IR 05000324/1988015

Insp Repts 50-324/88-15 & 50-325/88-15 on 880401-30. Violation Noted.Major Areas Inspected:Followup on Previous Enforcement Matters,Maint & Surveillance Observations, Operational Safety Verification & in Ofc LER Review

ML20154P978
Person / Time
Site:	Brunswick
Issue date:	05/20/1988
From:	Fredrickson P, Levis W, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154P965	List: ML20154P961 ML20154P978
References
50-324-88-15, 50-325-88-15, NUDOCS 8806060144
Download: ML20154P978 (20)

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0 UNITED STATES

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NUCLEAR REGULATORY COMMISSION yN

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p REGION il g ,j 101 MARIETTA STREET, * e ATLANTA, GEORGI A 30323

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Report No. 50-325/88-15 and_50-324/88-15

' Licensee: Carolina Pcwer and Light Company P. O. Box 1551 Raleigh, NC 27602

' Docket No. 50-325 and 50-324 License No. DPR-71 and=DPR-62 Facility Name: Brunswick 1 and 2 l Inspection Co uc ed: April 1 - 30, 19 8 Inspectors: . .

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g uTen s ~ ~ bate S'igned

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l y .EeNs LTite Slgned Accompanying Pe son el:. S. Schaeffer Approved by: C P. E. Fredrickson,3ec~tWChWf 24 /1M D6te 5TgneT~

Division of Reactor Projects

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SUMARY Scope: This routine safety inspection by the resident inspector involved the areas of followup on previous enforcement matters, maintenance observation, surveillance observation, operational safety verification, onsite review comittee, in office Licensee Event Report (LER) review, followup cn inspector identified and unresolved items, information meeting with local officials, plant startup from refueling, and onsite followup of event ]

Results: One violation. was identified - Improper change in operational condition during startup of Unit 2. Several examples of failure to follow procedure were included as part of the overall violation. Two unresolved items were identified: questions concerning procurement of comercial grade items for safety-related applications; questions concerning qualification of Victoreen Radiation Detector Cabl A significant safety issue concerning cracking of silicon bronze bus bar bolts, first identified in Inspection Report 324,325/88-05, was inspected furthe "

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REPORT DETAILS Persons Contacted

' Licensee Employees

- Biggs, Principal Engineer

• E. Bishop, Manager - Operations 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 P. Dorosko, Administrative Supervisor W. Dorman, Supervisor - QA

• R. Eckstein, Manager - Technical Support K. Enzor, Director - Regulatory Compliance I

R. Groover, Manager - Project Construction 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 '

L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC) l R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

J. McKee, Supervisor - QC ,

J. Moyer, Manager - Training l G. Oliver, Manager - Site Planning and Control

• J. O'Sullivan, Manager - Maintenance  ;

• B. Parks, Engineering Supervisor l

• R. Poulk, Senior NRC Regulatory Specialist j

• A. Richards, Project Engineer - QA l

• S. Scharff, Operations Engineer l

• J. Simon, Operations Engineer l J. Smith, Manager - Administrative Support l V. Wagoner, Director - IPBS/Long Range Planning R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)

D. Warren, Acting Engineering Supervisor B. Wilson, Engineering Supervisor

• T. Wyllie, Manager - Engineering and Construction .

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Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, office personnel, and security force member * Attended the exit interview

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' ExitInterview.(30703)

The inspection scope-and findings were summarized on April 29, 1988, with those persons indicated in paragraph 1. The inspectors described the areas- inspected and discussed in detail the inspection findings listed belo Dissenting coments were not received from the license ;

Proprietary information is not contained in this repor '

Item Number D e s cLip_t i o n1R eefe r_e_n c e_ Pa ra g ra p h 324/88-15-01 VIOLATION - Improper Change In Operational Condition (paragraph 11.b.(2)

325/88-15-02 & *URI - Procurement of Conrnercial Grade Items 324/88-15-02 Intended for Safety-Related Applications '

(paragraph 4.a)

324/88-15-03 URI - Qualification of Victoreen Radiation Detector Cable (paragraph 11.b.(1)

325/88-15-04 & IFI - Failures of GE 305 Auxiliary Contact Adder 324/88-15-04 Blocks (paragraph 12.a)

325/88-15-05 IFI - Normal Position for SW-V117, Nuclear Header to Vital Header Isolation Valve'(paragraph 4.a)

325/88-15-06 & IFI - Silicon Bronze Bolts in Safety-Related 324/88-15-06 Switchgear(paragraph 12.b)

Note: Acronyms and abbreviations used in the report are listed in ,

paragraph 1 ! Followup on Previous Enforcement Matters (92702)

(CLOSED) Violation 325/87-20-04 and 324/87-20-04, Failure to Implement Procedure The inspector reviewed the licensee's response dated September 25, 1987. The inspector verified that the licensee implemented i the required corrective action All procedure documentation changes as l

well as procedure adequacy reviews were complete In addition, for

example b, of this violation, the inspector _ also interviewed selected engineering and training personnel and concluded that appropriate emphasis l was placed on increased procedural training in the area of OP-19, High Pressure Coolant Injection System Operating Procedure, Section 5.3, Manual Startu *An UnresoWd'Ttem 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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l MaintenanceObservation(62703)

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

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codes and standard 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 implemented. The inspectors independently verified that selected equipment was properly returned to servic ;

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Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenanc The inspectors observed / reviewed portions of the following maintenance activities

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I o Oil changeout on HPCI booster pump - ticket 88-QMK20 J o Troubleshooting efforts of F004,1-SW-V117, F041 valve o Troubleshooting efforts of Unit 1 Recirc. MG set field break'e While watching oil changeout of Unit 1 HPCI booster pump on April 21, 1988, the inspector questioned licensee personnel as to which documents specified the lubricant type and what controls were in 1 place to ensure that the proper lubricant was used. Th'e inspector )

found that the oil type for this component is specified in OPM -

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LUB500ATT21 This document states that DTE 797 is the proper lubricant with an annual changeout' requirement. The' inspector then verified that DTE 797 was used and traced down the purchase order of the lubricant to determine how the lubricant was procure The inspector noted that purchase order 3720588A, dated February 2, 1987,

specified that the item be procured as an "Off the Shelf" item, which i

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rreant that the vendor was not required to be on the approved supplier's lis The receipt inspection performed en this purchase verified only that the part ordered matched the part received. No

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other quality requiremer ts were specifie ENP-42.2, Revision 1, dated January 6, 1988, Purchase Requis'ition and l

Data Base Review Procedure, defines a Q-0TS (Off the Shelf) item as a i comnercial grsde item used in safety-related applications. ENP-42.3, I Revision 1, aud January 6,1988, Material Engineering Evaluation  !

Procedure, 'tep / 2.3.4, requires that a documented evaluation, which includes idt< *it, ing the components critical characteristics, be perfonned, when it is decided to procure an item comercial grade l l

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that is. intended for safety-related applications. Step 7.2.9 of the same procedure requires that special receipt instructions,- which should verify the component's critical characteristics, be specifie The inspector then questioned licensee personnel concerning why these requirements were not followed with respect to the DTE 797 oil as no documented engineering evaluation was found and no special receipt inspection requirements were noted. The licensee responded that these requirements only went into effect in October 1987, and that any items procured prior to this time would not have been subject to these procedure However, the licensee .was able to provide information to the inspector which showed the lubricant used in the HPCI booster pump to be of the proper type. The possibility exists though, that there are commercial grade items presently in stores intended for safety-related applications in which a proper engineering evaluation has not been performed. The licensee had previously -identified programmatic procurement deficiencies in NCR No. A-87-023 which is still ope 1. This item is Unresolved pending-NRC review of the licensee's corrective action for the NCR:

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Procurement of Commercial Grade Items Intended for Safety-Related Applications,(325/88-15-02and324/88-15-02),

b. On April 20, 1988, during maintenance surveillance testing, the licensee experienced a failure of 1-E41-F041, the HPCI Outboard Suppression Pool Suction Valve. Under accident conditions this valve is required to open on low CST level or high suppression pool level '

to provide an alternate suction path for HPCI. Troubleshooting revealed that the series field had been shorted to ground. No ,

evidence of actuator or valve problems were found that may have caused the failure. The motor was then sent to Harris E&E for analysi It should be noted that the licensee has experienced other DC motor failures (see LER 1-87-23) and has committed to a followup report on ,

the above LER by July 1,1988, to explain any root cause of these

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motor failures. The inspectors will continue to followup on licensee tctivities in this are c. On April 23, 1988, while performing maintenance activities on the Nuclear Service Water Vital Header, the Vital Header Motor Operated Isolation Valve, 1-SW-V117, failed to open when manually activated by the control switc Under accident conditions this valve opens to provide Nuclear Service Water to the Core Spray Pump room cooling units, the RHR Pump room cooling units and to each RHR Pump Seal Cooling Heat Exchanger. Troubleshooting efforts by the licensee which included motor checks and control logic checks showed a limit switch problem. The contacts on limit switch rotor No.1 were .

cleaned after which the valve was satisfactorily cycled from the RTGB and also with the remote keylock switc The inspectors will

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)(' 'j followup licensee activities withlespect to their determinatdin if i

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the valve should be open rather than closed, duping nonnal oMratio s is an Inspector Followup Item: Nonnal Positinn for 5If117, / '

Nuclear Header to Vital Header Isolation Val 77(ry/%-154 '

No signif cant.s'afety matters, violations or devit dgns dre identifie '

> . h 4 i , ., Surveillarge Jhservation (61726). j

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l The inspectors observed surveillance testing required by Technical '

Specifications. Through observation, intfrviews, and record review, the inspectors verified that: tests conformed ^.w Tych leal Specification , ., 3

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requirements; adQistrative1 controls were' t

- pe(sonnel were ^f c qualified; lipstruuqtation was calibrated; 'an .dS vas accurate and ~

complete. lhe inspectors independently verified he ected pf esults and-proper return to service of equipmen .,,, . .

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The inspectors witnessed / reviewed portions of the' following ^ t'est t activities:

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PT-0 HP'CI Operability Test * '

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, . 7 PT-10. RCIC' System Operability \Tes:' N Flow Rates at 150 PSIG  !

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PT-14.1A Control Rod Coupling Check and CRD Testing .' s

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PT-14. ', .(?gquence r, Critical Shutdown MarginMlculation, Rev. 4

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AI,N.rosen' Backup System Operability Test s

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PT-7 ' /,General Atomic Stack Radiatija Monitor Channel dlibration ,,(f s

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No significant safety matters, violations, or deviations were 1/jptifie . OperationalSNetyVerification(71707) Ij ,'

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! The inspectors perified that Unit 1 and Unit 2 were operated in ,cabYiance i withTechnical,.SpecificationsandotherreguJatoryrequirementsbydirect observations of lactivities, ucility tours.; discussions with personne '

reviewingofrecordstrv}independentverificationofsafetysystemstatus, l's * , ,

The inspectoh verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met. Control operator, shift

. supervisor, clearance, STA, daily and landing instructions, and jumper / bypass logs were reviewed to obtait information concerning operating trends and out of service safety systems to ensure that there

~ were no conflicts with Technical Specifications Limiti79 Conditions foh Operat' ions . Direct observations were conducted of co rol room panels, ,

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rinftrtdientation and recorder traces important to safety to verify

operability and that operating parameters were within Technical Specification limit The inspectors observed shift turnovers to verify

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that continuity of system status was maintained. The inspectors verified

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the status of selected control room annunciators.

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g 'F Operabilit weekly by'y of a selected insuring that: Engineered each accessible Safety valve Featur'e in the division flow pathwas wasverified-in its correct- position; each power supply and breaker was closed' for -

components that must activate upon an initiation signal; the RHR subsystem cross-tie valve- for cach unit was closed with the power removed from the 4 valve operator; there was no leakage of major components; there was proper-y ') lubrication and cooling water available; and a condition did not exist 3" >;[ which might prevent fulfillment of the system's functional requirement :

Instrumentation essential to system actuation or performance was verified '

operable by observing ori-scale indication and proper instrument valve

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lineup, if accessibl '

The insrctors -verified that the licensee's health physics policies / procedures were followe This included observation of HP practices and a review of area surveys, radiation work permits, posting, and instrument calibratio ^

The inspectors votified 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 tN protected area; vehicles were properly authorized, searched and escorted within the PA; persons within the PA displayed photo identification a 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 '

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,e The inspector found the limit switch cover not completely screwed down for valve 1-CAC-V7, the Inboard Suppression Pool Purge Exhaust Valve. This installation is not in accordance with the design drawings or normal

installation practices which require the cover to be tightened t sufficiently to ensure the "0" ring is seated. The licensee has generated i NCR A-88-012 to track and resolve this issue. The inspector will continue to followup in this area in future routine inspection No significant safety matters, violations, or deviations were identifie . Onsite Review Comittee (40700)

The inspector attended -pre-startup Plant Nuclear Safety Comittee meeting 'l 88-047 conducted on Apdl 13, 1988. The inspectors verified that the i meeting was conducted in accordance with Technical Specification l requirements regarding quorum membership, review process, and personnel 1

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f qualification lieeting minutes were reviewed to confirm that decisions /reconmendations were reflected in the minutes and followup of-corrective actions was documente No significant safety matters, violations, or deviations were identifie . In Office Licensee Event Report Retiew (90712).

The listed LER was reviewed to verify that the information provided met NRC reporting requirements. The verification included adequacy of event description and corrective action taken or planned, existence of potential

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generic problems and the relative safety significance cf the event.

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(CLOSED) LER 1-88-09, Accidental Deenergization of Unit 2 Process Off-Gas

< Radiation Monitor and Reactor Building Ventilation Exhaust Radiation Monitors During Maintenanc '

No significant safety matters, violations, or deviations were identifie . Followup on Inspector Identified and Unresolved Items (92701) (CLOSED) Unresolved Item 325/87-11-02, Mispositioned Equalizing Valve for IB RHR/SW HX DP The inspector reviewed OER-87-20, approved on June 26, 1987, which adequately detailed the events concerning the mispositioned valve. The licensee has revised AI-58 and ENP-03 in order to improve clearance tag sheet verification as ,

, well as requiring independent verification in the acceptance test section of the plant modification, as opposed to BESU's past practice of taking credit for the clearance and/or system lineup at the completion of the unit outage. The inspector concluded that the root

causes of the event have been identified and corrected, including action to prevent recurrence. Since the licensee has met all the -

conditions of 10 CFR 2, Appendix C, regarding licensee identified ,

violations, no notice of violation will be issued, i (CLOSED) Inspector Followup Item 325/86-24-05 and 324/86-24-05, '

Review RHR Room Cooler Operation. The inspector reviewed EER N , completed on July 24, 1987, which included detailed RHR Room Cooler Operation Analysis. The licensee has issued TSI-87-02, dated ,

August 28, 1987, which provides operator instruction for i administrative equivalent TS LC0 implementation. The inspector has no further concerns with this issu (OPEN) Inspector Followup Iten 325/86-33-01 and 324/86-33-01, IRM Fuse Testing and Subsequent Required Modifications. The inspector reviewed the content and results of SP-86-068, Revision 2 (Unit 1),

which was performed on January 19, 198 The test results concerns in GE SIL No. 445 were valid. SP-86.073 (Unit 2), proved was not the run because the licensee concluded that plant modifications were i

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required.. To assure pre-modification operability, SP-87-014 was performed on February 12, 1987, and verified that applicable portions of surveillance procedures MST-IRM11W and 12W would-detect a blown -

-24 volt DC fuse (F2) in the IR EWR-04527 and PID No. 5400A have

been -issued and completed to- develop the plant modification. The licensee plans to modify the system by December 8, _198 The inspector concluded that the _ licensee's actions to address the SIL issues have been appropriate. This item will remain open pending completion of the modifications, (CLOSED) Inspector Followup Item 325/87-03-03, Review 'of IMST-DG12R Procedure Violation OER. The' inspector reviewed- the licensee's OER-87-09, dated -' April 28, 1987, which addressed the root cause of-I 'the communication failure between the I&C personnel in charge of the test and the control operator. The licensee concluded that the test  :

director had been given too many tasks. The DG Load Test MSTs 11R

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through 14R for Units 1 and 2 were revised on May 4,1987, to incorporate lessons learned and a ~ redistribution of the-procedural steps to allow verifications to be done in other locations separate from the Control Room. The inspector concluded that the licensee's corrective actions were appropriate, (OPEN) Inspector Followup Item 325/87-03-04, Inadequate Board -

Walkdown and Review. The licensee has issued Standing Instruction 87-014, dated February 13, 1987, which requires the shift foreman or the shift operating supervisor to walk the Control Board with the respective control operator in order to double check RTGB indication and enhance SR0 awareness of plant status. The licensee has also ,

, secured general business activity at the shift foremar, window for an hour during shift turnover, to reduce distractions.- The initial cause of the mispositioned valve has been corrected via procedure i revision OP-17, Revisions 12 and 68, respectively. for Units 1 and l This item will remain open until the requirements specified in ,

SI-87-014 are incorporated into the licensee's permanent procedure i

. , (CLOSED) Inspector Followup Item 325/87-06-01. Review Motor-Driven Fire Pump Breaker Misalignment OER. The inspector reviewed the completed OER-87-10 and concluded that the root cause~ determination and corrective actions taken were appropriate. The licensee issued i an LC0 declaring the motor-driven fire pump inoperable for over 38 l hours. No LC0 Action Statement time limitation was exceeded during the event. The 4160 breaker misalignment occurred because OP-41, {

' i Fite Protection and Well Water System, was unclear. The positions in OP-41 have been revised as of May 19, 1987, to include a more i

specific description of nonnal breaker position. This problem appears to be unique at Brunswick to the motor-driven fire pum (OPEN) Inspector Followup Item 325/87-06-02, Repair of Diesel Generator Exhaust Silencers. The inspector examined the rusted i bottom of the exhaust silencers and does not consider the condition i to adversely affect the operability of the diesels at this time. The i l

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licensee plans to replace the silencers by December 1,1988, to prevent any potential problems. This item remains open pending -

installation of the new silencer (CLOSED) Inspector Followup Item 325/87-06-03, Documentation of Welding Associated with LER 1-84-02. The inspector was unable to obtain any information regarding additional. welding records associated with the repair of the EH The inspector reviewed the as-is LER package which contained the EWRs, trouble tickets, design drawings, and certain portion; of the welding documentatio Appendix'B does mot apply to the non-safety EHC system, (CLOSED) Inspector Followup. Item -325/87-20-01 and 324/87-20-01, Enhancement of PID Trackin The licensee has developed a .PID tracking program which includes identification through acceptance -

milestones and the accompanying research of- late itens. The inspe'ctor reviewed DI-LRP-21, Revision 0, dated December 21, 1987, which provides the procedural documentation of the administrative controls relative to the PID tracking system, and found them to be 4 adequate. The new system also identifies and tracks PIDs that have

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been accepted, but remain open for completion of the defined scop Site procedure BSP-14, BSEP Project Identification, is currently being revised to incorporate the changes to the PID process. This revision is to be completed by June 30, 198 No significant safety matters, violations or deviations were identifie . Information Meeting With Local Officials (94600)

The inspectors and the project section chief explained the NRC's role and inspection program to local officials. Meetings were held with members of the New Hanover County Commission on April 14, 1988, and with the Chairman of the Brunswick County Commissioners on April 15, 198 The project section chief described the NRC inspection progrco to the City of .

Southport board of aldermen at their regularly scheduled monthly meeting on April 14, 1988.

The inspectors and section chief visited the local public document room at

the University of North Carolina at Wilmington. The collection was well

maintained and in order. The cognizant librarian stated that collection use averaged 1 person per month. The inspectors had no further question . Plant Startup from Refueling - Unit 2 (71711)

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The inspectors reviewed / observed activities associated with Unit 2 startup

, after the refueling outage to determine if activities were conducted in accordance with approved procedure The inspection included a review of the changes to GP-01, Revision 105, Startup Checklist; direct observation of startup and approach to criticality; observation of selected

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surveillance -tests; . review of licensee's drywell closecut inspection ' '

results with independent drywell inspection; examination of selected systems to assure startup readiness; and review of selected modification training package ;

, Specific inspection items included:

(1) Training During the refueling outage for Unit 2, the Alternate Rod .

Injection system was installed under PM-86-035 to comply with 10 CFR 50.62, the:ATWS rule. The purpose of the ARI system is to initiate a reactor _ scram by' a means independent of 'the Reactor Protection Syste Four_ additional vent paths have been

~ installed on the scram outlet valve air header, with each of the four vent paths consisting of two solenoid valves in serie The system is automatically initiated from ARI logic or can be manually initiated, if require The inspector reviewed the modification-training package, sample-examinations and examination test results to determine the adequacy and completeness of the training conducted for those operators licensed on Unit (2) Valve Lineups The completed valve lineup sheets for the Control Rod Drive and the Standby Liquid Control System were reviewed to assess their completenes In addition, the inspector physically verified the valve positions of the SLC system valve (3) Inspection

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Several areas were inspected that are normally high radiation !

areas to look for housekeeping and general material condition The areas inspected included the HPCI roof, RCIC steam tunnel i and the 66 foot penetration roo ,

(4) Drywell Closeout j

The inspector reviewed licensee preparations for closing out the-drywell. Administrative Procedure, AP-96, Drywell Closecut, was

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the governing procedur The inspector reviewed those discrepancies identified by the licensee along with their corrective action. In addition, a physical inspection of the i drywell was performed to verify that licensee actions were i adequat )

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b. Inspection findings were as follows:

(1) Victoreen Radiation Detector Cable During the drywell inspection closecut, the inspector noted that i.he installed configuration of the Victoreen high range radiation monitor differed from the tested configuration as

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documented in Victoreen report 950.30 Specifically, the method of terminating the cable to the detector, the cable manufacturer and lack of a sealed conduit system for the detector cable, all differ from that which was teste The termination method and use of a sealed conduit system by Victoreen in their test program was crucial to the successful completion of the test as evidenced by the numerous - test failures experienced in previous testing without -this configuration. The licensee did provide some information to the inspector describing the differences between their termination means and that which was tested and the specific measures taken to preclude moisture entry to the connector are No information has been provided which demonstrates that the installed cable which is supplied by a different manufacturer and is not installed in sealed conduit, is acceptable for this applicatio This item is unresolved and is listed as Unresolved Item: Qualification of Victoreen Radiation Detection Cable,(324/88-15-03).

(2) Improper Change In Operational Condition l While observing startup preparations being made for Unit 2 startup on April 26, 1988, the inspector first noted at about 2:45 p.m. that the Reactor Mode Switch was in the Startup/ Hot Standby position. Licensee personnel explained that the switch was in this position because of previous testing performed on the Rod Sequence Control System and the Rod Worth Minimizer and that it was left in this position because their procedures -

allowed this if reactor startup would commence shortly. The  ;

testing was completed at 9:45 a.m. Startup commenced at 4:00 p.m. A detailed sequence of events for both the control room '

personnel and the NRC inspector is contained in enclosure This enclosure also addresses appropriate procedures and steps utilized during the evolutio The Rod Worth Minimizer Periodic Test 1.6.2-2, Step 7.1.23, states to place the mode switch in SHUTDOWN unless the shif t  :

foreman verifies that the prerequisites are met for leaving the {

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mode switch in START /H0T STANDBY pending reactor startu The 1 Rod Sequence Control System Operability Periodic Test 1.6.1, '

which was performed after the Rod Worth Minimizer Test, did not contain this step. - At 9:45 a.m., following completion of the ]

I testing, the following prerequisites had not been met for i etartup:

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o RHR Division II not lined up for automatic LPCI initiatio The "B" RHR loop was in the shutdown cooling mode at this time and not lined up for automatic LPCI initiatio At 2:03 p.m., the licensee stopped the RHR "B" pump to secure the shutdown cooling mode of RHR and line it up for LPCI initiation in accordance with Section 7.2 of OP-17. The licensee completed steps through B.4 in this procedure (starting recirculation pumps) at 2:53 p.m. Steps 5 through 16, which would have completed the restoration and lined up the system for automatic LPCI initiation, were not continued at this time. At 3:03 p.m., GP-1 was signed off by the SF and SOS stating that all prerequisites for startup had been met. A PA announcement was made that primary and secondary containment was in effect and that reactor startup was commencing. GP-2 was entered and steps 5.2.4, 5.2.5 and 5.2.6 were completed. The next step, step 5.2.7, states to withdraw control rod At this time the NRC inspector noted that the RHR system had still not been l restored. Specifically, he noted that the suppression pool suction valves F0208, F004B and F004D were shut. These valves do not receive an open signal during LPCI initiatio When questioned about this configuration, the SF directed his people to restore the lineu Steps 7.2.B.5 through 16 of OP-17 were completed at 3:46 GP-1 was then signed off at 3:46 p.m. and reactor startup commenced at 4:00 i o Nitrogen backup system inoperabl The nitrogen backup system supplies a pneumatic source to selected safety-related load Following a LOCA and subsequent containment isolation, the normal air supply to the drywell will be isolated and supply will be from the ;

nitrogen backup system to the suppression pool to reactor !

building vacuum breaker and the SRVs. This system was tagged out in accordance with the requirements of OG-3, Primary Containment Access Control, which describes the i requirements for allowing personnel entry into the drywel l This procedure requires that RNA-SV-V5251 and RNA-SV-V5253 be closed and placed under SF clearanc- if the plant is in Condition 1, 2, or 3 or that the nitrereus backup isolation valves, RNA-V347 and RNA-V348 be tagged closed if the plant is in Condition 4 or 5. Clearance 2-13578 was hung at 4:25 a.m. on April 25, 1988, placing a tag on the control switches (closed) to 2-RNA-SV-V5251 and 2-RNA-SV-V525 These valves supply air to the non-interruptible air header which supplies the SRV These valves open on loss of powe The plant was in Condition 1 at the time the clearance was hun . . . . . .. .- . . . - . -.

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Although the safety significance of this clearani.e being in effect during the mode change is small, since the plant was not at pressure and the'SRVs were not required, the system *

was part of an active LC0 (A2-88-0716) and should have:been cleared prior to any mode-change .

o Primary Containment - Airlock Door Surveillance .Not Completed.

, PT-2.6.6 was completed at 11:11 This test verifies

that the drywell airlock is operable. It was necessary to 1

perform this test since the airlock was previously made-inoperable to support welding work . performed in the drywell. During this mode change period, therefore, primary

containment was not in effect as required by the plant's Technical Specification ;

! This startup showed several examples of failure to follow

! procedures, e.g., signing off OP-1 before it was complete, not J finishing 0P-17, not following the' instructions of PT-1,6.2- In addition, it showed an inccrrect interpretation of their t Technical Specifications regarding testing the Rod Worth Minimizer and the Rod Sequence Control Syste The Technical Specifications, Section 3.1.41 and 3.1.4.2, state that "Entry into Cor.dition 2 and withdredal of selected control rods is permitted for the purpose of determining operability of the RWM i (RSCS) prior to withdrawal of control rods for the purpose of .

bringing the reactor to criticality." The licensee interpreted this statement as allowing the mode switch to be placed in

"startup" for testing purposes. This requirement addresses

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l entry into Condition 2 which means that the prerequisites for t meeting that mode change must be satisfie ,

The licensee's position was that, although procedurally f permitted, they would not have pulled control rods with' the

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Division II RHR system not lined up for automatic LPCI s i initiation. During startup evolutions, the licensee stations an ,

j additional SR0 in the control room to oversee startup

activitie Although his tasks are not administratively ,

defined, the licensee states that this individual was aware that l

the LPCI lineup had not been restored and he would have i prevented pulling control rods without the system lined up for  !

automatic initiation.

1 This matter is a Violation of the licensee's Technical i

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Specifications: Improper Change In Operational Condition,  !

(324/88-15-01).  !

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One violation and no deviations were identified. A failure in the l licensee's administrative control of startup prerequisites was identifie .___ _ - _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ - - - - - _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ - - _ _ - _ _ _ - _ - _

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- 14 12. Onsite Followup of Events (93702) Auxiliary Contacts in Motor Control Centers Breakers GE _has found a potential-failure +. ode of an auxiliary contact block assembly used with contractors 'in safety-related motor. control '

center The licensee had previously -identified a problem with auxiliary contAt blocks (CR-205 device) in early _1987 (see LER ,

1-87-01). The CR-205 ' devices were sticking,-preventing

. motor-operated valve operation. The licensee replaced all CR-205-devices in important safety-related applications with a newer design '

CR-305 device which was not susceptible to the CR-205 failure mod However, .the licensee has found three sticking--305 devices since -

April 6,1988; two failed during replacement installatio GE -

.L Bloomington's initial assessment indicates that- the manufacturing process was creating small burrs on the moving parts, resulting in the failures. GE has modified the process and slightly re-designed the part to fix the proble ,

The licensee immediately inspected .133 important safety-related

, breaker compartments and verified that no installed 305 device was '

sticking. On April 29, 1988, in a conference call.with Region Il management, the licensee reported that they had again inspected (a week from previous inspection),133 important safety-related breaker compartments and verified that no installed 305 device was stickin In addition, they described their test program of the- 305 devices being conducted jointly with General Electric. The testing began on April 28 to verify the failure rate of the old and new 305 device The testing results will determine if the licensee's intention to -

, replace the old 305 devices with new 305 devices on an "as failed" basis or during preventative maintenance should be modifie The inspectors will continue to follow the licensee's actions regarding auxiliary contractors. This is an Inspector Followup Item:

Failures of GE 305 Auxiliary Contact Adder Blocks (325/88-15-04 and 324/88-15-04). Silicon Bronze Bolts in Safety-Related Switchgear The licensee infomed the inspector on April 19, 1988, that the broken and cracked bolts (see report No. 325,324/88-05) had failed'

due to intergranular stress corrosion cracking instead of excessive torque. This determination was made through metallurgical analysis conducted by the Harris E&E Center. The licensee, in a conference call with NRC personnel on April 29, 1988, stated that they would be replacing the 5/16 inch silicon bronza carriage head bolts with mild steel bolts in all safety-related electrical panels except the DC switchbeards by May 17, 1988; the DC switchboard plan would be ready in 30 days. This is an Inspector Followup Item: Silicon Bronze BoltsinSafety-RelatedSwitchgear(325/88-15-06and324/88-15-06).

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No significant safety matters, violations or' deviations were identifie . List of Abbreviations for Unit 1 and 2 AI Administrative Instruction A Auxiliary Operator ARI- Alternate Rod Injection

, ATWS Anticipated Transient Without Scram BESU Brunswick Engineering Sub Unit BSEP Brunswick Steam Electric Plant +

BSP Brunswick Site Procedure C0 Control Operator CP&L Carolina Power and Light Company CRD Control Rod Drive CST Condensate Storage Tank Direct Current

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DC DG Diesel Generator DPT Differential Pressure Test -

E8E Energy and Environmental EER Engineering Evaluation Report EHC Electro Hydraulic Control System ENP Engineering Procedure ERFIS Emergency Response Facility Information System ESF Engineered Safety Feature EWR Engineering Work Request ,

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F Degrees Fahrenhtit '

i GE General Electric j GP General Procedure ,

HP- Health Physics lj

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HPCI High Pressure Coolant Injection HX Heat Exchanger I&C Instrumentation and Control

, IE NRC Office of Inspection and Enforcement IFI Inspector Followup Item IPBS Integrated Planning Budget System i IRM Interrediate Range Monitor i LC0 Limiting Condition for Operation LER Licensee Event Report LOCA Loss of Coolant Accident LPCI Low Pressure Coolant Injection

, MG Motor Generator

. MST Maintenance Surveillance Test

NCR Non-Conformance Report NRC Nuclear Regulatory Conrnission OER Operating Experience Report j OP Operating Procedure i OPM Operating Procedure Manual PA Protected Area

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PID Project Identification

. PM Plant Modification

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PNSC Plant Nuclear Safety Committee PSIG Pounds per Square Inch Gauge PT Periodic Test QA Quality Assurance QC Quality Control RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RSCS Rod Sequence Control System RTGB Reactor Turbine Gauge Board RWM Rod Worth Minimizer RX Reactor SF Shift Foreman SI Standing Instruction SIL Service Information Letter SLC Standby Liquid Control SOE Sequence of Events SOS Shift Operating Supervisor SP Special Procedure SR0 Senior Reactor Operator STA Shift Technical Advisor S/U Startup SW Service Water TS Technical Specification TSI Technical Specification Interpretation URI Unresolved Item j

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ENCLOSURE _2 U-2 Startup Sequence of Events for 4/26/88 Initial Conditions: All rods in, Reactor temperature 180 degrees F Reactor Coolant System depressurized with manual head vents ope Procedure _ Step Time Action 0435 Mode switch 1.6.1 and in S/U 1.6.1-2 forC0's (from testing)

log pts 2041 PT-1.6.1 complete (RSCS - from SF log)

(0945 from GP-1, Attachment 1)

1046 PT-1.6.2-2 complete (RWM - from SF log (0912 from GP-1, Attachment 1)

1406 Secured Shutdown Cooling (from C0's log)

1419 Drywell LC0 cancelled (SF log)

1430 NRC inspector asks U-2 SF about status of RHR Loop "B" and why it was not lined up for auto LPC1 initiati o Informed by SF that procedure was in progress and that following start of

"B" Recirc. Pump the lineup would be restored. (Note: OP-17, Step 7.2. shows this to be the correct sequence.)

1445 NRC inspector asks U-2 SF why mode switch is in S/ Informed that it was placed in that position earlier for testing and that their procedures allow them to keep it there if S/U expected shortl (Note: Step 7.1.23 of PT-1.6.2-2 states to place mode switch in shutdown, unless the SF verifies that the prerequisites are met for leaving the mode switch in START-HOT STBY pending reactor startup.)

Step 7.2.B.4 of OP-17 1453 Started 2B Rx Recirc. MG set (from Step 5.1.16 of GP-1 C0's log)

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Enclosure 2 2 Procedure _S,tep Time Action 1503 Primary and Secondary Containment in effect(CO' slog)

1503 GP-01 Startup Checklist complete (time later changed to 1546)

Step 5.1.13 of GP-1 1505 PA announcement that Primary and Secondary Containment in effect, commencingRxS/U(fromChemistrylog)

Step 5.2.4 of GP-2 1507 Verified Rx Vessel Shell temperature to 5. right of criticality line (C0's log)

1510 (After PA announcement) NRC inspector again asks about status of RHR "B" Loop. U-2 SF talks with U-2 operators and then flush begins on RHR "B" Loop per Step 7.2.B.5 of OP-17, 1516 Commenced Reactor Startup (SF log, time i later changed to 1600) l Step 5.2.6 of GP-2 1519 Chemistry informed of U-2 mode change to mode 2 at 1516 (time later changed to 1600 - Chemistry log)

1530 When leaving Control Room, NRC inspector overhears SOS asking U-2 operators why rods are not being pulled. He was told that RHR flush was in progress, to which he asked, why can't one operator do the flush while the other pulls rods?

The operators explained that the flush had to be done in order to restore the RHR "B" Loop for auto LPCI initiation and rod pull would begin after LPCI "B" was lined u Step 7.2.B.13 of OP-17 1546 RHR B Loop in standby per OP-17 (CO's Step 5.1.15 of GP-1 log)

Step 5.2.4 of GP-2 1559 Verified Rx Yessel Shell temperature to right of criticality line (C0's log)

Step 5.2.7 of GP-2 1600 Cormenced Rx S/U, first Rod out (SF, C0's log)

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Enclosure 2 3 APPLICAB_LE _PROCEDU_R_ES_

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PT-1.6.2-2 Rod Worth Minimizer System Operability Test PT-1. Rod Sequence Control System Operabilit OP-17 Residual Heat Removal System Operating Procedure GP-01 Startup Checklist GP-02 Approach to Criticality and Pressurization of the Reactor

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