IR 05000445/1998003

Insp Repts 50-445/98-03 & 50-446/98-03 on 980329-0509. Violations Noted.Major Areas Inspected:Operation,Maint, Engineering & Plant Support

ML20236G950
Person / Time
Site:	Comanche Peak
Issue date:	07/02/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20236G912	List: IR 05000445/1998003 ML20236G910 ML20236G932
References
50-445-98-03, 50-445-98-3, 50-446-98-03, 50-446-98-3, NUDOCS 9807070027
Download: ML20236G950 (32)

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ENCLOSURE 2

! . U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

L l Docket Nos.: 50-445

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Ucense Nos.: NPF-87 NPF-89 Report No.: '30-445/98-03 50-446/98-03 Licensee: TU Electric L_ Facility: Comanchs Peak Steam Electric Station, Units 1 and 2 l

Location: FM-56 Glen Rose, Texas -

Dates: March 29 through May 9,1998

_ Inspector (s): Anthony T. Gody, Jr., Senior Resident inspector l

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Harry A. Freeman, Resident inspector Thomas R. Meadows, Reactor Engineer, Operations Branch n. , Approved By: Joseph I. Tapia, Chief, Branch A l Division of Reactor Projects

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EXECUTIVE SUMMARY Comanche Peak Steam Electric Station, Units 1 and 2 NRC Inspection Report 50-445/98-03,50-446/98-03 The resident inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspectio Ooerations

• Control room operators used good self-verification techniques, were attentive to the control boards, and with one exception, consistently used good communications (Section 04.1).

a During reduced inventory operations, operators were knowledgeable of system pump vortex and cavitation timits and criticallimits such as time to boil. Unit supervisors routinely quizzed reactor operators on these limits. Operators carefully monitored the reactor coolant system drain rate and frequently checked the redundant reactor vessel level indications for deviations (Section 04.1).

. An inadequate procedure revision in conjunction with poor coordination resulted in a partial loss of reactor coolant system level indication while in reduced inventor Temporary pressure gauges were inadequately vented and placed in service at an inappropriate time. During the procedure revision safety screen, the licensee incorrectly assumed that the change was administrative and that a safety evaluation was not necessary. The inadequate procedure revision was a violation of Technical Specification 6.8.1 (Section O3.1).

. Licensee corrective actions have not been completely successful in resolving level indication problems encountered during a vacuum fill of the reactor coolant system. This has resulted in reduced operator confidence in midloop inventory level instruments (Section O3.1).

MaintenanC2

. Maintenance personnel adhered to procedures, work orders, and radiation work permit Good foreign material exclusion controls were observed and equipment cleanliness was maintained. Maintenance personnel used good safety practices (Section M1.1).

. The licensee's investigation and corrective actions following a reactor trip breaker's failure to close during testing were appropriate (Section M2.1).

. Overall, the Unit 1 fafueling outage was well conducted. However, several coordination problems were noted. An error associated with the tensioning of reactor vessel studs resulted in an additional 1.8 rem to the stud tensioning crew. A partialloss of reactor vessel level indication occurred while the reactor coolant system was in reduced inventory because of poor coordination and configuration management. Poor I

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! coordination of reactor coolant pump maintenance resulted in a spill of reactor coolan ' And poor coordination of containment closeout activities resulted in a senior reactor operator cleaning' the containment sumps following repairs (Section M6.1).

Enaineerina

.- ~ Design modifications were clear and straightforward. Safety evaluations were complete-and engineering reviews were consistent with the plant design and licensing bases

'(Section E2.1).

. The joint engineering team inappropriately used a maintenance troubleshooting

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' procedure, intended for electronic equipment, to install temporary pressure gauges for

' altemate reactor vessel level indication. An engmeering supervisor approved a maintenance troubleshooting plan which was contrary to procedural recommendation The licensee failed to adequately evaluate the effect that temporary pressure gauges j would have on level instrumentation. These errors resulted in a partial loss of level 1 instrumentation while the reactor coolant system was in reduced inventor l-(Section E4.1).

.. Four examples of a violation of 10 CFR Part 50, Appendix B, Criterion lll were identified -

during the resolution of an unresolved item concoming discrepancies between the

"s design bases and the emergency procedure for switchover of the emergency core cooling system pumps to cold leg injection (Section E8.1).

. - One violation of 10 CFR 50.5g was identified for changes to a facility operating procedure as described in the FSAR, which was found to involve a USQ. The USQ was

._ introduced by adding steps to the procedure. These changes increased the probability of occurrence of a malfunction of equipment important to safety; namely, the increased

. probability of gas binding emergency core cooling system pumps due to the increased time required to complete switchover. The licensee inappropriately concluded, in a 73 .

4 safety evaluation performed to evaluate the changes, that no USQ existed ,

(Section E8.1).

. A noncited violation associated with switchgear breaker cubicles being placed in a seismically unanalyzed condition by placing them in the " remove" position was identified (Section E8.8). ,

J Plant Supfiort

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-.~ EThe licensee continued to minimize the amount of temporary radiological drip containments. Excellent radiological support of maintenance activities in the containment building during the Unit i refueling outage was noted. The licensee continued to closely control water chemistry and take aggressive actions to correct any readings that were outside prescribed guidelines (Section R1.1).

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l 3-l l . ' The licensee failed to identify and correct conditions adverse to quality involving the fire doors for the Unit i uninterruptible power supply rooms in that the ability of the doors to ..

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close fully was impaired from March 1997 to April 1998 and the licensee failed to identif that the inability of the fire doors to remain open on the fusible links impaired the operation of the tornado doors (Section F2.1).

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Report Details Summarv of Plant Status At the beginning of the inspection period, Unit 1 was in Mode 6 with the sixth refueling outage ongoing.. Major work accomplished during the outage included refueling, inspection of the main generator, heater drain system modifications, steam generator eddy current testing, reactor coolant pump motor replacement, and safety injection system check valve replacement. While at reduced inventory, conducting a vacuum fiil of the reactor coolant system on April 19, level indication fluctuations of the installed instruments extended the amount of time spent at reduced l , inventory by a few hours. The level indicators were restored to an acceptable level of operation

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and the RCS was filled. On April 29, the generator output breakers were closed, ending the -

l refueling outage. At the end of the inspection period, the unit was in Mode 1 at 100 percent

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f l Unit 2 was operated at approximately 100' percent power throughout the inspection perio . Operations 01 Conduct of Operations L .n 01.1 General The conduct of operations observed was characterized by good command, control, and

- communications with some isolated exceptions noted below. The inspectors observed operators use good self-verification techniques.

L l- On March 27 and on April 18, the inspectors commenced continuous reduced inventory observations. Both evolutions were characterized by conservative decision making and

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. good management oversite, in general, operators conducted the activity well. However.

l - during the reduced inventory following core reload, several reactor vessel level indication I

problems occurred.' These problems were addressed and vacuum fill of the reactor-coolant system (RCS) was completed on April 1 '

O3 ' Operations Procedures and Documentation 03.1 Midlooo Level Indication Problems

. Inspection Scope (71707)

Following reduced inventory level indication problems on April 18, the inspector reviewed the justification for a revision to the reduced inventog procedure, which contributed to the level indication problems. Corrective actions from past level instrument problems and the impact that poor level instrument reliability had en operators were also reviewe Discussions were held with the operations support manager and the shift manager that had requested the procedure change, Observations and Findinas

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2-Several months before the Unit i refueling outage, a shift manager requested that Integrated Plant Operating Procedures IPO-010A, " Reactor Coolant Sys+em Reduced Inventory Operations," be revised to incorporate temporary reactor coolant level instrumentation. Revision 9 to Procedure IPO-010A, dated March 5,1998, added a step i to Section 2.1 which stated, "The PROMPT TEAM has installed the temporary pressure {

indicators on the top of the Pressurizer, Reactor Vessel Head, and the reactor coolant l system (RCS) loop to provide an alternate means oflevelindication." The safety I evaluation screen contained no reference to the additional equipment to be installed on the reactor coolant system nor, did it require a safety evaluation. The justification for not ,

performing a safety evaluation was, in part, that the revision was administrative in natur The inspector discussed this with the operations support group and found that the procedure writers incorrectly assumed that the temporary pressure gauges were controlled by Maintenance Department Administrative Procedure MDA-111

" Troubleshooting Activities," and maintenance work orders. The inspector found that Procedure MDA-111 did not address the installation of pressure sensing devices and the work order did not adequately control the installation of the temporary pressure gauge A review of licensing bases documentation, including the plant Technical Specifications and the FSAR, revealed no reference to reduced inventory level instrumentatio Accordingly, the inspector concluded that the licensee did not make a change to the

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facility as described in the FSAR. Since the installation of the pressure gauges was not considered a test or experiment, no violation of 10 CFR 50.59, " Changes, tests, and l experiments," occurred. Additionally, since no safety evaluation was required by the safety evaluation screen, the inspector noted that Procedure IPO-010A, Revision 9, had ,

not been reviewed by the Station Operations Review Committe ]

Technical Specification 6.8.1 requires that written procedures be established in accordance with the recommendations of Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Regulatory Guide 1.33 states, in part, that procedures .

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for operating safety-related pressurized systems, such as the reactor coolant system, include instructions for filling, venting, and draining. Procedure IPO-010A did not include instructions for filling, venting, and placing in service the temporary gauges. As a consequence, a partial loss of reactor coolant system ievel indication occurred during midloop operations. The procedure revision was inadequate and was a violation of l Technical Specification 6.8.1 (50-445/9803-01).

The shift manager's request to include additional level instruments was an attempt to l provide operators additional information on reactor vessel level and was considered to  !

be proactive. However, the implementation of the shift manager's request was  !

problematic. As described above, Revision 9 to Procedure IPO-010A was not adequate l to preclude a partialloss of reactor vessellevelindication. The inspectors found that the l licensee's configuration controls, associated with plar' - the pressure gauges in service,  !

were poor. This aspect is discussed in Section M ally, the inspectors found several problems and a procedural violation associated with the engineering review and approval of the temporary pressure gauges. This aspect is discussed in Section E i l

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The inspectors performed a review of past level instrument problems and found that

! reactor vessel level indication reliabihty problems occurred during several of the past l reactor coolant system drain evolutions on both units. These problems, documented in Operations Notification and Evaluation (ONE) Forms, involved inadequate configuration controls, poor venting, inadequate scaling calculations, faulty installation, and inadequate procedures. Several of these issues were previously discussed in past NRC inspection l reports and at least one was found to be a violation of requirements. Each issue was

adequately addressed by the licensee during corrective action implementation with the

! exception of the corrective actions associated with level instrument problems l encountered during RCS vacuum fill, as described belo Reactor Coolant System Vacuum Fill L

L On April 19, preparations were made to begin a vacuum fill of the RCS which included

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lowering RCS level to 49 inches above the core plate and heating the RCS to 140"F.

l The purpose of lowering RCS level to 49 inches above the core plate was to ensure that l an adequate vent path was maintained between the loops, the pressurizer, and the

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reactor vessel.. The purpose of raising RCS temperature in conjunction with th,e vacuum

. was to reduce lhe amount of dissolved oxygen in the reactor coolant. Once a vacuum  ;

was drawn, the RCS could be filled with only a minor chemistry transient from dissolved l oxygen. This technique has been shown to establish the proper corrosive layer, minimize RCS corrosion during the operating cycle, reduce the potential for fuel corrosion damage, and reduce radiation dose rate To accomplish the RCS heatup, operators throttled residual heat removal (RHR) system flow and then bypassed the heat exchanger in accordance with Procedure IPO-010A. In theory, bypassing the heat exchanger should result in a measurable heatup rate, however, it did not. Operators then throttled the component cooling water flow to the l heat exchanger which resulted in a measurable heatup rate. The shift manager appropriately initiated a change to Procedure IPO-010A to note the option of throttling l component cooling water flow. However, the inspector questioned whether it was L appropriate to throttle residual heat removal system flow prior to bypassing the heat exchanger as directed in Procedure IPO-010A. Reducing RHR flow to an amount that

! was equivalent to the leakage the isolation valve allowed appeared contrary to good l ' operating practices. The inspector noted that the licensee's response to Generic

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Letter 88-17," Loss of Decay Heat Removal," in Letter TXX 89041 dated l February 10,1989, stated, in part, that reduced inventory operating procedures were

. being prepared that would address required minimum RHR flow for decay heat removal.

i This is an inspection followup item (50-445/9803-02).

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Historically, once vacuum is drawn, RCS level instruments tend to drift. The licensee has written several ONE forms for previous occurrences. The licensee has postulated that the level drift may be the result of gasses coming out of solution in the instrument reference and variable legs as vacuum is drawn. This issue will be further reviewed with the licensee. As discussed above, the licensee indicated that the temporary pressure gauges were installed to help determine what occurs to the level instruments when L_ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _

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-4-vacuum is drawn. As vacuum was drawn on April 19, both the narrow range and wide range instruments drifted high. The licensee stopped drawing a vacuum, vented the RCS, and then vented the level instruments. As vacuum was drawn a second time, both the narrow and wide range level instruments drifted slightly. Since both the temporary ultrasonic level instruments mounted on the loop and the extended wide range instrument remained fairly stable, and therefore reliable, throughout the evolution, the shift manager ordered that the RCS be filled. This was the appropriate decision and it also minimized the time spent in the midloop conditio ' Conclusions A shift manager implemented operating procedure changes to install temporary pressure

gauges for additionallevelindications. The inspectors found the procedure change, engineering involvement, and installation to be problematic, resulting in a partial loss of RCS levelinstrumentation while in midloop. Several barriers to prevent such problems failed. The procedure change was not discussed in the procedure revision safety screen, a safety evaluation was determined to not be necessary and, as a result, the procedure was not reviewed by the Station Operations Review Committee. The temporary pressure gauges were not adequately vented and were placed in service at an inappropriate time, resulting in the partial loss of RCS level indication while RCS level was within the midplane of the loop. Licensee corrective actions have not been completely successful in resolving level indication problems encountered during vacuum fill of the RCS. This has resulted in reduced operator confidence in reduced inventory levelinstruments. Section M6.1 below discusses the maintenance issues, and Section E 4.1 discusses the engineering barriers that faile Operator Knowledge and Performance 04.1 Unit 1 - Reduced Inventerv Ooerations Insoection Scooe (71707)

The inspectors continuously observed the Unit 1 control room activities during reduced reactor coolant system inventory operations both prior to fuel off-load and following the fuel reload. Prior to reducing reacts r vessel level, the inspectors reviewed the licensee's risk assessment, calculations for time to boil, and operator training. Inspectors verified that equipment was available for decay heat removal, that operators continuously monitored the residual heat removal system, and that residual heat removal system indications were adequate. Reactor vessel level instrumentation, operator knowledge, past operating history, and instrument calibrations were also reviewed. Technical Specification requirements for reactor coolant makeup sources, electrical power, decay i heat removal, and containment integrity were reviewed and verified by the inspectors to be satisfie . .. - - - - _ __ --__--___ - __ ______

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b; Observations and Findings

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Operator Attentiveness and Communicahons The inspectors observed that operators were attentive to the control boards and ~ H maintained an awareness of all ongoing outage activities which could potentially affect l plant operation. - Communications were good with consistent use of three-wa communications (command, repeat back, and acknowledgment). However, one miscommunication was identified by the inspectors just prior to entering reduced inventory operations.- The inspector asked the reactor operator what RCS level indications were available. The reactor operator believed that only the extended wide'

range instrument was available when both the extended wide range and wide range level :

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instruments were in service. ' After informing the unit supervisor that the reactor operator believed that only one level instrument was available, the unit supervisor informed the reactor operator that both level instruments were available. The inspectors found that the unit supervisor had not effectively communicated that the wide range level instrument ;

was placed in service eariier in the watc Plant Knowledge

The inspectors confirmed that operators were familiar with RHR pump vortex and cavitation limits. Operators were also aware of critical limits such as the amount of time it would take for boiling to occur if cooling were lost. One good practice noted was that

senior reactor operators and unit supervisors routinely quizzed reactor operators on system limits and their base Reactor Coolant System Drain

During RCS draining, the inspectors noted that operators carefully controlled the drain

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rate, often checking the redundant reactor vessel level indicators for deviations. When -

L - reactor coolant level approached the loop midplane, shift management appropriately stationed.a plant equipment operator at the RHR pump to listen for cavitation. Audible indications from the RHR pump had been determined to be the first sign of cavitation

- during pre-startup testing.. Operators demonstrated a healthy skepticism of reactor -

, vessel level indications because of past problems and at the first sign of level deviation, the RCS drain was appropriately slowed or stoppo One significant problem occurred during the RCS drain following fuel reload. At

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approximately 57 inches above the core plate, with RCS level within the reactor coolant

['l' pipe, both the extended wide range and wide range level instrument readings dropped to

$ zero.- Indications retumed to their previous levels in about 2 minutes. Operators immediately recognized the problem, stopped the RCS drain, verified that the narrow range instrument was reading properly and then investigated the cause of the level . l

. instrument problem. The extended wide range and wide range level instrument failures occurred at the same time that a temporary pressure gauge attached to a common 3 variable leg was placed in service. The narrow range instrument had separate level )

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-6-taps. The licensee vented the affected level instruments, verified that all three level

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instruments were within severalinches, compared the permanent instruments with the temporary instrument, and continuet tha RCS drain to 49 inches above the core plate in preparation for the vacuum fill. Both the c."andad wide range and wide range level instruments required further venting to indicate accurately. The licensee postulated that )

placing the pressure gauge in service may have introduced an air bubble into the variable legs for the extended wide range and wide range level instruments, resulting in a decrease in indicated level. Issues related to this failure are discussed in Sections 03.1, M6.1, and E4.1. Although the licensee experienced several problems with some of the permanent reduced inventory level instruments, a temporary ultrasonic level instrument on the reactor coolant system piping and the narrow range level instrument was reliabl As a result, operators had two reliable levelindication Conclusions During reduced inventory operations,' operators were knowledgeable of system oump vortex and cavitation limits and criticallimits such as time to boil. Unit supervisors routinely quizzed reactor operators on these limits. Operators carefully monitored the reactor coolant system drain rate, frequently checking the redundant reactor vessel level indications for deviations'.

OS Operator Training and Qualification ,

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05.1 'Outaae Trainino .l l Insoection Scone (71707)

The inspector reviewed the preoutage operator training, including the lessons learned )

from previous outages and event Observations and Findinos l

The inspector verified that specialized drain down evolution and specialized outage, "just l in time," training was provided to the appropriate crews by the following training simulator

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. LO44.G98.lPO, " Rapid Shutdown"

. LO44.G98.X05, "Cooldown to solid operations"

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During Requalification Cycle 98-04, the inspector verified that operating crews were properly trained and evaluated in fuel handling and refueling operations. In addition, the licensee conducted midloop simulator training that included severalloss of decay heat removal scenario !

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08' Miscellaneous Operations issues i

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[ 08.1 - (Closed) Insoection Followuo item 50-445/9714-01:, reactor makeup water storage i tank (RMWST) draining evolution valve lineup error. The waste holdup tank was over0"3d by approximately 1000 gallons due to operator error and an inadequate l: procedure. The reactor operator failed to ensure that a flow path for the RMWST was

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l routed only to the floor drain and that there was no flow path unaccounted for. While L' draining the RMWST, an incorrect lineup' caused water to drain to the wasts holdup tank c l which overflowed. Contributing to the event was that the waste holdup tank annunciator (' did not respond at the 30 percent h;gh level mark to wam the operators in time to prevent the overflow, since it was a shared annunciator with the low level alarm and had already

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been tripped The licensee enhanced the applicable' draining procedure and distributed appropriate

. lessons leamed material to all operating crews. The annunciator was also modified to L allow reflash capability without operator intervention to reset the alarm. The inspectors -

determined that these corrective actions were appropriate.'

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08.2 - (Closed) Violation 50-445(446)/9717-01: three examples of operations personnel failing_

to follow procedures.- The inspector verified the corrective actions described in the licensee's response letter dated October 20,1997, to be reasonable and complete.- No l' similar problems were identifie >

08.3 '- (Closed) Licensee Event Raoort (LER) 50-445/98001-00: entry into Technical Specification 3.0.3 when feedwater isolation valves were declared inoperable due to low nitrogen pressure. This event was discussed in NRC Inspection Report

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' 50-445(446)/9801. ' No new issues were revealed by the LE )

i 08.4 - (Closed) Violation 50-445(446)/9617-01: failure to implement adequate corrective j j actions associated with plant transients following initial main turbine loading during plant - l H startup. The inspectors reviewed the licensee's response letter dated April 14,1997, and p found it responsive to the issues.- A verification of the licensee's corrective actions revealed that they wcre complete, encompassing all aspects of the violation.

l 08.5 ~ (Closed) Insoection Followuo item 50-445(446)/9705-01:' problems maintaining control'

t room work request and instrument out-of-service tags current. During a board walkdown on February 24,1997, the inspectors found a number of work requests and instrument i

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out-of-service tags on both the Unit 1 and 2 control boards that either had the wrong ;

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work order number on them or were already closed. The licensee immediately corrected 1

. the deficiencies identified by the inspectors and determined that changes were needed in j the process for tracking work requests written on degraded plant equipment. As a result, j the licensee modified Procedure ODA-401, " Control of Annunciators, instruments, and j Protective Relays." The inspector noted that the changes resulted in increased  ;

ownership by both operations and work control. The status of work requests and i

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l instruments taken out-of-service is tracked more frequently by operators and is periodically reviewed by work control. The licensee's corrective actions should prevent further problems in this area.

l 11. Maintfinance M1 Conduct of Maintenance M1.1 General Insoection Scoce (61/26. 62707)

The inspector observed the conduct of maintenance and surveillance activities to determine if the plant was being maintained as described in its design basis, to evaluate the impact that maintenance had on plant operations, and to determine if the licensee adhered to procedures and requirements. Specific rdivities observed are listed belo Unit 2. Train B Safeguards Slave Relay K631 Actuation Test

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Unit 1 Emergency Core Cooling System (ECCS) Check Valve Operability Test

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Unit 1, Low Power Physics Testing

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Unit 2, Turbine-Driven AFW Pump 2-01 Operability Test

- Unit 1, Main Feedwater Pump Maintenance

- Unit 1, EDG Maintenance

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Unit 1, Calibration of Extended Wide Range, Wide Range, and Narrow Range RCS LevelIndicators 'x . . .

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Unit 1, RHR Testing

- SI System Check Valve Replacements

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Reactor Coolant Pump (RCP) Motor Replacement Observations and Findinos Overall, the inspectors found that maintenance personnel adhered to procedures, work orders, and radiation work permits. Good foreign material exclusion controls were observed and equipment cleanliness was maintained. P/ ntenance personnel used good safety practices. Issues and noteworthy observations are detailed in the following section _ - - _ _ _ _ _ _ _

-9-l M1.2 ECCS Check Valve Operability Surveillance Test l

l Insoection Scooe (61726)

in accordance with the outage inspection plan, the inspector observed the licensee test the operability of the ECCS check valves in accordance with Procedure PPT-S1-8200,

"ECCS Check Valve Operability Test," Revision 4. The inspector reviewed selected data calculations and compared the results to the ccceptance criteri Observations and Findinas The inspector found that the licensee performed the surveillance test in accordance with the procedure. Generally good communications were observed between the test engineer, the licensed operator, the surveillance test coordinator (senior reactor operator), and the on-shift crew. The test coordinator closely observed and interacted with the engineer and operator as the test was conducte The inspector noted one instance where the test engineer and the reactor operator did not identify that the as-read safety injection pump recirculation flow was outside the listed range of 30 to 50 gallons per minute. When questioned by the inspector, the test coordinator concluded that the procedure steps for sti.rting the pump had been taken directly from the standaro operating procedure which assumed that only one pump was running instead of two. With two pumps running, the recirculation flow would be highe The test engineer annotated the work package and informed the inspector that the procedure would be revise The inspector confirmed that the licensee had accurately input the test parameters into the calculations and that the check valves met the acceptance criteri M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Reactor Trio Breaker Failure to Close a. Insoection Scoce (61726)

The inspector reviewed the licensee's determination of the cause of a reactor trip breaker's failure to close during testing. The inspector also reviewed the licensee's determination of the significance of the failure and the extent of the corrective action b. Observations and Findinas On April 27, a Westinghouse DS-416 reactor trip breaker failed to close during rod drop testing. Licensee investigation revealed that a washer, which holds the trip latch on the >

power operating mechanism, had been placed on the wrong side of the trip latch pivot pin retaining clip. The washer migrated across the pivot pin to a point where it jammed against the close cam when the breaker motor attempted to charge the closing sprin .. - _ _ _ _ - _ _ _

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.,10 The motor failed to charge the closing spring and the breaker would not clos The licensee replaced the defective power operating mechanism and verifed that the mechanisms in the other three reactor trip / bypass breakers and those in the warehouse were correctly assembled. Additionally, the licensee added a statement to their breaker

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refurbishment procedure to inspect this key arrangement. The power operating mechanisms on the reactor trip /t,ypass breakers had been replaced by the vendor with new ones when the breakers were refurbished. The defective mechanism was returned to the vendo The licensee concluded that the condition did not impact the safety function of the reactor trip breaker. The reactor trip breakers at Comanche Peak are designed to charge the closing spring after the breaker trips open (known as a 6a wiring scheme).

Because of this scheme, the improper washer placement would not prevent a shunt trip of the breaker if the motor completes charging of the closing spring, then the physical arrangement of the pivot pin, the close cam, and the washer would ensure that the shunt trip pivot pin is free to trip open the breaker. The inspector agreed with the licensee that the improper assembly did not affect the reactor trip breakers capability to tri However, the DS-416 breaker is also used in numerous other 480 voit applications both safety and nonsafety. In a few of these other applications, the breaker motor charges the closing spring after the breaker is shut (7a wiring scheme). In those cases, the equipment may not start due to this type of problem. The inspector also noted that an

' improperly positioned washer could bind against the close cam and bind the trip latch pivot pin after the breaker was shut. In that case, a shunt trip of the breaker may not occur. The licensee informed the inspector that there were only a few nonsafety-related breakers with the 7a wiring schem Conclusions The licensee's investigation and corrective actions following a reactor trip breaker's failure to close during testing were appropriat M6 Maintenance Organization and Administration M6.1 Coordination of Unit 1 Refuelino Outaae Maintenance Activities Insoection Scone (92902)

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. The inspector reviewed the causes and impact of several outage coordination prob'3ms l- involving: the need to retension several reactor vessel head studs, a reactor coolant spill

[ ~ during reactor coolant pump recoupling, the partial loss of midloop level instrumentation during reduced inventory operations, and the containment sump closeout inspectio ,

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-11- I l b. Observations and Findings Notable performance issues related to coordination problems are documented below for trending purpose Midloco Level Instruments As discussed in Section 04.1 above, on April 18, with reactor coolant level at about 57 inches above the core plate both extended wide-range and wide-range level instrument readings dropped to zero and stabilized to their previous levels in about 2 minutes. The extended wide-range and wide-range level instrument failures occurred at the same time a temporary pressure gauge, which was attached to the common extended wide range and wide range instrument variable leg, was placed in service by {

plant equipment operators. The licensee vented the affected level instruments, verified I that all three level instruments were within several inches, compared the permanent instruments with the temporary instrument, and continued the RCS drain to 49 inches above the core plate in preparation for the vacuum fill. Both the extended wide range and wide range level instruments required further venting to indicate accurately. The l licensee postulated that placing the pressure gauge in service may have introduced an air bubble into the common variable legs for the extended wide range and wide range level instruments, resulting in a decrease in indicated leve The inspector reviewed the circumstances and found that the licensee did not adequately control the installation of the temporary pressure gauges. A lack of coordination resulted in the temporary gauges being placed in service before they were ,

vented and after the RCS was already in a reduced inventory condition. In addition, the licensee indicated that long runs of tubing were used to connect the temporary pressure gauges to the RCS. This was a bad practice which had the potential of introducing significant amounts of air into the affected system.

I Reactor Vessel Head Tensioning On April 18, a problem was encountered with the reactor vessel head hoist which was ,

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being used to facilitate reactor vessel head tensioning. While troubleshooting was being conducted on the hoist, the Westinghouse coordinator recommended to the outage I control center that the Westinghouse conoseal crew be allowed to begin installation of conoseals. The outage control center authorized the conoseal crew to start work To facilitate installation of the conoseals, the crew installed platforms on top of the reactor vessel studs. In order to install the platforms, however, the crew found it necessary to remove the elongation measuring rods from a number of the stud When the Westinghouse tensioning crew returned to reperform stud tensioning, they I discovered that the elongation measurement rods were not returned to the same studs l they were removed from. As a result, the baseline tensioning measurements were lost ,

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[ on eight studs. In order to facilitate retensioning, all eight studs had to be detensioned and then retensioned. This required a change from Mode 5 to Mode 6 and was a

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-12-significant administrative burden on operators. Once operators completed the mode  !

change requirements, Westinghouse detensioned and retensioned the eight studs. The I whole process of detensioning and retensioning resulted in an additional 1.8 rem to the l

Westinghouse cre I Reactor Coolant Pumo Backseat Soill On April 16, while a Westinghouse crew was recoup!ing the motor on reactor coolant pump RCP 1-02, a coordination problem resulted in spilling approximately 50 gallons of l

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reactor coolant onto the floor. The licensee's investigation of the work control problem disclosed: a failure to include a prerequisite to close a drain valve on the RCP seal injection piping prior to recoupling, a failure to conduct a prejob briefing of the infrequent l evolution, poor instructions and/or procedures, and a lack of oversight. In addition, the l licensee found that the immediate response of the field support supervisor and radiation l protection personnel to isolate and stop the spill of reactor coolant, was good. The i inspector attended the performance enhancement meeting and found the communications to be open and forthright. The inspector agreed with the licensee's

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conclusion Containment Sumo Closeout j The inspector conducted a thorough inspection of the Unit 1 containment sumps along with the senior reactor operator assigned to the task. Both the inspector and the senior reactor operator noted debris in the sumps. The inspector found metal slag, weld wires, and other materials left from containment sump screen repairs. The senior reactor operator found similar materials in addition to a ladder and a portable light. Since this was a final closecut of the containment sumps, the senior reactor operator, demonstrating excellent ownership, proceeded to clean the sumps including vacuuming debris. Although the sumps were clean after the senior reactor operator was done, the coordination of sump cleaning and containment closecut was ineffective. The licensee indicated that the containment sump closeout inspection was moved ahead of schedule by the outage control center apparently without moving the sump cleaning activity along with i c. ConglyEQD1 Overall, the Unit i refueling outage was well coordinated. Several coordination errors resulted in varying degrees of impact. A coordination error associated with the tensioning of reactor vessel studs resulted in an additional 1.8 rem to the Westinghouse stud tensioning crew contrary to good ALARA practices. A partialloss of reactor vessel

level indication occurred while the reactor coolant system was in a reduced inventory condition, partially because of poor coordination and configuration management. Poor coordination of reactor coolant pump maintenance resulted in a spill of reactor coolant, and, poor coordination of containment closeout activities resulted in a senior reactor

operator cleaning the containment sumps following repairs.

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e I -13-M7 Quality Assurance in Maintenance Activities M7.1 Human Performance Improvement Insoection Scoce (62707)

On April 7, the inspectors met with members of the nuclear overview department to discuss the licensee's plans for human performance improvement. Following the outage, the inspectors reviewed a comparison of errors committed during the third Unit 2 outage and the sixth Unit 1 outag Observations and Findinas The inspectors found that the licensee had a thorough plan to reduce human performance errors. This included two planned human performance stand down days, ;

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close management interaction, and an effort to eliminate schedule pressure. Compared to the previous refueling outage in Un.t 2, there was a major reduction in signific .

personnel errors (1 versus 6) and an overall 31 percent reduction in errors. The inspectors concluded that the licensee's efforts had been effectiv )

M8 Miscellaneous Maintenance issues M8.1 (Closed) Insoection Followun item 50-446/9611-02: steam plant pump failures. The {

inspector reviewed the licensee's balance-of-plant reliability improvement efforts and subsequent equipment performance data and found that condensate pump reliability was maintained at a high level. In addition, materialimprovements were made to the j condensate pump rotating elements. Several significant reliability improvements were i made by the licensee to both units feedwater pumps which included the installation of a digital control system, tilting pad bearing installation (Unit 1 only), replacement of carbon <

steel piping with stainless steel piping, the installation of a new diaphragm coupling, and l vibration probe changes. Heater drain system reliability was improved by installing !

modifications to eliminate several operator work-arounds from design problems. Based i on direct observations during several startups, a review of equipment reliability, and j direct observations of maintenance, surveillance, and testing activities, the inspectors concluded that the licensee adequately monitored the performance of balance-of plant pumps and implemented effective corrective action ,

M8.2 (Closed) Insoection Followuo item 50-445(446)/9708-02; root cause followup on turbine-driven auxiliary feedwater steam admission valve diaphragm failures. This issue will be reviewed as part of the followup to IFl 50-445(446)/9712-03.

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! 111. Enaineerina E2' Engineering Support of Fac811 ties and Equipment l E2.1 - Review of Desian Modifications Insoection Scone (37551)

. The inspector reviewed the safety evaluation and the appropriate sections in the FSAR l for selected design modifications implemented during the Unit 1 Spring 1998 refueling outag Design Modification 93-061, " Install Woodward 701 Govemor System"

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Design Modification 96-086, " Install Automatic Safety injection Letdown isolation on Five Valves"

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Design Modification 97-061, " Change Average Coolant Temperature Determination from Auctioneered to Averaged" Observations and Findinas Woodward 701 Govemor Design Modification 93-061 installed the Woodward 701 digital governor on th emergency diesel generators. The Woodward 701 digital govemor provides capabilities for slow starting of emergency diesel generators during maintenance and certain surveillance to reduce engine wear and increase reliabilit The inspector found the design modification engineering input consistent with plant design requirements. Electromagnetic interference, equipment qualification, operating procedures, failure modes and effects, software verification and validation, and reliability were all properly considered in the design. Safety Evaluation 94-48 determined that the u design change did not involve a USQ. A Technical Specification amendment was issued by the NRC to allow slow starts of the emergency diesel generator.

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Design Modification 96-086 installed automatic letdown isolation during a safety injectio Securing letdown upon the initiation of the safety injection rather than upon pressurizer low level would save RCS inventory, reduce operator burden during the accident, and reduce exposure caused by overflowing the pressurizer relief tan The inspector found that the design modification was straightforward and added safety

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benefit during a postulated accident. The inspector confirmed that Safety i Evaluation 97-31 accurately determined that the design change did not involve a USQ.

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-15-un taae Coolant Temperature Ceterminatino Oc:agn Modification 97-061 changed the method for determining average coolant temperature from an auctioneered high average to an actual average temperature. The change was made to facilitate operation of rod controlin automatic. This change resulted in an overall small increase in the bulk coolant temperature. As a result, the heat transfer across the fuel cladding decreased slightly, slightly increasing the possibility of reaching the fuel hot channel factor limit in one or more ast,emblie The inspector reviewed the modification package and compared the temperature values included in the package with the values used in the FSAR and core operating limits report. The inspector discussed the assumptions and methods used to evaluate the error introduced by the modification, and how it impacted the safety analysis and core reload calculations, with the responsible engineer. The inspector found that the error analysis was completed in a satisfactory manner. Although the modification reduced the amount of administrative margin incorporated in the original calculations, inclusion of this margin was a conservative measure and was not required by regulations nor credited in the safety analysis. Thus, the small temperature increase expected during plant operation following this modification was bounded by the original safety analysis and core reload calculatio Conclusion Design modifications were appropriately implemented. Safety evaluations were clear and the engineering review was consistent with the plant design and licensing base E4 Engineering Staff Knowledge and Performance E RCS Reduced Inventorv Level Instrumentation lj Insoection Scooe (92903. 37551)

The inspector reviewed the circumstance surrounding the engineering involvement in the installation of temporary pressure gauges in the reduced RCS level instrument variable and reference leg Observations and Findinas As discussed in Section O3.1 above, a shift manager's request to include additional reduced RCS inventory level instruments was an attempt to provide operators additional

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information on reactor vessel level and was considered by the inspectors as proactiv However, implementation of the shift managers request was problematic. As described in Section O3.1 above, Revision 9 to IPO-010A was not adequate to preclude a partial loss of reactor vessel level indication. In Section M4.1, the inspectors found that the

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licensee's timing and configuration controls associated with placing the pressure gauges in service were poor.

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l A review of the joint engineering team involvement revealed several important problems

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associated with the engineering review and approval of the temporary pressure gauge l The engineering involvement concerning the temporary pressure gauges was limited to implementatbn of a troubleshooting plan in accordance with Maintenance Department Administrative Manual, Procedure MDA-111 " Troubleshooting Activities." The inspector found numerous problems associated with the use of the MDA-111 procedure: (1) the procedure applied to troubleshooting activities performed on all systems, equipment, and components under the cognizance of the maintenance department, however, the RCS level instruments were under the cognizance of the operations department; (2) MDA-111 troubleshooting plans are required to be approved by a maintenance department j I

supervisor and the temporary pressure gauge troubleshooting plan was approved by engineering; (3) MDA-111 lists several good maintenance practices to be observed while troubleshooting, however, the inspector found that all of these related to electronic equipment and that venting of pressure sensing devices was not mentioned; (4) MDA-111 focused on the installation of electronic monitoring equipment and did not mention the installation of pressure sensing devices, therefore, it was not readily applicable to the installation of pressure sensing devices; (5) Section 6.3.5 of MDA-111 required that test equipment and test leads be listed as pre-approved for use or their use be evaluated using a technical evaluation; however, the pressure sensing devices were not pre-approved for use and no technical evaluation was performed by engineerin The failure to adhere to the provisions of MDA-111 during the use of pressure sensing devices while troubleshooting and the failure to perform a technical evaluation regarding the adequacy of the temporary pressure gauge design led to the loss of level indication and were in violation of Technical Specification 6.8.1 (50-445/9803-03).

In addition, the troubleshooting pian contained a table that provided correlation between  !'

the pressure gauge reading and the corresponding reactor vessellevel which was used by shift management in the control room. The inspector found that the joint eng:neering team performed calculations to develop this table which were informal and undocumente Conclusiorn The joint engineering team inappropriately used a maintenance troubleshooting procedure intended for electronic equipment to install temporary pressure gauges for alternate reactor vessel level indication. In addition, engineering approved the  ;

maintenance troubleshooting plan contrary to the procedure recommendations. The l licensee failed to perform a technical evaluation to determine the impact that installing temporary pressure gauges in the .RCS would have on reactor vessellevel instrumentation as required by MDA-111. These failures contributed to a partialloss of {

levelinstrumentation while the RCS was in a reduced inventory condition. In addition, 1 calculations that were performed to support the troubleshooting plan were informal and undocumente )

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E-17-E8 . Miscellaneous Engineering issues E (Closed) Unresolved item 50-445(446)/9616-06: discrepancies between the operating

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procedures and the FSAR on the switchover of the ECCS from injection to recirculatio Subsequent to NRC 5pection Report 50-445(446)/9616, several additional issues were identifiedi These issues were: (1) the design bases calculations did not support the plant operating procedures; (2) the licensee's initial engineering review to support the operability decision was not technically correct; and (3) the 10 CFR 50.59 safety l)

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evaluation supporting the inspector identified licensing basis / plant operating procedure Discrepancy involved a US Design Bases

! ~As discussed in NRC Inspection Reports 50-445(446)/9616 and 50-445(446)/9617, the

, inspector identified that the design bases for the RWST level instrument setpoint, (

contained in Calculation 16345-ME(B)-389, "RWST Setpoints, Volume Requirements and Time Depletion Analyses," Revision 1, included outflow requirements during

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s itchover as contained in Westinghouse Letter WPT-3358, dated July 16,198 Westinghouse Letter WPT-3358 specified these RWST outflow requirements based on L

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manual operator and equipment response times for a sequence of six steps required to' -

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shift suction from the RWST to the containment sumps for the ECCS pumps. The

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inspector found that the licensee failed to subject changes to emergency Procedures

. 1 EOS-1.3A, Revision 6, and EOS-1.38, Revision 1. " Transfer to Cold Leg Recirculation,"

L to control measures commensurate with those applied to the original design in that Calculation 16345-ME(B)-389 did not account for additional water volume required to

, , . complete the switchover which was added by nine additional steps in the procedures.

l, -This was an example of a violation of 10 CFR Part 50, Appendix B, Criterion Ill.

i The licensee's subsequent investigation identified two other design bases issue The first issue also involves Westinghouse Letter WPT-3358 which specified that sufficient volume be provided below the empty alarm to allow time for operators to shut off any pump still taking suction from the RWST prior to level dropping below C the level where ECCS pump net positive suction head was inadequate or vortex formation could occur. The inspector concluded that the licensee failed to verify the adequacy of design prior to receiving their operating license in that neither Calculation 16345 ME(B)-320, " Vortex Potential at Charging Pump Outlet in RWST," Revision 0, nor Calculation 16345-ME(B)-389 accounted for operator response time in determining the empty alarm setpoint. This was an example of a violation of 10 CFR Part 50, Appendix B, Criterion Il

The second issue involved Calculation 16345-ME(B)-389, Revision 1, which established the empty _ level alarm such that the level was as low as possible while'still maintaining enough water to switchover the containment spray system as well as to prevent the possibility of vortex formation. The calculation used

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q 1 minute as a design basis to determine the setpoint. The inspector concluded

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-18-l that the licensee failed to verify the adequacy of design prior to receiving their operating license in that the design opening and closing times for the containment l

spray valves to the RWST and for the containment spray valves to the containment sumps was 120 seconds; therefore, switchover of the containment spray system could not be completed in 1 minute as assumed in the design basis. This was an example of a violation of 10 CFR Part 50, Appendix B, Criterion Il A final design control issue was identified by the inspector during the licensee's corrective actions to the original issue. In Calculation 16345-ME(B)-389, Revision 2, the licensee reduced the analyzed instrument uncertainty from total loop (13 inches) to rack l (7 inches) in their determination of RWST volume available to complete switchover to

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recirculation. This reduction in uncertainty was based on the erroneous assumption that instrument setpoint drift was adequately considered because the same instrument would be providing both the RWST low level switchover signal and the empty alarm. The inspector found that this assumption was not valid. This was an example of a violation of j 10 CFR Part 50, Appendix B, Criterion Il The inspector found that the four examples above were contrary to 10 CFR Part 50, Appendix B. Criterion Ill, " Design Control," which requires that design control measures provide for verifying or checking the adequacy of design, and that design changes be subject to design control measures commensurate with those applied to the original design (50-445(446)/9803-04).

Licensina Bases The licensee had developed Procedure EOS-1.3 subsequent to developing the FSAR sections describing the manual procedure steps required to conduct switchover (Tables 6.3-7, and 6.3-11). The procedure was revised five times prior to Unit 1 operating license receipt in 1990. At that time, EOS-1.3 deviated significantly from the procedure described in FSAR Table 6.3-7. Subsequent to 1990, but prior to the inspector identification of the issue in December 1996, one additional revision was made and a functionally identical procedure was developed for Unit 2. Each of the procedure i

changes indicated through a 10 CFR 50.59 safety evaluation screen that a safety I evaluation was not required.

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10 CFR 50.59, " Changes, tests and experiments," in part, permits the licensee to make changes to its facility and procedures as described in the safety analysis report and conduct tests or experiments not described in the safety analysis report without prior Commission approval provided the change does not involve a change in the technical specifications or a USQ. The licensee is required to maintain records of changes and the records must include a written safety evaluation which provides the bases for the determination that the change does not involve a US l

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O D-19-The inspector found that, as of December 19,1996, the facility was not as described in the FSAR in that Table 6.3-7 did not include the time and outflow requirements to perform all of the manual actions required to switcht. $r the ECCS pumps from injection to recirculation, as implemented by Emergency Response Guidelines Procedure ,

EOS-1.3, " Transfer to Cold Leg Recirculation." Therefore, the facility was changed b no evaluation existed which provided the bases for the determination that the changes did not cunstitute a USQ. As a result, Table 6.3-11 was not accurate in that it )

underestimated the water volume required to complete the manual switchover. This was I a first example of a violation of 10 CFR 50.59 (50-445(446)/9803-05). I

_UlQ1i The inspector found that FSAR Table 6.3-7 listed six manual operator steps required to switchover the emergency core cooling system from injection from the refueling water storage tank to recirculation from the containment sumps. Final Safety Analysis Report j Table 6.3-11 stated that 90,166 gallons were required to complete the switchover and j that the time requiod to complete the required action included a conservative {

30 seconds for operator response time for each manual procedure. Section 6.3. stated that 94,179 gallons were available for switchover. The inspector identified that Procedures EOS-1.3A, Revision 6, and EOS-1.38, Revision 1, " Transfer to Cold Leg Recirculation," required 15 manual operator steps to switchover the emergency core cooling system pumps. This meant, according to the margin specified in the phnt design, that operators had about 4,000 gallons to complete the additional nine step After the licensee declared the RWST and ECCS operable during their backup engineering review of tha operability decision, an evaluation of the existing discrepancies between the design / licensing bases and plant operation in accordance with 10 CFR 50.59, " Changes, tests and experiments," was completed on January 3,199 :

10 CFR 50.59, " Changes, tests and experiments," allows a licensee to make changes in the procedures described in the safety analysis report, without Commission approval, unless the proposed change involves a USQ. A proposed change shall be deemed to involve a USQ: (1)if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (2) if a possibility for an accident or malfunction of a different l type than any evaluated previously in the safety analysis report may be created; or (3) if the margin of safety as defined in the basis for any technical specification is reduce In their 10 CFR 50.59 safety evaluation (Evaluation 97-001, Revision 0), the licensee

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concluded that the existing ECCS switchover procedure, described in Procedures l EOS-1.3A and EOS-1.3B, did not involve a USQ. In January 1997, the inspectors l questioned the licensee's conclusions in Evaluation 97-001, Revision 0, believing that at least one USQ existed. _ - _ _ - - _ _ _ _ -- . _ _ _ _ -

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l After a detailed review of the circumstances surrounding the licensee's conclusions in '

Evaluation 97-001, and with input from the NRR staff, the inspectors concluded that the additional steps added by the changes and the reduced response time required by these steps could reasonably increase the probability of occurrence of a malfunction of equipment important to safety; namely, the increased probability of gas binding of emergency core cooling system pumps due to the increased time required to complete switchover. The licensee inappropriately concluded in a safety evaluation performed to evaluate the changes that no USQ existed. This was a second example of a violation of 10 CFR 50.59 (50-445(446)/9803-05) .

Current Operability Since the time that the inspector identified discrepancies between the operating procedures and the FSAR in December 1996, the licensee has revised Calculation 16345-ME(B)-389 and Procedure EOS-1.3. The net effect of these changes has been to restore the original FSAR procedure steps and the 30 second per operator response assumption by modifying the operating procedure, refining assumed pump ;

flows, and raising the low-low level setpoint. Two steps were still at the beginning of Procedure EOS-1.3 which were not described in the FSAR, however, the time and 3 required . vater volume for these steps were included in the latest calculation and the licensee indicated that an FSAR change was in p ogress. Additionally, the licensee also verified that there was sufficient water volume available when the instrument uncertainty en ;omponents were correctly reanalyzed. The inspectors found that the licensee's ope; ability determination was correc Conclusion Four exarrp!ec of a failure to verify the adequacy of design in violation of 10 CFR Pan 50, Appendix B, Criterion Ill, were identified during the resolution of an unresolved item concerning discrepancies between the FSAR and the emergency procedure for switchover of the ECCS pumps to recirculatio Two examples of a violation of 10 CFR 50.59 were identified related to a change to the .

facility as oescribed in the FSAR when a USQ that was introduced by adding steps to a l procedure and another that was introduced by reducing the operator response time per step were not correctly identified. The changes introduced the possibility of increasing the consequences of an accident, and therefore, constituted an US E8.2 (Closed) Violation 50-445(44619718-04: failure to use temperature instrument with the accuracy required by procedure. This violation was issued for performing a test using a ,

temperature instrument with an accuracy of 12.2*C when the procedure required an !

instrument with an accuracy of 12*F. The inspector verified the corrective actions j described in the licensee's response letter, dated December 19,1997, to be reasonable !

and complete. No similar problems were identifie l l

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E8.3 (Closed) Violation 50-445(446)/9718-03: inadequate procedures for hydrogen

' recombiner surveillance and control room pressurization surveillance. The inspector verifed that the corrective actions described in the licensee's response letter, dated December 19,1997, were reasonable and complete. The inspector verified that the

' applicable procedures had been revised to prevent recurrence of the violation.'

E8.4 '(Closed) LER 50-445/97006-00: 50-445/97006-0j: control room pressurization unit flow

' found outeide of Technical Specification limits. This event was discussed in inspection

' Report 50-445(446)/9718._ No new issues were revealed by the LE p , E8.5 _ (Closed) LER 50-445/96004-00: potential failure of personnel airlock control systems in both units. This LER was written when the licensee discovered that non-class 1E controls could cause a breach of containment integrity during a design basis accident by allowing the inadvertent opening of equalizing valves. The hydraulic pumps which supplied the equalizing valves were originally powered from a breaker which received an

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S-signal during a loss of coolant accident. A design modification changed the power -

source to one which did not receive an S-signal. Consequently, the harsh environment inside containment during a loss of coolant accident could cause the non-class 1E core ois to open the equalizing valves._ The licensee corrected the problems by providing

. an S-signal to trip the power supply to the hydraulic pump E8.6 (Closed) Insoector Followuo item 50-445(446)/9801-04: followup on the regulator compliance with diesel generator load shed surveillance requirements. This item will be

- reviewed as part of the review of Unresolved item 50-445(446)/9802-01, Untested contacts identified during licensee's NRC Generic Letter 96-01, " Testing of Safety-Related Logic Circuits."

. E8.7 (Closed) Violation 50-445/9612-03: failure to perform relief valve testing in accordance

- with the ASME Boiler and Pressure Vessel Code Section XI, Subsection IWV, " Inservice Testing of Valves in Nuclear Power Plants." The inspector verified that the corrective a

. actions described in the licensee's response letter, dated December 20,1996, were I

, reasonable and complete. Trse inspector verified that the applicable procedures had been revised to prevent recurrence of the violation and observed the testing of relief valves with the modified relief valve test stan . E8.8 (Closed) LER 50-445/96009-00: 50-445/96009-01: switchgear breakers racked out to

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,' the " remove" position resulted in a seismically unqualified condition outside the plant

. design bases. The inspector reviewed the licensee's corrective actions and found them L' - adequate to prevent recurrence. This nonrepetitive, licensee.-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the L NRC Enforcement Policy (50-445/9803-06). j h~

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E8.9 . (Closed) Insoection Followuo item 50-445(4461/9612-02: calculation review and I

approval process. - This. item was opened based on an inspector identified error in determining the potential concentration of ammonia released during steam generator chemical cleaning. The licensee controls for engineering calculations were specified in l

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-22-Engineering Procedure ECE 5.03, " Calculations." The procedure specified the level of -

verification expected for calculations. The inspector found that the licensee had not formally conducted a calculation to determine a concentration value. The licensee explained that management had directed that the' concentration value be handled as a

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formal' calculation. However, the licensee stated that they had an action item to review when " informal' calculations were acceptable in followup to a 1997 engineering self-assessment item. Although there has been one licensee-identifed calculation error in the last two years, no significant breakdown in the calculation process has been identifie E8.10 - (Closed) Violation 50-445(446)/9715-03 . inadequate area temperature monitoring surveillance procedure. In the enclosed Notice of Violation to NRC Inspection Report 50445/97-15l 50-446/97-15, the inspectors concluded that the information regarding the reason for the violation and the corrective actions taken and planned were already adequately addressed on the docket and that a response was not necessary unless the licensee concluded that the descriptions or corrective actions did not accurately reflect their position. The licensee did not respond, therefore, this item is close IV. Plant Support R1 ' Radiological Protection and Chemistry Controls

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R1.1 - General Insoection Scone (71750)

The inspectors observed radiological protection activities during routine tours and observation of maintenance and surveillance activities. The inspectors also observed radiological protection support of containment activities during a planned Unit i refueling outage. The inspectors reviewed the primary and secondary water chemistry and l radiation protection department log Observations and Findinas -

-  : The inspectors found that the licensee continued to minimize the amount of temporary, radiological drip containments. The inspectors noted excellent radiological support of maintenance activities in the containment building during the planned Unit i refueling

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outage. The inspectors found that the licensee continued to closely control water .

L ~ chemistry and take aggressive actions to correct any readings that were outside the

. prescribed guideline R8 ~ Miscellaneous Radiological Protection and Chemistry issues R (Closed) Violation 50-445(4461/9802-07: failure to control a locked high radiation area in accordance with technical specification requirements due to removal of containment access hatch locks for three days.'In the enclosed Notice of Violation to NRC Inspection

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.e-23-Report 50-445(446)/9802, the NRC concluded that the information regarding the reason for the violation and the corrective actions taken and planned were already adequately addressed on the docket and that a response was not necessary unless the licensee concluded that the descriptions or corrective actions did not accurately reflect their position. The licensee did not respond, therefore, this item is close P8 Miscellaneous Emergency Planning issues F (Closed) Insoection Followuo item 50-445(446)/9708-04: the issue of minimum manning was documented as part of an Emergency Plan review in NRC Inspection Report 50-445(446)971 F,1 Conduct of Security and Safeguards Activities Insoection Scoos (71750)

~ The inspectors observed security and safeguards activities during routine tours, at protected area access facilities, and at compensatory posts throughout the inspection perio Observations and Findinas  ;

The inspectors found that security officers were attentive and conducted their duties in a professional manner. Officers were knowledgeable of their post requirement S8 Miscellaneous Security and Safeguards issues S (Closed) LER 50-445/98002-00 inoperable security search detector allowed unauthorized access to the protected area. This event was discussed in Inspection ;

Report 50-445(446)/9811. No new issues were revealed by the LE F2 Status of Fire Protection Facilities and Equipment j F Inocerable Tornado Blowout Doors Inspection Scooe (71750)  !

The inspector reviewed the licensee's corrective actions for two impaired fire doors. The inspector reviewed the FSAR and design basis documents to determine the tornado l venting requirement ) Observations and Findinas While touring the control building on November 17,1997, the inspector identified that the fire door separating the Unit 1, Train B, uninterruptible power supply from the hallway would not stay open. While the fire door was performing its safety function (separating

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e e-24-two different fire zones), it was designed to stay open by a fusible link. In the event of a fire, the fusible link would melt and a spring closer would shut the fire door. The inspector questioned the licensee whether the closed door could impair the operation of the tornado door on the other side of the door frame. The licensee informed the inspector that they would look into the issue. The inspector noted that the issue appeared to be minor and that the licensee was investigating. However, on April 13,1998, the inspector identified that the fire door was, again, close The inspector contacted the engineer responsible for tomado venting icsues. The engineer informed the inspector that the tornado venting design calculations assumed that the fire door was open. After inspecting the condition of the door in the field, the engineer also identified that the Unit 1, Train A, uninterruptible power supply room fire door would not stay open. The engineer wrote a ONE form and recommended that both doors be tied open. Since fire impairments were already 'i place due to problems with the door closure mechanisms, the doors were tied open niat afternoo On April 18, the inspector again identified that the two fire doors were closed. The inspector informed the outage duty manager and questioned the adequacy of previous corrective actions. The duty manager informed the inspector that they would tie the doors open with aircraft wire and hang caution tags indicating that they must remain open to provide adequate tornado ventin The uninterruptible power supply rooms are required to vent into the hallway, and are fitted with specially designed, normally shut, tornado doors which open into the hallwa To meet fire protection requirements, these rooms are also fitted with fire doors which ,

are held open by fusible links. Because the doors had not been shutting fully, the i licensee had installed stronger door closer springs on the fire doors. However, toe j fusible links were no longer strong enough to hold the fire doors open. These doors J have been problematic since early 1997, and have had a work order or a fire impairment written against the door problems since that tim CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," requires, in part, that conditions adverse to quality are properly identified and corrected. Failure to correct the problem of the fire doors not shutting fully and failure to identify or correct the impaired tornado doors is a violation of Criterion XVI (50-445/9803-07). Conclusion l The licensee handled the issue of tornado venting and fire doors for these rooms poorl This issue is being cited due to the untimeliness and ineffectiveness of the licensee's corrective action i

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.e-25-V. Management Meetings X1 Exit Meeting Summary The inspectors presented the results of the inspection to members of licensee management on May 14,1998. Mr. Walker, the regulatory affairs manager, stated that they believed that the

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i- potential violations conceming the midloop level indications met the criteria for enforcement discretion. Mr. Walker indicated that the event was self-revealing, that they documented the problem on a ONE form in a timely manner, and that the discovery of the issues would have occurred independent of the NRC's identification. Mr. Walker also stated that they did not understand what 10 CFR 50.59 USQ criteria was violated. The inspectors acknowledged these l comments. No proprietary information was identified.

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ATTACHMENT SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED l Licensee l

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M. R. Blevins, Vice President, Nuclear Operations R. D. Bird, Jr., Nuclear Planning Manager J. J. Kelley, Vice President, Nuclear Engineering and Support D. R. Moor 4, Operations Manager R. D. Walker, Regulatory Affairs Manager INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: Plant Support Activities ,

IP 92902: Followup - Maintenance )

IP 92903: Fo!!owup - Engineering i l

ITEMS OPENED AND CLOSED Ooened 50-445/9803-01 VIO Inadequate reduced inventory operating procedure (Section 03.1)

50-445/9803-02 IFl Minimum decay heat removal flow during reduceo'

inventory operations (Section O3.1).

50-445/9803-03 VIO Failure to perform adequate engineering review which  !

contributed to partial loss of reactor vessel level indication (Section E4.1).

50-445(446)/9803-04 VIO Four examples of failure to verify or check the adequacy of design of the ECCS switchover procedure (Section E8.1).  !

50-445(446)/9803-05 VIO Failure to maintain records of safety evaluations which provided bases for the determination that the changes to the procedures for ECCS switchover did not constitute a USQ and failure to identify a USQ during a subsequent l

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evaluation (Section E8.1).

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50-445/9803-06 NCV Switchgear breakers racked out to the " remove" position l resulted in a seismically unqualified condition outside the plant design bases (Section E8.8).

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50-445/9803-07 .VIO Failure to correct the problem of the fire doors not shutung fully and failure to identify or correct the impaired tomad-)

doors (Section F2.1).-

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50-445/9714-01 IFl Reactor makeup storage tank draining evolution valve

lineup error (Section 08.1).

50-445(446)/9717 01 VIO Three examples of operations personnel failing to follow procedures (Section 08.2).

50-445/98001-00' LER Entry into Technical Specification 3.0.3 when feedwater isolation valves were declared inoperable due to low nitrogen pressure (Section 08.3).

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50-445(446)/9617-01 - VIO Failure to implement adequate corrective actions

, associated with plant transients following initial main turbine loading during plant startup (Section 08.4).

50-445(446)/9705-01' IFl Problems maintaining control room work request and instrument out of service tags current (Section 08.5).

50-446/9611-02 - IFl Steam plant pump failures (Section M8.1)..

50-445(446)/9708-02 IFl . Root cause followup on turbine-driven ' auxiliary feedwater

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steam admission valve diaphragm failures (Section M8.2).

50-445(446)/9616-06 URI Discrepancies between the operating procedures and the FSAR on the switchover of the emergency core cooling system from injection to recirculation (Section E8.1).

50-445(446)9718-04 VIO Failure to use temperature instrument with the accuracy required by procedure (Section E8.2).

, ' 50-445(446)/9718-03 . VIO Inadequate procedures for hydrogen recombiner surveillance and control room pressurization surveillance 1 (Section E8.3).

50-445/97006-00 LER Control room pressurization unit flow found outside of .

50-445/97006-01 Technical Specification limits (Section E8.4).

i 50-445/96004-00 LER Potential failure of personnel airlock control systems in both ;

units (Section E8.5). i

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50-445(446)/9801-04 IFl Followup on the regulatory compliance with diesel lt generator load shed surveillance requirements l ' (Section E8.6).~  !

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z !50-445/9612-03? VIO Failure to Meet ASME/ ANSI Code Requirements For Relief

. Valve Testing (Section E8.7).

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50-445/96009-00 LER Switchgear breakers racked out to the " remove" position l 50-445/96009-01 resulted in a seismically unqualified condition outside the .

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50-445/9803-06 NCV plant design bases. (Section E8.8).

50-445(446)/9612-02 IFl Calculation review and approval process (Section E8.9).

50-445(446)/9715-03 VIO Inadequate area temperature monitoring surveillance procedure (Section E8.10).

i 50-445(446)/9802-07 VIO Failure to control a locked high radiation area in accordance with technical specification requirements due to removal of containment access hatch locks for three days (Section R8.1).-

50-445(446)/9708-04. IFl Minimum shift manning requirements (Section P8.1).

50-445/98002-00 LER Inoperable security search detector allowed unauthorized access to the protected area (Section S8.1).

LIST OF ACRONYMS USED ALARA As Low as Reasonably Achievable ASME American Society of Mechanical Engineers

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ECCS Emergency Core Cooling System Final Safety Analysis Report

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FSAR ONE Operations and Notification Evaluation PDR Public Document Room RCS' Reactor Coolant System RCP- Reactor Coolant Pump RHR Residual Heat Removal RWST Refueling Water Storage Tank l USQ Unreviewed Safety Question l

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