions of the following maintenance and surveillance activitie ~ Procedure 9015.01, Standby Liquid Control [SLC) System Operability" Procedure 9054.01,. "RCIC System Operability Check" AR F05033, " Troubleshoot Turbine Bypass Valve" PEMSYM004, . " Perform 10 Year Inspection of Switchyard Breaker 45-22"

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'PCISCM008 'and 9, "SLC Pump A and B Suction Pressure Indicator" Specific observations pertaining to these maintenance and surveillance activities are discussed in the sections belo . M1.2 Review of Minor Maintenance

' Inspechon Scope (62707)

The inspectors reviewed the minor maintenance procedures and selected activities completed as minor maintenanc Observations and Findinas Between June 1 and June 16,1999, the licensee completed approximately 70 quality related minor maintenance activities. The inspectors randomly selected 9 activities for

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review (six Fix-it-Now, two electrical maintenance, and one C&l). The inspectors identified concems with each of the minor maintenance activities involving electrical maintenance (I-M) and C&l personnel. No discrepancies were identified with minor maintenance activities completed using the Fix-It-Now proces Technical Specification 5.4.1.a requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in RG 1.33, Revision 2, Appendix A, dated February 1978. Section 9 of Appendix A to RG 1.33, recommended procedures be implemented for conducting maintenanc Procedure' 1029.01, " Action Requests and Maintenance Work Orders," is a procedure

' used for conducting maintenance. Section 3.1 of Appendix C, " Implementing Minor Maintenance," of Procedure 1029.01, specified that during minor maintenance the safety-related function of any component or system will not be affected and entry into a

. TS LCO or Operational Requirement Manual (ORM) required action will not be caused or required. Section 4.2 specified that quality control personnel would review work

. Instructions for quality-related components. In the following three instances, although

an ORM or TS LCO was entered, the work was performed as minor mai'itenanc On June 6, C8il personnel completed work on safety-related hydrogen analyzer 1NGG-N012A per Action Request (AR) F05014. The work involved entry into ORM 2.2.11; however, C&l personnel completed the task as minor maintenanc * On June 9, EM personnel completed work on safety-related control room ventilation damper.0F2VC014 per AR F06243. The work involved entry into TS LCO 3.7.3;- however, EM personnel completed the task as minor maintenanc *- On June 10, EM personnel completed work on safety-related control room

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' ventilation damper OVC11YA per AR F06213. This work involved entry into TS LCO 3.7.3; however, EM personnel completed the task as minor maintenanc During a subsequent review of the AR by the EM supervisor, the licensee determined that the work instructions did not receive the required quality control revie The inspectors determined that the failure to ensure that a TS entry was not required before conducting minor maintenance, and that the safety function of a component was 11 n. .

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j ln not disabled as part of minor maintenance; in these three instances, were a violation of TS 5.4.1.a. However, this Severity Levei IV violation is being treated as a Non-Cited

Violation, consistent with Appendix C of the NRC Enforcement Policy

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(NCV 50-461/99013 02). This violation is in the licensee's corrective action program as CR 1-99-06-14 The inspectors reviewed CR 1-99-02-446 wliich described the removal of a power supply fuse for the P630 panel as part of a minor maintenance activity. When the power supply fuse was pulled, several annunciators in the MCR unexpectedly alarme Several issues associated with this work activity were identified, including a lack of sufficient detail in the work package, the need for an engineering work request to enable maintenance personnel to determine the functionality of power supplies on the P630

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panel, and the need to conduct a self-assessment on maintenance work-around practices. The inspectors considered the licensee's review of CR 1-99-02-446 appropriate and the associated corrective actions reasonabl I The inspectors reviewed CR.1-99-02-284 which described the installation of an incorrect power supply associated with the turbine runback signal. The licensee determined that the part number in the computer data base incorrectly related to a 124 VAC power supply instead of a 24 VDC power supply for the runback signal. Consequently, during the preventive maintenance (PM) activity to replace the power supply in December

.1998, C&l personnel installed the incorrect power supply. The licensee determined that

. the CR database did not contain any other examples of incorrectly specified power supplies and revised the PM tasks to reflect the correct part numbers. However, the inspectors determined that the CR did not address why the technicians replacing .the power supplies did not notice that the power supplies were different. In response to this issue, the Maintenance Mana9er stated that in reviewing CR 1-99-02-284. the licensee

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should have determined why the technicians did not recognize that the parts were not like for like. As a result, the Maintenance Manager committed to supplement CR 1-99-02-284 with additional information regarding why the technicians replacing the power supplies did not notice the power supplies were different. The supplement, which included the additional information, was provided to the inspectors and the CR group on July 9,199 I Conclusions

! Three examples of a Non-Cited Violation were identified for EM and C&l personnel !

using the minor maintenance process to conduct work which required entry into a TS LCO or ORM required action and/or resulted in disabling the safety function of the componenti M1.3 ' Review of Standbv Liould Control (SLC) System Surveillance Testina

. Inspection Scoon (61726)

' The inspectors observed preparations for and the completion of SLC system testing on l July 13,199 o

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. Observations and Findings As part of the preparations for SLC system testing, operations personnel developed a contingency plan in accordance with Procedure 1151.12, "On Line Risk Assessment."

In developing the contingency plan, the licensee considered a letter dated May 27,1999, from the probabilistic risk assessment (PRA) group stating that the method for conducting SLC system surveillance testing isolated both SLC system trains, placed the plant in risk condition red (unacceptable risk), and increased the containment failure frequency by a factor of 180. To address these concems, the PRA group recommended that Procedure 9015.01, "SLC System Operability," be revised to ah a

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the surveillance tests referenced in the procedure to be completed as one activity rather than three, which wouH minimize plant ris In a letter dated July 8,- 1999,'(also attached to the contingency plan), the PRA group again stated that Procedure 9015.01 should be revised. However, this letter also stated that adequate time was not available to revise the procedure since the SLC system surveillances were required to be completed by July 20,1999. The inspectors

. questioned operations services personnel to determine what actions were taken after the PRA group issued the May 27 letter. The inspectors were informed that, as of July 8,1999, no action had been taken. This was of concem to the inspectors in that

. the May 27 letter was received by multiple management member To complete the SLC system surveillances by July 20, the PRA group agreed that the SLC system surveillances could be completed as three activities as long as operations '

personnel: 1) walked through all surveillance test and restoration steps prior to actually

. performing each surveillance test,2) ensured that all equipment credited in the PRA

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was in an available status, and 3) limited the total time for completing all three activities to less than or equal to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Procedure 1151.12 was also revised to allow activities previously classified as " risk condition red" to be considered " risk cond!* ion orange" as long as a detailed PRA ana'ysis was completed to support this char,ge in risk classification. The inspectors observed portions of the SLC system surssillance tests and determined that operations personnel adhered to the limitations invo' ed by the PRA grou On July 15, the inspectors questioned the SLC system manager to determine why both SLC system trains were required to be isolated during surveillance testing. The system manager was unable to provide the inspectors with a technical explanation for this condition. The inspectors considered the inability of the system manager to provide an explanation for isolating both SLC system trains during surveillance testing to be a weakness in system specific knowledge. Subsequently, operations personnelinformed the inspectors that the original SLC system design had not included cross tie valves to allow one train to be isolated at a time. As a result, operations personnel isolated both

. SLC system trains to prevent the migration of sodium pentaborate into other portions of the system and to minimize the potential for sodium pentaborate hardening in SLC system pipin At the conclusion of the inspection, operations and engineering personnel were determining if the SLC system surveillances could be completed without isolating both trains. In addition, operations personnel were revising Procedure 9015.01 to al;ow the SLC system surveillances to be completed as one activity. Operations personnel expected to complete the revision to Procedure 9015.01 by July 28,199 :

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. The licensee failed to promptly implement recommendations developed by the PRA group to minimize plant risk during SLC system surveillance testing even though the

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_ testing configuration and methodology placed the plant in an unacceptable risk condition and increased the containment failure frequency by a factor of 18 M3 Maintenance Procedures and Documentation M3.1 Review of RCIC System Cold Quick Start Testina Schedule Inspection Scooe (61726)

l The inspectors reviewed the licensee's decision to delay the completion of RCIC system cold quick start testin Observations and Findings in June 1999, the inspectors planned to observe RCIC system cold quick start testing as q part of the licensee's startup activities. Due to the extended staitup, the licensee deferred the test to the next RCIC system outage window. Following the RCIC system outage, the licensee experienced problems with the RCIC system gland seal compressor and the cold quick start test was rescheduled for Fall 199 In mid-June, the inspect >rs questioned work management personnel to determine why it was acceptable to keep rescheduling the RCIC system cold quick start test. This test had not been completed in more than 2.5 years and the impact of delaying the test on the RCIC system had not been evaluated. Work management personnelinformed the inspectors that cold quick start testing was required to be completed once per operating cycle; therefore, the test could be conducted anytime during the next 18 months regardless of when the test was last complete ' During a subsequent review, the inspectors determined that other RCIC system testing had been completed which tested similar functions as those tested during the cold quick start test. However, the inspectors identified a deficiency in the licensee's testing and PM programs. The licensee's testing and PM programs required that the deferral of any safety-related test or PM task be evaluated for potential operability impacts. However,

. the RCIC system cold quick start test was not considered a surveillance or a PM tas As a result, the decision to defer the test and an evaluation of the potential impacts was not require The inspectors discussed this issue with work management personnel. After an additional review, work management personnel agreed with the inspectors' assessment and identified several other activities which were not part of the testing or PM programs but should be evaluated if deferred. As a result, work management personnel were developing a desk top guide for activities that were not part of the testing or PM programs to ensure that any deferral of these activities was appropriately evaluate l l

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. Conclusions The inspectors concluded that the failure of work management personnel to recognize that several activities that were not part of the surveillance test or PM task programs, such as RCIC system cold start testing, should have been evaluated for potential

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operability impacts, was a weakness in the licensee's operability determination progra M8: Miscellaneous Maintenance issues (92700)

t-M8.1 (Closed) Licensee Event Report 50-461/99-008. Failure of the motor-driven reactor feedwater pump regulating valve results in a level transient and the insertion of a manual scram. This issue was discussed in Section 01.15 of Inspection

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Report 50-461/99010. No new issues were identified during the inspectors' review of this event during this inspection perio Efatpineering

'E1 LConduct of Engineering

= E Bayi!tEcf Primary Coolant Sources Outside Containment Proaram Insaadion Scoce (37551. 71707)'

The inspectors reviewed Procedure 1019.07, " Leakage Reduction and Monitoring Program," to determine whether the procedure adequately implemented the -

_ requirements of TS 5.5.2, " Primary Coo! ant Leakage Sources Outside Containment." Observations and Findinas

' Technical Specification 5.5.2 requires each licensee to ' maintain leakage as low as

' practicable for those portions of systems outside containment that could contain highly

, ' radioactive fluids during a serious transient or accident. Procedure 1019.07 stated that a leakage was maintained by conducting periodic PM tasks and visual inspections, and by completing integrated leak tests once per refueling cycle or sooner. -

The inspectors determined that the licensee was completing the required testing, maintenance, and inspection activities. However, what constituted "as low as practicable" was not well defined or understood. Since the licensee had not quantified

"as low as practicable," the inspectors asked the licensee if compliance with the dose l limits specified in 10 CFR Part 100, " Reactor Site Criteria," and General Design Criteria j (GDC).19, ? Control Room," was determined after conducting maintenance or testing j

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activities or upon identifying new leaks. ;The licensee stated that, while a calculation to _ !

verify continued compliance with 10 CFR Part 100 and GDC 19 was not completed, the )

intent of the TS was met as long as system leakage was low. At the conclusion of the inspection period, the licensee had completed a calculation which verified compliance with Part 100 and GDC 19. The inspectors reviewed the calculation and had no concern '

s 1 & . Conclusions -

The inspectors' identified one weakness in the licensee's program to control primary .-

coolant leakage sources.outside containment. Specifically, the licensee had not defined limits for maintaining leakage "as low as practicable" for those portions of systems outside containment that could contain highly radioactive fluids during a transient or accident (a Technical Specification requirement). As a minimum, operations and engineering personnel were not verifying continued compliance with 10 CFR Part 100 '

and GDC 19 dose limits following testing or the identification of new leakage source E1.2 Review of Reactor Recirculation (RR) Pumo Seal Anomalies '

Insoection Scope (37551)

The inspectors reviewed three RR pump seal anomalies that occurred between June 26 and July 2,199 Observations and Findinat i

On June 26, operations personnel identified that outer seal cavity pressure on the "B" RR pump decreased approximately 75 pounds and seal cavity temperature dropped 10 degrees Fahrenheit ('F). After three hours, the seal cavity parameters retumed to

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. normal. Four days later, the "B" RR pump outer seal cavity pressure and temperature

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dropped by 10 pounds and 10'F, respectively, for several minutes. On July 2, another

"B" RR pump seal anomaly occurred. During this event, seal cavity pressure decreased by 3 pounds and temperature dropped approximately 4' The inspectors discussed the behavior of the "B" RR pump seal with the system manager. The system manager explained that he was unsure what had caused the anomalies, but it appeared that the faces of the "B" RR pump seal were momentarily coming apart which caused a decrease in pressure and temperatur . The inspectors reviewed Procedure 3302.01, " Reactor Recirculation," to determine if guidance was provided to operations personnel regarding reductions in seal cavity pressure and temperature. The inspectors determined that no guidance was provided on this type of anomaly. The inspectors questioned engineering personnel to determine if Procedure 3302.01 needed to be revised to include the appropriate guidance. Initially, engineering personnel stated that Procedure 3302.01 was acceptable as written but that

. a review of the information gathered during the three anomalies would be completed and that the procedure would be revised as needed

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The licensee appeared to have identified the cause for three RR system flow anomalie E8 > Miscellaneous Engineering issues (92703) '

E (Closed) Unresolved item 50-461/99003-07: Review of 10 CFR 50.59 Safety Evaluation 98-066.. As discussed in NRC Inspection Report 50-461/99003, the NRC inspection team reviewed 10 CFR 50.59 safety evaluation 98-066, Revision 0," Updated Safety i Analysis Report (USAR) Change to Substitute the Automatic Depressurization System l l

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. [ ADS) for RCIC." Following that review, the team questioned the licensee's classification of the feedwater line break outside containment as a special case of a

loss-of-coolant-accident (LOCA), and the licensee's determination that the planned revisions to the USAR to resolve contradictory information regarding the systems designed to mitigate this event, was not an unreviewed safety question. As a result, Task interface Agreement 99-004 was forwarded to the Office of Nuclear Reactor Regulation for technical review.-

During this inspection, the technical staff in the Office of Nuclear Reactor Regulation completed their review and concluded that the change to the Clinton USAR to substitute ADS for RCIC to mdigate a postulated feedwater line break outside containment was not an unreviewed safety question since the systems designed to mitigate the feedwater

~ line break outside containment are HPCS and ADS in conjunction with the low pressure r emergency core cooling systems. The NRR staff further concluded that the classification of the feedwater line break outside containment as a special case of a LOCA was appropriate, as long as the consequences were bounded by another LOCA analysis.- The inspectors considered this item resolve E8.2 (Closed) Violation 50461/99011-02: Failure to assure design basis information for the hydrogen mixing compressors was transisted into a safety-related calculation. The inspectors identified that an incorrect horsepower rating for hydrogen mixing

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compressors was used in a safety-related calculation conducted to evaluated diesel generator loads. This issue was brought to the attention of engineering management and a new calculation was performed. However, while the calculation was revised to address other concems, the horsepower value was not corrected in the revised '

calculation. The licensee determined that several engineers were aware that calculation

.19-AK-05, " Calculation for D;esel Generator Load Monitoring," required a revision to include updated information regarding the horsepower rating for the hydrogen mixing -

compressors. However, this violation occurred because engineering personnel did not document the need for a revision in the corrective action plan for CR 1-97-07-105 or as part of Nuclear Station Engineering Department (NSED) Procedure E.1, " Calculations."

When this violation was ' initially identified, the documentation of required calculation revisions via a CR corrective action plan or as pad of NSED Procedure E.1 was not required. However, NSED Procedure E.1 now requires that all calculation revisions be documented on a calculation impact assessment form. Other corrective actions for this issue included revising calculation 19-AK-05 to include the correct horsepower rating and conducting an additional review to identify any other calculations impacted by the change in horsepower rating. The inspectors reviewed calculation 19-AK-05 and verified that the horsepower rating was correct. The inspectors considered the licensee's other corrective actions to be adequat .

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IV. Plant Support P1 Conduct of EP Activities P Emeroency Preparedness Trainina Drill Observations Insoection Scooe (71750)

On June 16,1999, the inspectors observed a site-wide emergency preparedness training drill from the technical support center (TSC), operations support center (OSC),

and the emergency operations facility (EOF). The simulator was not used as part of this drill due to hardware problems with the simulator computer, Observations and Findinas Technical Support Center Observations by licensee controllers and evaluators in the TSC included late emergency response organization pager activation, slow activation of the TSC, confusion regarding the information to be placed on the " major problems" board, lack of communications within the TSC and to other emergency facilities, conflicting information on status boards, the ineffective use of urgent teams to evaluate equipment problems, and the

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communication of inconsistent dose informatio Positive observations by the licensee controllers and evaluators in the TSC included the timely initiation of required emergency notifications and protective action recommendations and a thorough tumover between station emergency directors. The inspectors observations were consistent with the licensee's observation Operations Support Center Observations by licensee controllers and evaluators in the OSC included a lack of timeliness in reporting to the OSC, confusing prioritization of repair teams early in the scenario, the use of inconsistent plant terminology, and the inefficient dispatch of emergency teams on some occasions. No other concems were identified by the inspector Positive observations by licensee controllers and evaluators in the OSC included good role playing by RP personnel, the limitation on the number of priority activities late in the scenario, and the control of noise levels in the OSC. The inspectors observations were consistent with the licensee's observation Emergency Operations Facility l

The EOF controllers and evaluators observed that EOF team members were not proactive in questioning data provided on status boards or communicating to other EOF team members. Specific coordination problems were identified between the dose assessment supervisor, the dose assessor, and the radiological controls coordinator. In addition, EOF controllers and evaluators did not observe a sense of urgency to resolve specific issues by the EOF team members in that issues such as containment status,

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radioactive release status, and the restoration of offsite power were not aggressively pursue Positive observations by the EOF evaluators and controllers involved the response to an inquiry regarding estimated core damage and the completion of a detailed critique. The t inspectors observations were consistent with the licensee's observation Controllers and Evaluators  ;

The TSC, OSC, and EOF controllers and evaluators recorded detailed field observations of activities as the emergency preparedness training drill progressed. These detailed observations aided in the completion of an effective post-drill critique and in developing recommendations for continued improvement. The emergency preparedness organization planned to conduct additional training to address the areas for improvemen Conclusions Licensee controllers and evaluators in the TSC, OSC, and EOF provided an accurate assessment of activities during the June 16,1999, emergency preparedness training drill. The licensee's observations and post-drill critiques were effective in recognizing strengths, weaknesses, and areas for continued improvemen V. Manaaement Meetinos I'

X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on July 28,1999. The licensee acknowledged the findings <

presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie X3 Management Meeting Summary On July 7,1999, Commissioner Nils Diaz and Mr. James Dyer, Regional Administrator, visited the Clinton site. Items discussed during the site visit included recent licensee challenges and improvement initiative On July 14,1999, a public meeting was held on-site to discuss recent plant startup activities as well as NRC activities associated with implementation of NRC Manual Chapter 0350, " Staff Guidance for Restart Approval." Specific topics included the results of the NRC's startup inspection and the status of development and implementation of the licensee's long-term improvement pla V: .,

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' PARTIAL LIST OF PERSONS CONTACTED Licensee I

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G. Baker, Manager - Nuclear Support Services K. Gallogly , Director - Corrective Action J. Goldman, Manager - Work Management P. Hinnenkamp, Plant Manager - Clinton Power Station W. Maguire, Director - Operations J. McElwain - Chief Nuclear Officer R. Phares, Manager - Nuclear Safety and Performance improvement R. Schenck, Manager- Maintenance J. Sipek, Director- Licensing D. Smith, Director - Security and Emergency Planning D. Warfel, Manager - Nuclear Station Engineering Department E. Wrigley, Manager- Quality Assurance INSPECTION PROCEDURES USED IP 37551: Engineering Observations IP 61726: Surveillance Observations IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support and Observations IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92903: Followup - Engineering t

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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-461/99013-01 NCV Failure to critically review work schedule and barricade access to protected equipment during high risk activities 50-461/99013-02 NCV Failure to ensure that a TS entry was not required or the safety function was not disabled during the performance of minor maintenance Closed 50-461/99-003 LER Manual reactor scram required to insert control rod caused by a failed transponder circuit board 50-461/99-008 LER Failure of the motor driven reactor feedwater pump regulating valve results in a level transient and the insertion of a manual scram 50-461/99003-07 URI Review of 10 CFR Part 50.59 safety evaluation 98-066 50-461/99011-02 VIO Failure to assure design basis information for the hydrogen mixing compressors was translated into a safety related calculation 50-461/99013-01 NCV Failure to critically review work schedule and barricade access to protected equipment during high risk activities 50-461/99013-02 NCV Failure to ensure that a TS entry was not required or the safety function was not disabled during the performance of minor ]

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Discussed None l

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, LIST OF ACRONYMS USED EADS? l Automatic Depressurization System .

0 .AR : Action Request -

. C&l ) Controls and instrumentation .

s: . CR . Condition Report :

CRS- Control Room Superviso CW Circulating Water DRP Division of Reactor Projects .

-EDG, Emergency' Diesel Generator EM Electrical Maintenance EO Emergency Operations Facility GDC' General Design Criteria

HPCS High Pressure Core Spray LCO Limiting Condition for Operation LOCA Loss of Coolant Accident MCR- Main Control Room MWi Megawatt Thermal NCV- Non-Cited Violation-NSED- Nucisar Safety Engineering Department

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2OD Operability Determination O Operability Evaluation ORM- Operational Requirements Manual OSC . Operations Support Center .

P Preventive Maintenance PRA' Probabilistic Risk Assessment -

RCIC Reactor Core Isolation Cooling RG ' Regulatory Guide RP Radiation Protection RR Reactor Recirculation .

SLC Standby Liquid Control SM Shift Manager

.SRM Source Range Monitor TS Technical Specification TSC Technical Support Center -

USAR. Updated Safety Analysis Report VG Standby Gas Treatment System

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