ML20204D981

Insp Rept 50-298/98-09 on 981227-990206.Violations Noted. Major Areas Inspected:Operations,Maint & Engineering

ML20204D981
Person / Time
Site:	Cooper
Issue date:	03/18/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20204D949	List: ML20204D945 ML20204D981
References
50-298-98-09, 50-298-98-9, NUDOCS 9903240321
Download: ML20204D981 (15)

See also: IR 05000298/1998009

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ENCLOSURE 2

l U.S. NUCLEAR REGULATORY COMMISSION

l REGION IV

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Docket No.: 50-298 l

License No.: DPR 46

Report No.: 50-298/98-09

. Licensee: Nebraska Public Power District'

Facility: Cooper Nuclear Station j

Location: P.O. Box 98 I

Brownville, Nebraska I

Dates: December 27,1998, through February 6,1999 l

Inspectors: C. Skinner, Acting Senior Resident inspector

M. Miller, Senior Resident inspector

B. Smallridge, Resident inspector, Wolf Creek Nuclear Station

Approved By: C. Marschall, Chief, Branch C

Division of Reactor Projects

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ATTACHMENT: Supplemental Information I

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9903240321 990318

PDR ADOCK 05000298

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EXECUTIVE SUMMARY

Cooper Nuclear Station

NRC Inspection Report No. 50-298/98-09

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Operations

The inspectors identified that on two occasions operators failed to perform an operability l

determination or evaluation when required, despite test data that failed to meet '

established acceptance criteria. Additionally, one operability determination was

inadequate to ensure that the reactor equipment cooling system would continue to be

operable under all conditions. Inspector involvement was required to assure proper

licensee resolution of the potential safety aspects of these three issues (Section O2.1). l

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• Following overspeed trips of both steam turbines, the licensee identified that the fill and ;

vent procedures for the high pressure coolant injection and reactor core isolation cooling

systems were inadequate. The procedures incorrectly stated that the instructions would

vent the entira system. This is a noncited violation of Technical Specification 5.4.1

(Section O3.1).

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Maintenance

A number of equipment problems occurred following the refueling outage. Two of the l

equipment problems were caused by inadequate skill of the craft and procedural

adequacy, and four do not have a root cause identified. These equipment problems i

have not directly resulted in significant safety concerns (Section M2.1).  !

Technicians improperly installed steam supply flanges, resulting in a high pressure

coolant injection steam supply valve flange leak. Licensee investigators determined that

technicianu improperly installed the flanges because of inadequate skill of the craft

(Section M4.1).

A 1997 licensee event report documented a degraded lubricating oil pump on the diesel

generator. Maintenance personnel had incorrectly installed the pump during

maintenance caused by an inadequate procedure. The pump clearance information

from the vendor manual was not correctly incorporated into the work instructions. This

is a noncited violation of 10 CFR Part 50, Appendix B, Criterion V (Section M8.2).

Enaineerina

Engineering personnel, who wrote and reviewed an engineering evaluation on the

reactor equipment cooling time delay relays, failed to recognized that data included in

the evaluation indicated that one of the time delay relays was inoperable. Also, the

inspectors identified that the acceptance criteria for surveillance procedures written to

test the subject relays did not include repeat accuracy of the relays. The licensee wrote

a problem identification report to enter this issue into the corrective action program.

This is a noncited violation of Technical Specification 5.4.1 (Section E2.1).

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Report Details  :

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' Summarv of Plant Status

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At the beginning of the inspection period, the plant was at 100 percent power. On January 16,

l 1999, licensed operators reduced power to 70 percent to perform a control rod line adjustment.

Operators restored power to 100 percent that same day. Later, on January 16, Feedwater

Pump A tripped and power was stabilized at 70 percent. On January 23 following repairs,

licensed operators returned Feedwater Pump A to service and returned the plant to 100 percent j

power. The plant remained at 100 percent power throughout the remainder of the inspection i

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period.

1. Operations

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02 Operational Status of Facilities and Equipment

O2.1 Inadeauate Operability Determinations  !

a. Insoection Scope (71707 and 37551)

The inspectors reviewed a problem identification report that documented accuracy

problems with time delay relays for the reactor equipment cooling system, an

engineering evaluation on the same time delay relays, and an operability determination

on the failure of service water temperature control valves. Discussions were held with

operators, engineers, and management personnel. j

b. Observations and Findinos

Failure to Perform an Operability Determination or Evaluation For Relavs Outside

Acceptance Criteria

On January 18,1999, system engineers wrote Problem identification Report 4-00383 to

document that test results from all four time delay relays used for sequential loading of

heavy loads on the diesel generators during a loss of offsite power had an increasing

trend. The engineers projected that increasing timing of the relays could result in the

relays not meeting the procedure acceptance criteria by the end of a cycle. The system

engineers further documented that the acceptance criteria was i 10 percent, and the

accuracy of the time delay relays was also * 10 percent. The engineering supervisor

who reviewed the problem identification report documented a need for an operability

determination and an engineering evaluation. However, operators had indicated in the

operation's review section of the report that no operaLuity determination nor operability 1

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evaluation were needed because the issue was programmatic.

Technical Specifications Surveillance Requirement 3.8.1.10 requires verification that the

interval between each sequenced load is to be within 10 percent of nominal timer

i setpoint. The nominal timer setpoint was 20 seconds, based on the Updated Safety

! Analysis Report. Calibration Procedure 6.1(2) REC.302," REC Pumps Time Delay Relay

i Testing and Setting," established a nominal setpoint of 20 seconds with an acceptance

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criteria of 18 to 22 seconds.

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The inspectors reviewed the problem identification report and determined that, if the

acceptance criteria and the accuracy of the time delay relays were the same, then the

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only acceptable as-left value would be exactly 20 seconds. If the relay was not set at I

20 seconds, the accuracy listed in the vendor documentation indicated to the inspectors l

that there was no assurance that the relay would meet the Technical Specification i

requirement of 110 percent the next time the relay actuated. The inspectors identified  ;

that none of the four relays had been set at exactly 20 seconds. Therefore, the  ;

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inspectors determined that an operability determination or evaluation was needed. The

operators issued another problem identification report documenting the inspectors' l

concern and performed an operability determination. This demonstrated the time delay

relays were operable based on an evaluation using historical as-found test data. The

licensee showed that, historically, the repeat accuracy was 15 percent and this better

accuracy was corroborated by verbal information from the vendor. The inspectors

reviewed the operability determination and no other concerns were identified.

Failure to Comolete an Operability Determination or Evaluation When Reauired l

On January 21, the inspectors identified that the time delay relay for Reactor Equipment

Cooling Pump D appeared to be inoperable based on information in Engineering

Evaluation 1999-002. The details of this example were documented in Section E2.1 of

this inspection report.

Inadeauate Operability Determination

On January 31, Reactor Equipment Cooling Heat Exchanger B Outlet Valve SW-TCV-

451B failed to control temperature as designed. Operators placed the valve in local

manual centrol. Licensee personnel developed Operability Determination 4-00603,

dated February 1. In this document, engineers concluded that the valve was operable

provided that a minimum service water flow through the reactor equipment heat

exchanger of 400 gallons per minute was established and maintained. The flow was

established by manually throttling Valve SW-TCV-451B.

The inspectors reviewed the safety function of Valve SW-MOV-651, another valve in the

system downstream of Heat Exchanger B. Valve SW-MOV-651 performs an automatic

function on a Group 6 isolation. The valve's function is to ensure that sufficient flow is

available through Heat Exchanger B and the residual heat removal service water system

simultaneously. The valve is designed to reposition to 10 percent open if the valve was

less than 10 percont open prior to the Group 6 isolation. The valve fails in its prior

position if the valve was greater than 10 percent open prior to the group isolation.

Emergency procedures direct the operators to manually reposition the valve to

10 percent open after 10 minutes. This position provides a flow of 400 to 1200 gallons

per minute to Heat Exchanger B provided that Valve TCV-451B is fully open,

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The inspectors determined that, if the temperature control valve were failed in any

position other than fully open, then the flow through the heat exchanger may be less

than 400 gallons per minute. Depending on valve position, operator action may be

required to open Valve TCV-451B following a Group 6 isolation. Operators identified

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that emergency procedures addressed the need to fully open the temperature control

valve, but were not clearly indexed for operator use and should be more clearly '

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referenced.

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l The licensee concluded that the safety significance of this finding was low, because the

temperature control valve had not been closed sufficiently to allow significant reduction

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in service water flow. Licensee managem'ent acknowledged the need to have

j performed a more thorough operability determination and documented the concern in a

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problem identification report. A night order was issued to document this concern and

i reference portions of relevant procedures in the case of an emergency. Procedure

! changes were in process to more clearly reference system operating constraints under

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emergency conditions. Operators found that procedures adequately constrained system l

configuration once they were clarified, and an operability determination was no longer

necessary.

c. Conclusion

The inspectors identified that on two occasions operators failed to perform an operability

i determination or evaluation when required, despite test data that failed to meet

established acceptance criteria. Additionally, one operability determination was

inadequate to ensure that the reactor equipment cooling system would continue to be

operable under all conditions. Inspector involvement was required to assure proper

licensee resolution of the safety aspects of these three issues. 4

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l 03 Operations Procedures and Documentation l

03.1 Failure of Hiah Pressure Coolant Iniection (HPCI) and Reactor Core Isolation

Coolina (RCIC) Systems Tests Caused by inadeauate System Ventina

a. Inspection Scoce (71707)

On December 17,1998, the control room crew performed required surveillance testing

of the RCIC system. The RCIC turbine tripped on overspeed 26 seconds after starting. 1

Licensee technicians performed troubleshooting on the overspeed emergency trip

spring tension and made minor adjustments. The test was performed again with similar

results. During the next shift, licensed operators performed testing of the HPCI system,

and the pump turbine tripped on overspeed less then 1 minute after starting.

The inspectors reviewed the licensee's actions and evaluation in response to these

, failures. Discussions were held with operators, engineers, and management personnel.

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• Surveillance Procedure 6.RCIC.309, "RCIC (150 PSIG) Bepbning of Cycle Test"

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• Surveiiianca Pmecoure 6.HPCI.313,"HPCI (160 PSIG) L. ginning of Cycle Test"

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• Procedure 2.2.33, "High Pressure Coolant Injection System," Revision 46

l * Procedure 2.2.67, " Reactor Core isolation Cooling System," Revision 4

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b. Observations and Findinas

The HPCI and RCIC systems have common suction lines from the emergency

condensate storage tanks. Operators had drained both systems to facilitate

maintenance during the refueling outage. Upon completion of the maintenance

activities, ope rators had vented the common suction line in accordance with

Procedures ft.2.33 and 2.2.67. However, these procedures did not provide sufficient

guidance to vent the entire suction line. Following the pump failures, operators vented

the systerr.s again, but this time they used additional vent valves on the suction line that

were not 'Jsted in the procedures. Af ter the venting process was complete, the

operators repeated the surveillance testing, and both systems performed satisfactorily.

Licensee personnelinitiated Problem identification Report 3-53339. Through '

investigation, they determined that the root cause was inadequate procedures. The

system operating procedures did not require the operators to vent the suction lines from

the emergency condensate storage tanks. The procedures documented that the

instructions would vent the entire system, but following the procedural steps mly

resulted in venting the discharge piping. This procedural inaoequacy is ir, uolation of

Technical Specification 5.4.1 (50-298/98009-01). However, this Severity Level IV

violation is being treated as a noncited violation, consistent with Appendix C of the NRC

enforcement policy. This violation is in the licensee's corrective action program as

Problem Identification Report 3-53339.

Additionally, the inspectors identified that the operators did not formally control the

system configuration when the additional valves were cycled to vent the suction line.

The only procedure addressing formal control of valve configuration applicable to this

circumstance was Procedure 0.9, " Tagging Orders," Revision 22c3. Steps 2.3.1.2 and

8.4.2 required the use of caution tags for identifying and tracking components that were

not in their normal position and were not being controlled by any other station procedure

or process.

Operator's could have initiated a revision to correct the fill and vent procedures to

include the three valves prior to implementing the procedures again. The operators did

not use any alternative method to control or document that the valves were opened

and/cr closed. The failure to use a formal method of control of the system configuration

such as a caution tag order or another method of administrative control and failure to

document the specific valve manipulations were examples of informal control of

configuration and less than rigorous immediate corrective actions. The valves were

restored to the correct position after venting. Therefore, safety significance was minor.

c. Conclusion

Following overspeed trips of both steam turbines, the licensee identified that the fill and

vent procedures for the high pressure coolant injection and reactor core isolation cooling

systems were inadequate. The procedures incorrectly stated that the instructions would

vent the entire system. This is a noncited violation of Technical Specification 5.4.1. The

inspectors identified that, while appropriately venting the systems, operators failed to

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follow a formal process to control the system configuration when three valves were

manipulated, although two formal processes were available. The safety significance of

this action was minor.

08 Miscellaneous Operations issues (92700 and 92901)

08.1 (Closed) Licensee Event Report 50-298/98-012: Overspeed trip of HPCI and RCIC

systems during the 150 psig vessel pressure test. This event was discussed in

Section O3.1 of this inspection report and is closed.

08.2 (Closed) Inspector Followuo item 50-298/97003-04: continuing problems with

procedural adherence and adequacy. In 1997, during the review process for closing

Violations 50-298/95008-01 and 50-298/95017-02, written for procedural adherence and

adequacy issues, the inspectors identified a continuing problem in these shme areas.

This followup item was written to evaluate the licensee's actions to address procedure

performance problems.

The licensee had opened an item in their tracking system to ensure that the issues

would be addressed. This item was still open at the end of the current inspection period

with a due date of March 6,1999. The due date for this open item has been extended

five times since the original due date of June 30,1997.

NRC Inspection Reports 50-298/98-07 and 50-298/98-08 issueo December 11,1998,

and January 22,1999, respectively, documented procedural adherense problems.

Violation 50-298/98008-01 required a response to state the licensee's corrective actions. ,

The licensee's actions to address procedural adherence issues will be reviewed during  ;

the closure process for the subject violation. Therefore, the issue regatiing procedural  !

adherence addressed by this inspector followup item is closed.

Inspectors routinely review licensee's procedures to determine their adequacy to

facilitate safety-related activities in all functional areas. These inspection actaities are

included and directed by the core inspection program. As such, there is no need to I

specifically track an issue to evaluate the licensee's overall program for evaluation and

improvement of procedural adequacy. Therefore, this inspection followup item is

administratively closed.

11. Maintenance ,

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M1 Conduct of Maintenance j

l M1.1 General Comments

a. Insoection Scope (62707 and 61726)

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The inspectors observed the performance of various maintenance activities. The

following procedures and work packages were reviewed:

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Maintenance Work Request 98-2622 Emergency Lighting Upgrade

Surveillance Procedure 6.HPCI.103 High Pressure Coolant injection in-service  ;

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Test and 92-Day Test Mode

Surveillance Procedure 6.RCIC.705 Reactor Core isolation Cooling Turbine

High Exhaust Pressure Channel Functional

Test

Surveillance Procedure 15.RF.601 Reactor Feedwater Pump Turbine

Overspeed Test

Preventive Maintenance 02125 Reactor Pressure Local Gage Preventive l

Maintenance .

Preventive Maintenance 08874 Reactor Core isolation Cooling Steam

Supply Valve Mechanical Examination

Preventive Maintenance 08905 Reactor Core Isolation Cooling Steam

Supply Valve Electrical Examination

b. Observations and Findinas

During direct observation and/or review of the maintenance documentation, the

inspectors did not identify any significant negative or positive findings. The technicians i

followed work instructions, equipment was properly tagged out of service, and the

correct Technical Specifications were entered when required. Health physics and

security requirements were followed.

c. Conclusion

Routine observation of maintenance activities were properly conducted.

M2 Maintenance and Material Condition of Facilities and Equipment

M2.1 Eautoment Problemq )

a. Inspection Scope (62707)

The inspectors noted a number of equipment pioblems that had occurred since plant

startup from the recent refueling outage. Discussions were held with operators,

engineers, maintenance technicians, and management personnel,

b. Observations and Findinas

l A number of plant equipment problems were identified by inspectors, the licensee,

and/or were self-revealing. The more significant problems included:

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l * The inspectors noted numerous lubricating oil leaks on the diesel generator

engines that appeared to have existed for an extended period. Oil and dust had

, built up on the lower portions of the engines and on the adjacent lubricating oil 1

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system piping, along with an accumulation of old rags, strips of plastic, old )

absorbent pads, and small articles of trash. The housekeeping associated with I

these less accessible areas of the diesel generators was inadequate. The

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l licensee issued work requests to repair the oil leaks. The direct safety I

significar> w was low because the combustible loading was bounded by the

100 gallon transient combustible loading analysis.

• The temperature recorder for Safety Relief Valve 71HRV recorded 13 spikes

from 110 degrees to 150 degrees (maximum) from December 20-28,1998. The

temperature indication then stabilized at 115 degmes. The Safety Relief l

Valve 71GRV temperature recorder indicated 125 degrees, rose to 190 degrees I

for 6 days, then decreased to and stabilized at 140 degrees. The temperature '

recorder for Safety Relief Valve 71FRV increased from 165 to 200 degrees and

continued to indicate 200 degrees through the remainder of the inspection

period. Licensed operators continued to monitor the recorders and enginearing

personnel planned to correlate the temperature to a leak rate and then establish

an upper temperature limit. The root cause for the temperature indications were

unknown rt the end of the inspection period.

• The licensee identified that both service water outlet valves for the residual heat

removal heat exchanger, Valves SW-MOV-MO89A and SW-MOV-MO89B,

leaked and could not isolate flow. Similar leakage has historically resulted in silt

intrusion into the residual heat removal service water system and the residual

heat removal heat exchangers. Licensee maintenance technicians had rebuilt

both valves in the last two outages. While in an operational mode, operators

have run the service water booster purrps for 15 minutes weekly and have

monitored the discharge pipe temperature to avoid silt buildup. Also, licensee

personnel have monitored the valve body thickness to detect erosion.

• Maintenance personnel rebuilt Sump Pumps G 1 and G-2 during the last

refueling outage. These pumps provide indication of the identified equipment

leakage in the drywell, as required by Technical Specifications. Af ter the outage,

the licensee identified that Sump Pump G-1 failed to operate properly. On one

occasion, the pump failed to automatically operate on demand. When the pump

was manually started, the running amperage was higher than normal and tne

flow rate was lower than normal. Near the end of the inspection period, the

pump tripped on electrical overloads immediately after a start signal. Outage

management personnel added work on the pump to the forced outage list. The

root cause for the Sump Pump G-1 problem was unknown at the end of the

[ inspection period.

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= The licensee identified a steam leak on H@h Pressure Coolant injection Steam

I Inlet Valve HPCI-MOV-MO 14. The licennee a evaluation determined that the

workers failed to align the flanges during rcassembly, causing the tongue flange

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to improperly fit into the groove flange. The licensee replaced the gasket and ,

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reassembled the flange. This was discussed more fully in Section M4.1 of this l

inspection report.

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The control rod drive pump lower bearing failed because the inboard bearing oil

bubbler was improperly replaced after oil addition. The licensee replaced the

pump and injected too much grease into the coupling, increasing pump vibration.

The grease was removed and the vibration returned to normal. The root causes

for these problems were inadequate training for operators and an inadequate

procedure, respectively.

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Reactor Recirculation Pump B and Reactor Feedwater Pump B have indicated

intermittent high vibrations. Licensee perse,1nel determined that the vibration

probes for both of these pumps had fr.ileo or intermittently given false indications

of high vibrations. The licensee concluded that vibrations were actually normal. l

Following a decrease in plant power, a low flow annunciator for Reactor

Recirculation Pump B Seal Cavity 2 alarmed. The seal cavity pressure

fluctuated between 350 to 600 psig and the temperature dropped 2 degrees.

When reactor power was restored to 100 percent, seal cavity pressure and

temperature returned to normal indications. The licensee continued to monitor

the pressure and temperature. The licensee ordered new seal parts and

scheduled the seal work in their forced outage schedule.

On January 16,1999, Feedwater Pump A tripped on an overspeed trip signal.  ;

Maintenance personnel determined that the overspeed device had a defective i

part. Additionally, during troubleshooting, technicians identified that the main '

pump bearing was worn. Maintenance personnel replaced the overspeed device .

and main pump bearing. A contractor using sensitive vibration equipment l

monitored the main pump bearing when the feedwater pump was returned to

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service.

c. Conclusion

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A number of equipment problems occurred following the refueling outage. Two of the

equipment problems were caused by inadequate skill-of-the-craft and procedural

adequacy, and four do not have a root cause identified. These equipment problems

! _have not directly resulted in significant safety concerns.

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M4- Maintenance Staff Knowledge _and Performance

M4.1 Hiah Pressure Coolant Iniection Steam Supolv Valve Flanae Leak j

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a. Insoection Scope (62707)

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l The inspectors reviewed the licensee's evaluation for a leak that developed on the high

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pressure coolant injection steam supply valve during a surveillance test. Discussions t

were held with maintenance and engineering personnel and with licensee management. I

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b. Observations and Findinas

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On December 18,1998, operators identified a steam leak on High Pressure Coolant

Injection Steam Supply Valve HPCI-MO-14 during a surveillance test. The leak

developed on the downstream flange of the steam supply valve. The control room

- operators tripped the high pressure coolant injection turbine and isolated the leak.

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! The cause of the steam leak was personnel error by maintenance personnel. The

steam leak occurred because of a misalignment of the outlet flange of the steam supply

L valve and the piping flange. Licensee reviewers determined that the misalignment was

attributed to inattention to detail during the flange alignment process. The work that

resulted in the misalignment was performed during the recent refueling outage.

Licensee personnel incorporated this issue into the lesson learned for future refueling l

, outages, general employee training, continuing training for maintenance personnel, and

l initial training for contractor craft personnel. Corrective actions taken, but not ,

. documented in the problem identification report, included placing guidance into the i

maintenance writer's guide.

c. Conclusion

Technicians improperly installed steam supply flanges resulting in a high pressure .

coolant injection steam supply valve flange leak. Licensee investigators determ'aed that

technicians improperly installed the flanges because of inadequate skill of the craft.

'M8 ' Miscellaneous Maintenance issues

M8.1 (Closed) Inspection Followuo item 50-298/98005-02: diesel generator lubricating oil j

. pump failures. In paragraph M2.2 of NRC Inspection Report 50-298/98-05, the

inspectors determined that Emergency Diesel Generator 1 was degraded because

maintenance personnel had failed to correctly set the engine-driven lubricating oil pump

clearances. The inspectors opened an inspection folbwup item to determine the extent

of the degradation. Licensee personnel determined that the degradation of Emergency

Diesel Generator 1 was limited to the engine-driven lubricating oil pump, and the

inspectors found this reasonable. This issue was discussed further in Section M8.2 of

this inspection report. No additional followup is required.

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M8.2 (Closed) Licensee Event Reood 50-298/97015-00 and -01: Diesel Generator 1 was

inoperable because of a degraded lubricating oil pump. In Revision 1 of this licensee

event report, the licensee documented that Emergency Diesel Generator 1 failed to slow

start because the engine-driven lubricating oil pump had been installed with incorrect

- clearances between the pump idler assembly and the pump head plate. Subsequently,

a fast start of the emergency diesel generator was conducted, as required by the .

Technical Specifications. Therefore, no operability concern with Emergency Diesel

Generator 1 existed.

However, the incorrect installation of the engine-driven lubricating oil pump and the

l. failure of the emergency diesel generator to slow start resulted from the use of

inadequate maintenance procedures. Pump clearance information from the vendor

manual was not correctly incorporated into the work instructions. The failure to ensure

that design requirements were appropriately translated into work instructions is a

violation of 10 CFR Part 50, Appendix B, Criterion V (50-298/98009-02). This Severity

Level IV violatiori is being treated as a noncited violation, consistent with Appendix C of

the NRC enforcement policy. This violation is in the licensee's corrective action program

as Problem Identification Report 2-25033.

M8.3 (Closed) Violation 50-298/97010-01: a procedure used inappropriate methodology in

L determining if the acceptance criteria was met. The inspectors verified the corrective

actions described in the licensee's response letter dated February 11,1998, were

reasonable and complete. No similar problems were identified.

l M8.4 (Closed) Violation 50-298/97011-03: a procedure allowed the installation of

nonessential fuses in essential fuse applications. The licensee previously provided an

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operability evaluation that concluded that the nonessential fuses that had b'aen installed .

in essential applications were operable. Licensee personnel changed the fuse

dedication and usage program. The use of nonessential fuses in essential fuse

applications now requires an engineering review. The inspectors found the corrective

actions described in the licensee's response letter, dated April 8,1998, to be reasonable

and complete.

M8.5 (Closed) Licencee Event Report 50-298/98011-00: steam leak at the high pressure

coolant injection turbine valve caused by a misalignment of the flanges. This event was

discussed in Section M4.1 of this inspection report.

Ill. Enaineerina

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j. E2 Engineering Support of Facilities and Equipment

E2.1 - Inadeauate Procedures for the Reactor Eauioment Coolina System Time Delav Relavs

. a. Inspection Scope (37551)

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j As documented in Section O2.1 of this inspection report, on January 21, the inspectors

I identified that the time delay relay for Reactor Equipment Cooling Pump D appeared to

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be inoperable based on information in Engineering Evaluation 1999-002. The

! inspectors reviewed Engineering Evaluation 1999-02, the vendor manual, and

Procedures 6.1 REC.302 and 6.2 REC.302, " REC Pumps Time Delay Relay Testing." In  !

addition, the inspectors held discussions with engineers and management personnel.

b. Observations and Findinos

in response to Problem Identification Report 4-00383, engineers wrote Engineering

Evaluation 1999-002 to address the failure of the reactor equipment cooling pump time

delay relays to stay within operability limits. As previously stated, engineers had i

concluded that a repeatability of 5 percent of the previous setting was more I

appropriate than the * 10 percent. Engineers had documented that the calibration l

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procedure should be revised to recalibrate the relays if the as-found value was outside

the tolerance band of 19.5 - 20.5 seconds. This would ensure a repeatability of

• 5 percent and that the next as-found values would be within Technical Specification

tolerances.

The inspectors reviewed the as-left test values for the four reactor equipment cooling

pump time delay relays documented during the last refueling outage. The inspectors l

noted that the as-left value for the Pump D time delay relay was 18.92 seconds. The

inspectors discussed this fin ung with the shift technical engineer and the system

engineer. A problem identification report was written to document that the relay had

been left outside the previously established tolerance band, and the relay was

conservatively declared inoperable. A clearance order was issued to disable the pump

from starting after a loss of power and subsequent restoration of power to the critical

bus. Licensee technicians recalibrated the time delay relay within the calibration

tolerance, declared the relay operable, and returned the system to a standby alignment.

During the research and review of the evaluation, the engineers had failed to identify

that one of the as-left values was outside the * 5 percent and operability was therefore

questionable. However, evaluation of the specific relay indicated th t it had remained

operable.

Dui;ng their review, engineers had not included all appropriate vendor information in

Procedures 6.1 REC.302 and 6.2 REC.302. The vendor manualincluded information on

the relay repeat accuracy. However, the test procedures did not use or include this

information in determining the required as-left values. The failure to maintain

Procedures 6.1 REC.203 and 6.2 REC.302 to include information on repeat accuracy is a

violation of the Technical Specification 5.4.1 requirement that procedures appropriate to

the circumstances be established (50-298/98009-03). Licensee personnel recalibrated

the relay, entered the concern into the corrective action system, and planned activitier

that included appropriate procedure changes and evaluations. This Severity Level IV

violation is being treated as a noncited violation, consistent with Appendix C cf the NRC

! enforcement policy. This violation is in the licensee's corrective action program as

Problem identification Report 4-00654.

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c. Conclusion

Engineering personnel, who wrote and reviewed an engineering evaluation on the

reactor equipment cooling time delay relays, failed to recognize .I that data included in

the evaluation indicated that one of the time delay relays was l* c perable. Also, the

inspectors identified that the acceptance criteria for surveillano procedures written to

test the subject relays did not include repeat accuracy of the ! Jaya. The licensee wrote I

a problem identification report to enter this issue into the corrective action program. l

This is a noncited violation of Technical Specification 5.4.1. '

E8 Miscellaneous Engineering issues

E8.1 (Closed) Violation 50-298/97011-06: inadequate surveillance requirement for the i

standby liquid control system relief valves. NRC Inspection Report 50-298/97-11 l

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documented that the specific corrective actions for this violation were adequate. The

licensee's corrective action for the general problem of failure to promptly identify and

correct conditions adverse to quality will be evaluated as part of the closure for

Escalated Enforcement Action 97-424 (NRC Inspection Report 50-298/97-12). ,

T herefore, this item is administratively closed.  !

VI. Manaaement Meetina j

Exit Meeting Summary

The inspectors presented the inspection results to members of licensee management at the

conclusion of the inspection on February 11,1999. The licensee acknowledged the findings

presented. The inspectors asked the licensee whether any materials examined during the

i.1spection should be considered proprietary. No proprietary information was identified.

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! ATTACHMENT

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PARTIAL LIST OF PERSONS CONTACTED

Licensee

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' O. Buman, Assistant Matager, Plant Engineering Department j

l 'J. Burton, Performance Analysis Department Manager . j

P. Caudill, Senior Manager Technical Services .  !

z T Chard, Radiological Manager - j

L. Dewhirst, Licensing Engineer  !

J. DeWitt, Maintenance Engineering

P. Donahue, Engineering Support Manager i

C. Fidler, Assistant Maintenance Manager- )

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T. Gifford,' Design Engineering Department Manager  !

! M. Gillan, Assistant to Plant Manager  !

J. McMauam, Work Control Supervisor l

L L. Newman, Licensing Manager l

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M. Peckman, Plant Manager

B. Rash, Engineering Manager

A. Shiever, Operations Manager

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.lNSPECTION PROCEDURES USED -

IP 37551: Onsite Engineering .

IP 61726: Surveillance Observation

IP 62703: Maintenance Observation

IP 71707: Plant Operations

IP 71750: Plant Support Activities

IP 92901: - Followup - Plant Operations

IP 92902: Followup - Maintenance

IP 92903: Followup - Engineering

IP 92700: LER - Onsite Review

IP 93702: Onsite Response  ;

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ITEMS OPENED AND CLOSED

Opened an' d Closed i

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50-298/98009-01 NCV Inadequate venting procedure for the high pressure coolant

injection and the reactor core isolation cooling systems ,

(Section 03.1). I

'50-298/98009-02 NCV Failure to have adequate work instructions for the diesel

generator lube oil pump (Section M8.2). )

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50-298/98009-03 VIO - Inadequate procedures for the reactor equipment cooling system

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j. time delay relays (Section E2.1).  ;

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Closed

50-298/98-012 LER . Overspeed trip of high pressure coolant injection and reactor core

isolation coolant systems during the 150 psig vessel pressure test

(Section 08.1).

50-298/97003-04 IFl Licensee continuing problems with procedural adherence and

adequacy (Section 08.2).

50-298/98005-02 IFl Diesel generator lubricating oil pump failures (Section M8.1).

50-298/97015-00

50-298/97015-01 LER Diesel Generator 1 inoperable due to degraded lubricating oil

pump (Section M8.2).

50-298/97010-01 VIO Procedure used inappropriate methodology in determining if

acceptance criteria was met (Section M8.3).

50-298/97011-03 VIO Procedure allowed installation of nonessential fuses in essential

fuse applications (Section M8.4).

.50-298/98011-00 LER Steam leak at the high pressure coolant injection turbine valve

caused by a misalignment of the flanges (Section M8.5).

50-298/97011-06 VIO Inadequate surveillance requirement for the standby liquid control

system relief valves (Section E8.1),

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