IR 05000324/1987020

Safety Insp Repts 50-324/87-20 & 50-325/87-20 on 870701- 0806.Violations Noted.Major Areas Inspected:Followup on Previous Enforcement Matters,Maint & Surveillance Observations & Operational Safety Verification

ML20237H306
Person / Time
Site:	Brunswick
Issue date:	08/26/1987
From:	Fredrickson P, Garner L, Ruland W, Schnebli G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20237H228	List: IR 05000324/1987020 ML20237H224 ML20237H271
References
50-324-87-20, 50-325-87-20, NUDOCS 8709030204
Download: ML20237H306 (15)

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. ~ A cteuq' UNITED STATES

~g 8 o NUCLEAR REGULATORY COMMISSION

$ E REGION 11 o, 101 MARIETTA ST.. N.W., SUITE 3100 b[ ATLANTA, GEORGI A 30303

%**..+o Report Nos. 50-325/87-20 and 50-324/87-20

' Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos. 50-325 and 50-324 License Nos. DPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Conducted: July 1 - August 6,1987 Inspectors: i At Brunswick \< U

. H. RuMnd Date Signed

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. W. Garher Date Signe *

ab ories: < 8/ Zd' / I7 G E. Schheb11 t Date Signed Approved By: )-

P. E. Fredrickson, Sectio'n Chief

%- N D6te S'igned 6h7 l Division of Reactor Projects

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SUMMARY l

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Scope: This routine safety inspection involved the areas of followup on I previous enforcement matters, maintenance observation, surveillance {

observation, operational safety verification, onsite Licensee Event Reports l (LER) review, followup on inspector identified and unresolved items, onsite I followup of events, ESF system walkdown, followup of events at Wyle I Laboratories, Huntsville, Alabama, and engineering request administrative

! contro Results: One violation was identified: two examples of failure to follow operating procedures and one example of failure to follow a surveillance procedure.

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i DETAILS 1. Persons Contacted  ;

Licensee Employees L. Eury, Senior Vice President - Operations Support P. Howe, Vice President - Brunswick Nuclear Project C. Dietz, General Manager - Brunswick Nuclear Project

't. Wyllie, Manager - Engineering and Construction J. Holder, Manager - Outages R. Eckstein, Manager - Technical Support E. Bishop, Manager - Operations L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)

R. Helme, Director - Onsite Nuclear Safety - BSEP J. O'Sullivan, Manager - Maintenance G. Cheatham, Manager - Environmental & Radiation Control J. Smith, Manager - Administrative Support K. Enzor, Director - Regulatory Compliance R. Groover, Manager - Project Construction  !

A. Hegler, Superintendent - Operations W. Hogle, Engineering Supervisor B. Wilson, Engineering Supervisor B. Parks, Engineering Supervisor R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2) .

R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1) {

W. Dorman, Supervisor - QA W. Hatcher, Supervisor - Security R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

C. Treubel, Mechanical Maintenance Supervisor (Unit 1)

R. Poulk, Senior NRC Regulatory Specialist Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, office personnel, and security force  !

member . Exit Interview (30703)

The inspection scope and findings were summarized on August 6, 1987, with the Vice President - Brunswick Nuclear Project, the Senior Vice President Operations Support, the General Manager, and the Manager of Engineering and Construction. The violation - three examples of f ailure to follow procedure (paragraphs 6 and 9), were discussed in detail. The licensee  ;

did not take exception to the violation. However, the licensee took exception to the conclusions made by the inspector as indicated in the above referenced paragraphs. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during the inspectio . Followup on Previous Enforcement Matters (92702)

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(CLOSED) Violation (324/84-31-07), Inadequate Modification Acceptance Test. The inspector verified by review of meeting minutes that Brunswick Engineering Sub Unit (BESU) Practice 512,. Plant _ Modification Acceptance

' Testing - Functional Testing, was discussed with~ lead engineers on January 11. 1985. BESU Practice 512 appears adequate to address the cause of the violatio .- MaintenanceObservation(62703)

The inspectors observed maintenance activities 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 were adequate; personnel were qualified *, correct replacement parts;were used; 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 requirements were implemente The inspectors independently verified that selected equipment was properly returned to servic Outstanding. work requests were reviewed to ensure that the licensee gave priority to safety-related maintenanc The inspectors observed / reviewed the'following maintenance activities:

OPM-BKR003 Preventive Maintenance of General Electric 480 VAC Motor Control Center (MCC) Compartment BQKS1 Perform Preventive Maintenance on 2E11-F004C MCC Compartmen Portions of other maintenance activities were observed on a routine basi l No violations or deviations were identifie . Surveillance Observation (61726)

The inspectors observed surveillance testing required by Technical Specifications. Through observation and record review, the inspectors verified that: tests conformed to Technical Specification requirements; administrative controls were followed; personnel were qualified; instrumentation was calibrated; and data was accurate and complete. The inspectors indeper.dently verified selected test results and proper return to service of equipmen The inspectors witnessed / reviewed portions of the following test activities:

IMST-APRM28Q Average Power Range Monitor (APRM) Flow Bias Flow Units A&B Channel Calibratio _ - _ _ _

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IMST-RCIC14M Reactor Core Isolation. Cooling (RCIC) Steam Leak Detector Channel Functional Tes MST-ADS 21M Automatic Depressurization System (ADS) Reactor Water LL1 Trip Unit Channel Calibratio MST-RCIC26M RCIC Reactor High Water Level Trip Unit Channel Calibratio PT-0 HPCI System Operability Tes No violations or deviations were identifie . Operational Safety Verification (71707) .

The inspectors verified conformance with regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system statu The inspectors verified that control room manning requirements of 10 CFR 50.54 and the . technical specifications were me Control room, shift-supervisor, clearance and jumper / bypass logs were reviewed to obtain information concerning operating trends and out of service safety systems l to ensure- that there were no conflicts with Technical Specifications Limiting Conditions for Operations. Direct observations were~ conducted of control room panels, instrumentation and recorder traces important to safety to verify operability and that parameters were within Technical Specification limits. The inspectors observed shift turnovers to verify .

that continuity of system status was maintained. The inspectors verified

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the status of selected control room annunciator Operability of.a selected Engineered Safety Feature (ESF) train was verified by insuring that: each accessible valve in the flow path was in its correct position; each power supply and breaker, including control room fuses, were aligned for components that must activate upon initiation signal; removal of power from those ESF motor-operated valves, so identified by Technical Specifications, was completed; there was no leakage of major components; there was proper lubrication and cooling water available; and a condition did not exist which might prevent fulfillment of the system's functional requirement Instrumentation essential to system actuation or performance was verified operable by observing on-scale indication and proper instrument valve lineup, if accessibl The inspectors verified that the licensee's health physics policies / procedures were followed. This included a review of area surveys, radiation work permits, posting, and instrument calibratio The inspectors verified that: the security organization was properly manned and security personnel were capable of performing their assigned functions; persons and packages were checked prior to entry into the

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protected -area - (PA); vehicles were properly authorized, searched and escorted -within the PA; persons within the PA displayed photo

' identification badges; personnel in vital areas were authorized; and effective compensatory measures were employed when require . The- _ inspectors also observed plant housekeeping controls, verified position of certain containment isolation valves, checked a clearance, and verified .the operability of onsite and offsite emergency power source On July 26, 1987, the inspector observed that N-BFV-RB and I-BFV-RB on Unit 2 control board indicated open_ Page 29 of Valve Lineup Pre-startup Checklist of Operating Procedure OP-10, Standby Gas Treatment System Operating Procedure, Revision 30, dated May 19, 1987, lists these valves as normally closed.- I-BFV-RB is the purge exhaust fans suction valv N-BFV-RB is the purge exhaust fans inboard exhaust header isolation valv The operator restored them to their normally closed position. Failure to

. maintain the valves in their specified position is the first' example of Violation of Technical Specification 6.8.1.a: Failure to Implement Procedure 2MST-ATWS22M and OP-10, (325/87-20-04 and 324/87-20-04).

Mispositioning of these valves has minor safety significance in that the valves would get a closed signal on a secondary containment initiatio An example of one violation and no deviations were identifie . Onsite Review of Licensee Event Reports (92700) l The listed Licensee Event Reports (LERs) were reviewed to verify that the information provided met NRC reporting requirement The verification !

included adequacy of event description and corrective action taken or J planned, existence of potential generic problems and the relative safety  !

significance of the even Onsite inspections were performed and ,

concluded that necessary corrective actions have been taken in accordance with existing requirements, licensee conditions and commitment The following reports are considered close Unit 1 (CLOSED) LER 1-84-35, Failure to Maintain Two Control Cell Separations for Bypassing _ Full-In Input Into the Rod Sequence Control System. The inspector verified via training records that operations and maintenance personnel training had been performed as committed. The inspector also verified that the caution signs described in the LER are still in place on Unit 1 and 2 backpanels.

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(CLOSED) LER 1-85-06, Fungi Growth Contributes to a Spurious Control i

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Building Heating, Ventilation and Air Conditioning (CB HVAC) System isolation. The root cause of this item is similar to that in LER 1-85-4 See closecut inspection of LER 1-85-43 in inspection report 87-02.

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l (CLOSED) LER 1-85-15, Multiple CB HVAC System Automatic Initiations and Manual Isolations. The inspector reviewed the events and has no questions. See LERs 1-84-04, 1-84-32 and 1-85-43 closeout paragraphs in inspection report 87-0 (CLOSED) LER 1-85-42, Failure to Compare Battery Cell Data to Pre-operational Baseline Values as Required by Technical Specification Surveillance. The subject requirement was replaced in Technical Specifications as proposed in the LER. Hence, the item is closed because it is no longer-applicabl (CLOSED) LER 1-85-58, Inadequate Surveillance Test Procedure Renders Both Core Spray and Residual Heat Removal (RHR) Inoperable During Refueling ,

Logic Functional Tes The inspector verified that Procedure Test PT-07.1.9-1 (Rev. 2) and PT-07.1.9-2 (Rev. 0), Core Spray Simulated i Automatic Actuation and Logic Functional Test, have been issued as i committed to, such that 3 out of the 4 reactor low pressure permissive remain operable while the testing is performe (CLOSED) LER 1-86-02, Surveillance Test Procedure Fails to Verify Automatic Starting Capability of Both Control Room Emergency Filtration Trains. The inspector reviewed the revised procedure, PT-46.4, Control Building HVAC Auto Initiation, Revision 6, issued February 10, 1986, and the latest revision, No. 8, issued April 3, 198 (CLOSED) LER I-86-14, HPCI Isolation Due to Failed Resistor in Temperature Module. A similar failure is discussed below in LER 2-83-58 closecut. In addition, a temperature module in the RCIC logic recently faile This latest event is reported in LER 1-87-0 Inspection of a potential common root cause and need for corrective action will be conducted as part of the followup to LER 2-87-0 (CLOSED) LER 1-86-15, Automatic Control Building Emergency Air Filtration System Actuations Due to Area Radiation Monitor Malfunction. In the LER, the licensee committed to perform a failure analysis to determine the ,

failure mod The inspector reviewed their analysi (CLOSED) LER 1-86-16, Securing Circulating Water Pumps With Chlorination System in Service Resulted in Minor Chlorine Release and Subsequent Contr'o1 Building HVAC Isolation. The inspector verified that the precautions and steps which were added are contained in current revisions of OP-29, Circulating Water System, Revisions 13 and 39 (Units 1 and 2, respectively), and OP-43.1, Chlorination System Operating Procedure, Revision 1 i Unit 2 (CLOSED) LER 2-83-58, Closure of HPCI Steamline Isolation Valve Due to Failed Temperature Sensor Circuitry. The LER indicated that the potential of a heat induced failure mechanism might have occurred. The inspector reviewed the licensee's findings as documented in Hill to Dietz memorandum

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l dated January 11, 1985. Their study, which included collection of logic cabinet temperature data and a history search of heat induced component failures, indicates that there is no generic problem in this area or a specific problem with the subject temperature sensor circui (CLOSED) LER 2-83-61, Improper Breaker Coordination Resulted in MCC Feeder Breaker Tripping Before Load Breaker. The inspector reviewed the licensee's closecut package designated as FACTS 84B0446. In summary, feeder breakers and associated major load breakers were compared with the 1979 Load Study and setpoints were changed, as necessary. Based on this effort, the item was close However, the licensee identified that modifications such as PM-79-108 and 109 (increase thermal overloads to 300%), had not been incorporated into the 1979 study. Thus, a project identification item, PID-001438, had been . issued by the plant manager on May 28, 1986, to request non plant engineering groups to update the 1979 Load Study by December 31, 198 On July 27, 1987, in response to an inspector's question, the licensee discovered that the non plant engineering group had no record of receiving the item. Work on PID-001438 ,

has now been re-initiated with the anticipated completion date being the  !

same as the originally specified completion date. See paragraph 12 for additional information concerning PID tracking problem Because failure to use an updated Load Study to verify breaker coordination is a potential safety concern, the inspector discussed the safety significance of this with the cognizant engineers and the technical support manage In summary, the licensee could not state unequivocally that breaker coordination between load centers and MCCs had been maintained during modifications, e.g. , change of setpoints, breaker types, etc. However, the technical support manager stated that they have a high confidence level that modifications installed since 1979 have not adversely affected the load center and MCC coordination. This was because coordination had been verified as adequate for several large loads on some of the MCCs and changes implemented by the modifications do not tend towards adversely affecting the coordination. However, to ensure that no problem exists, they plan to verify the coordination of the largest load on each safety-related vital MCC. This should be completed by September 198 The licensee committed to review the results and methodology with the inspector when available. This is an Inspector Followup Item: Review of Safety-Related MCC Breaker Coordination Results and Methodolo '

(325/87-20-02 and 324/87-20-02).

(CLOSED) LER 2-84-05, Failure to Recognize a Limiting Condition for l Operation Existed When Primary Containment Isolation Valve Packing Was Adjuste The inspector verified via training rosters that select individuals in operations received training on this event and associated procedure (CLOSED) LER 2-84-12, Spike on 40% High Steamline Flow Instruments Causes a Reactor Tri The LER indicated that an evaluation of the actuation

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setpoints would be performed. Subsequent investigation revealed that the flow spikes occur af ter long outages or when condenser cooling mode has been used. The root cause of the spiking appears to be residual water in the steamlines. Operating Procedure OP-32, Condensate and Feedwater System, has been revised to provide better drainage of the steamlines after these evolution j No violations or deviations were identifie . Followup on Inspector Identified and Unresolved Items (92701)

(CLOSED) Inspector Followup Item (324/84-31-08), Resolution of Non-Conformance Reports (NCRs) Associated with Plant Modification 83-26 The NCRs referenced in the inspection report were voided and consolidated on February 18, 1985, into one item, NCR S-85-01 The plant issued BSP-12, Construction Maintenance Interface Procedure, to address the issue. The inspector reviewed the latest revision, Rev. 2, dated April 10, 1986, of this procedure. It appears to adequately address the communication problem which contributed to the event. The inspector reviewed the NCR closecut package and attached records of personnel receiving training on BSP-1 (CLOSED) Unresolved Item (325/81-12-01), Procedure Changes to Prevent Water Hammer in RHR Service Water Header and Design Review to Eliminate Shell Growth Related Problems in Heat Exchanger The licensee has modified the service and circulation water chlorination system to make it more effectiv See Survey of Licensee's Response to Selected Safety Issues (Temporary Instruction (TI) 2515/77) in inspection report 50-325/86-17 and 50-324/86-18. During the last Unit 1 outage, the inspector examined the IB RHR heat exchanger and two of the four diesel generator service water heat exchangers for pluggage or excessive foulin None was observe The inspector verified that item 10 of Auxi;'ary Operator Daily Check Sheet, for Unit 1 and 2 reactor buildings, requires the RHR service water headers to be vented once per 7 day No violations or deviations were identifie . Onsite Followup of Events (93702)

On July 1, 1987, at 10:35 a.m., Unit I reactor experienced a turbine control valve fast closure trip from 100% of full power. Group 2, 6 and 8 isolations occurred on low vessel water level N From the safety parameter display system computer, water level decreased to between 118 ,

and 120 inches. The momentary spike down in water level caused a partial j actuation of the low level No. 2 instrumentation. This resulted in a full ;

, group 1 isolation, starting of both standby gas treatment trains, group 3 j l outboard isolation valve closure and tripping of the "A" recirculation ]

pump. Reactor pressure increased in response to the control valve closure '

and group 1 isolation. Safety Relief Valve (SRV) 821-F013A, auto lifted to control pressure. Operator action then controlled reactor pressure and water level by lifting of SRV A, B and E and manual int *.ation of HPCI and

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RCI At 11.01 a.m. , the main steamline isolation valves were opened, such that pressure and level could be maintained using the main condenser and the condensate and feedwater systems, i

SRV B21-F013J failed to manually open during the scram recover The licensee localized the problem to the SRV and it's solenoid. Late that same afternoon, the licensee again tried to manually lift SRV F013J using the control room switch with the unit in hot shutdown at 800 psi SRV F013J again failed to lift. The licensee removed and tested the SRV F013J solenoid valve in the hot shop and did not conclusively find the cause of j failur The SRV F013J solenoid was replaced with a factory supplied j solenoid. The SRV F013J valve was then tested satisfactorily at 250 psig ]

using decay heat. On July 3,1987, Unit 1 was restarted and critical at i 1:04 a.m. At about 3:45 a.m., the six remaining ADS valves were tested at

250 psig. Valves F013K, C, D, A and H operated satisfactorily. SRV F013L failed to lif At 4:15 a.m., a reactor shutdown was commenced to investigate the cause of the failure and resolve the potential generic concer At 1:11 p.m. , SRV F013L was retested at 250 psig with the reactor in hot shutdown. The licensee had installed a pressure gauge on the air operator test connection. SRV F013L failed to lift with no air reaching the operator, demonstrating that the solenoid valve was not openin The licensee removed the L and K valves about 10:00 p.m. for inspection and testing. A Target Rock representative was onsite to aid the troubleshooting effort. The L valve tested satisfactorily when power and air were applied. Internal 0-ring shavings were found in the solenoid valve internals but they did not appear to interfere with valve operatio The inspector had witnessed the removal of the L and K valve and the L valve testin The licensee replaced the remaining six ADS valves with new solenoid All the solenoids that had been installed in Unit I had been refurbished ;

at Wyle Laboratories this past refueling outage to maintain the valve's '

environmental qualificatio The licensee concluded that the failure mechanism was unknown but was a service sensitive failure related to rebuild at Wyle. The vendor representative concurred with the licensee's assessment. All seven ADS solenoids were then shipped to Wyle Labs for further testing (see paragraph 11).

The inspector had been informed by the plant manager of their conclusions and actions by telephone about 6:00 p.m. on July 4. At that time, the licensee informed the inspector of their decision to restart the unit and committed to retest the newly installed ADS solenoid valves within 30 days of synchronizing to the grid. The licensee restarted the unit and synchronized to the grid at 12.58 p.m. on July Based on the determination that the root cause was determined at Wyle, and that the error was caused by one individual not following procedure, the licensee requested relief from the 30 day retest commitment in a meeting with Region II management on August 4, 1987. On August 5, Region II management informed the licensee by telephone that SRV testing was no longer required.

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One operational problem occurred during the transient. In attempting to manually place HPCI into service, the turbine was oversped when the  !

operator performed stops of the operating procedure out of order, i. e., j starting the auxiliary oil pump before opening the steam admission valv '

As an interim measure, on July 13,1987, OP-19, High Pressure Coolant Injection System Operating Procedure, Revisions 11 and 55 (Units 1 and 2, ,

respectively), were issued to emphasis performing these steps sequentially. Per discussions with the operations manager, long term corrective actions, including changes to the training program, are be'ng developed. These actions are to be issued as part of. Operating Experience Report (0ER) 87-04 The lack of knowledge by operating personnel of the proper sequencing order of these components was identified on May 14, 1987, as part of the -

HPCI Safety System Functional Inspection (SSFI). Of 8 licensed personnel j tested on May 8, only 2 individuals knew the correct sequenc The 4 results were known by May 14, when an operations engineer graded the test papers. The operations manager became aware of the deficiency on May 26, 1987, during an Onsite Nuclear Safety (ONS) presentation to the Plant Nuclear Safety Committee (PNSC) of the most potentially significant SSFI items. The PNSC was convened to review readiness of Unit 1 to restart after it's refueling outage. The operations manager directed that a training package be developed and training be completed by June 30, 198 The operator who made the error on July 1,1987, had not completed the trainin Discussions were held between the inspectors and licensee personnel who attended the PNSC meeting. These personnel included the operations manager, technical support manager, plant manager, director of ONS and ONS engineer. From these discussions, the inspectors concluded that the timeliness of the recommended ONS actions were not explicitly addressed by PNSC relative to the problem's safety significance. There was little safety significance to the actual deficiency, e., HPCI overspeeds, then instantly resets and is operationa The licensee, in the exit, took exception to the inspectors' conclusio The general manager stated that the SSFI team had been staffed with a Senior Reactor Operator (SRO) to review such problems and bring them to management's attention as warranted. They further stated that the actual event consequences proved that the timeliness of the response was appropriat Failure to perform the steps in OP-19 sequentially is the second example of the f ailure to follow procedure violation of Technical Specification 6.8.1.a (325/87-20-04 and 324/87-20-04).  ;

Prior to the scram, Instrumentation and Control (I&C) personnel were repairing the auto voltage regulator. During this evolution, the manual voltage regulator was shorted such that it's output signal failed to zer This resulted in over-excitation of the main generator, a subsequent primary turbine generator lockout and turbine trip. The licensee is currently investigating ways of increasing the reliability of the voltage regulators. Three reactor trips have occurred in less than a year due to

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l- malfunctioning regulators or during repair activities on the regulator This is -an Inspector Followup Item: Monitor Licensee's ' Program to

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. Increase Reliability of the Main Generator Voltage Regulator Syste ~(325/87-20-03 and 324/87-2.0-03) . _In the interim, the licensee plans to reduce power below the turbine trip bypass power level if online repairs to the circuit is attempted. in the future. This will prevent a reacto scram, thereby reducing the number of challenges to safety system On July 22, 1987, at 2:03 p.m., while at.100% full power,- Unit'2 experienced. a "B" recirculation pump trip due to personnel error. The control operator reduced reactor power to 40% by inserting control rods and decreasing the speed of recirculation pump "A". The "B" pump was-restarted at 2:18 p.m. , at which time return to full . power operation was

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initiate The event occurred during the performance of 2MST-ATWS22M, Anticipated Transient Without Scram (ATWS) Reactor High Pressure Trip Instrument ,

Channel Calibration. Inadequate communications between the I&C technician in the field and the- one in the control room resulted in step 7.4.4 not being performed. . Step 7.4.4 requires the test switch to be placed in the

" Test B" position. Failure to perform this step prior to proceeding with the calibration caused the pump trip signal to actually occur. Failure to perform step 7.4.4 is the third example of Violation of Technical Specification 6.8.1.a: Failure to Implement Procedure 2MST-ATWS2M and OP-10 (325/87-20-04 and 324/87-20-04).

The inspector interviewed the technicians involved in the even It appears that failure to set adequate work standards is a major contributor to the even Based on previous observations of work in progress, the inspector has witnessed various practices involving communicating whose turn it is to complete procedure steps. The inspector believes that this lack of a standard way of doing business when coupled with individuals who had only infrequently worked together was the root cause of this even This was discussed with the maintenance manager who disagreed with the inspector's determination. He indicated that the major root caus'e was personnel error, a momentary lapse of attention to the job in progress by one of the technicians, not a lack of setting work standards by management. However, he did state that enhancements in communications practices are being evaluated as a result of this even The inspector acknowledged that the licensee's position is plausible and reasonabl Also, the licensee stated, during the exit, that the surveillance group performs literally hundreds of thousands of steps each year with nearly zero personnel errors. The inspector agrees with the stated performance record of the surveillance grou Two examples of one violation and no deviations were identifie . Engineered Safety Features (ESF) System Walkdown (71710)

An inspection was performed of the following batteries on both units:

250, 125, 48 and 24 VDC systems The inspection included verification of proper electrolyte level, appearance of tight connections, no indication

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of cell leakage and absence of battery rack deterioration due to loose components or corrosion. The inspector noted minor fastener problem Five rack to concrete floor anchor bolts appeared to have been improperly.

, tightened. One such bolt was located on battery racks 248-1, IA-2, and ,

L 24A- Two; such conditions were observed on battery rack 2A- I addition, an occasional randomly spaced empty hole indicated .that bolts might be missing. These conditions appear to be original installation problem The conditions were reported to the license Engineers performed a walkdown to evaluate the existing conditions. The licensee determined that the racks were operable at least per the short ter qualification criterion. Work requests were issued to tighten the. loos bolts. BPE-5663 was issued on August 3, 1987, to address long term seismic qualification of the racks with missing bolts. Failure to have bolts installed per drawings or specifications is a violation of

.10 CFR 50, Appendix B, ' Criterion V. However, no notice of violation is being issued because the licensee was requested in Inspection Report 87-17 to address in writing to the Regional Administrator this type of problem on a generic basis and the response to this report has yet to be issue With the assistance of operations personnel, the inspector verified that the electrical breakers associated with the DC distribution systems :re positioned in accordance with that specified in the electrical pre-stertap checklist of OP-51,- DC Electrical System Operating Procedure, _ Revision 7 and.15 (Units ~1 and 2, respectively). These check sheets were verifieu in their ' entirety except for Unit I transfer switches located in junction boxes L6A and L68, which were inaccessible, and. the breakers associated with the Caswell Beach pump station battery chargers, which are non-safety-related battery and distribution systems, e., not associated with .the above-mentioned safety-related batterie The inspection included verification that all loads were being powered from their normal source and no load breakers were in the tripped stat The associated battery chargers were also verified to be in service per the operating procedur I A check of the 250/125 VDC battery ground detector system confirmed that no grounds were present in sufficient magnitude to cause an annunciated condition. The battery voltage instrumentation on the main control boards were observed to be in working orde No violations or deviations were identifie . Followup of Events at Wyle Laboratories, Huntsville, Alabama (93702) )

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Huntsville, Alabama, to witness testing and troubleshooting of the seven ADS solenoids from July- 8 - 10, 198 The solenoids (SN 443, 323, 471, 32, 337, 445, 447) arrived at Wyle on July 8, 1987, and were receipt inspected for obvious external defects. Engineering personnel from CP&L, Wyle Labs., and Target Rock (the vendor) were present for this inspection and the subsequent testing and disassembly of the valves for a detailed examinatio !

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.The initial testing consisted of placing the solenoids, one at a time, on a slave relief valve body and pilot. The test assembly was then heated to simulate the valve's normal operating conditio Steam- pressure was adjusted to normal . operating pressure and the solenoids were actuated to lift the slave relief valve. All seven solenoids actuated.as require The solenoids were then removed and production . tested by . Target Rock technicians on -a bench test assembl Production testing checks all electrical, mechanical, and pneumatic circuits of-the valves. Again, ~ll a seven solenoids performed as required. The ' valves were .then disassembled for a detailed visual . inspectio This inspection revealed traces o " Loctite" inside the solenoid bonnet tube.and a corresponding mark on the plunger for solenoid valves 443 and 323, which were the two valves that failed to actuate at the . site. Target Rock engineering personnel stated the " Loctite" could .cause the plunger and bonnet tube to be bonded together, thus preventing the plunger from moving and' opening the various air ports in 'the valve.. The Target Rock representative further stated that this problem was identified in the early 1980s and was corrected procedurally with precautions to ensure any excessive " Loctite" was removed prior to final assembly. " Loctite" is used on the plunger-to-stem threads to prevent separation during normal operatio Disassembly of the remaining five solenoids revealed slight traces of

" Loctite" in the same location on valves 32 and 447. These two valves operated properly when cycled at the plant. Further research by Target Rock found that the same individual had refurbished the four solenoids with ' traces of " Loctite" and indicated that the matter would be looked into furthe CP&L stated that the four valves exhibiting signs of " Loctite" would be sent to the Shearon Harris Energy Center laboratories for further analysis. Region II followup will be tracked with the LER that will be issued in Jul No violations or deviations were identifie . Engineering Request Administrative Control (92700)

Because of the lost PID described in the LER 2-83-61 closecut in paragraph 7, the inspector selected PID 005092, Repair and/or Upgrade of Corroded

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Instrument Racks in the Reactor Building, as an example to walk through 1 the administrative controls governing PIDs. During the process, it was discovered that this PID had been received by the non plant engineering group but had not been accepted. It had been issued on March 5, 1987, and was still in an undefined statu The licensee indicated that the administrative control problems associated with the two PIDs reviewed by the inspector were caused by the following: PID 001438 had been unknowingly misplaced in transit between groups and PID 005092 had involved a communication problem between groups as to who had the action  ;

on resolving a question. If the problems had not been discovered by the

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inspector, the licensee would have most likely uncovered the _ _ _ _ - _ _ _ _

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PID 001438 would have been re-activated by the licensee when it's associated Engineering Work Request (EWR) would have appeared on the 4 week to due-date EWR lis PID 005092 which had no completion date assigned to it, would probably have had it's limbo status determined during the licensee's annual review of backlogged EWRs. This latter mechanism could have also re-activated PID 001438. The technical support manager indicated that these items revealed weakness in their present administrative system. The licensee plans to enhance their administrative controls in this area by developing and implementing a cradle to grave PID tracking system by the end of this year. This is an Inspector Followup Item: Enhancement of PID Tracking System (325/87-20-01 and 324/87-20-01).

The issue of corrosion of the instrumentation racks has been addressed by the licensee as requiring no immediate action, i . e. , operability of the equipment is not affected. Engineering periodically inspects these racks to determine if additional degradation is occurring. Since the condition which caused the corrosion has been nearly totally corrected, this item is being considared as a long term item. Normally, only items deemed by the licensee to have no immediate safety significance are susceptible to the above problems. Items involving immediate safety concerns or regulatory commitments are tracked as "PNSC" or " FACTS" item These already have formal controls in place which should preclude similar problems from affecting processing of identified significant safety issue No violations or deviations were identifie !

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