ML20126B852

Insp Rept 50-298/92-22 on 921004-1114.Violations Noted. Major Areas Inspected:Operational Safety Verification, Surveillance Observations & Licensee Event Repts

ML20126B852
Person / Time
Site:	Cooper
Issue date:	12/14/1992
From:	Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20126B835	List: ML20126B834 ML20126B846 ML20126B852
References
50-298-92-22, NUDOCS 9212220245
Download: ML20126B852 (15)

See also: IR 05000298/1992022

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APPENDIX B

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Inspection Report:

50-298/92-22

Operating License:

DPR-46

Licensee:

Nebraska Public Power District

P.O. Box 499

Columbus, Nebraska

68602-0499

Facility Name:

Cooper Nuclear Station

Inspection At:

Brownville, Nebraska

Inspection Conducted: October 4 through November 14, 1992

Inspectors:

R. A. Kopriva, Senior Resident Inspector

W. C. Walker, Resident Inspector

J. M. Keeton, Operator Licensing

Approv '

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. Gagliardo,' liTeT7T jects section C

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Inspection Summar_y

Areas Inspected:

Routine, unannounced inspection of onsite response to

events, operational safety verification, surveillance observations, followup,

and onsite review of licensee event reports.

Resulu:

Overall, the licensee operated the facility safety (paragraphs 2 and

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3.5).

The licensee's evaluation and corrective actions to address the water

hammer event in Residual Heat Removal System B on October 22,-1992, were

prompt and appeared to be good (paragraph 2),

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Housekeeping was improving.

Licensee management was addressing this

issue (paragraph 3.2).

A compressed gas cylinder was not properly controlled on the refueling -

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floor for an extended period of time.

This is a violation

(paragraph 3.2),

One example of improper control of visitors was identified. This is a

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violation (paragraph 3.4).

Surveillance tests were performed well.

The licensee personnel involved

were knowledgeable of the tasks required and their actions were good

(paragraph 4.3).

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The maintenance activity to repair and inspect the faulty diesel

generator fuse holders was good (paragraph 5.1).

The licensee appropriately addressed, from a safety perspective, the use

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of a process can in the spent fuel pool (paragraph 6.3).

Licensed operator training weaknesses were observed in command, control,

and communications; however, the licensee was aware of the problems and

was actively pursuing their corrective actions program.

The simulator

evaluators were very professional and exhibited good evaluation skills.

Examination material was very good and in accordance with the standard.

The licenset operators appeared to be safety-conscious and competent

(paragraph 6.5).

Summar.y of Inspection Findings:

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Violation 298/9222-01 was opened (paragraph 3.2).

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Violation 298/9222-02 was opened (paragraph 3.4).

Inspection Followup Item 298/9034-02 was closed (paragraph 6.1).

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Unresolved Item 298/9219-01 was closed (paragraph 6.2).

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Unresolved Item 298/9219-02 was closed (paragraph 6.3).

Licensee Event Reports92-008, 92-012, and 92-013 were closed

(paragraph 7).

Attachments (and/or Enclosures):

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Attachment 1 - Persons Contacted and Exit Meeting

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DETAILS

1 PLANT STATUS

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At the beginning of this inspection period, the plant was operating at

53 percent power and in single-loop operation. On October 1, 1992, Reactor

Recirculation Motor-Generator Set B had tripped due to a faulty resistor and

two faulty diodes.

The components were replaced and the motor-generator set

was restarted.

The unit returned to full power on October 5.

At the end of

this inspection, the plant was operating at 100 percent power.

2 ONSITE RESPONSE TO EVENT (93702)

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Residual Heat Removal System B Inoperable

On October 22, 1992, Residual Heat Removal System B was declared inoperable

during performance of Surveillance Procedure 6.3.5.1, "RHR Test Mode

Surveillance Operation Quarterly Inservice Test," Revision 35.

During the surveillance, Residual Heat Removal Pump B was run, determined to

be acceptable, and shut down.

Pump D was then aligned according to the

procedure, which took approximately 5 minutes.

Upon the starting of Pump D, a

loud noise was heard.

The licensee investigated the source of the noise and

located a leak on the 958 foot elevation of the reactor building, at the

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flange for Pressure Maintenance System Check Valve 19. The check valve was

located in a 4-inch line which is part of the auxiliary ccndensate system,

which provides pressure maintena e o r the residual heat removal system.

Approximately 50 gallons of wat e had leaked out of the system into the

reactor building.

Licensee empsoyees bserved that the bonnet gasket on Check

Valve 19 was unseated.

They pr3ce h

to walk down the remainder of the

pressure maintenance system and (Qsm ed two pipe supports which had been

deformed from-the event and als seve al pipe hangers which were misaligned.

The licensee determined that the n.

<et faih re and pipe damage were caused by

a water hammer.

The licensee reviewed past water hammer events that have occurred in boiling

water reactors, conducted system walkdowns, and assov.ed the impact the water

hammer had on the residual heat removal system. The licensee repaired the

pipe supports that were damaged and the check valve which was found to be

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leaking due to the ';ent.

The check valve was functionally -tested and found

to be satisfactory.

Documentation was provided which showed that the event

had not compromised the system pressure boundary integrity in its repaired

configuration.

The licensee determined that the event was caused by valving out the pressure

maintenance system when switching over from Pump B to Pump D during the

surveillance test.

A procedure change had been made which requires that the

pressure maintenance system remain in service during pump changeover.

The

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inspectors reviewed the licensee's corrective actions and found them

appropriate.

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Conclusions

The licensee's evaluation and corrective actions were prompt and appeared to

be good.

3 OPERATIONAL SAFETY VERIFICATION (71707)

3.1 Control Room Observations

The inspectors observed operational activities throughout this inspection

period to verify that proper control room staffing and control room

professionalism were maintained.

Control room shift supervisor log book, tag

out log book, and control room balance-of-plant log book entries were reviewed

to verify that appropriate entries were made. The licensee's control of these

activities was good.

3.2 Plant Tours

The inspectors toured various areas of the plant to verify that proper

housekeeping was being maintained.

Housekeeping was found to be improving,

but su4e areas remained where additional improvement was needed.

The

licensee's increased efforts for improving housekeeping were evident and

management was continuing to review this activity.

On October 5, the inspectors found an unsecured, wheeled fire extinguisher in

the reartor building on the 958-foot elevation and questioned the licensee as

to what effect a seismic event would have on the unsecured fire extinguisher.

Approximately 15 feet separated the fire extinguisher cart from Fuel Pool

Cooling Instrument Rack 25-16 containing essential equipment. The licensee

performed a seismic analysis to determine whether the subject fire

extinguisher could have interacted with essential equipment.

The analysis

concluded that it would be unlikei) that the extinguisher would topple during

a seismic event. However, if it did tip over, there was no essential

eauipmerit located where it could interact with the extinguisher.

As a conservative measure, the licensee secured the extinguisher.

In

addition, the licensee reviewed six other wheeled fire extinguisher locations

within the plant to determine possible interaction of those extinguisbers with

essential equipment.

The licensee concluded that no concerns existed with the

six other wheeled fire extinguishers. The inspectors reviewed the licensee's

actions and considered them to be appropriate.

On October 6,1992, during a walkdown of the reactor building, the inspector

identified a gas cylinder in the northwest quadrant of the refueling floor

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which was roped to the two-wheel cart used for transporting the gas cylinder.

The gas cylinder was not secured to a fixed restraint, the cart was not a

wheeled cart of approved design for storage or use, and the wheels of the cart

were not blocked or locked. At the time of discovery the inspector could not

identify a use for the cylinder or the status of the cylinder (i.e., whether

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it was full or empty). Under certain conditions, the cylinder could become a

missile and damage equipment or personnel on the refueling floor or equipment

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in the fuel pool.

The licensee determined that the gas cylinder was helium and that it had been

used on April 11 to leak test the reactor pressure vessel . surveillance

specimen shipping cask in accordance with Special Procedure 92-022.

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special procedure did not include specific precautions or instructions for-

handling, storage, or removal of the gas cylinder- The licensee removed the

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gas cylinder from the refueling floor.

The protective cap was in place on the

cylinder and it was partially, if not completely, depressurized.

Title 10 CFR Part 50, Appendix B, Criterion V, states that activities

affecting quality shall be prescribed by documented instruction,

3rocedures,

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or drawings of a type appropriate to the circumstances and shall

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accomplished in accordance with these instructions, procedures, or drawings.

Procedure 0.7, Revision 8, " Flammable, Combustible, and Chemical Material

Control," paragraph-8.3.2.2.0, states that, during storage and use, gas

cylinders shall be individually secured to a fixed suppor.t by a restraint, and

paragraph 8.3.2.3 states that use of wheel-mounted carts of approved design

are permitted for certain uses of gas cylinders. .The helium gas cylinder had

been on the refueling floor since approximately April 11 and on October 6 was

not secured to a fixed support and was not on a wheel-mounted cart of approved

design.

This is a violation.(298/9222-01).

3.3 Radiological Protection Observations

The inspectors verified that selected radiological protection activities were

in conformance with facility policies, procedures, and regulatory

requirements. Radiation and/or contaminated areas were properly posted and

controlled.

3.4 Security Program Observations

On October 5, the inspectors observed a repairman, with a visitor's badge, on

the_ first floor of the administration building, in a room with two separate

access points, and he was not within the line of sight of his escort.

One

access point would have allowed the repairman to leave the-work area-unseen by

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the escort and obtain access to other areas within the protected area. ..The-

inspectors asked the repairman about his escort. The repairman thought he

could identify his escort, but was uncertain where the escort was. The

inspectors located the escort. The escort had assumed that the access door

leading from the work room to other areas within the plant was closed. After

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being questioned by the inspectors,=the door was closed.

However, there was

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no way to lock this door which would prevent the repairman from exiting

unobserved. The inspectors promptly reported the situation ~to station

security and a security officer was dispatched to review the situation.

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The inspectors reviewed the licensee's escort training and training

documentation.

The escort training lesson plan and Visitor / Tour Station

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Access Procedure 1.15 provided instructions to escorts to maintain positive

control of visitors.

The individual responsible for escorting the repairman

had received the training. The licensee counseled the individual responsible

for escorting the visitor to ensure understanding of proper escort procedures,

On October 5, Security Event Report 92-224 was completed, which outlined the

event details.

immediate corrective actions included providing an escort for

the repairman and sending a security guard to the incident location to review

the situation.

The licensee also counselled the individual, emphasizing

instructions regarding visitor control requirements. The licensee was

reviewing the procedures to determine their adequacy, and long-ierm corrective

actions had not been established at the end of this report period.

Title 10 CFR 50.34(c) requires that each application for a license to operate

a production or utilization facility shall include a physical security plan.

The Cooper Nuclear Station Physical Security Plan, Section 1.5.2, requires

that escorts exercise and maintain control of their visitors at all times.

Cooper- Nuclear Station Operations Manual, Plant Services Procedure 1.15,

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" Visitor / Tour Station Access," Revision 8, Section 4.2.1, states that an

escort is responsible to exercise and maintain control of the visitor at all

times,

The failure to exercise and maintain control of a visitor (i.e., an

individual not authorized by the licensee to enter protected areas without an

escort) while the visitor was working within the protected area on October 5,

1992, is a violation of NRC requirements (298/9222-02).

3.5 Conclusions

Overall, the licensee operated the facility safely,

Housekeeping was improving.

Licensee management was addressing this

issue.

A compressed gas cylinder was not properly controlled on the refueling

floor for an extended period of time.

This is a violation.

One example _of- improper control of visitors was identified.

This is a

violation.

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SURVEILLANCE 0BSERVATIONS (61726)

4.1 Undervoltage Relays and Rela _y Timers Functional Test

On October 16, 1992, the inspector observed the performance of Surveillance-

Procedure 6.2.2.1.10. "4160V Buses If and 1G Undervoltage Relays and Relay

Timers Functional Test," Revision 18.

Operators appeared to be following the surveillance procedure both locally and

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in the control room.

Good communications were noted between the control room

operators and individuals performing the surveillance.

In reviewing the

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procedures the inspector noted that proper signatures and approvals were

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evident.

During the surveillance the inspector observed that the conditions

inside the 4160V breaker cabinets were clean.

4.2 Reactor Core isolation Coolina Steam line Hiah Flow Calibration and

Functional Test

On October 28 the inspectors observed performance of Surveillance

Procedure 6.2.2.6.1 " Reactor Core Isolation Cooling Steam Line High Flow

Calibration and Functional Test," Revision 21. The inspector observed an-

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instrument mechanic performing the calibration of the differential pressure

switches which are used to monitor reactor core isolation cooling steam line

fl ow. The instrument mechanic was adhering to the procedure and-m..intained

good communications with the control room operators-throughout the

surveillance.

The instrument mechanic was conscientious in complying with

good radiological practice as he routinely changed protective gloves during

his manipulation of the valves associated with the differential pressure

switches.

The surveillance was completed satisfactorily with no anomalies

encountered.

4.3 Conclusions

The surveillances observed were performed well.

The licensee personnel:

involved were knowledgeable of the tasks required and executed these tasks

sufficiently to comply with the procedures. The inspectors found the licensee

actions, as they pertained to these surveillances, to be good,

5 MAINTENANCE OBSERVATION (62703)

On November 10, 1992, during a routine surveillance run of Emergency Diesel

Generator 1, it was noted that.the air start solenoid to_one bank of air

cylinders had not actuated.

Upon'further investigation, the licensee found

the fuse holder for that solenoid to be loose.

The inspectors observed the corrective: maintenance activity to repair the fuse

holder and the panel inspections-to check other fuse holders that may have-

experienced similar problems. The licensee did not -identify any addition

examples.of this deficiency. The inspectors verified ~that the workers

obtained proper authorization to perform the work, _that control room operators

were cognizant.of the maintenance activity, that workers followed'the

maintenance instructions, and that appropriate safety. precautions were taken

for work in energized panels. The inspector observed the postmaintenance

functional check of the solenoid and verified proper operation.

The

inspectors noted that the electrical cabinets were clean.

No unacceptable-

conditions were identified.

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5.1 Conclusion

The maintenance activity to repair and inspect _ i -

>olders was good.

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6 FOLLOWUP (92701)

6.1

(Closed) Inspection Followup Item 298/9034-02:

Entry into a Technical

Specification Limittna Condition for Operation During the Performance of

Surveillance Testing

The resident inspectors reviewed a licensee memorandum dated April 17, 1991,

which outlined proposed technical guidance and reflected existing policy on

the subject of entering Technical Specification. action statements during the

performance of surveillance testing.

The licensee had identified several

cases where procedures could disable a safety function during the performance

of a routine test.

As a result, several procedures were revised.

Certain

Technical Specification surveillance requirements have been amended to change

the test frequency to allow the performance of the st/veillance procedures

during refueling shutdowns instead of performing these at power.- Also, a

Technical Specification amendment eliminated the testing of certain systems

and components following the failure of a redundant system or component, a

practice which could result in the removal from service of the only operable

system or component. The licensee has taken further action to address the

issue of operability during the performance of surveillance procedures by

organizing a task force to identify additional required changes in the

surveillance program and Technical Specifications.

6.2 _(Closed) Unresolved Item 298/9219-01:

Implementina Organizational Change

without Having Amended the Technical Specifications

The licensee implemented a site reorganization on July 20, 1992, and had not

revised their Technical Specifications to reflect the changes in the

reorganization.

On October 8, the licensee submitted their-Technical

Specification amendment to the Commission.

Inspectors reviewed, for the time

between reorganization and submittal of the amendment, the person assigned

full. time responsibility for the operation of the facility as specified in

Technical Specification 6.1.1.

The inspectors concluded that the licensee met

Technical Specification 6.1.1 during this time period.

6.3 LClosed) Unresolved item 298/9219-02:

Potential Failure to Perform a

10 CFR 50.59 Review for Eauipraent Placed on Ten of Empty Spent Fuel Racks

On September 25, 1992, during a plant walkdown, the inspector identified a

process can located on top of empty spent fuel racks.

The process can was

used as part of the licensee's spent fuel pool cleanup project.

The

inspectors questioned whether a 10 CFR 50.59-evaluation for the process can

pertaining to its location on the spent fuel racks had been performed.

The process can was 2 feet in diameter by 4 feet long with a fully loaded

weight of.approximately 800 pounds. An engineering evaluation had been

performed prior to placement of the can on the spent fuel racks, to ensure

that the racks would handle the fully loaded weight of the can.

Also, the

licensee considered the possibility of damaging fuel assemblies should a

seismic event or industrial accident happen.

Interaction between the can and

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spent fuel assemblies was not deemed to be credible _because of the 20-foot

distance between the can and the storage racks containing spent fuel.

The licensee concluded that the calculated design basis seismic force would

overcome the friction between the process can and the spent fuel rack before

tipping the can, therefore, the can would remain upright-and horizontal

movement would be limited because of the oscillating nature of a seismic

event.

lhe can was submerged in water which had a dampening effect on any

movement of the can.

If the process can were to slice or roll far enough to

impact fuel bundles, damage to the fuel assemblies would not be expected.

The

fuel manufacturer estimated that it would take 250 foot-pounds of downward

impact loading to damage one fuel rod.

further, the licensee's_ Refuel

Accident Radiological Effects Calculation (No, NEDC 88-171), which assumes

111 rods to be broken, concludes that the resulting lifetime-thyroid and whole

body dose would be less than 1 percent of the NRC 10 CFR Part 100 reactor

siting criteria. The relationship between a vertical drop loading and a side

loading (assuming the process can moves horizontally) would not be one to one.

The 800 pound process can would have to free-fall approximately 34 feet- to

damage 111 fuel rods.

This amount of energy would not be attainable for the

configuration and controls the licensee had in place for the process can.

If

the can were to move in a direction away from the spent fuel, it could

possibly fall into the cask pad area of the fuel pool.

This accident would be

significantly less severe than the shipping cask drop accident analyzed in

Burns & Roe Calculation 2520-02.

Concerns for loose parts-(i.e., if the can were to topple over) falling into

the spent fuel pool or even potentially being transported into the reactor

have been addressed in bounding analysis previously completed for the site.

The licensee concluded that, with the procedures being used, the location of

the process can in the spent fuel pool, and the previous analysis performed,

all safety questions / concerns pertaining to the process can had been

addressed.

The inspectors concluded that the licensee's evaluation of the use of the

process can was appropriate.

6.4 Licensed Operator Requalification Program Evaluation

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On November 4 and 5, 1992, the resident inspector and a Region-based inspector

observed some requalification examinations, interviewed on-shift' supervisors,

and reviewed training and testing material. . Also, the licensed operators were

observed during the simulator examinations to determine if they were

conducting activities in a manner conducive to protection of the public health

and safety.

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The following previously identified weaknesses (from NRC Inspection

Report 50-298/9102) were specifically addressed either by direct observation,

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interviews, or by reviewing training program records:

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Crew command, control, and communication

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Adequacy of simulator scenarios

Operators' ability to establish shutdown cooling

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Operators' ability to diagnose conditions

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Some communications weaknesses were seen during this inspection:

During one scenario, the supervisor directing panel activities was not

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concise in his directives. A lack of uniformity in communication among

crews was seen,

During another scenario, a supervisor directed an operator to establish

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torus spray.

The operator could not get torus spray started and did not

inform the supervisor, who assumed that the torus was being sprayed.

The communications problems observed were compensated by actions of the

operators such that safety problems did not develop and mitigation strategies

were not degraded. The facility managers stated that initiatives were in

progress to improve communications.

This was primarily being done in the

evaluation sessions during the requalification training.

There was no formal

classroom presentation geared to defining a communications policy.

Training Guide NTG 318, " Command and Control" and operations directive, "CNS

Communications," were developed to address command and control.

However,

there did not appear to be a formal method to define -their interrelationship.

Command and control training had been incorporated into the evaluation

sessions during requalification training, but there were no formal classroom

presentations scheduled to address this area.

A review of the training and testing material used for this requalification

cycle showed that the material was current-and that mechanisms were in place

to update the material.

The simulator scenarios developed for this. evaluation

were in accordance with the guidelines stated in NUREG-1021, " Operator-

L.icensing Examiner Standards," Revision 7.

Critical task identification and

task standard definitions were very good.

A review of the graded written-

examinations indicated that they were developed based on the sample-plan' and

that they discriminated at the proper level.

During the simulator scenarios and walkthroughs, conditions existed that

required establishing shutdown cooling.

The operators were able to perform

all operations necessary to ' accomplish shutdown cooling.

No errors were

noted.

The licensed operators observed during the simulator and walkthrough

examinations demonstrated the ability to diagnose events and conditions.

No.

errors were observed.

The facility evaluators conducted the dynamic simulator and walkthrough

examinations professionally and in accordance with the standards.

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evaluators were able to function autonomously without management interference

or visible constraints. During simulator evaluation sessions that were

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observed, the lead examiner elicited full participation from all evaluators.

Facility evaluations were consistent with their program guidance, and the

licensee took appropriate measures to preserve examination integrity.

Other observations made by the inspectors and connunicated to the licensee

include:

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Shift technical advisor rotation policy and involvement during

requalification examinations was not fully understood by the shift

Crews.

Simulator difficultly with P-1 printout has contributed to negative

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training.

Rather than following up when a P-1 was not obtained, the

crew assumed it was a simulator problem and simulated having a. printout.

At one point during a shift crew scenario, both reactor operators were

behind the control panels at the same time.

Based on inspectors' observations, the licensee has made progress to

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increase operations' sense of ownership in training.

Areas of strength that were identified include:

Evaluators were very professional and exhibited good evaluation skills,

Examination material was very good and in accordance with the standard.-

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Licensed operators took a serious professional approach to the annual

evaluation.

Although weaknesses were seen in command, control, and communications, the

licensee was aware of the problems and was actively pursuing their corrective

actions program.

The licensed operators appeared to be' safety-conscious and

competent.

6.5 Conclusions

The licensee appropriately addressed, from a safety perspective, the use

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of a process can in the spent fuel pool.

Licensed operator training weaknesses were observed in command, control,

and communications; however, the licensee was aware of the problems and

was actively pursuing their corrective actions program. The simulator

evaluators were very professional and exhibited good evaluation skills.

Examination material was very good and in accordance with the standard.

The licensed operators appeared to be safety-conscious and competent.

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7 ONSITE REVIEW 0F LICENSEE EVENT REPORTS (92700)

7.1

(Closed) Licensee Event Report 298/92-008:

Inoperability of the High

Pressure Coolant Injection System Due to Stem Nut Wear of a Motor-

Operated Valve

This licensee event report documented the licensee's determination that the

high pressure coolant injection valve, HPCI-MOV-58, which is the pump suction

valve from the torus, was not stroking properly.

During the running of

Surveillance Procedure 6.2.2.3.4, "HPCI Suppression Chamber and Emergency

Condensate Storage Tank Water Level Calibration and Functional / Functional Test

and Water Initiation," Revision 25, both HPCI-MOV-58 and HPCI-MOV-17,-the pump

suction valve from the emergency condensate storage tank, could have been

closed. The system logic for these two valves is such that one of them should

always remain in an open position to provide suction for emergency core

cooling through the high pressure coolant injection system.

The licensee concluded that, had the high pressure coolant injection system

been required, it would have functioned as designed for as long as 10 minutes

before tripping off on low suction pressure.

The most limiting accident

requiring operation of the high pressure coolant injection system is a small

break loss-of-coolant accident and, for accident analysis purposes, high

pressure coolant injection is considered inoperable.

The response of the

plant to the small break loss-of-coolant accident has been predicted in~ the

latest accident analysis.

The licensee replaced the worn stem nut and reset the limit and torque switch

settings.

The licensee established acceptance criteria for stem nut thread

inspection, but had not yet revised the maintenance procedure.

The licensee

committed to provide detailed instructions for performing stem nut inspections

in the Limitorque maintenance procedures. Also, all_ Generic Letter 89-10

safety-related motor-operated valves with rising stems which have original

stem nuts installed are being identified.

Following the above activities, a

representative sample of the motor-operated valves identified will have their

stem nuts inspected to determine whether a potential motor-operated valve stem

nut wear problem. exists.

The inspector reviewed the documentation of the completed corrective' actions

and concluded that the licensee's actions were appropriate.

7.2

(closed) Licensee Event Report 298/92-012:

Inoperability of Reactor

Core Isolation Coolina Motor-Operated-Valve Due to Water Intrusion into

the Motor Operator

This_ event involved the surveillance testing on_the outboard _ steam supply

isolation valve to the reactor core isolation cooling system.

As part of the

surveillance, the outboard isolation valve was closed but failed to reopen

when required. Upon investigation, moisture was discovered in the limit

switch box which caused the valve to not open. A hair-line crack was found in

the flexible conduit installed to protect the wiring between the limit switch

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compartment and the terminal box. This crack was near a steam packing leak

which allowed moisture to enter the conduit line and travel into the limit

- switch box. The inboard and outboard isolation valves were both normally

open. The inboard isolation valve was operable.

The licensee reduced power for ALARA purposes so that entry into the steam

tunnel for repair of the valve could be made safely.

The corrective actions

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included drying out limit switch internals and replacing the valve motor

degraded terminal blocks. A tee drain was installed n- the limit switch

compartment cover to provide a drain path for any fF

moisture

accumulation, and a shield was installed around the

.uit in the immediate

vicinity of the motor-operated valve.

The licensee pians to replace the-

cracked conduit during the 1993 refueling outage and to inspect other motor--

operated valve installations where flexible conduit containing motor-operator

leads may be in close proximity to valve packing glands.

The inspectors reviewed the documentation of the completion of the licensee's

corrective actions and concluded that the licensee appropriately addressed

safety.

7.3

(Closed) Licensee Event Report 298/92-013:

Error in Limiting Single

Failure Assumption for the Emergency Core Coolina System Performance

Analysis

This event involved the discovery of a nonconservativo assumption in the

emergency core cooling system performance analysis, under postulated design

'

basis loss-of-coolant accident conditions. The nonconservative assumption was

that the most limiting single failure was the_ failure-of one low pressure

coolant injection subsystem injection valve.

During the licensee's review of

their design basis reconstitution program, they determined several failure

modes existed for the 125-Vdc power system which would result in a more

limiting single failure condition than previously analyzed. The licensee's

immediate corrective action was to. reduce power toward hot shutdown in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

and cold shutdown in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> as required by Technical Specifications,-and a

Notification of Unusual Event was declared.

Prior to achieving hot shutdown,

a vendor analysis indicated that meeting the design basis for emergency core

cooling systems was possible with certain operating restrictions. An

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- operating restriction of 90 percent power was imposed and remained in effect.

';

until modifications were completed which restored the validity of the original

assumptions-used in the emergency core cooling system performance loss-of-

coolant analysis.

On September 14, 1992, the licensee completed Design Change 92-141B which

allowed control of low pressure coolant injection and -reactor recirculation

discharge valves to be independent of the 125-Vdc battery system and, thus,_

not subject to failure due to loss of one 125-Vdc battery system.

The inspector observed changes made to the 250-Vdc control power and verified

documentation for completion of the design change.

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8 MANAGEMENT MEETINGS (30702)

On September 25, 1992, the Region IV Regional Administrator and members of his

staff accompanied the resident inspectors on a site tour and attended a

presentation by the licensee.

The licensee presentation included site

communications, quality assurance training, and their deficiency reporting

program, followed by an open discussion between the licensee and the NRC

staff.

On October 1 and 2, the Division Director for the Division of Reactor Projects

was onsite for a site tour and discussions with select members of the

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licensee's staff.

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ATTACHMENT 1

t

1 PERSONS CONTACTED

1.1

Li_censee Personnel

R. L. Beilke, Radiological Support Manager

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L. E. Bray, Regulatory Compliance Specialist

'

R. Brungardt, Operations Manager

M. A. Dean, Nuclear Licensing and Safety Supervisor

J. W. Dutton, Nuclear Training Manager

C. M. Estes, Senior Manager of Operations

J. R. Flaherty, Engineering Manager

R. L. Gardner, Plant Manager

M. D. Hamm,_ Security Supervisor

H. T. Hitch, Plant Services Manager

R. A. Jansky, Outage and Modifications Manager

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E. M. Mace, Senior Manager Site Support

J. H.-Meacham, Site Manager

C. R. Moeller, Acting Technical Staff Manager

S. M. Peterson, Senior Manager of Operations

G. E. Smith, Quality Assurance Manager

M. E. Unruh, Maintenance Manager

R. L. Wenzl, NED Site Engineering Manager

The personnel listed above attended the exit meeting held on November 16,

-1992.

In addition to the personnel listed above,_the inspectors contacted

other personnel during this inspection period.

2 EXIT MEETING

An exit meeting was conducted on November 16, 1992.

During this meeting, the

inspectors reviewed the scope and findings of this report.

The. licensee did-

not identify as proprietary any information provided to,_or reviewed by, the

inspectors.

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