IR 05000324/1988018

Insp Repts 50-324/88-18 & 50-325/88-18 on 880501-0604. Violations Noted.Major Areas Inspected:Previous Enforcement Matters,Maint Observation,Operational Safety Verification, LER Review & Silicon Controlled Rectifier Controllers

ML20151A251
Person / Time
Site:	Brunswick
Issue date:	06/29/1988
From:	Fredrickson P, Levis W, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20151A237	List: ML20151A234 ML20151A246 ML20151A251
References
50-324-88-18, 50-325-88-18, NUDOCS 8807190302
Download: ML20151A251 (21)

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UNITED STATES

[- p Riog s

D h!UCLEAR REGULATORY COMMISSION o

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p3 REGION 11

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j 101 MARIETTA STREET.N.W.

't AT L ANTA, GEORGI A 3o323

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Report No. 50-325/88-18 and 50-324/88-18 Licensee: Carolina Power and Light Company P. O. Box 1551

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Raleigh, NC 27602 Docket No. 50-325 and 50-324 License No. DPR-71 and DPR-62 Facility Name:

Brunswick 1 and 2 Inspection Conducted: May 1 - June 4, 1988 bh4 Inspectors:

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. H. 8) land Da'te Signed E

'N chq)d kW. Levis Date Signed

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Accompanying Personnel S. Shaeffer Approved Bg:

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i g. E. W edrickson, Section Chief Date Signed (J ' Division of Reactor Projects

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SUMMARY Scope:

This routine safety inspection by the resident. inspectors involved

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the areas of followup on previous enforcement matters, maintenance observation, surveillance observation, operational safety verifica-tion, onsite Licensee Event Report (LER) review, in office LER review, followup on inspector identified and unresolved items, Standby Gas Treatment (SBGT)- Silicon Controlled Rectifier (SCR)

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controllers, and inadvertent heatup.

Results:

In the areas inspected, 4 violations were identified:

failure to follow a plant modification test procedure; withdrawal of a control rod during condition 5 with the Reactor Protection System (RPS)

shorting links installed; failure to adequately control reactor coolant system temperature; and High Pressure Coolant Injection (HPCI)/ Reactor Core Isolation Cooling (RCIC) high steam line flow instrument setpoints greater than Technical Specification (TS)

setpoints.

8807190302 880629

PDR ADOCK 05000324 O

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Three unresolved items were identified:

control room fire detectors'

affect on control building emergency air filtration (CBEAF) system operability; information provided to NRC regarding silicon b anze bolts; and environmentally qualification of a non-safety pu, r'an (. the SBGT system whose failure could have caused system failure.

No deviations were identified.

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.-l REPORT DETAILS i

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Persons Contacted Licensee Employees W. Biggs, Engineering Supervisor

• E. Bishop, Manager - Operations

"J. Brown, Resident Engineer - Engineering

• S. Callis, On-site Licensing Engineer T. Cantebury, Mechanical Maintenance Supervisor (Unit 1)

• G. Cheatham, Manager - Environmental & Radiation Control R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)

• C. Dietz, General Manager - Brunswick Nuclear Project W. Dorman, Supervisor - QA

• R. Eckstein, Manager - Technical Support

• K. Enzor, Director - Regulatory Compliance W. Hatcher, Supervisor - Security A. Hegler, Superintendent - Operations R. Helme, Director - Onsite Nuclear Safety - BSEP J. Holder, Manager - Outages P. Howe, Vice President - Brunswick Nuclear Project

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• L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)

• M. Jones, Principal Engineer - On-site Nuclear Safety R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

• J. O'Sullivan, Manager - Maintenance B. Parks, Engineering Supervisor

• R. Poulk, Senior NRC Regulatory Specialist

• J. Smith, Manager - Administrative Support R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)

B. Wilson, Engineering Supervisor

• T. Wyllie, Manager - Engineering and Construction Other licensee employees contacted included construction craftsmen, i

engineers, technicians, operators, office personnel, and security force members.

• Attended the exit interview

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2.

Exit Interview (30703)

The inspection scope and findings were summarized on June 3,1988, with

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those persons indicated in paragraph 1.

The inspectors described the areas inspected and discussed in detail the inspection findings listed t'e l ow.

Dissenting comments were not received from the licensee.

Proprietary information is not contained in this report.

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Item Number Description / Reference Paragraph 325/88-18-01 VIOLATION - Failure to Follow a Plant Modification Test Procedure (paragraph 6.b).

324/88-18-03 VIOLATION - Control Rod Withdrawn During Condition 5 With the Shorting Links Installed (paragraph 7, LER 2-88-06).

324/88-18-04 VIOLATION - Failure to Adequately Control RCS Temperature (paragraph 11),

325/88-18-05 &

VIOLATION - HPCI/RCIC High Steam Line Flow 324/88-15-05 Instruments Inoperable (paragraph 9.b).

325/88-18-02 &

• URI - Failure to Environmentally Qualify 324/88-18-02 SCR Controllers for the SBGT System (paragraph 10).

325/88-18-06 &

URI - Control Room Fire Detectors' Affect on 324/88-18-06 CBEAF Operability (paragraph 5).

325/88-18-07 &

URI - Adequacy of Action to Identify and Correct 324/88-18-07 Silcon Bronze Bolt Problem (paragraph 9.d).

Note: Acronyms and abbreviations used in the report are listed in para-graph 13.

3.

Followup on Previous Enforcement Matters (92702)

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(CLOSED)

Violation 324/86-15-01, Failure to Maintain a Service Water.

Valve Motor Operator Breaker in the Correct Position.

The. inspector reviewed the Notice of Violation response dated July 24, 1986.

The training documentation for I&C/ Electrical Maintenance personnel was reviewed by the inspector.

(CLOSED)

Violation 324/86-25-01, Inadequate Procedure to Control DG

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Jacket Water Cooler Service Water Outlet Valves.

The inspector reviewed the Notice of Violation response dated November 26, 1986, and OP-39, Diesel Generator Operating Procedure.

OP-39, Revision 29, dated.

October 20, 1986, now incorporates steps allowing for the throttling of the subject valves during diesel generator operation and for returning them to the locked open position (with independent verification) upon securing the diesels.

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• An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.

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(CLOSED)

Violation 325/87-36-03, Allen Bolt Placed in Variable Hanger Preset Hole.

The inspector reviewed the Notice of Violation responses dated December 21, 1987, and February 11, 1988, along with the corre-sponding documentation package.

The licensee believes that the subject allen bolt came from the limit switch compartment cover of the adjacent motor operator, a RHR heat exchanger inboard vent, which was missing a bolt of the same type; however, the period of time when the bolt was placed in the hanger could not be determined.

The limit switch internal components were inspected and no apparent damage was found due to moisture intrusion resulting from the missing bolt.

The inspector reviewed all currently completed VERS which document the results of the initial in-service inspections of Unit 2's spring can supports.

A total of 61 supports were examined, the remaining 7 were deferred to the next Unit 2 outage. None of the VERS identified any obstructions or pins installed in the preset pin holes.

(CLOSED) Violation 325/88-01-01 and 324/88-01-01, Failure to Perform DG Surveillance Within TS Time Limits. The inspector reviewed the Notice of Violation response dated April 13, 1988. The licensee concluded that the cause of the violation was due to a high level of activity resulting in personnel error. The inspector verified that communications made with the plant shift operating supervisors concerning the root cause of the violation contained an adequate amount of emphasis on not continuing this kind of personnel error trend.

In addition, timers have been purchased and distributed for use by shif t foremen end control operators to assist in keeping up with TS related significant events which have time limita-tions.

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No significant safety matters, violations or deviations were identified.

4.

Maintenance Observation (62703)

The inspectors observed maintenance activities, interviewed personnel, and reviewed records to verify that work was conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The inspectors also verified that:

redundant components were operable; administrative controls were followed; tagouts are adequate; personnel were qualified; correct replacement parts were ssed; radiological controls were proper; fire protection was adequate; quality control hold points were adequate and observed; adequate post-maintenance testing was performed; and independent verification require-

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ments were implemented.

The inspectors independently verified that selected equipment was properly returned to service.

Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenance.

The inspectors observed / reviewed portions of the following maintenance activities:

WR/JO 88IAB211 Torque Switch Inspection for 2-SW-V294

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WR/JO 881AC211 Torque Switch Inspection for 2-SW-V295 WR/JO 88ALQ61 MCC 2 XC Bolt Replacement

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WR/JO 88PZP225 SLC 2B Accumulator During the performance of the torque switch inspection for 2-SW-V294, conducted in accordance with MI-10-25, the maintenance personnel noted that the torque switch contacts were c'orroded and that the motor leads were terminated with switch lock wire nuts and black electrical tape.

Trouble ticket 002564 was written to correct these deficiencies.

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inspector questioned why the limit switch contacts were also not inspected

at this time.

In view of recent valve failures attributed to dirty limit switch contacts (1-SW-V117, 2-E11-F003A) and the fact that the limit switch contacts are readily accessible during the torque switch inspec-tion, it seemed appropriate that the limit switches also be inspected at this time.

The licensee had previously taken action in this area associated with the failure of the 1-SW-V117 valve.

As stated in LER

,88-013, dated May 23, 1988, the expected implementation d te for the

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revised MI-10-25, which will include inspection and cleaning of the limit switch contacts in addition to the torque switch contacts, is June 1, 1988.

No significant safety matters, violations, or deviations were identified.

5.

Surveillance Observation (61726)

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i The inspectors observed surveillance testing required by TS.

Through observation, interviews, and record review, the inspectors verified that:

tests conformed to TS requirements; administrative controls were followed; personnel were qualified; instrumentation was calibrated; and data was accurate and complete.

The inspectors independently verified selected test results and proper return to service of equipment.

The inspectors witnessed / reviewed portions of the following test activi-ties:

IMST-APRM12W APRM CH B, D, and F Channel Functional Test RPS Inputs 1MST-CS21M Core Spray Pump Discharge Pressure' ADS Permissive

1MST-RCIC21M RCIC Steam Line Break High D/P Trip Unit Channel Calibration PT-34.4.1.3 Control Building Fire Detection System Operability Test During the performance of PT-34.4.1.3, Control Building Fire Detection.

Instrumentation Operability Test, the inspector observed that in step 7.0.4 the disconnect switch in the Control Building local alarm panel was

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placed in the disceanect position.

This switch was added in a plant

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modification to allow for testing of the Control Building fire detection system without fear of a spurious initiation of the CBEAF system. Placing this switch in disconnect disables the automatic start of CBEAF for both units if smoke is detected in the control room area.

In the licensee's response to NUREG 0737, Item III.D.3.4, Control Room Habitability, dated March 2,1983, Section 4.1 and 4.2.2, the licensee states that the CBEAF will automatically initiate upon the detection of smoke in the control room area. This design requirement was incorporated into the licensee's TS Surveillance Requirements, Section 4.7.2.d.2, which verifies that "on a smoke detector signal, the Control Building ventila-tion automatically diverts its inlet flow through the HEPA filters and charcoal adsorber banks of the emergency filtration system."

With the switch in disconnect, one of the design functions of the CBEAF (i.e.,

automatic initiation on a smoke detection system signal) is inoperable.

TS 3.7.2.b requires that the units be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee has stated that they do not consider the automatic start feature of the CBEAF system, due to a fire detection system signal, to be part of the design basis of the system.

It is their position that this design feature is not required to satisfy Criterion 19. In addition, the licensee is pursuing a change to their TS which would remove the automatic start feature of the CBEAF from a fire detection system.

This item is Unresolved pending NRR review and obtaining additional information from the licensee:

Control Room Fire Detectors' Affect on CBEAF Operability (325/88-18-06 and 324/88-18-06).

No significant safety matters, violations, or deviations were identified.

6.

Operational Safety Verification (71707)

The inspectors verified that Unit 1 and Unit 2 were operated in compliance with TS and other regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records, and independent verification of safety system status.

The inspectors verified that control room manning requirements of 10 CFR 50.54 and the TS were met. Control operator, shift supervisor, clearance, STA, daily and standing instructions, and jumper / bypass logs were reviewed to obtain information concerning operating trends and out of service safety systems to ensure that there were no conflicts with TS Limiting Conditions for Operations. Direct observations were conducted of control room panels, instrumentation and recorder traces important to safety to verify operability and that operating parameters were within TS limits.

The inspectors observed shif t turnovers to verify that continuity of system status was maintained.

The inspectors verified the status of selected control room annunciators.

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r Operability of a selected Engineered Safety Feature division was verified weekly by ensuring that:

each accessible valve in the flow path was in

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its correct position; each power supply and breaker was closed' fon

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components that must activate upon initiation signal; the RHR subsystem

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cross-tie valve for each unit was-closed with the-power. removed frcm the

valve operator; there was no leakage of cajor components; there was proper

lubrication and cooling water available; and a condition _did not exist which might prevent fulfillment of the system's functional requirements.

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Instrumentation essential to system actuation or performance was verified operable by observir.g on-scale indication and proper instrument valve lineup, if accessible.

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t The inspectors verified that the licensee's health physics policies /

f procedures were followed. This included observation of HP nractices and a i

review of area surveys, radiation work permits, posting, and instrument i

calibration.

Additionally the inspectors verified '. hat:

the security

organization was properly manned and security personnel were capable of performing their assigned functions; persons and packages were checked

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prior to entry into the protected area; vehicles were properly authorized, c-searched and escorted within the PA; persons within the PA displayed photo

identification badges; personnel in vital areas were authorized; effective

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compensatory measures were employed when required; and security's response l

to threats or alarms was adequate.

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The i nspectors also observed plant housekeeping controls, verified

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position of certain containment isolation valves, checked a clearance, and

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verified the operability of onsite and offsite emergency power sources.

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The following items were identified:

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Evidence of Smoking Found in DG Building Basement j

The inspectors found numerous cigarette butts in the DG building basement during a routine tour on May 20, 1988, at 10:00 a.m. -The

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butts, with empty cigarette packs, were found in uni-struts above eye

level and on top of the DG transformers. The inspectors also found a l

paperback book above the No. 3 DG transformer.

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inspection in the area found additional cigarette butts and several magazines.

All of the safety-related AC power cabling passes through the DG basement.

The area has an Appendix R Halon system installed but not yet operational.

Consequently, a roving fire watch has been stationed in the area. To date no individual has been found actually smoking in the area.

This issue had been previously identified by the inspector in viola-

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tion 325/86-17-02; 324/86-18-02 and by QA in NCR S-87-024P and l

S-87-024C. Recently, on October 15, 1987, QA issued NCR S-37-065 for i

failure to take adequate corrective actions for the previously issued

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violation and NCRs.

QA issued a Notice of Deficiency (inadequate response) to NCR S-87-065 on May 23, 1988, as a result of the-inspector's findings.

No Notice of Violation is being issued regarding this issue now since the licensee had previously identified this problem in the above NCRs.

The inspectors will continue to-follow the licensee's corrective actions during future routine inspections.

This is a

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Licensee Identified Item:

NCR Issued for Smoking in DG Building (325/88-18-07 and 324/88-18-07), and will be opened and closed for documentation, b.

During a walkdown of the Unit 1 back pannel area on May 17, 1988, the inspector noted that the keys for the drywell drain isolation logic test switches A718-S56A and A71B-S56B were insertej with the B switch in the test position. No maintenance or surveillance activities were being performed at the time.

When informed of the situation and

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after verifying that plant conditions were prc;se, the SF had -the B switch returned to normal and both keys removed.

The licensee's investigation into the matter revealed that PM-85-061-W (ERFIS modification) step 40.3.9 required that section 7.3.2 of IMST-PCIS38R be performed, Step 7.3.2.4 of this test places the B switch in test and step 7.3.2.8 places the A switch in test. Step 7.3.2.11 returns the A switch to normal while step 7.3.2.15 returns the B switch to normal.

The procedure was signed off as complete at 3:30 p.m.

Although the safety significance of this event is minimal, since leaving the switch in test provides one of the two required trip signals to the logic system for the affected inboard containment

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isolation valves, it does constitute a failure to follow procedure.

Accordingly, it is classified as a Violation:

Failure to Follow Plant Modification Test Procedure (325/88-18-01).

It shouid also be noted that two shift changes occurred without

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realizing that this switch was in the wrong position.

Other contributing factors to the event include the followin; o

Procedure did not require that the keys be returned. Had this been the case, the switch would have to have been returned to normal as that is the only position in which the key can be removed.

o The switch position (Normal / Test) was not marked on the P611 panel for the B switch.

No significant safety matters, two violations, and no deviations were

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identified.

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7.

Onsite Review of Licensee Event Reports (92700)

The below listed LERs were reviewed to verify that the information

provided met NRC reporting requirements.

The verification _ included

adequacy of event description and corrective action taken or planned,

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existence of potential generic problems and the relative safety signifi-cance of the event. Onsite inspections were performed and concluded that necessary corrective actions have been taken in accordance with existing requirements, licensee conditions and commitments.

UNIT 1 (CLOSED)

LER 1-86-10, Automatic Reactor Scram on Low Level No. 1 Following Loss of Electrical Bus 10.

Following an investigation of the UAT 1-A07 breaker to bus ID trip, the licensee could not reveal the cause of the trip. Breaker testing and certain procedure revisions were made in efforts to identify / prevent future occurrences.

The inspector. reviewed the completed documentation package, enhancements to preventive mainte-nance precedure PM-BKR001 relative to breaker' compartment checkouts of ITE 4 KV switchgear, and completed work request / job orders initiated to resolve minor problems identified during the event evaluation.

(CLOSED)

LER 1-86-11, Automatic Reactor Scram Due to Upscale Trips of Intermediate Range Monitors A and H.

The licensee has completed adjust-ments which lower the IRM upscale alarm annunciation and range for both units to allow adequate operator response time to transient IRM range

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power levels.

Appropriate operations personnel have received training regarding the event.

The inspector reviewed-the proceeding corrective

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actions and the documentation package.

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(CLOSED) LER 1-86-19, Failure to Perform Technical Specification Surveil-lance Requirement 4.3.5.7.1.

The inspector reviewed the completed documentation package and internal correspondence regarding the event.

The licensee performed a review of surveillance test completion / exception

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form RCI-02.4, and identified the need for a revision-regarding "partial completion satisfactory" notification. However, upon review, the licensee

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regarded the one identified deficiency in over a three year period in completing the form, as inadequate justification for change.

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licensee's final root cause determination for the event was personnel

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error.

(CLOSED) LER 1-86-21, Automatic Reactor Scram Resulting from Main Turbine Master Trip Signal.

Other inspection effort related to this event was

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documented in report No. 325/86-21 and 324/86-22. The inspector reviewed the corrective actions taken and the completed documentation.

(CLOSED)

LER 1-86-22, Reactor Scram Due to Upscale Tripping of Inter-mediate Range Monitors.

The licensee completed Plant Modifications PM-87-120 (Unit 1) and PM-87-182 (Unit 2), and declared them operable on i

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May 22, 1987 and April 5, 1988, respectively.

A GE representative was contracted by the licensee to evaluate the occurrence of noise spikes in the source and intermediate range monitors.

The inspector ccapared and

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reviewed the completed modifications design basis with GE letter G-KB1-6-169, dated March 30, 1987. The majority of the GE recommended modifica-tions were evaluated to be beneficial by the licensee and incorporated into the modification packages.

One difference noted was that GE i

recommended the rerouting of certain cables.

However, after a detailed analysis, the licensee concluded that this rerouting was unnecessary. The inspector interviewed engineering personnel with regards to this and otner discrepancies and concluded that adequate analysi s. was exhibited in justifying the currently completed PMs in order to reduce SRM/IRM noise spikes which may occasionally cause a half or full scram.

(CLOSED) LER 1-86-24, Automatic Reactor Scram Resulting from Loss of Main Generator Output Voltage Control. See report No. 325/86-24 and 324/86-25 for further event details. The inspector reviewed the internal documenta-

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tion of the event and the licensee's subsequent corrective actions.

Corrective actions included the replacement of the automatic and manual voltage adjustment potentiometers, an evaluation of on-line maintenance, and operational review of the voltage regulators.

The licensee has instituted weekly cycling of the manual controller and a momentary swap from the automatic to the manual controller to preclude similar events in the future.

(CLOSED) LER 1-86-26, Manual Reactor Scram Resulting from Loss of Main l

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Condenser as Heat Sink.

The licensee has completed implementation of procedures OPM-CDU500, OPM-CDU501, and OPM-CDU502, dsted January 29, 1987.

These procedures provide for surveillance and inspection of the ball collector for the Amertap condenser tube cleaning system. The inspector reviewed the procedures for completeness and quality and found them to be

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adequate in helping to preclude similar events in the future.

(CLOSED)

LER 1-86-27, Late Performance of TS Surveillance Requirement 4.11.2.7.2 Due to Perso,nel Error.

The inspector reviewed the LER l

package, which included appropriate procedural changes.

The inspector also reviewed the lesson plan in the Real Time Training Package 87-1-1 for subject adequacy and proper training emphasis.

No discrepancies were

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noted.

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(CLOSED) LER 1-86-31, Aito Isolation of RX Building Ventilation System and Auto Starting of SBGT During Cable Pulling in Distrubution Panel.

The licensee has revised Construction Procedure WP-217, Cable Pulling, Revision 0, Deviation 2, which provides specific guidance for construction craft foremen to follow pri r to pulling cable into an energized panel or box. The inspector revievred work request 87-AACM2 which was used to ascertain whether any exposed cable section(s) exist which may nave

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resulted from the event.

No damage was found to cables, conduit, or the

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distribution panel.

Each cable was separately pulled, inspected, and meggered. No discrepancies were noted.

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UNIT 2 (CLOSED) LER 2-82-83, Drywell to Torus Vacuum Breakers - X18A, C, and E -

Leakage Probitm. This item was previously addressed in inspection reports 324/86-30 and 324/87-35.. The licensee has changed PT-20.6, Drywell to Torus Leak Rate Test, adding visual inspection of vacuum breaker seats for defects which might prevent adequate sealing of the vacuum breaker. The~

inspector reviewed Revisions 5 and 16 of PT-20.6 and the previously issued LER documentation.

(CLOSED) LER 2-83-33, Main Steam Line Radiation Monitors A and D Out of Calibration.

This item was previously addressed in Inspection Report No. 324/87-35.

The licensee has installed the new NUMAC monitors in both units. All problems identified in the pre-installation checkout have been corrected. The inspector verified. operability of the new monitors through observation and personnel interviews.

(CLOSED) LER 2-86-17, Automatic Scram on Low Water Level Resulting from Failure of Reactor Feedwater Penp 2B Discharge Check Valve to Close..The inspector reviewed the completed work package and corrective actions taken to prevent recurrence of this event. The licensee has determined that the control logic for the feedpump discharge valves does allow the stroke direction to be reversed at any time during valve travel. Along with this, OP-32 for feedwater pump operation has been revised as of June 20, 1986, to minimize the potential for a defective check valve to cause a level transient while placing reactor feedwater pumps in service.

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LER 2-86-23, Primary Containment Group 4 Isolations of High Pressure Coolant Injection System.

The licensee has completed modifica-tion work involving the installation of lugs on those safety relt ed thermocouple circuits where ERFIS was recently installed.

Appropriate

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tests were conducted to verify the temperature readings from this equipment were not deg.>aded by the lugging.

The addition of these lugs providas a secure means of terminating the thermocouple wires mentioned

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above in the RHR, HPCI, RWCU, and RCIC systems.

The inspector reviewed

the corrective actions and the records of their implementation.

(CLOSED)

LER 2-86-25, Failure to Perform keactor det Pump Surveillance Testing. The licensee has completed Real Time Training for all applicable operations personnel concerning the subject event. The inspector reviewed the completed work package along with verifying that appropriate cautions were included in PT-13.1, Reactor Recirculation Jet Pump Operability, to preclude operating personnel from improperly performing jet pump surveil-

lance testing when normally used equipment is not in service.

(CLOSED)

LER 2-86-26, Late Performance of Required Hourly Fire Watches Due to Radiological Spill in the Unit 2 Reactor Building. The inspector reviewed the documentation package.

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(CLOSED)

LER 2-87-05, Inoperability of High Pressure Coolant Injection System (E41) Due to Closure of E41-F002 in Order to Comply with TS 3.6.3.

The inspector reviewed the work package and the analysis of the equipment failure.

The licensee considers the event a random end-of-service life failure that could not have been detected without destructive testing.

(CLOSED)

LER 2-88-06, Control Rod 10-39 Unknowingly Withdrawn With Reactor Protection System Shorting Links Installed.

This event was described to NRC in an Enforcement Conference held on May 27,.1988.

As identified in the LER, the licensee violated TS 3.3.3, which requires that the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be operable.

Taole 3.3.1-1 includes the neutron flux-high functional units of the intermediate and average power range monitors.

These units are required operable in operational condition 5 (refueling)

with the shorting links removed from the RPS circuitry prior to and during the time any control rod is withdrawn.

Control rod 10-39 was withdrawn

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with the shorting links installed in the refueling condition from 3:50 a.m. to 8:52 p.m. on March 8, ICB8.

This is a Violation:

Control

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Rod Withdrawn During Condition 5 With the Shorting Links Installed (324/88-18-03).

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The' inspector reviewed the documentation of the operator counselling that was performed subsequent to the event and had no questions.

This LER is closed for administrative purposes.

No significant safety matters, one violation, and no deviations were identified.

8.

In Office Licensee Event Report Review (90712)

The below listed LER was reviewed to verify that the information prov1ded met NRC reporting requirements.

The verification includes adequacy of event description and corrective action taken or planned, existence of

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potential generic problems and the relative safety s'gnificarce of the event.

(CLOSED) LER 2-88-09, Full RPS Trip While Selecting a Cor. trol Rod for

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Withdrawal with RPS Shorting Links Removed During Ref ueling/ Maintenance Outage.

No significant safety matters, violations, or deviations were id9ntified, l

9.

Followup on Inspector Identified and Unresolved Item (92701)

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a.

(CLOSED)

Unresolved Item 325/87-36-01, Review of Licensee's Root Cause Determination for RCIC Problems.

The inspector reviewed OER 87-083, dated December 4,1987, which listed a detailed Sequence of Events, root cause determination, and corrective actions to be taken

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concerning the Unit 1 RCIC unavailability from October 13 through October 30, 1987.

The root cause of each evert appears to be unrelated.

The inspector reviewed all procedural revisions which implemented the corrective action (3) for each occurrence and found them to be adequate.

b.

(CLOSED)

Unresolved Item 325/87-39-05 and - 324/87-40-05, Erroneous

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Setpoints, High Steam Line Instruments.

The inspector reviewed the following licensee documents to verify appropriate corrective action for the above matter:

OER-87-088 Potential HPCI/RCIC High Steam Flow Instrument Line Problems, December 23, 1987.

EER-88-0074 Setpoints for HPCI/RCIC High Steam Flow Isolation (Unit 1), February 12, 1988 EER-88-0184 Setpoints for HPCI/RCIC !iigh Stsam Flow Isolation (Unit 2), March 31, 1988

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01-3.1 C0 Daily Surveillance Report (Unit 1), Rev. 10, May 17, 1988 0I-3.2 C0 Daily Surveillance Report (Unit 2), Rev. 15, April 26, 1988 PID-06156 A&B HPCI/RCIC High Steam Flow Instrument Line Re-route The inspector found that:

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o Loop seals existed in the following lines:

High Pressure Line for 1-E41-PDT-N004 Low Pressure Line for 1-E41-PDT-N005

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High Pressure Line for 1-E51-PDT-N017 Low Pressure Line for 1-E51-PD1-N018

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High Pressure Line for 2-E41-PDT-h004 Soth High & Low Pressure Lines for 2-E51-PDT-N017 o

Licensee established new setpoints for the above instruments in an attempt to ensure that isolations would occur at less than-or

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equal to 300% of rated flow considering the loop seal errors, while still not causing spurious isolations during turbine starts.

o The nominal setroints i r, inches of water - were revised as follows:

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Unit Instrument Current New

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N004 205 125.5

N005 205 141.75

'N017 387 322.5

N018 387 534.0

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2-NOC4

.207 122.81

Wr17 362 336.5

o Assuming that the new sotpoints are correct with the piping as presently configured, the licensee has been in violation of TS 3.3.2 for years prior to the re-adjustment.

(S 3.3.2 requires that isolation actuation instrumentation channels. for RCIC and HPCI. steam line flow-high be operable with their trip setpoint less than or equal to 300% of rated flow. The above corrections indicate that the setpoints of 6 instruments were in excess of'

300% of. rated flow. No historical determinatior, of operability was provided to the inspector, o

No record wts shown to the inspector that indicated any determination of reportability was made.

o Proposed modifications to the instrument piping were deleted from the current fiscal year.

Accordingly, the inspector relayed his concerns to the licensee on May 26, 1988, and during the exit interview. The licensee agreed to provide additional information to the ir.spector early in the week of June 6 regarding past operability of the high steam -line flow instruments.

The above matter is a Violatinn:

HPCI/RCIC High Steam Line Flow Instruments Inoperable (325/88-18-05 and 324/88-18-05).

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c.

(CLOSED)

Inspector Followup Item 325/87-42-10, HPCI F001 Motor Failure.

The inspector reviewed the licensee consultant's failure

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analysis contained in a Jeffries to Harness, April 26, 1988, memorandum.

The inspector also interviewed selected plant personnel regarding the progress of root cause determinatio.t.

The probable root cause of motor failure was "development of a short-circuit between the series and the shunt field windings and the resulting open-circuiting of the shunt field winding This fault could

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have been caused by an initial flaw in the winding insulation or by a voltage transient in either the shunt or serie.s field winding.

A voltage transient is induced in the shunt field winding every

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time the local circuit breaker is opened. This occurs because no

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discharge path exists for the energy stored in the shunt field for the BSEP OC motors.

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The licensee issued a Part 21 report on this issue dated May 6, 1988.

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In that report, the licensee identified the following corrective actions:

o Operability checks of DC valve motors in both units, o

Implementation of a standing instruction requiring verification of DC valve operability after power restoration to operator.

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o Installation of surge protection by January 20, 1989 (antici-pated).

Based on the licensee's actions, the inspector has no further questions regarding the licensee s plans.

This item is closed; any further inspection in this matter will be conducted as followup to LER 1-87-023 and the Part 21 report.

d.

(CLOSED)

Inspector Followup Item 325/88-15-06 and 324/88-15-06, Silicon Bronze Bolts in Safety-Related Switchgear.

The licensee continued their program of replacing the 5/16-inch silicon bronze carriage head bolts with steel bolts. On May 10, 1988, the technical support manager informed the inspector. that they found recently-replaced 5/16-inch silicon bronze carriage head bolts cracked.

Further, that certain of the above bolts had not been changed within the past few months as previously thought.

This included 5/16-inch silicon bronze carriage head bolts in the incoming lines to each MCC or switchboard and in the shipping splices, and several sizes of silicon bronze hex head bolts. The licensee issued EER-88-0258, JC0 for Unit 1 with silicon bronze MCC carriage head bolts, on May 11, 1988.

Unit I was at power while Unit 2 was in an outage; thus, Unit 2 posed no immediate concern.

Region II had conference calls with the licensee on May 11 and 12 regarding the issue.

On May 13, the licensee revised the JCO Test results showed that approximately 2 month old cracked 5/16-inch silicon bronze carriage head bolts net the strength requirements of the ASTM specification.

p The licensee also developed a sample and testing program for other silicon bronze bolts.

The licensee shutuown Unit 1 on May 21 to replace silicon bronze bolting in the MCCs and switchboards.

Some exceptions were made.

Certain shipping splices.were analyzed individually where the bolts were inaccessible.

e The results of the licensee's sampling program for other silicon bronze bolts in the switchgear will be provided to NRC by June 20, 1988. The licensee also provided the inspector with a silicon bronze bolt balance-of plant replacement schedule. The last switchgear bolt replacement is scheduled for November 3,1989.

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This matter is now Unresolved pending NRC review of operability considerations for the failed bolts and review of the adequacy of ccrrective action to solve this problem, from the first bolt failures in 1986 to the present time:

Adequacy of Actions to Identify and Correct Silicon Bronze Bolt Problem (325/88-18-07 and 324/88-18-07).

No significant safety matters, one violation, and no deviations' were identified.

10.

SBGT SCR Controllers (25576)

During the procurement process for replacement SCR controllers for the'

SBGT trains, the licensee questioned the specified quality requirements.

Evaluation by licensee personnel revealed that although the device is not required for proper operation of the SBGT system, its failure:(short to ground) would make the SBGT train inoperable. Accordingly, its qualifica-tion is required by 10 CFR 50.49(b)(2).

This item was not previously identified as requiring environmental qualifications.

After this deficiency was identified and the inspector informed on May 13, 1988, the licensee prepared a JCO, documented in EER-88-0255, to allow the continued operation of both units with the present SCR controllers until they can either be replaced with qualified components or properly electrically isolated. The completion date for this fix is June 30, 1988.

The JC0 was developed after evaluating the probable failure modes, the.

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specific function to be performed, and the environment-in which the device is required tc operate.

During standby operation, the SCR controller controls the air intake heaters based upon a temperature signal input from thermocouples in the prefilter compartment.

Upon SBGT initiation, the power to the SCR controller is electrically bypassed.

However, the

controller is not electrically isolated and a failure that would result in a short to ground could trip the supply breaker and shutdown the blowers and heaters.

The licensee ha'd previously been issued a violation in this area (Inspec-tion Report No. 325,324/87-22) which noted problems with the qualification of skid mounted components, including SBGT. As a result of this report, the licensee reviewed again the qualification of their skid mounted equipment. The review for SBGT was completed on September 15, 1987. Page 3 of this review incorrectly concludes that the SCR controllers are not

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required to be environmentally qualified.

The licensee has not yet responded to the violation and their corrective action.

Therefore, this item is classified as a Unresolved Item:

Failure to Environmentally Qualify SCR Controllers for the SBGT System (325/88-18-02 and 324/88-18-02).

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11.

Inadvertent Heatup (71707)

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On May 31, 1988, while shutdown in mode 4, Unit 2 experienced an inadver-tent heatup of 35 degrees F from 170 degrees _ F to 205 degrees F.,The "A"-

RHR heat exchanger was mistakenly isolated during troubleshooting efforts on the E11-F003A RHR "A" heat exchanger outlet valve which had f ailed to open earlier in the shif t.

The specific; sequence of events is shown in Enclosure 3.

Some contributing factors noted by the inspector include the following:

OP-17, Revision 76, RHR System Operating Procedure, requires that the.

F003 valve be only in an opened or closed position when the unit is in the shutdown cooling mode.

During this evolution, the valve was being throttled to maintain temperature control.

Operators were trained to monitor RHR beat exchanger inlet tempera-ture when in mode 4 and in shutdown cooling (Standing Instruction 88-033).

If this valve were shut to limit cooldown, it was opened to record temperature and then closed. Other parameters which could have been used to monitor temperature were not used.

A "Caution" in OP-17 specifically alerts the operator that RHR heat

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exchanger i nl et tempe-a'.ure is an invalid indication of reactor coolant temperature when the F003 valve is closed.

o No temperature band was specified in the Daily Instructions. Had the 160 - 180 degrees F band been explicitly stated, the 190 degrees F value noted at shift turnover would have caused earlier action.

o Lack of sensitivity to work in progress. The operators should have been more cognizant of troubleshooting efforts on this valve and its potential for affecting plant conditiors.

This incident was addressed by the licensee in an Enforcement Conference htid on May 27, 1988, in Region II. This matter is a violation:

Failure to Adequately Control RCS Temperature (324/88-18-04).

No significant safety matters, one violation, and no deviations were identified.

I 12.

List of Abbreviations for Unit 1 and 2 AC Alternating Current ADS Automatic Depressurization System A0 Auxiliary Operator APRM Average Power Range Ponitor i

ASTM American Society for Testing Materials BSEP Brunswick Steam Electric Plant j

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CBEAF Control Building Emergency Air Filcration C0 Control Operator DC Direct Current DG Diesel Generator D/P Differential Pressure

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EER Engineering Evaluation Report ERFIS Emergency Response Facility Information 5ystem

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ESF Engineered Safety Feature F

Degrees Fahrenheit GE General Electric HEPA Hiah Efficieacy Particulate Air HP Health Physics HPCI High Pressure Coolant Injection H/V Heatup I&C Instrumentation and Control IE NRC Office of Inspection and Enforcement IFI Inspector Followup Item IPBS Integrated Planning Budget System IRM Intermediate Range Monitor JC0 Justification for Continued Operation KV Kilovoit LER Licen,ee Event Report MCC Motor Control Center MI Maintenance Instruction NCR Non-Conformance Report NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation OER Operat;ng Experience Report

Operating Instruction OP Operating P ocedure OPM Operating Procedure Manual PA Protected Area PID Project Identification PM Plant Modification PNSC Plant Nuclear Safety Cimmittee PT Periodic Test QA Quality Assurance QC Quality Control RCIC Reactor Core Isolation Cooling RCS Reactor Coolant Sy;;am RHR Residual Heat Removal

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RPS Reactor Protection System RTGB Reactor Turbine Gauge Board RWCU Reactor Water Cleanup RX Reactor SBGT Standby Gas Treatment j

SCR Silicon Controlled Rectifie.

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SF Shift Foreman i

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SJAE Steam Jet Air Ejector SLC Standby Liquid Control SOE Sequence of Events SOS Shift Operating Supervisor FRM Source Range Monitor STA Shift Technical Advisor TS Technical Specification UAT Unit Auxiliary Transformer URI Unresolved item VER Visual Examination Report WR/JO Work Request / Job Order i

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ENCLOSURE 3

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SOE FOR INADVERTENT HEATVP FOR UNIT 2 ON. MAY_12, 1988 1255 E11-F003A (RHR A Heat Exchanger _ Outlet-Valve), would not open by the control switch at the RTGB.

It was being opened so that the operator could record RHR inlet temperature which they use to monitor reactor coolant temperature. If this valve were shut, RHR inlet would not be a true measure of RX coolant temperature as this line had no flow; An operator was then sent to manually open the valve 15% open.

A t ot. ole ticket was written to have I&C troubleshoot.

'ote: This above sequence is logged at 1645 in the CO's log.

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1357 WR/JO AMTL1 generated to troubleshoot the F003A valve.

1440 (Information taken from WR/JO_ AMTL1. )

Troubleshooting begins on F003A valve. C0 closed valve from RTGB. C0 attempted to open valve while I&C monitored logic.

I&C found no continuity across rotor No. 1 contact No. 4.

They suspected dirty contacts as the cause.

The breaker was left in the off position and I&C waited until clearance could be hung on valve to begin work.

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1700 (Shift change.) Reactor temperature logged as 190 degrees F in SF log.

2019 Clearance No. 1514 signed to allow work to begin on F003A valve.

2040 Work commenced on F003A valve.

Valve found closed by local indica-tion.

2115 While obtaining 2100 temperature, the C0 noted thac a temperature differential of 40 degrees F existed-across A RHR heat exchanger with inlet temperature reading 165 degrees F and outlet 205 degrees F.

The SF was informed. At this time it was recognized that the F003A s alve was shut and that the recorded inlet temperatures were not a true measure of reactor coolant temperature.

An operator was dispatched to open the valve 25% open.

After valve was opened, temperature was verified to be decreasing.

Other temperatures recorded at this time include the following:

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A Recire. Suction 185 degrees F B Recirc. Suction 190 degrees F PWCU Inlet 190 degrees F (RWCU takes suction from Recire.

Suction line)

2140 Work completed by I&C on F003A valve.

OPS stroked valve open, closed and back open.

Dirty switch contacts were found to be the problem.

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