IR 05000324/1997011

Insp Repts 50-324/97-11 & 50-325/97-11 on 970819-0927. Violations Noted.Major Areas Inspected:Licensee Operations, Engineering,Maint & Plant Support

ML20199C260
Person / Time
Site:	Brunswick
Issue date:	10/27/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20199C229	List: IR 05000324/1997011 ML20199C227 ML20199C238
References
50-324-97-11, 50-325-97-11, NUDOCS 9711190282
Download: ML20199C260 (51)

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U. S, NUCLEAR REGULATORY COMMISSION- 1 i

REGION 11 .l

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Docket Nos: 60-325. 50-324 l Litense Nos: DPR-71. DPR 62  ;

. . t Report No:- .50-325/97-11. 50-324/97-11 l l

Licensee: Carolina Power & Light (CP&L) q

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l Facility: Brunswick Steam Electric Plant. Units 1 & 2 r

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Location: 8470 River Road SE  !

Southport. NC 28461  ;

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Dates- August.19 - September 27, 1997

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inspectors: C. Patterson. Senior Resident inspector j E. Brown. Resident Inspector >

E. Guthrie, inspector in Tra Ming E. Girard, Reactor Inspector (Sections E1.1. E8) .'

W. Rankin. Regional Inspector (Sections R1.1. R4 R8)

J. Coley. Reactor Inspector (Sections M2.3 - M !

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M7.2) '

R. Aiello. Regional Inspector (Section 0.5) .

Approved by: M. Shymlock. Chief. Projects Branch 4 Division of Reactor Projects

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EXECUTIVE SUMMARY Brunswick Steam Electric Plant. Units 1 & 2 NRC Inspection Report 50-325/97-11. 50-324/97-11 This integrated inspection included aspects of licensee operations.

engineering, maintenance. and plant support. The report covers a 6-week period of resident inspection; in addition, it includes the results of a training inspection maintenance inspection, health physics inspection, and motor operated valve (MOV) inspection by regional inspectors.

Doeration1 e lhe inspector concluded tnat Brunswick was able to set a new Boiling Water Reactor dual unit continuous operation record due to excellence in op2 rating and naintaining the plant. (Section 01.1)

e During controi room observation of fuel loading. operators stopped fuel movement as required once conounications were los Good control of refueling activities was observed on the refueling bridge. (Section 01.2)

e The control copies of Technical Specifications (TSs) were being maintained. One missing page was identified which was promptly corrected. (Section 01.3)

e The inspector identified a weakness in the power transfer to the E8 emergency bus DC control power. The operator did not verify th the transfer of the DC control power source was successful and the p.ocedure did not specify how the operator was to ensure that the power transfer was successful. (Section 01.4)

e The difference in labeling of Single Scram Point equipment between the Unit 1 and Unit 2 control boards demonstrates a lack of consistency from a Human Factors perspective and a responsible organization to maintain consistency. (Section 02.1)

e Upon losing breaker control the licensee failed to enter into an alternate safe shutdown (ASSD) impairment for the Residual Heat Removal (RHR) Shutdown Cooling Suction Valve. This failure was the result of an inaccurate listing of ASSD components located in the Equipment Database Systems. This failure was identified as a violation. (Section 03.1)

e The inspector identified no cultural differences between License Operator Requalification and Hot License training. (Section 05.1)

e The inspector reviewed 001-01.02. Conduct of Operations. Revision 8 and found it to be satisfactory. (Section 05.2)

e After reading the World Association of Nuclear Operators report the inspector concluded that there were no new significant findings unknown to the NRC. (Section 07.1)

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e The inspector concluded that a violation of TS occurred because the Plant Nuclear Safety Committee meeting was not conducted with the recuired membership quorum. The chairman assigned an action item for an incustry experience review deficiency. (Section 07.2)

e The facility's philosophy of safe reactor oparation was supported and reinforced by training and constant conservative decision makin (Section 08.1)

Maintenance

e The high watec level Remote Shutdcoi Panel trip channel calibration was completed satisfactorily with no identified concerns. (Section M1.1)

e The licensee changed out 30 control rod drives. The changeout was necessary due to high insert and/or withdrawal stall flows duruig drive movemen Satisfactory supervisory oversight. communications. and health physics technician oversight-was observed during the changeou (Section M2.1)

e The inspector concluded that. modification work on the RHR and Core Spray suction strainers was progressing smoothly due to excellent planning of the job. f,adiation control activities in the torus and drywell were well controlle (Section M2.2)

e Evaluation of anomated ultrasonic data revealed that examinations cnnducted on tne Reactor Pressure Vessel vertical welds were performed in accordanc.: with the approved examination procedure by skillful and knowledgeatae ultrasonic test examiners. (Section M2.3)

e Evaluation of automated ultrasonic data revealed that examinations conducted on recirculation system welds were performed in accordance with approved examination procedures by skillful and knowledgeable ultrasonic test examiners. (Section M2.4)

e Evaluation of automated ultrasonic data revealed that examinations conducted on the reactor core shroud welds were performed in accordance with approved examination procedures by skillful and knowledgeable ultrasonic test examiners using state-of-the-art examination equipmen (Section M e The inspector concluded that " pen and ink" changes were made to a safety-related surveillance test procedure contrary to the TS requirements for processing a temporary revision. This was identified as a vialation. This occurred due to guidance in a plant procedure and direction given by I & C supervisors. (Section M3.1)

e The inspector concluded that several recent chlorine system leaks could have been prevented by fully implementing post-maintenance testing for all parts of the chlorine syste Inadequate attention to detail in assembly of components attributed to the leakage problems. Additional

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i manag2 ment attention is needed in the maintenance and testing of this l system. (Section M7.1)

e Nuclear assessments, performed by the licensee ir maintenance and inservice ins)ection, which identified findings in hot work practices, radiation worter practices, equipment storage and control, and failure to issue condition reports on identified problems was considered a strength. (Section M7.2)

Enaineerina e Generic Letter (GL) 89-10 had not been satisfactorily implemente (Section El.1)

e One violation was identified for inadequate corrective actions for motor operated valves. (Sections El.1)

e The inspector concluded that Technical Support Memorandums (TSMs) were being referenced although TSM's wer in a nonactive statu The full impact of this requires further review. (Section E3.1)

e The inspector identified improper rigging of the torus suction strainer mock-up to safety-related piping. Subsequent reviews by the licensee of other areas revealed that non-seismically mounted temporary piping was installed over a safety-related motor control center without a temporary rigging release. The improper rigging of the strainer mock-up and suspending piping without d temporary rigging release was identified as a violatio (Section E4.1)

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Plant Suonort e The licensee's radiation control 3rogram in the areas of radiation surveys. RWP controls, and pre-jo) planning and ALARA briefings, was determined to be implemented effectively and in accordance with procedures. (Section R1.1)

e The radiological controls program was effectively implemented with good occupational exposure controls demonstrated during outage cond:tion Good radiological control performance was evident in the occupational exposure control activities observed. A violation was identified for failure to properly label containers of radioactive materials. (Section R1,2)

e A violation was identified for a contract worker's repeated failure to log or, to a radiation work permit and obtain the required monitoring dosimetry. Contractor knowledge of general radiological control requirements was poor as evidenced by several radiological control noncomformances and noted as a weaknes (Section R4.1)

e The inspector concluded that the licensee conducted a thorough review of a situation where a contract employee had a positive whole body count

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indicating internal radioactivity of 43 nanocuries upon termination of site duties. The information provided to the NRC was consisten (Section R7.1)

e The licensee's ALARA program was adequately controlling collective dose and collective dose was on a favorable reducing trend. (Section R8.1)

e Appropriate actions were observed during morning peak access activitie A weatness was identified which had the potential to allow items to '

bypass normal monitoring methods. (Section S4.1)

e The inspector concluded that the licensee was still struggling with control of transient combustible material, but had accelerated their corrective actions to clearly identify areas where transient combustibles were not allowe (Section F2.1)

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I Reoort Details Summary of Plant Status

_ Unit 1 operated continuously during this period. Two control rods remain inserted around an identified fuel pin leaker. During this period, two additional control rods were inserted around the existing fuel pin leaker. Indication of a second fuel pin leaker developed and two control rods were inserted around the second leaker. A mid-cycle shutdown outage is planned for November 5. 1997, to remove the leaker At the end of the period the unit had been on-line 324 day Unit 2 operated cor.tinuously duiing this period until September 1 , completing 363 days of operation before starting a planned 35-day refueling outag The mechanical vacuum ptmps remained tagged out on Unit I due to cencerns about control room dose in the event of a Rod Drop Acciden The licensee, in a letter to the NRC dated February 13. 1997, committed to upgrade the mechanical vacuum pump trip function to implement a vacuum pump trip from the main steam line radiation monitor prior to the next startup. This modification was completed on Unit 2 during the current refueling outag Due to concerns about the control room dose the licensee imposed an adminittrative limit on Iodine until a Technical Specification (TS)

amendment submitted was approve The licensee made a procedure change to Administrative Procedure OAI-81. Water Chemistry Guidelines, setting the limit at 0.1 microcurie per gram dose equivalent Iodine 131 compared to the TS value of 0.2 microcurie 3er gram. Also, the licensee has been aro'ciding weekly data to NRR and t1e resident inspector for revie lone of the data reviewed has exceeded the administrative limi Due to a reassessment of the Environmental Qualification (EO) program and items identified, there are 21 of 23 Justification for Continued 03eration (JCO) that remain open for both units. The following provides tie status of the EO JCOs and associated Engineering Service Requests (ESRs):

1) ESR 96-00425. Evaluation of E0 sealants was previously closed by the licensee but was reopened - closure date to be determined (TBD).

2) ESR 97-00331 (old ESR 96-00503). Associated Circuit E0 was closed by the license ) ESR 96-00426. Evaluation Quality class and E0 classification of Post Accident Sample System (PASS) valves was scheduled for completion June 6.1997, closure date TB ) ESR 97 00330 (old ESR 96-00501). Motor Control Center (MCC) EQ was previously closed by the licensee, but was reopened - closure date TB ) ' 97-00329 (old ESR 96-00625). EQ Type JC0 for EQ Fuses Without a ..uolification Data Package (ODP) was closed by the licenre __ - __

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6) ESR 97-00340 (old ESR 96-00627). ODP for Marathon 300 Terminal Blocks was scheduled for completion December 31, 1997 but revised to August 1. 1997, but closure date is now TB >

7) . ESR 97-00087. E0-Type JC0 for Improperly configured Conduit Seal was previously closed by the licensee but was reopened - closure dated TB ) ESR 97-00229. JC0 for GE Condition Report (CR) 151 B Terminal Blocks was scheduled to be completed September 1. 1997 closure date TB ) ESR 97-00238. JC0 for Standby Gas Treatment Motor Operated Valve (MOV) Position Indicator Rheostat, was closed by the licensee but was reopened - closure date TB ) ESR 97-00250. Conduit Union in EQ Boundary, was scheduled for completion December 31, 1997, but closure date is now TB ) ESR 97-00256. Main Steam Insulation Valve (MSIV) Hiller Actuator JCO was scheduled for completion September 2. 1997 but closure date is now TB ) ESR 97-00289. PASS Valve Limit Switch Panel Wiring, scheduled for completion September 15, 1997, but closure date is now TB ) ESR 97-00343. Qualification of Kulka Model 600 Terminal Blocks was scheduled for completion September 1. 1997, but closure date is now TB ) ESR 96-00587. PASS Valves. closure date TB ) ESR 97-00449. Degraded Junction Boxes. closure date TB ) ESR 97-00435. MCC Fittings, closure date TB ) ESR 97-00446. GE Radiation Detectors, closure date TB ) ESR 97-U0500 Grayboot Connectort closure date TB ) ESR 97-00535. Target Rock Solenoids Terminal Block Spray closure date TB ) ESR 97-00529. Failure of Unit 1 Drywell Motor, closure date TB ) ESR 97-00534. GE EB-5 Type Terminal Strips closure date TB ) ESR 97-00523. High Pressure Coolant Injection (HPCI) Auxiliary 011 Pump Motor Unit 1. closure date TB ) ESR 97-00513. In-Board Drywell Electrical Penetrations, closure date TB In summary. Unit 1 operated continuously during this report period and Unit 2 started a refueling outage, However, there were 21 outstanding JCOs in the E0 area identified for both nit I. Operations 01 Conduct of Operations 01.1 Dual Unit Ooeratino Record Insoection Scoce (71707)

The inspector reviewed the plant continuous operating run time for both unit i l

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3 Observations and Findinas On August 20. 1997. Brunswick set a new world record for continuous dual unit Boiling Water Reactors (BWR) operating run time. The old record set by another facility was 285 days 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />. and 22 minutes. On August 20. 1997. Unit 1 had operated 286 continuous days and Unit 2. 340 continuous days. This was attributed to the excellence in operating and maintaining the plants that kept both units on line for this extended period of time, Conclusions The inspector concluded that Brunswick was eble to set a new BWR dual w unit continuous operation record due to excellence in operating and maintaining the plan .2 Unit 2 Core Off-load Insoection Scone (71707)

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The inspector reviewed Unit 2 fuel movement for Unit 2 core off-loa Observations and Findings On September 18, 1997. the inspector observed fuel movement on the refuel floor and in the control room. Core off-load began at 5:10 on September 18. 1997. The inspector observed that a senior reactor operator (SRO) was in charge of the refueling operation and was on the bridge crane during fuel movement. The inspector observed that no one on the bridge was using a headset for communication. Communication was by a s]eaker on the bridge and bridge personnel picked up a microphone to talc to the control roo Later, while in the control room at the refueling operation desk, the inspector observed that connunication from the control room to the refuel bridge was being conducted by using a hand-held radio. The inspector observed that several attempts by the control room to reach the refuel bridge were not successful. The operator expressed concern to the control room supervisor about the communication difficulty. The SRO stoaped the refueling activities until constant communications could be esta31ishe Af ter 10 minutes, communications were re-established with the refuel bridge using the Gaitronics system. Core alterations were recommence The SRO initiated CR 97-03177. Loss of Refuel Radio Communications, to document this proble The inspector reviewed procedure 0FH-11. Refueling, and no specific type of communication equipment was specified for establishing communications between the control room and refueling bridge. The procedure indicated that fuel movement be stopped if communications were los _- _____ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - __ _ _ - _ _ _ - _ _ _ _

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On September 19. 1997, the inspector observed fuel unloading movements ~~

from the refueling bridge. The inspector observed good procedural adherence pertaining to the required communication between the operator

. and refueling SRO the requirements for the control of fuel movement sequence and placement, and requirements for independence between the SRO and spotter on the refueling bridge, Conclusions During control room observation of fuel offload, operators stopped fuel movement as required, during an instance where communications was los Good control of fuel offload activities was observed on the refueling bridg .3 Audit of Technical Soecifications insoection Scone (71707)

The inspector conducted a page-by-Jage audit of the TS's maintained in the Control Room. The inspector clecked that all of the pages with the correct amendments were presen Observations and findinas The inspector audited controlled copics numbered 019 and 020 for Unit 1 and Unit 2 respectively. No discrepancies were found in the Unit 2 1Ss. However, the inspector found that a page was missing in the Unit 1 TSs. The missing page was 3/4 2-4. Amendment 159. Page 3/4 2-4 is a continued page which has the Action Statement and Surveillance Requirements for the Minimum Critical Power Ratio (ODYN OPTION B).

Limiting Condition For Operation. This is an infrequently used sectio The operator replaced the page immediately after the deficiency was identifie Conclusions The control copies of TS's were being maintained. One missing page was identified which was promptly correcte .4 Loss of E8 Emeraency Bus DC Control Power Insoection Scoce (71707)

The inspector reviewed the sequence of events associated with a loss of DC control power to the E8 emergency bus which occurred on September 23, 1997. The inspector reviewed TS's, plant 3rocedures. DC control power alignment, operator logs, and the updated ;inal Safety Analysis Report (UFSAR) to verify that the actions taken by the licensee were in compliance with the requirement l

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5 Observations and Findinas The E8 emergency bus DC control power was lost on September 23. 1997, when an automatic bus transfer (ABT) switch did not transfer to the alternate source of power due to a malfunction in the ABT. The ABT was to l'e shifted to its alternate position by an operator manually repositioning the supply circuit breakers. Control power transfer to the alternate power source was necessary since the normal power source was going to be secured, which, in this case, was the N1 battery system. The transfer of DC control power was directed by 0)erating Procedure 00WP-51/1. Removal of the 125 VDC Battery System rom Service including DC Control Power Alignment. Approximately seven hours after the control power transfer, it was determined that the E8 emergency bus

.in service did not have DC control power. This condition was found during a control room panel walkdown during shift turnover. The o)erator identified that the indicating breaker status light was out on t1e E8 emergency bus supply circuit breaker. A weakness was identified in control of the power transfer process. During the initial transfer, the operator did not verify that the transfer had been successful. The procedure did not require any indication be checked or verified to ensure that the transfer was successfu This condition rendered the E8 emergency bus inoperable and in accordance with TS 3.8.2.1 recuired that the bus be restored within eight hours or be in hot shutcown within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The ABT switch was repaired and DC control power was restored to the E8 emergency bus prior to exceeding the TS action statemen c, Conclusions The inspector identified a weakness in the power transfer to the E8 emergency bus DC control power. The operat6r did not verify that the transfer of the DC control power source was succ':ssful and the procedure did not specify how the operator was to ensure that the power transfer was successfu .5 Soecial UFSAR Review

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A recent discovery of a licensee o)erating the facility in a manner contrary to the UFSAR description lighlighted the need for a special focused review that compares plant practices. procedures, and/or parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practice;, procedures, and/or parameter During this period the inspector reviewed operation of the control power for emergency buses as discussed in section 0 No problems were identi fie i

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02 Operational Status of t'acilities and Equipment 02.1 Disoarities in the labelina of the Sinale Scram Point Identification Plates Inspectico Scoce (71707)

On September 19. 1997, the inspector observed the Single Scram Point identification label plates af fixed to the Unit 1 and Unit 2 Control Room control boards. The inspector looked for continuity of labeling between the two units' control board Observations and Findinas The inspector identified five disparities in the labeling of the Single Scram Points (SSPs) between Unit 1 and Unit 2 control boards. The SSPs were identified on each unit by red label plates which indicate " Single Scram Point." The labeling was ap)arently used to identify a control function such as a switch or push-;utton which, when actuated, could potentially either directly or indirectly cause ac inadvertent reactor scra The five differences between the two units' labeling were that Unit I had the SSP label plates on the following equipment. whereas Unit 2 did not have the same ec,uipment identified:

1) Reactor Feed Pump Turbine 1A 2) Reactor Feed Pump Turbine IB 3) Reactor Building Closed Cooling Water to Drywell Isolation Valves 4) Condenser Vacuum Breakers 5) Condensate Deep Bed Demineralizer/ Condensate Filter Demineralizer Valves Conclusions The difference in labeling of Single Scram Point equipment between the Unit 1 and Unit 2 control boards demonstrates a lack of consistency from a Human Factors perspective and a responsible organization to maintain consistenc Fire Protection Procedures and Documentation 03.1 Alternate Safe Shutdown Impairment Not Entered Inspection Scone (71707)

The inspector reviewed the circumstances surrounding a breaker malfunction during the performance of Periodic Test OPT-8.1. LPCI/RHR System Component Test - Loop _ - - - - _ -- --- - - _ - - - - - - - - - - - - - - --- - - - - - - _ - - _

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7 Observations and Findinas On September 18, 1997, with Unit 1 operating at 100%. the inspector reviewed the operatcr logs as part of the routine inspection activitie It was noted by the inspector that on Se)tember 17, 1997 at 9:40 p.m. a technician was unable to energize a breater to verify valve position for the 1-E11-F006A. Residual Heat Removal (RHR) Pump 1A Shutdown Cooling Suction Valve during the performance of OPT-8.1.3.a. The inspector determined that the associated breaker. DE9 on MCC 1XA was a com)onent used for alternate safe shutdown (ASSD). Further review of the Jnit 1 operator and fire impairment logs revealed no active or tracking limiting conditions for operations or fire impairment The inspector reviewed the September 17, 1997 performance of OPT-8.1.3.a. Fire Protection Procedure 0FPP-20. Impairment Notification and Alternate Safe Shutdown Procedure OASSD-01. Alternate Safe Shutdown Procedure Index. The General Comments section of OPT-8.1.3.a noted that the breaker handle for the 1-E11-F006A was broken internally. Valve position was subsequently verified closed using continuity readings across the closed limit switches. The inability to energize breaker DE9 ccased the inability of the unit to complete those actions outlined in 0ASSD-0 Procedure OASSD-01 required the energizing of breaker DE Licensee review determined that the Equipment Database System (EDBS)

improperly listed the 1-E11-F006A as a non-ASSD componen Procedure 0FPP-20 required that a 14-day ASSD impairment be entered when ASSD equipment is found to be inoperable, not accessible, or not available as required to perform safe shutdown evolutions. The failure to enter into an ASSD fire impairment in accordance with 0FPP-20 is a violation. This violation is identified as VIO 50-325/97-11-0 Failure to Initiate Alternate Safe Shutdown Impairment. The licensee initiated an ASSD im)airment on September 23. 1997. starting the 14-day for restoration clocq on September 17, 1997. In addition the licensee is reviewing the accuracy of the EDBS to assure that no other ASSD components are left off the EDBS ASSD lis Conclusions When the licensee identified that they could not manually closed a valve breaker. they failed to enter into an alternate safe shutdown impairment for the RHR Shutdown Cooling Suction Valve. This failure was the result of an inaccurate listing of ASSD components located in the Equipment Database Systems. This failure was identified as a violatio Licensed Operator Reqaulification (LOR) Program Evaluation and Training and Qualification Effectiveness

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05.1 Consistency Between Initial License and Recualification Trainina (41500). (/1001) Insoection Scope The inspector reviewed portions of the licensed operator training program to verify consistency of Licensed 0)erator Requalification (LOR) and Hot Licensed (HL) training with tlat of actual licensed operator duties: determine how well the individual operators and crews had mastered the training objectives Observations and Findinas The inspector identified that the only difference between LOR and HL training was that the facility trains with six persons in the simulator during LOR and three during H The facility trains with six in LOR to be consistent with how they operate. The facility trains with three in HL to model f!RC standards for NRC administered initial examinations as outlined in Nureg 1021, ES-302, " Administering Operating Tests to Initial License Applicants". Senior reactor operator c6ndidates have to be examined by the NRC in all disciplines for which they are licensed. The maximum number of candidates that can be examined during any one NRC examination scenario is three. Therefore, the SRO candidate may be asked to perform multiple duties in the interest of the examination in order to satisfy the criteria in Nureg 1021, ES-301, " Preparing Initial Operating Tests", Conclusions The inspector identified no cultural differences between LOR and HL trainin .2 Initial and Reaualification Trainina Reaardina the Acoropriateness of Insertina a Reactor Scram Before Tricoina the Turbine (41500). (71001) Insoection Scone (41500. 71001)

The inspector reviewed the licensee's philosophy and determined the appropriateness of inserting a reactor scram before tripping the turbine, and reviewed Operating Instruction. 00I-01.02,

" Conduct of Operations", Revision 8 for conservative decision making, Observations and Findinas Paragraph 5.3.5 of 001-01.02, Conduct of Operations, Revision 7 stated, "If automatic actions fail to occur with valid initiation signals present, operators SHALL initiate associated manual actions." This could imply ad lend one to believe that a manual turbine tri) w3uld be required even though a reactor scram was imminent . 3rocedure 001-01.02 has been revised. Revision 8, l.

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paragraph 5.3.5 now reads, "If automatic actions fail to occur with valid initiation signals present operators SHALL initiate associated manual actions. If the manual actions will result in an automatic reactor scram. THEN INSERT a manual reactor scram

PRIOR T0 performing the manual actions."

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! The inspector reviewed 001-01.02. Conduct of Operations. Revision l 8 and found it to be satisfactor e 07 Quality Assurance in Operations

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07.1 Third Party Reviews Jnspe: tion Scooe (71707)

The inspector reviewed the '.Jorld Association of Nuclear Operators (WANO)

report for the 199 assessmen ; Observations and Findinas The inspector read the WANO report for the assessment conducted on-site during the weeks of April 21 and April 28. 1997. This peer review was conducted by WANO and took the place of the normal Institute of Nuclear Power Operations revie Conclusions After reading the WANO report the inspector concluded that there were no new significant findings unknown to the NR .2 Plant Nuclear Safety Committee Insoection Scope (71707)

On September 25, 1997, the inspector attended a Plant Nuclear Safety Committee (PNSC) meeting and reviewed the activities of this committe Observations and Findinas PNSC is a requirement of TS 6.5.3.3. The committee reviewed a proposed TS change, debrief of a PNSC subcommittee for Improved Technical "

Specification. Drywell' to Torus Bypass issue, and two CRs. The meeting lasted two hours and each item was discussed in detai The chairman provided a good safety focus on the Drywell to Torus Bypass issue. It was noted that on two occasions the licensee failed to properly evaluate industry experience information and to identify that the same issue existed at Brunswick. It was only after questioning by the NRC resident inspector concerning this issue at another BWR facility that the licensee concluded that the same problem existed at Brunswic _ - _ - _ _ _ - _ - _ - _ _ _ _ _ - _ - _ _ _ _ _

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The chairman assigned an action item to further review the failure to properly evaluate industry experience informatio The inspector noted during the meeting that the Plant General Manager, the normal chairman, and the Operations Manager were not in attendanc Several alternates were filling members positions. The inspector further reviewed TS 6.5.3.3 and noted that the PNSC shall be composed of a chairman and seven to nine members. The members shall be from the following areas:

Operations Maintenance Health Physics / Chemistry Regulatory Affairs ,

Nuclear Assessment Engineering .

Members shall be unit level or above and alternates must be designated in writing. A quorum shall consist of the chairman (or his designated'

alternate) and four members (including alternates). No more than two alternates shall be counted toward meeting the quorum requirements at any one tim In this m eting. Operations was an alternate Maintenance was a member but was the alternate chairman, Engineering was an alternate. Health Physics / Chemistry was a member, Regulatory Affairs was an alternate, and Nuclear Assessment was an alternate. Thus, for this PNSC meeting there were only 'wo members, one of which was the chairman, and the rest were alternat The inspector reviewed a letter dated September 24, 1997, designating PNSC members, alternates and alternate chairman. The letter indicated two additional organizations for the Brunswick PNSC, the Training and Outage & Scheduling groups. The licensee initially stated that another member of PNSC was from Outage & Scheduling providing the required quorum However, this organization was not listed in T Accordingly, this was a violation of TS 6.5.3.b for having more than two alternates as quorum members for a PNSC meeting. This violation is identified as V10 50-325(324)/97-11-02, PNSC Quorum Too Many Alternate Conclusions The inspector concluded that a PNSC meeting was conducted without the required quorum. The chairman assigned an action item for an industry experience review deficienc _ _ - _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _

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08 Miscellaneous Operational Issues (92901)

0 (Closed) VIO 50-325(324)/96-10-02: Failure to Disposition Repeated LCR Failures The inspector identified in report 50-324.325/96-10. that licensee management had not taken action to correct repeated LOR failures as required by 1M-1.05. Student Counseling and Remedial Training. TM-1.0 Paragra)h 4.5.1.b. stated that an Academic Review Board is mandatory after t1ree test failures within a 12 month period in a "rogram or course. The inspector identified seven different operators from 1994 through 1996 that failed at least three examinations within a 12 month period. At no time was an Academic Review Board convened to determine whether the student should have been retained in the program with no conditions, retained with special conoitions or dropped from the program entirely as stated in TM-1.05. Attachment 3. Academic Review Board Findings. The inspector reviewed TAP 1.05. Revision 3. (previously TM-1.05 revised). Student Counseling and Remedial Training. The inspector also reviewed several operators' performance records, remedial training plans and academic review board findings. The inspector identified no significant errors or incidences that would be indicative of poor or inadequate remedial training, lhe inspector had no other concerns and this item is close II. Maintenance M1 Conduct of Maintenance M1.1 Remote Shutdown Panel Channel Calibrdtion Insoection Scone (61726)

The inspector cbserved portions of the performance of the Unit 2 Reactor Core Isolation Cooling (RCIC) High Water Level Remote Shutdown Panel Trip Channel Calibratio Observations and Findinas On September 5. 1997 the inspector observed portions of the Unit 2 performance of Maintenance Surveillance Test 2MST-RCIC270. RCIC High Water Level Remote Shutdown Panel Trip Channel Calibration. This procedure verified that high level switch B21-LSH-N017D-3 was properly calibrated and sends a "close" signal to the 2-E51-F045. RCIC Steam Supply Valve. Comunications was satisfactorily established and conducted. The inspector verified that the procedures in use were properly verified and met the TS frequency and acceptance criteria rec uirements. The inspector noted that during this performance no " pen anc ink" changes were eviden Calibration instrumentation used was found to be within the calibration frequency requirements. The test was completed satisfactorily with no concerns identifie l

_ . _ _ _ _ _ _ _ _ - - _- _ _ - _ _ - - - _ _ . _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ - _ - - - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _

_ _ _ _ _ __________

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%

12 Conclusions The high water level Remote Shutdown Panel trip channel calibration was completed satisfactorily with no identified concern M2 Maintenance and Material Conditien of Facilities and Equipment M2.1 Control Rod Drive Exchanae insoection Scoce (62706)

The inspector observed portions of the control rod drive (CRD)

replaccment activitie Observations and Fiaginas On September 24. 1977, the inspector observed portions of the CRD changeoat. Thirty CRDs were removed and replaced with new or rebuilt drives. The renlacements were necessary due to high insert and/or withdrawal stall flow during CRD movement. The inspector noted that adequate communication was maintained between the under-vessel workers, work supervisors, and the control room. Upon loss of communication with an under-vessel worker, the worker exited the work area until communications could be restored. The inspector observed that the reference use procedure. Corrective Maintenance OCM-CRD500. CRD Removal and Installation Inspection using Slim Line Drive Exchange System (SLDES). was present at the work site. CRD uncouoling was verified as well as CRD position and serial number upon removal and installation of the drives. Health Physics technicians were present both under-vessel and in the changeout area. Area dose rates were routinely determined and clean-up activities were being performed concurrent with drive removal to control area contamination. The inspector noted that a satisfactory level of supervisory oversight was present and the job was completed satisfactorily with no identified concern c Conclusions The licensee changed out 30 control rod drives. The changeout was necessary due to high insert and/or withdrawal stall flows during drive movement. Satisfactory supervisory oversight, communications, and health physics technician oversight was observed during the changeou M2.2 Drywell and Torus Inspection Tesoection Scone (62707)

< September 24. 1997, the NRC inspectors and Branch Chief reviewed work activities in the Unit 2 torus and drywel _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-_-_____-_ _ -___ - - __ __ -

..

. _ . . ..

- . . -. _

i 13 Observations and Findinas The torus work on the RHR and Core Spray (CS) suction strainer ,

modifications was observed. This modification was well planned and performed with the torus in a dry condition. On the day of the observations, the new larger strainers were in place with supports being .

welded to the torus lining. The licensee had successfully rigged these large strainers into the torus without difficulty. The inspector attrii;uted this success to thorough planning and using a full-size mockup to test the rigging installatio Activities in the drywell included E0 walkdowns and under-vessel wor Access to the upper elevations was restricted due to fuel movements in progress. The general condition of the drywell was clean for this stage of the refueling activitie Environmental & Radiation Control personnel did an excellent job of controlling drywell activities, pre-entry briefing, and general containment control. Activities were being controlled in a proactive state rather than just reacting to problem Conclusions The inspector concluded that modification work on the RHR and CS suction strainers was progressing smoothly due to excellent planning of the jo Radiation control activities in the torus and drywell were well controlle M2.3 Observation of Data Evaluation Activities for Unit 2 Reactor Pressure Vessel Vertical Welds Examined Usino the GERIS 2000 UT System Insoection Scone (73753)

Brunswick Unit 2 is presently in the second outage of the third period of the second 10-year inservice inspection interval. During pre)aration for scheduling the completion of all 2nd interval examinations t1e licensee requested relief (Serial No. BSEP 97-0387. dated August 22, 1997), from performing the reactor pressure vessel (RPV) examination requirements of the ASME Code.Section XI 1980 Edition, with Winter 1981 Addenda, and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) for the Brunswick Electric Plant (BSEP) Unit Specifically aursuant to provisions of 10 CFR 50.55 a(a)(3)(i), and consistent wit 1 information contained in NRC Information Notice 97-6 Status Of NRC Staff's Review Of BWRVIP-05, relief was requested from the examination of the RPV circumferential shell welds for two operating cycles. The basis for this relief was documented in the re) ort. BWR Vessel and Internals Project. BWR Reactor Pressure Vessel (RPV) Shell Weld Inspection Recommendations (BWRVIP-05), that was transmitted to NRC in September 1995 and a risk-informed independent assessment performed by the NRC Staf . _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

i During the RPV examinations. the inspector held discussions with cognizant inservice inspection (ISI) personnel- to review the current vessel examination plans. General Electric Nuclear Energy's (GENE)

Ultrasonic Examinatior Procedure for the GERIS 2000 Ultrasonic OD Examination for RPV Assembly Welds (No. UT-BRU-702VO Revision 1), and also observed a GE ultrasonic (UT) data analyst tvaluate three vertical RPV welds, and reviewed examiners certification records for all GE examiners used during outage 13 of Brunswick Unit Observations and Findinas The inspector observed a GE Level III analyst evaluate UT data for the three vertical RPV welds listed below:

. Weld No. B11-RPV-E2A

. Weld No. B11-RPV-E3A a Weld No. B11-RPV-E2B The evaluation of the automated UT data was observed to determine whether calibrations were performed correctly and test sensitivity was adequate. whether setu) parameters were set to obtain proper inspection coverage, and whether JT reflectors were properly dispositioned and recorded. Review of the UT inspection procedure revealed that it was technically adequate and had been approved by the licensee and the Authorized Nuclear Inspector. Examiner certification records for all GE examiners were also reviewed and found to be satisfactor Conclusions Evaluation of GERIS autcmated UT data revealed that examinations conducted on the RPV vertical welds were performed in accordance with the approved examination procedure by skillful dnd knowledgeable UT test examiner M2.4 Observation of Data Evaluation Activities for Unit 2 Recirculation System Pioina Welds Examined Usina the Smart 2000 UT System Inspection Scone (73753)

The inspector reviewed two UT examination procedures and observed a GE Level III data analyst evaluate UT data for two recirculation system welds to determine if the examinations were performed and the data evaluated in accordanca with the approved examination procedure Observations and Findinas GE's Ultrasonic Procedure Nos. GE-UT-225 Version 2. Procedure for Automated Ultrasonic Examination of Austenitic and Ferritic Welds In Accordance with PDI, and UT-BRU-2250V Revision 0. Procedure for Automated Ultrasonic Examination of Dissimilar Metal Piping Welds, were

. _ _ _ _ _ _ _ _ _ - .

reviewed for technical content. weld applicability and proper approval In addition, evaluations of automated JT data for nozzle to safe-end weld No. 2BllN2CRPV-FWABA and sweep-o-let to pipe weld No. 2B32FF-12-FWRRB14A were observed to determine whether calibrations were performed correctly and test sensitivity was adequate. whether the scanning fixture was )roperly set and adequate inspection coverage was obtained, and whether J1 reflectors were properly dispositioned and recorde Conclusions Evaluation of Smart automated UT data revealed that examinations conducted on recirculation system welds were performed in accordance with approved examination procedures by skillful and knowledgeable UT test examiner M2.5 Observation of Data Evaluation Activities for H6B Reactor Core Shroud Weld Usina the Smart Ultrasonic System Insoection Scope (73753)

In 1996 UT examinations were performed on a total of 78.4% of accessible weld on H6B and a total of 69.6% of this weld was flawed. If flaw growth was subsecuently experienced, the licensee was prepared to re-examine this welc using UT phase array techniques. The inspector observed a GE UT data analyst evaluate and size crack indications on the H68 reactor core shroud weld. However, at the conclusion of this inspection (September 25, 1997), insufficient data had been evaluated on this weld to compare the examination results with previous UT examination results in order to determine whether additional crack growth had been experjence Observations and Findinas The inspector observed a GE UT data analyst evaluate and size crack indications on the H68 reactor core shroud weld. Discussions and observations were also conducted to determine how knowledgeable the .

analyst was of the inspection technology. In addition. GE's Ultrasonic Procedure No. GE-UT-524 Version 0. Procedure for Automated Ultrasonic Examination of the Shroud Assembly Welds Using Phased Array, was also reviewed for technical content and proper approval The inspector also observed Electrical Power Research Institute (EPRI)

engineers demonstrate the phase array techniques on a flawed specimen with discrepancies of known depth and lengths. The result of the demonstrated techniques indicated that sizing error was reduced and signal resolution was excellen Conclusions Evaluation of Smart automated UT data revealed that examinations conducted on the reactor core shroud welds were performed in accordance __ . _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _

._ _ _ _ _ __ _-___

with approved examination procedures by skillful and knowledgeable HT test examiners using state-of-the-art examination equipment.

M3 Maintenance Procedures and Documentation M3.1 Syrveillance Observation Inspection Scone (62707)

The inspector reviewed the aartial completion of a surveillance test for the scram discharge volume ligh water level function test during performance of the test on August 29, 199 Observations and Findinas The inspector toured the Unit 2 Reactor Building and observed the setup of test equipment for performance of Maintenance Surveillance Test 2MST-RPS270, RPS Scram Dish Vol Hi Wtr Lvl Chan Funct Test and Chan Ca Testing had been completed on the north side Scram Discharge Volum The inspector reviewed the partially completed procedure and noted several " pen and ink" changes to the procedur The inspector observed that a "]en and ink" change had been made because a vent and drain valve were num)ered the same in the procedure. Another

" pen and ink" change had been made to delete the step that checked the reset level. This change referenced Procedure Action Request 96-66 The inspector questioned how the " pen and ink" changes were made. The technicians stated that these changes had been discussed and approved by their superviso Also, the ins wctor observed that some steps were not initialed as complete in t1e procedure. The technicians stated that this was the field copy of the procedure and the master copy was in the control roo The inspector went to the controi room and reviewed the other co)y of the procedure. This copy had all the procedure steps completed aut did not contain the " pen and ink" change The inspector indicated to the shift superintendent (SS) that the procedure changes were not correctly made. The SS reviewed the procedure and concluded that the changes were made in accordance with Administrative Procedure OAP-010. Procedure Use and Adherence. This procedure allows changes to be approved by the supervisor. The performance of the maintenance surveillance was continu The inspector reviewed procedure OAP-010. Section 4.6 titled. Minor Editorial Enhancements and Field Corrections. Minor editorial enhancements were defined as clarifications a] proved by the supervisor that will not affect the intended result of tie procedure step in questicn. Field corrections were defined as changes a] proved by the supervisor to correct obvious typographical errors suc1 as misspellings or procedural sty misnumberings. This procedure guidance did not

. . . ' follow the guidance for procedure changes in TS 6. Temporary changes to procedure may be made provided:

_ (a) The intent of the original procedure is not altere (b) The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator License on the unit affecte (c) The change is documented, reviewed, and approved by the general manager within 14 days of implementatio This change to the procedure as a " pen and ink" change was atrary to TS 6.8.2 concerning temporary changes to procedures. Accordingly, this was a violation of TS 6. This will be identified as violation 50-325(324)/97-11-03. Pen and Ink Changes to I & C Procedur In addition, the adequacy of the plant procedure OAP-010. Procedure Use and Adherence, shoulu be addressed in response to this violation. This procedure under section 4.6. Minor Editorial Enhancements and Field Corrections, gives guidance that allows " pen and ink" changes. This violation was of particular concern because technicians felt the guidance was incorrect but were directed to make the changes by I & C supervision. After further discussion with plant management, the licensee issued CR 97-02893. Use of AP-10 Field Changes. to address this proble Conclusions The inspector concluded that " pen and ink" changes were made to a safety-related surveillance test procedure contrary to the TS requirements for )rocessing a temporary revision. This was identified as a violation. T11s occurred due to guidance in a plant procedure and direction given by I & C supervisor M7 Quality Assurance in Maintenance Activities M7.1 Chlorine System Leak _s Inspection Scone (62707)

The inspector reviewed the post-maintenance testing (PMT) on the chlorine system after leaks developed following maintenance activitie Observations and Findinas On July 24, 1997, a Control Building Emergency Air Filtration system actuation occurred when the chlorination system was placed in service following maintenance. The licensee made a four-hour notification of this event to the NRC. The licensee issued CR 97-02568. Chlorine Leak Number Two Chlorinator, to document this proble Operations was attempting to place the Chlorinator in service following repairs to correct a low vacuum condition when a chlorine leak develope The leakage occurred at a flange connectio . -

. _0n September 19.-1997. shile nlacing the chlorination system in service, an unexpected chlorine building alarm was re 'ved. The licensee determined that a one-fourth inch plug was missing on the aressure gas regulating valve on Chlorinator number four. Maintenance lad just been performed to replace this valve by work request / job order (WRJ0) 97-AETA The inspector entered the chlorine building on September 19. 1997. with a maintenance worker, maintenance supervisor, and engineering supervisor to review the leak locations and discuss PMT activities. Liquid chlorine enters the building from a chlorine tank car. Once in the building the chlorine goes into an evaporator where the chlorine changes-to a gas and passes through a pressure relief valved (PRV) before entering the control panel. After passing through several regulating valves in the control panel the chlorine gas enters the circulating water system or ser_vice water service by an ejecto The licensee *s review of this item determined that PMT activities are only required to be performed on the liquid side of the system. PMT requirements of Plant Procedure OPLP-20. Post-Maintenance Testing Program, only required a PMT for the liquid portion of the system. The licensee's practice was to only test the system on the liquid side up to the PR In NRC Inspection Report 50-325(324)/95-03 the inspector had discussed the licensee actions to correct problems with the chlorine system. The Chlorine Institute provided recommendations for testing for leaks using nitrogen and drying out the system to a minus 40 degrees dewpoint after maintenance to ensure all the moisture was out of the system. From discussion with the licensee it was learned that these recommendations were only partially implemented due to the difficulty of pressure testing with the PRV in the system. In both of these cases where leaks developed no PMT was performed. The inspector also observed the valve that was replaced that had the plug missing. The orientation of the valve in the system was such that the missing plug was easily seen during installation of the valve. The licensee attributed this problem to a bad part supplied by.the vendor. The inspector's review of this problem concluded that this was an easy item to catch and was more of a problem of attcation to detail and the lack of a questioning attitude when installing a valve with a visible hole in it in a poisonous chemical syste Conclusions The inspector concluded that several recent chlorine system leaks could have been-prevented by fully implementing PMT for al' parts of the chlorine system. Inadequate-attention to detail in assembly of components attributed to the leakage problems.. Additional management

-attention is needed in the inaintenance and testing of this syste . . ,.. . , , , - . - ,. .-.- -.-..-..-.-. -

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A E

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fM7.2? Effectiveness of Licensee Self Assessments- -

.

'

a InsDeCtion SCoDe (62707)  !

'

_

1The inspector reviewed three CP&L nuclear: assessments conducted at the a'

1 Brunswick Nuclear Plant in maintenance and inservice'

inspection / inservice testing, ?These audits were reviewed to' determine ,

the effectiveness.of.the assessments and the_ status of program areas E examined as viewed by the licensee's assessment team . Observations and Findinas

The1following CP&L-Nuclear Assessments were reviewed by the inspector:

,

e Brunswick. Maintenance Assessment B-MA-97-01 dated April 7,1997 4

.- - Brunswick Engineering Support Section Assessment B-ES-96-03 dated November-6. 1996 3

. Draft Brunswick Engineering Support Section Assessment B-ES-97-02

- The' inspector reviewed the above engineering support assessments only in

-the areas where they. addressed inservice inspection activities. The-

'

. review revealed that inspection attributes for each of the assessments-were very good for determining the effectiveness of the 3rogra Findings and subsequent corrective actions addressed in iaintenance Assessment'B-MA-97-01 which dealt with hot work practices, rad-worker ,

practices.and equipment storage and control findings should improve maintenance adherence in these area Engineering Support Assessment B- *

.

ES-96-03 identified that lessons learned and associated corrective -

actions from previous outages related to a vendor inservice inspection have not been sufficient to prevent recurrence of similar event Specific examples were given where condition reports were not consistently written to identify problems, events and areas of improvements in the vendor-inservice inspection related work. During

. the inspection, the licensee was in the process of finalizing the

- -corrective < action Engineering Support Assessment B-MA-97-01 did not have. any negative issues involving inservice inspection and listed engineering-initiatives-related to inservice testing as a strength,

.

lc; Conclusions ,

Nuclear assessments, performed by the licensee in maintenance and

-

i inservice ins)ection, which identified findings in hot work practices.

.

radiation worcer: practices, equipment storage-and control, and failure

.to issueLcondition. reports on: identified problems was considered a strength-.

- b

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I u

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i

M8 Miscellaneous Maintenance Issues (92902)

M (Ocen) IFI 50-325(324)/96-15-04: Material Condition of Remote Shutdown Panels The inspector reviewed corrective acH ^m taken by the licensee for the ren.ote shutdown panels and found that the actions taken to resolve electrical equipment inconsistences had been completed on Condition Report 96-2819. However, actions to complete panel cleanup and Jainting were scheduled to be com)leted on April 15. 1998. This work is )eing tracked by WR/J0s 97-ABC11 for Unit 1 and 97ABCS1 for Unit 2. This item will remain open until all work is complete on these panel III. Enaineerina El Conduct of Engineering El.1 Generic Letter (Gl> 89-10 Program Imolementation Insoection Scone (Temocrary Instruction 2515/109)

This ins)ection evaluated the licensee's implementation of GL 89-1 " Safety-Related Motor-0perated Valve Testing and Surveillance." GL 89-10 requested the implementation of a program to ensure that safety-related motor-operated valves (MOVs) were capable of performing their design-basis functions. The GL and seven supplements provided recommendations for this program which included establishment of MOV switch settings, design-basis testing to demonstrate the capabilities of the MOVs. and actions to assure that design-basis capabilities were maintained. The licensee notified the NRC that implementation of GL 89-10 was complete in letters dated June 21, 1995 (Unit 1) and April 1 (Unit 2).

The inspection was conducted through reviews of documentation and interviews with licensee personnel. The documents reviewed included:

o Procedure OMMM-032. " Generic Letter No. 89-10 Motor-Operated Valve Overview and Guidance Procedure." Revision 0 e Standard Procedure EGR-NGGC-0203. " Motor-Operated Valve Performance Predication. Actuator Settings, and Diagnostic Test Data Reconciliation". Revision 2 e Calculations BNP-MECH-MOV-VF BNP-MECH-MOV-ROL and BNP-MECH-MOV-SF (all Revision 0) establishing MOV valve factors, rate of loading, and stem factors e Engineering Evaluation Report 94-0145. " Evaluation of BNP Safety-Related MOVs for GL 89-10 Requirements." Revision 0 e Calculation BNP-MECH-SW-V19/20. " Mechanical Analysis and Calculations of 1-SW-V19/20 and 2-SW-V-19 Nuclear Header Pump Discharge Valves." Revision 0 e Self Assessment Report ESS-9' ! "hRC Generic Letter 89-10 MOV Program Assessment for brund .k Nuclear Plant." dated April 1997

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!

e Calculation BNP-MECH G31-F004, " Mechanical Analysis and Calculations of 1 & 2G31-F004 RWCU Inlet Outboard Isolation Valves," Revision 0 e Calculation BNP-MECH-B32-F031A/B. " Mechanical Analysis and C':1culations of 1 & 2-B32-F031A/B, Reactor Recirculation Pump Discharge Valves," Revision 0 e Work Request / Job Orders referred to in the following paragraphs e Summary tabulations of MOV information and calculation result including a list of MOV "a'ailable valve factors" calculated using formulas described in prev;aus NRC inspection reports (e. Inspection Report 50-338. 339/97-01)

e Setup calculations, dynamic test evaluation packages, and associated test reports for the following list of Unit 1 and 2 MOVs:

Valve Functional Name B32-F031A Reactor Recirculation Pump Discharge Valve E41-F002 High Pressure Coolant Injection (HPCI) Steam Supply Inboard Isolation Valve E41-F003 HPCI Steam Supply Outboard Isolation Valve E51-F007 Reactor Water Cleanup (RWCU) Steam Supply Inboard Isolation Valve E51-F008 RWCU Steam Supply Outboard Isolation Valve G31-F001 RWCU Inlet Inboard Isolation Valve G31-F004 RWCU Inlet Oetboard Isolation Valve SW-V19 Service Water Pump Discharge Valve Other valves were included in the 6ssessment, as described in subsequent paragraphs of this repor Observations and Findinos S_ cone of MOVs Included in the Proaram The scope of the licensee's GL 89-10 program currently included 88 gate valves. 40 globe valves, and 62 butterfly valves for a total of 190 MOVs. This scope was based on an evaluation documented in Engineering Evaluation Report (EER) 94-0145. In their review of EER 94-0145, the inspectors questioned the omission of so,ne MOVs from the Brunswick GL 89-10 program, such as certain containment isolation valves. A self-assessment performed by the licensee in April 1997 identified similar issues regarding the scope of the GL 89-10 program at Brunswick. The inspectors referred the licensee to information provided in NRC Temporary Instruction 2515/109 and an NRC safety evaluation prepared on the scope of the GL 89-10 program for the Hatch plant. In a letter

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dated October. 2,1997, the licensee committed to resolve the issue 01 GL 89-10 program scope by December 15, 1997, Sizino and Switch Settinos The licensee used the standard industry equations in calculating the predicted thrust and torcue requirements used in sizing and establishing settings for its gate anc globe valves, Vendor predictions of the torque requirements were used in sizing and setting butterfly valve Recently, the licensee had initiated a complete revision to the GL 89-10 program's calculations. This revision was reportedly undertaken tc incorporate improvements in MOV setup methodology identified from NRC inspection 50-261/96-12 (Robinson Nuclear Plant) and the licensee's internal assessment, and to incorporate changes to address Generic Letter 96-05 and a 1996 Brunswick load study. The revision activity resulted in the setup calculations for many valves being unavailable for this inspection. In its letter dated October 2,1997, the licensee committed to complete the revision of the setup calculations for the MOVs in the current program by December 15, 1997. Additionally, the licensee stated that the setup calculations for any valves added to the scope of the program would be completed by February 1,199 The inspectors' findings regarding the assumptions which the licensee had used in determining the sectings and sizing of MOVs are as follows:

Grouoino and Valve Factors for Gate Valve Thrust Calculations The licensee calculated the thrust requirements for gate valves that were not dynamically tested using valve factors determined through an analysis of in-plant and industry dynamic test reiilts documented in Calculation BNP-MECH-MOV-VF. The analysis divided gate valves into two groups and established a bounding valve factor for each group. The grouping was based solely on disk type and divided the valves into:

1) double-disk gate valves and 2) flex-wedge gate valves. The licensee's analysis gave no consideration to other design, application, or manufacture differences which might impact valve factor in establishing the grou)s. Supplement 6 to GL 89-10 indicated that consideration should )e given to size, manufacturer, performance during static testing, etc., in. grouping valves. The inspectors found the licensee's grouping and valve factor determinations insufficient or inadequate, as follows:

Flex-Wedge Gate Valves -

Calculation BNP-MECH-MOV-VF determined that a minimum valve factor of 0.70 was acceptable for use in calculating the thrust requirements of flex- wedge gate valves. The calculation justified this valve factor through an analysis of the licensee's in-plant dynamic test result The inspectors found that the licensee's dynamic test results supported this value for many of the valves but that there were no dynamic test results to justify ap)lication of the 0.70 value to certain sizes and to one manufacturer of t1e valves. The size / manufacturer combinations

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l 23 lacking supporting test results were as follows: 20-inch 600#

Anchor and 2-inch / Darling 600 # Velan valves..16-inch The licensee 150# Anchor had not / Darlin dynamically 6-inch 150# Anchor / D tested valves from these combinations (or closely similar combinations) and had not obtained industry information to su) port the 0.70 valve factor for these valves. In its letter dated Octo)er 2. 1997, the licensee stated that it had initiated a survey of industry information to verify the adequacy of its 0.70 valve factor assum) tion for non-dynamically tested flex-wedge gate valves at Brunswick. T1e licensee committed to complete this effort by the December 15. 199 Double-Disk Gate Valves -

Several safety systems at Brunswick used Anchor / Darling double-disk gate valves in safety-related ap)lications. These applications included containment isolation for t1e High Pressure Coolant Injection. Reactor Core Isolation Cooling, and Reactor Water Cleanup systems under potential blowdown flow conditions. The licensee set these MOVs to operate with limit switch control. such that electric power to the motor would be removed after flow isolation but prior to full wedging of the disks. The licensee verified sufficient wedging of the two valve disks against the seat by local leak rate testing. The licensee had not dynamically tested any Anchor / Darling double-disk gate valves at Brunswick but had assumed a 0.40 valve factor in calculating the thrust required to achieve the closure position based on: 1) an Anchor / Darling steam blowdown test (Report CTS-27) performed in 1991. 2) double-disk gate valve friction coefficient data from the Electric Power Research Institute (EPRI). and 3) a reference to the use of a 0.40 valve factor for double-disk gate valves in the MOV program at the Callaway nuclear plant. For the following reasons. the inspectors determined that these sources did not provide adequate justification for the assumed valve factor in verifying the design-basis capability of Anchor / Darling double-disk gate valves at Brunswic First, an Anchor / Darling representative indicated in a caper included in NUREG/CP-0137 (July 1994). " Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing." that the Anchor / Darling double-disk valve testing focused on the stroke position at convergence of valve body and upstream pressures (referred to in the Anchor / Darling paper as the point of flow isolation) when calculating a valve factor from the test data. He also noted that the test results revealed an increasing trend for the valve factor with subsequent test strokes. The licensee had not evaluated its use of a stroke position that provides sufficient wedging to meet local leak rate testing requirements to determine whether more thrust might be required than Anchor / Darling's selection of a flow isolation Joint. The licensee also had not evaluated whether the Anclor/ Darling test results would be applicable to the Anchor / Darling double-disk gate valves and their various service conditions at Brunswic _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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e Second. the licensee referenced s?ecific language regarding flow test results in EPRI Proprietary Report TR-103237 "EPRI MOV Performance Prediction Program - Topical Report-Rcvision 1." EPRI conducted separate effects testing of valve seat materials and flow testing of an Anchor / Darling double-disk gate valve in developing a hand-calculation model for predicting the thrust required to operate these valves. However, the use of individual data )oints from EPRI testing might not be reliable in predicting the tarust required to operate an MOV. In the supplement (dated February 20, 1997) to the NRC Safety Evaluation on the EPRI Topical Report TR-103237 the NRC staff noted that it had reviewed the EPRI model as a complete package and cautioned licensees regarding the use of selective test data or methods from the EPRI progra Further. EPRI discussed other available test information for Anchor / Darling double-disk gate valves in its non-proprietary response to NRC staff questions in a letter dated April 5.1996, from John F. Hosler. EPRI. to Richard H. Wessman. NR In that discussion. EPRI indicated that its model might predict greater thrust requirements than seen in some individual test data from various sources including Anchor / Darling. However. EPRI maintained its position that the model was appropriate. Furthe guidance prepared by Commonwealth Edison Company for sizing its Anchor /Daring double-disk gate valves in its White Pa]er 164 (Revision 1. dated October 9. 1995). " Anchor / Darling Jouble-Disk Gate Valve Factors." suggests that the valve factors for various service conditions might be greater than the 0.40 value assumed at Brunswick. The licensee did not evaluate such applicable information in determining the adequacy of its use of test data from a single Anchor / Darling double-disk gate valve tested by EPR e Third, the Anchor / Darling double-disk gate valves in the GL 89-10 program at Callaway (about 5 valves) were each dynamically tested to verify their thrust requirements. The Callaway licansee could l not determine a reliable valve factor from the test date because of the type of diagnostic equipment used in the tests. In addition to dynamically testing each of its Anchor / Darling double-disk gate valves. the Callaway licensee had established a plan to dynamically test each MOV where a thrust margin of at least 25%

could not be maintained as part of its long-term MOV program. See NRC letter dated June 8. 1994, to Union Electric Company closing NRC re"iew of GL 89-10 at Callaway. The Brunswick licensee did not consider these additional attributes of the Callaway MOV program when relying on a valve factor used at that facilit Based on the above. the inspectors found that the licensee had not taken sufficient action to provide an adequate basis for its valve factor assumption for Anchor / Darling double-disk gate valves in light of its corporate and industry-wide information. For example, the licensee's April 1997 Self Assessment rReport ESS-97-04 identified that even a more conservative 0.50 valve factor might not be justifiable and the licensee's " Improvement Plan & Schedule for the Brunswick Motor-Operated l

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- Valve' Program." dated-May 5,1997, proposed dynamically testing two .of - ,

Brunswick-s Anchor / Darling double-disk gate valves in the" future to_ .

address the'issuc. . In addition to the -industry-wide information *

discussed above..the NRC alerted licensees to the higher than expected'

valve factors in Information Notice'97-07 (March-6. 1997). The licensee's failure to promptly resolve the inadequately justified valve factor assumption used in the MOV setting calculations for its double-disk gate valves was considered to represent inadequate corrective

- action for industry concerns regarding valve settings and capability requirements and is identified as the first example of violation 50-325(324)/97-11-04. Inadequate Corrective Actions for MOVs. In its letter dated October 2,1997, the licensee committed to 3rovide short-

- term justification for its valve factor assumption for t1ese MOVs by s December 15, 1997. using information to be obtained from an industry

. survey. In' addition, the licensee indicated that long-term justification would be developed based on information from Joint Owner *s

- Group testing and from testing that Brunswick will perfora after installation of SmartStems on two valves. The long-term actions are to be completed by September 1. 199 Valve Factors for Globe Valve Thrust Calculations

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Calculation BNP-MECH MOV-VF determined that a minimum valve factor of 1 10 was acceptable for- use in calculating the thrust requirements of =

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globe valves that were not dynamically tested. The calculation justified this valve factor through an analysis of the licensee's in-plant globe valve dynamic test results. As for the flex-wedge gate valves discussed above. the inspectors found that the analysis was

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irsufficient because it included no test results supporting the applicability of the 1.10 valve factor to some sizes and to one manufacturer of the valves. The valves lacking supporting test data were the 2-inch 600# Velan. 10-inch 900# Valtek. 3-inch 600#

Anchor / Darling. 3-inch 900# Schutte & Koerting, and 0.5-inch 600# Velan globe valves. In its letter dated October 2, 1997, the licensee stated that it had initiated a survey of industry information to verify the adequacy of its valve factor assumption for non-dynamically tested globe valves at Brunswick. The licensee committed to complete this effort to verify the _1;10 valve factor assumption by December 15.199 Seat Area-for Globe Valve Thrust Calculations The inspectors questioned whether the licensee had used appropriate area teans for seat or guide based globe valves. In its letter dated

- October-2. 1997, the licensee stated that the procedural requirements would:be reviewed against EPRI guidance and revised, as appropriate by )

December 15. 1997. Additionally, the licensee stated that any changes

would be1 included in setup calculation revisions now in-proces Toraue Reauirements Calculated for Butterfly Valves The Brunswick.GL 89-10 butterfly valves were manufactured by Fisher and

, Jamesbury and included sizes ranging from 4- to 30-inches (all class

?

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. #). The licensee did not obtain any test data to support the accuracy of the torque predictions for design basis operation of its butterfly valves. The licensee's setup calculations (e.g. . BNP-MECH-SW.

Vl9/20) relied on the predictions of torque requirements provided by vendors in establishing GL 89-10 butterfly valve settings and s1zes for design basis operation. GL 89-10 was initiated because vendor predictions of setting and sizing requirements had been found unreliable, further. Supplements 1 and 6 of GL 8910 and Information Notice 97 07 indicated concern as to the accuracy of sendor torque predictions for butterfly valves. The licensee's failure to resolve inadequately justified torcue predictions used in MOV calculations for butterfly valves was consicered to represent inadequate corrective action for industry concerns regarding valve settings and capability requirements and is identified as the second example of violation 50-325(324)D7-ll 04. Inadequate Corrective Actions for MOVs. In its letter dated October 2. 1997, the licensee committed to obtain data for its butterfly valves from an industry survey and to incorporate appropriate chane.s into calculations by December 15. 1997. Where no l indastry data can be obtained, the licentae 3roposad to evaluate the ,

EPRI model and data to determine its applica)ility to Brunswick !

butterfly valves. Further, the letter stated that additional data to assess its butterfly valve assumptions would be obtained from static testing of similar butterfly valves located in the licensee's test facility and from torque tests to be initiated on its butterfly valves in 1999 using Teledyne smart couplings.

Rate of loadina The licensee statistically analyzed the rate of loading values obtained on dynamically tested valves to establish values that could be assumed

determining thrurt requirements for all non-dynamically tested gate and globe valves. This analysis, documented in Calculation BNP MECH-MOV-ROL. determined a mean value of 0% and a two-standard deviation value of 17.3%. These were used as bias and random variables for use in calculating thrus The inspectors revicwed the licensee's data and found that they supported these values. However, the inspectors noted the following weaknesses in the licensee's Justification for assunling these values could be applfed to all of its non dynamically tested gate and globe valves

o The rate of loading andlysis contained a limited amount of glcbe valve data and relied primarily on gate valve data. Industry testing has shown that globe valves can exhibit greater rate of loading than gate valves, o The licensee had a number of MOVs with ball-screw stem nuts and assumed that MOVs with this mechanism would not exhibit rate of loading. However, the licensee did not have any data to verify this assumptio .

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e The licensee's rate of loading data was entirely from MOVs with standard actre-thread stems. However, some of the licensee's MOVs had stub-acme threaded stems. The licensee indicated that discussions with another utility had not revealed any differences between performance of stub acme thread and standard acme-thread stem nuts regarding rate of loading.

In its letter dated October 2.1997, the licensee committed to implement the following actions to address the weaknesses:

e The licensee indicated that an evaluation of globe valve test cnd operating information to determine appropriate rate of loading would be completed by December 15, 1997 as part of ongoing efforts to revise and improve MOV setup calculations.

e An industry survey would be conducted to establish rate of loading for valves with ball-screw stem nuts by December 15. 1997. In the long-term, an improved basis for the rate of loading value would be obtained by installing Teleo ne SmartStems on valves with ball-screw stem nuts and measuring the rate of loading. Data from the first of these tests would be evaluated by September 1, 1998.

e Conduct an industry survey to establish the rate of loading for valves with stub acme threads and incorporate the results into calculations by December 15, 1997. Also, evaluate the a]plicability of a]propriate EPRI data to valves with stub acme t1readed stems by Jecember 15. 1997.

Stem Friction Coef ficient The iicensee generally assumed that a stem friction coefficient of 0.20 would be bounding for gate and globe MOVs in the GL 89-10 program at Brunswick- This value was justified through a statistical analysis documented in Calculation BNP-MECH-MOV SF. The inspectors identified the following weaknesses in the licensee's stem friction coef ficient analysis and in the licensee's application of the bounding value:

e The licensee used an assumed stem frlction coefficient of 0.15 in thrust calculations prepared for three valves with marginal capabilit e i The measured stem friction coefficients of these valves wa. less than 0.15, but it was not clear how the licensee would assure that a stem friction coefficient of 0.15 or less would be maintained long-term.

e The licensee did not have any test data to j'.stify the stem friction coef ficient assumed for MOVs with ball-screw stem nut Rather than the bounding 0.20 stem friction coefficient the licensee used vendor-provided stem factor information for these stem nut The licensee had initiated Condition Report 97-000586 to resolve this issu o The licensee assumed that a stem friction coef ficient of 0.20 was applicable to MOVs with stub-acme threads. Howevt . the licensee did not have data obtained from testing of actuators with stub-acme threads to support this assumption, in its letter dated October 2. 1997, the licensee committed to implement the following actions to address the weaknesses:

e The licensee stated that the calculations for the three valves would be revised to use a stem friction coefficient of 0.20. The letter also indicated that topical calculation (PNP-MECH-MOV-SF)

would be revised to address mini-groups of valves with lower friction coefficients. The letter stated that these actions would be completed by December 15. 1997. The inspectors found the letter discussion somewhat unclear in regard to use of mini-groups and stem friction coef ficients Delow 0.20 and contacted the licensee by phone on October 3. 199 In that call. the inspectors informed the licensee of concern that the values used for mini-groups be justified based on sufficient tests (including periodic verification for the mini groups) and design or thread condition difference e An industry survey would be conducted to establish stem friction coefficients for valves with ball-screw stem nuts by December 1 . In the long-term, an improved basis for the friction coefficient values would be obtained by installing Teledyne Smarts +. ems on valves with ball-screw stem nuts and measuring the friction coef ficients. Data from the first of these tests would be evaluated by September 1. 199 e Conduct an industry survey to establish stem friction coefficients for valves with stub acme threads and incorporate the results into calculations by December 15.199'/. Also, evaluate the a]plicability of a)propriate EPRI data to valvet, with stub acme tireaded stems by )ecember 15. 199 Extranolatie' of Test Data The inspectors noted that Section 9.6.2 of Standard Procedurn EGR-NGGC-0203 provided recommendations for evaluation of test data when dynamic test conditions are less than 80% of design-basis differential pressur However, the recommendations did not address the minimum disk loads that are necessary to ensure that test the results are reliabl For example. EPRI has provided guidance for the extrapolation of test data (response to NRC staff questions in a letter dated April 5. 1996. from John F. Hosler. EPRI. to Richard H. Wessman. NRC). In its letter dated October 2. 1997, the licensee committed to resolve this issue by December 15. 1997.

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The licensee's GL 89-10 program included a large number of actuators l with de motors. The inspectors noted that the licensee assumed an application factor of 1.0 in its output torque calculations for selected .

dc MOVs that had marginal output capability (e.g., Calculation BNP. MECH- '

E51-F007). The licensee had not verified that this value could be applied to dc powered MOVs. although it appeared a logical extension of ,

criteria used for ac valves. In its letter dated October 2. 1997. the licensee indicated it would resolve this issue and prepare anj necessary i calculation and procedure changes by December 15. 199 . Desian Basis Canability ,

As the licensee wa', u) dating many of its MOV calculations. the inspectors were noc a)le to review the final status of the capabilities determined for the licensee's MOVs. At the inspectors' request, the licensce calculated the available capability of the MOVs which were considered to have the least amount of margin. The inspectors reviewed  :

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this information and did not identify any immediate operability concern However, they noted several instances where MOVs appeared -

marginal in their aerformance capability. In its letter dated October 2. 1997, tle licensee committed to prepare a matrix to clearly document the design basis capability of its current GL 59-10 MOVs by January 15. 199 Valves added to the scope of the licensee's program would be incorporated into this matrix by rebruary 1, 1998.

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in their review of available licensee calculations and data. the -

inspectors identified the following additional concerns regarding MOV capabilities:

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D oabilities of Coti Soray System Bynass Valves As documented in Condition Re) ort 97-02546. the licensee had identified that Core Spray Mini'num Flow 3ypass Valves 8 E21-F031A and B had torque switches that were set above the degraded voltage output capability of their motors. If electrically closed under degraded voltage conditions and maximum dif ferential 3ressure for system flow requirements, breaker tri) or motor failure mig 1t occur. Three of the valves were determined to 3e capable of achieving sufficient closure for a subsequent '

containment isolation requirement prior to the motor stall or breaker tri The fourth. 2 E21 F031B had insufficient capability to ensure it would fully close before the trip or stall. As documented in Licensee Event Report 97-002-00. this valve was declared inoperable and Technical Specification requirements were implemented. The licensee subsequently 3

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lowered the torque switch setting for the MOV and declared it o>erabl '

The new setting provided the capability to achieve sufficient ()ut not complete) closure for system flow requirements without causing the MOV to-trip or stall, and sufficient capability to subsecuently close for a ,

containment isolation requiremcnt (during which the cifferential pressure would be lower). In its letter dated October 2. 1997, the licensee stated that an engineering evaluation had determined that the

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current switch settings were acceptable for the short-term. The '

licensee cnmmitted to determine if long-term modification is required by '

July 31. Ik ^.

Doenina Thrust Measurement Uncertainty

The licensee used Liberty Technclogies* VOTES diagnostic test equipment to measure MOV operating thrust and establish thrust settings. The <

inspectors found that the licensee had not made adjustments to pen thrust measurements to account for the measurement uncertainty identified by the licensee's V0TES diagnostic equipment verdor. Liberty Technologies, in Customer Service Bulletin 31 (issued November 1 ). The inspectors were informed that a review to determine the i impact of this measurement uncertainty on Brunswick thrust measurements had been contracted to Liberty Technologies. The licensea received and ,

performed a preliminary evaluation of a draft re> ort of Liberty Technologies findings during this inspection. _icensee personnel indicated that they identified no operability concerns. They also noted that the report only addressed measurements from differential pressure testing: static test measurements were still under revie The licensee's failure to adjust its thrust measurements for the above uncertainties was cited by the NRC at the licensee's Robinson site in October 1996 (Inspection Report 50-261/96-12) with Brunswick and other licensee MOV personnel present. Further, the licensee identified this finding for Brunswick in its April 1997 Self Assessment. ESS-97-04. The Self Assessment stated that, like Robinson. Brunswick "did not correctly ap)1y the VOTES system open thrust measurement error as sp?cified in Li)erty Technologies CSB-31." The license's failure to resolve the uncertainty surrounding its open thrust measurements was considered to represent a failure to take prompt corrective action ano 15 identified as the third example of violation 50-325(324)/97-11-04. Inadequate Corrective Actions for MOVs. In its letter dated October 2. 1997, the licensee stated that it had aerformed a preliminary review and had not identified any immediate pro)lems resulting from the open calibration issue. In its letter, the licensee committed to incorporate the results into the MOV calculations by December 15. 199 MOV Stroke Time In its Self Assessment Report ESS-97-04, the licensee noted that its current analysis method for determinina de MOV stroke time did not account for motor load profile that would be aresent at design-basis conditions. This analysis is performed to ensure that Technical Specification and Safety Analysis response times are met. The inspectors reviewed current calculations and identified several MOV stroke times that were close to the respective Technical S)ecification limits. For example, the estimated stroke time for 1-G31 2004 was 3 seconds as com)ared to the Technical Specification time limit of 35 seconds. If tle licensee's method is nonconservative. the Technical Specification limit may be exceeded. The licensee's self assessment *

- expressed concern regarding the stroke time determinations and l

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recommended that a methodology deveioped by the Boiling Water Reactor Owners Group be a) plied. In its letter dated October 2. 1997 the licensee stated tlat it would provide short-term resolution of this issue by January 15. 1998. The inspectors understoort this to mean that the a>propriate method of determining stoke time would be selected and opera)1lity re-verified. The licensee statad that any associated procedure / calculation revisions would be completed by September 1. 199 ,

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Differential Pressures for Reactor Recirculation Pumo Discharae Valves The licensee's GL 89-10 program documentation specified that the maximum design basis differential pressure that the Reactor Recirculation Pump Discharge Valves 1/2832-F031A and B valves will experience during closing is 65 psid. The inspectors questioned this value, as it was lower than typically assumed for the closing of these valves during a Recirculation System line break scenario. The licensee's analysis determined that these MOVs receive an automatic signal to close when reactor pressure decreases to 325 psig. The licensee's analysis then assumed that these valves would have a closing stroke time of ,

approximately 27 seconds. Using reactor system and containment pressure profiles for a rupture in the Recirculation line, the licensee estimated that 65 psid would exist at valve closure. Given that a gate valve will experience significant disk loading under blowdown flow conditions well before valve closure, a higher differential pressure would be ex)ected for design-basis conditions. Further. the pressure profiles mig 1t be conservative for the rate of depressurization of the reactor coolant system, bd might be nonconservative for valve operating requirement In its letter dated October 2. 1997, the licensee stated it wou1J '

further evaluate the design-basis accident conditions for these MOVs and determine their impact on the differential 3ressure conditions previously established, This effort is to )e completed by December 15, 199 The NRC will continue to review this issue as part of the evaluation of the Brunswick GL 89-10 progra Toraue Switch ReDeatabi1ity During review of Section 9.6.2 of Standard Procedure EGR-NGGC-0203, the inspectors noted that one of the licensee's equations used to calculate closing margin did not include consideration of torque switch repeatability. In its letter dated October E.1997, the licensee notd thtt equation which had been questioned was used for information only but committed to 3rovide appropriate changes to the procedural requirements by t1e December 15. 199 Actuators Doerated Above the Oriainal Thrust Ratinas The inspectors noted that the licensee routinely applied extended thrust ratings to some of Brunswick's actuators, based on a Kalsi Engineering actuator test program. However, the MOV maintenance procedures did not s)ecify any additional inspections for MOVs that used the extended tarust ratings. The NRC staff discussed extended actuator ratings in Supplement 6 to GL 89 10. In response to the inspector questions. the

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licensee initiated a change to procedure OPM-M0504 to provide increased attention in inspection of actuators to help ensure that the use of extended ratings does not result in actuator damage. This was

_ accomplfshed through Action item 97-02879-1. The issue was resolve . Periodic Verification of MOV Canabilities The laspectors verifled that the licensee's current periodic verification requirements were implemented in the Brunswick Preventive Maintenance database. A current printout of the database indicated that

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the valves would receive diagnostic testing in accordance with procedure OPDM M0005 The database showed the date of the list test for each valve and the due date for the next test. The periods between tests were valve specific and varied between four and about six years (up to three refueling outages). The licensee indicated it was a member of the inNstry Joint Owners Group and that its testing would incorporate dynamic testing for t.m Owners Group plus others considered appropriat *

The NRC may further assess the licensee's long term periodic verification program as part of its review for GL 96-05. Periodic Verification of Design Basis Capability of Safety-Related Motor-Operated Valves, dated September 18, 199 . Post Maintenance and Post Modification Testina

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The inspectors reviewed the licensee's post maintenance test guidanc which was )rovided in procedure OPLP-20. " Post Maintenance Testing Program." Revision 16, The licensee's post modification test requirements were determined using post maintenance testing requirer =nts as guidanc Diagnostic testing was specified for acti"ities that could affect thrust. For packing adjustment, use of the packing " Enforcer" was permitted as an alternative to static diagnostic testing. Dynamic diagnostic testing was to be considered for work on seating surface To assess the licensee's implementation of the test requirements. the inspectors reviewed the post modification and post maintenance testing recorded for Work Request / Job Orders (WR/J0s) 96-AFIP2. 97-ACN01. 96- ,

AHZil 97-AATC1. 95-AGQE2. 96-AHDA1. 96-ABDN1, 96-ADNA1. 97-AATC1. 96-AAJBl. and 95-AJYHl. These WR/J0s involved packing replacement and adjustment, limit switch adjustmer' spring pack inspection and replacement, a breaker trip, mo+ nspection and replacement, replacement of a broken valve /i dtor bolt, and valve replacemen Generally, the inspectors found inat the licensee performed appropriate testing. However, two examples were questioned:

Eq, Thrust Measurement After Packina Adiust' ment WR/JO 96-AHZIl documented that the packing gland nuts on valve 1-G31-F004 (RWCU Inlet Inboard Isolation Valve) were retorqued following a leak. No post maintenance thrust measurement was documented. The ins)ectors questioned the lack of a thrust measurement after this ,

paccing adjustment. as the licensee had identified this valve as having marginal thrust capabilities, Licensee personnel stated that it had not

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been practical to measure thrast on this valve. as the leak occurred during operation when testing of the valve was impractical. They further stated that the retorquing of the nuts had not exceeded the original torque value and this was not expected to increase the packing thrust load. In response to the inspectors' concerns, the licensee contacted the manufacturer of the packing who indicated that the packing load would not be expected to increase when retorquing te the original value. Additionally, the licensee initiated WR/JO 97-AFEG1 to verify the packing load during the Spring 1998 refueling outage. In its letter dated October 2.1997. the licensee committed to evaluate its packing adjustment procedural requirements and work history to aid assuring there are appropriate controls for packing adjustment load No Structural Evaluation After Motor Failure WR/JO 96-ABDN1 described two breaker trips that occurred February 3, 1996, while attempting to open MOV 2 E11-F009 (Shutdown Cooling Inboard Suction isolation Valve). The licensee's subsequent investigation was documented in Condition Report (CR) 96-00341, which found that the valve had failed to open due to a shorted motor. The investigation concluded that the motor shorting was due to magnesium rotor degradation and hydraulic locking of the spring pack due to excess grease present, Based on an overload condition documented in 1994 (WR/JO 94-AJFJ1), the CR concluded that the failure most likely occurred at that time (just 3rior to plant start-up in 1994) and remained undetected until the 1996 areaker trip ;

The inspectors noted that the CR had not addressed the possibility of structural damage to MOV components that were not replaced following the failur In response, the licensee evaluated the component loads caused by the failure, assuming that the motor stalled due to the hydraulic locking due to grease in the spring pack. The evaluation found that the structural capability of the weakest link, the yoke clan.p. would have been exceeded at stall. The licensee postulated that any damage to the clamp would have been detected in routine MOV inspections subsequently performed. The inspectors agreed that severe deformation or cracks would likely have been seen. However, as no specific attention had been directed to the clamp in the licensee's inspection, they were concerned that less obvious damage might not have been detected. In its letter dated October 2. 1997 the licensee committed to inspect the yoke clamp -

for damage in the Fall 1997 refueling outage. The licensee's initial failure to determine whether any MOV structural weak links had been exceeGed when the valve failed to open was considered indicative of inadequate corrective action measures. This was identified as the fourth example of violation 50-325(324)/97-11-04. Inadecuate Corrective Actions for MOVs. Other examples of this violation is cescribed in E above and in E8.2 belo .

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34 Pressure Lockina/ Thermal Bindina l Further review of the adequacy of the licensee's actions to address  ;

pressure locking will be addressed through corres>ondence with the licensee regarding its submittal in response to G. 95-07. " Pressure tocking and Thermal Binding of Safety-Related Power-0perated Gate Valves." Self Assessment

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The inspectors found that the licensee's Self Assessment ESS 97-04 of the Brunswick GL 89-10 program was thorough and resulted in important findings. However, the effort could have been more timely. Although .

the licensee considered implementation of GL 8910 complete for both Brunswick units in April 1996 and problems had been identified with the implementation at its Robinson plant in October 1996 (documented in NRC i inspection Report 50-261/96-12), the assessment was not performed until April 1997. The Robinson findings were sufficiently serious to result in two violations and an NRC conclusion that GL 89-10 had not been adequately implemented for Robirso As reported in previous paragraphs, several of the inspectors' findings in the current inspection were identified by the licensee (e.g.. assumed valve factor for Anchor / Darling douole-disk gate valves not justifiable, open thrust measurements not in accordance with Customer Service Bulletin, etc.), These findings were alse similar or identical to findings identified by the NRC for the licensee's Robinson piant (NRC Inspection Report 50-261/96-12).

On May 5. 1997. the licensee issued its "1997 Imarovement Plan &

Schedule for the Brunswick Motor-Operated Valve 3rogram" as a result of its self assessment of the Brunswick MOV program and the NRC inspection at Robinson. On July 2.1997, the licensee issued Revision 1 of the

"CP&L GL 89-10 Corporate Improvement Plan" to help direct resolution of the identified issue As described above, the inspectors found that the licensee had not resolved issues regarding MOV performance that were similar to issues identified earlier for Robinso In its letter dated October 2.1997, the licensee indicated that the improvement alan would be revised to include corrective actions to address each of t1e self assessment findings and that the identified corrective actions would be tracked via the Brunswick Corrective Action Program. The date given for

. completing the improvement plan and instituting the tracking was December 15. 1997. The licensee also committed to provide more specific indication of its review and consideration of the MOV issues discussed in NRC Information Notices 96 48. " Motor-0perated Valve Performance issues." and 97-07. '"oblems Identified During Generic Letter 89-10 Closeout inspections by January 15. 199 .

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c. Conclusions The licensee had not satisfactorily implemented GL 89-10. Assumptions used in setting and capability calculations had not been adequately justified for some valves and uncertainties in valve o)ening thrust measurements reported to the licensee in 1993 had not aeen resolve Various other issues described in the preceding paragraphs also require resolution including issues identified in a licensee self assessmen Following the on site inspection, the licensee responded to the issues with planned actions as described in a letter dated October 2.199 .

The actions were considered satisfactory, subject to further review in a subsequent NRC inspectio .

The licensee's failures to respond to corporate and industry-wide MOV performance information to promptly resolve inadequately justified design assumptions and uncertainties in valve opening thrust measurements were considered indicative of inadequate corrective actions and are identified as examples of violation 50 325(324)/97-11-04 Inadequate Corrective Actions for MOVs. Important examples of the inadequately justified assumptions incluced the valve factors used in calculating thrust settings for Anchor / Darling double-disk gate valves and the torque requirements used to establish capabilities for butterfly valve The licensee's corrective actions for two MOV failures were inadequate, as the licensee did not determine the forces generated during the failures and did not determine whether structural (weak link) limits were exceede This was considered indicative of inadequate measures to assure prompt identification and correction of conditions adverse to quality and is identified as a further example of violation 50-325(324)/97 11-04. Inadequate Corrective Actions for MOVs.

E3 Engineering Procedures and Documentation E3.1 Use of Technical Support Memorandums insoection Scone (37551)

The inspector reviewed the use of Technical Support Memorandums (TSMs)

by plant personne Observations and Findinas In NRC IR 50-325(324)/97-09. section El.1 concerning E0. it was determined that no procedure existed for control of temperatures in the reactor building. The inspector noted that Operating Instruction 201-03.4.2. Auxiliary Operatur Daily Check Sheet, referenced TSM 89-39 Tha inspector inquired as to what was a TSM and asked to see TSM 89-39 The operations shift superintendent (SS) could not locate a copy of the TSM in the control roo _ +,.g.. p ..y-,-4 ~

-mm , ymr

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follow-up discussions with Document Control revealed that because of adverse condition report ACR 92-306. the use of TSMs was stopped. All TSMs were in a nonactive status. The licensee issued CR 97-0335 Procedures Referencing TSMs. to document this problem. The CR stated that numerous site procedures reference TSMs. The TSM process was deleted in Novenber 1992. The TSMs were utilized to transmit design information obtained from approved design documents or source The licensee's review of site procedure will look for other inappropriate uses of TSM Site procedures should not reference TSMs and the licensee was considering revising all procedures that reference TSMs to reference the correct applicable design documents. This issue will be tracked as URI 50-325(324)/97-11-05. Use of Technical Support Memorandum Conclusions

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The inspector concluded that TSMs were being referenced although TSM's were in a nonactivt statu The full impact of this requires further revie E4 Engineering Staff Knowledge and Performance E Imoroner Rioqino to Safety Related Pinina inspection Scone (37551)

During routine inspection activities. the inspector noted rigging to safety related piping. The inspector reviewed the associated WR/J ESRs. related plant procedures, and other evidence to determine whether an appropriate safety evaluation was conducted for the rigging activit Observations and Findinas l

l On August 18. 1997, the inspector was performing routine inspection activities in the Unit 2 Reactor Building. Connected to the Reactor Building Purge Vent discharge and the Reactor Building Service Water (SW) discharge piaing the inspector observed lines connected a two ton pulley supported ]y an installed rigging point. Subsequent questioning of licensee personnel determined that the rigging was being used for the upcoming torus strainer replacement. Upon review of the rigging planned, no instructions to rig to safety-related piping could be identi fied. This deficiency was recorded in CR 97-2787. Placement of Temporary Load Control of plant rigging was contained in Maintenance Management Manual OMMM-022. Instructions for Placement of Temporary loads (e.g., Riggin Scaffolding. Ladders. Personnel). Section 5 of OMMM-022 required that a temporary rigging release be obtained prior to rigging to or above any safety-related component that would be required to perform its safety-related function regardless of the magnitude of the loa . . --.

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l The inspector determined that the licensee had built a mock-up of the 3400 pound torus strainer. The 300 pound mock-up was usei August , to test the evaluated heavy load path, rigging plan, and whether there was enough clearance for the suction strainer to pass through the torus equipment Fatc During the movement of the mock-up from the 20 ,

foot elevation to the nine foot torus equipment hatch elevatio ;

initially the rigging used was configured in accordance with the rigging plan. DJring the lifting, it was determined that additional rigging around the safety-related piping would be necessary. No temporary '

rigging release was obtained. Upon the discovery, the inspector was concerned that the rigging around the safety-related piping would have r

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been used to rig the actual 3400 pound suction strainer without having '

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a properly evaluated the heavy load's stress on the safety related piain The failure to obtain a temporary rigging release as required by OMiM-022 is the first example of violation VIO 50-324/97-11 06. Rigging to -

Safety Equipment Without a Safety Evaluation. Upon notification of the i nonconformance, the rigging was removed from the safety-related piping, t a-rigging release was completed, and a standdown conducte In addition the licensee reviewed existing plant rigging for any other '

unreviewed temporary loads over safety ecuipment. On September 8. 1997 Unit 2 was at 95 percent power coasting cown to the planned outage on September 13. 1997. The licensee determined that air piping suspended ',

over the HPCI MCC 2XDA was huna without a temporary rigging release and was not seismically mounted. $ubsequent engineering evaluations .

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determined that during a seismic event the possibility existed t1at the piping could damage the MCC. The piping was subsequently removed. The licensee issued CR 97-2962 for this 3roblem. The licensee made a four-

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hour report in accordance with 10 CFl 50.72 on September 8, 199 The inspector reviewed the CR, WR/J0, associated evaluations and procedures, and discussed the occurrence with the licensee. There was no temporary rigging release initiated 3rior to installation of th?

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temporary air 31 ping. The failure to o)tain a rigging release in I accordance witl OMMM-022 is the second example of violation VIO 50-324/97-11-06. Rigging to Safety Equipment Without a Safety Evaluatic ,

c. Conclusions The inspector identified improper rigging of the torus suction strainer

_ mock up to safety-related piping. Subsequent review by the licensee af other areas revealed that non seismically mounted temporary piping was installed over a safety-related MCC without a temporary rigging releas The improper rigging of the strainer mock up and suspending of air piping without a temporary rigging release was identified as a

, violatio ,

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E8 Hiscellaneous Engineering Issues (92903)

E (Closed) Lnspector Followun item 50 325(324)/94-20-01: Electrical lorque Ca'culations 1or DC MOV Motor t This item identified that the licensee was not using the dc motor's nameplate rated torque in its actuator capability calculations. Rather, the licensee used the stall torque at degraded voltage obtained from the zero speed point of the generic motor curve. Licensee Action Item 94-01623 resolved the irisue by revising the Brunswick dc MOV motor torque calculations to use the de motor rated torque values. The action item also documented that these changes did not involve any operability concerns. The inspectors sampled de MOV actuator capability calculations and verified that motor rated torque values were use E8.2 ! Closed) InsDettor Fol10 woo item 50-325(324)/94-20-02: MOV 2-E51.F046 lhermal Overload Tri This issue involved a failure of valve 2-E5;3046 (Reactor Core Isolation Cooling Water Su) ply) that occurrca on June 9, 199 The limit switch control for tais valve was not correctly set and the valve had run open into the backseat, stalled the motor and tri) ped the thermal overload. The followup i'em questioned whether t1e licensee had adequately evaluated the stall of this MOV to assure there had been no adverse effect on the motor and the output torque. The inspector s'

review during the current inspection found that the licensee had not adequately evaluated the effects of the stall event on valve structural component The licensee performed an evaluation during the inspection and determined that " weak link" ftructural thrust limits had been exceeded for the valve backseat, operator bolts, stem. disk and lifter, and yoke. The licensee identified the issue for resolution in Condition Report 97-02846. Based on discussion v!+ a representative of the valve manufacturer, the licensee believes , rt is no internal damage to the valve. Additionally, the valve has performed satisfactorily for several years following the stall. In its letter dated October 2. 1997, the licensee committed to inspect the valve for damage at the next refueling outage (starting September 1997). Additionally. the licensee stated it would provide a procedure revision by January 15. 1998, to ensure that any future stall is correctly evaluated. The licensee's initial failure to determine whether any MOV structural weak links had been exceeded when the valve stalled was considered indicative of inadequate corrective action measures. This was identified as an example of violation 50-325(324)/97-11-04. Inadequate Corrective Actions for HOV Other examples of this violation is described in El.1 abov . . - . .-

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IV. Plant Support

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R1 Radiological Protection and Chemistry Controls RI.1 General Radiolooical Controls Insoection Scone (83750)

The inspector evaluated the licensee's radiation control program in the area of occupational exposure controls during outage operations consistent with the requirements of 10 CFR Part 20. Areat evaluated included adequacy of pre-job radiation control planning. As Low As Reasonably Achievable (ALARA) pre-job briefings, effectiveness of the Radiation Work Permit (RWP) process, and adequacy of radiation surveys to support outage work activities, Observations and Findinos The inspector reviewed licensee controls for radiological exposures, such as RWP controls, radiation surveys, and pre-job ALARA briefings, to determine if they met applicable regulatory requirements and were designed to maintain exposures ALARA. The inspector reviewed select RWPs util12ed to control ongoing outage work within the radiation control area (RCA), with emphasis on high dose activities, and noted that the radiation controls observed were appropriate for described taiks and radiological conditions. Several specific Brunswick Nuclear Plant (BNP)

B213R1 outage related RWPs were reviewed to determine if supporting radiation survey data was current and sufficient to sup3 ort work to be conducted under the RWP with no discrepancies noted. RW)s reviewed-included those developed for control rod drive uncoupling and exchang RHR system work & isolation valves, and main steam isolation valve maintenance. Radiological control requiremtnts specified for the specific RWPs reviewed were determined to be adequate for the work scopes identified for these RWP The inspector reviewed the RWPs being utilized on the refuel floor for specific high dose tasks as well as general maintenance tasks. routine job coverage and inspection activity. Based on the inspector's review of these RWPs and discussions with licensee personnel, the inspector determined that the RWPs being utilized for general refuel floor work were aapropriate and adequate for the tasks that were permitted under these RWPs. The licensee was also able to demonstrate that RWPs had been prepared for those situations requiring a special RWP in accordance with licensee procedure. No discrepancies in implementation of the licensee's RWP procedure or with regulatory requirements were identifie The inspector evaluated the adequacy of the licensee's pre-job ALARA and radiation control briefing program to ensure that briefings were

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conducted in full compliance with procedure and were conducted in a

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manner sufficient to address radiological concerns of ongoing work. The

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inspector attended pre-job briefings. including those for ongoing work l

in the drywell, during this inspection and consistently observed

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brief'ngs that were thorough, indepth, and sufficient to minimize  ;

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unnecessary exposure and sufficient to identify radiological risks to radiation workers. Observed during these briefings was good specific planning as to how to minimize personnel exposure as well as full consideration of ALARA objectives with no procedural discrepancies identifie The inspector evaluated the adequacy of the licensee's radiation survey program to ensure that sufficient surveys were being conducted at the needed frequency to identify potential radiological hazards that may be :

1 resent. The inspector selected at random a sample of surveys in the Jnit 2 reactor building. Unit 2 turbine building, and radwaste buildin Each of the selected surveys was determined to be in compliance with the '

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licensee's procedure with respect to being up to date of adequate detail and completeness per procedure to fully characterize radiation hazards, and sufficient to support work planning needs with no discrepancies note ' Conclusions  ;

The licensee's radiation control 3rogram in the creas of radiation i surveys. RWP controls, and pre-jo) planning and ALARA briefings, was determined to be implemented effectively and in accordance with procedure R1.2 Soecific Radioloaical Controls Insnection Scone (83750)

Specific radiation controls evaluated in accordance with the requirements of 10 CFR Parts 19 & 20 included internal and external exposure controls locked high and very high radiation area controls, radiation area and contaminated area posting, general contamination controls, effectiveness of radiation worker training for radiation hazards, and labeling of radioactive material, Observation and Findinas The inspector made frequent tours of the radiation control are observed compliance of licensee personnel with radict. ion protection procedures for high dose B213R1 outage work evolutions, and conducted interviews with licensee personnel with respect to knowledge of radiation controls and specific radiological working condition During plant walkdowns within the RCA, the inspector conducted brief interviews at random with radiation workers inside the RCA in order to determine the level of understanding of RWP requirements and radiation working conditions. All of the workers interviewed were verified to have signed onto an RWP. were wearing dosimetry appropriate to their work activities within the RCA. ano were performing specific work activities on ap)ropriate RWPs. The questions asked included the RWP number of the RW) signed in on, electronic dosimetry alarm setpoints for

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dose and dose rate limits and general radiological working conditions for the areas worked in. For the workers interviewed, a good knowledge of RWP requirements and a good knowledge of radiation hazards and working conditions was demonstrate The inspector reviewed whole body exposures for all radiation workers at the site and determined that all personnel exposures assigned since the beginning of 1997 through September 24, 1997 were within 10 CFR Part 20 limits. A review of licensee personnel exposure records indicated the following maximum individual exposures at the plant during this period:

Total Effective Dose Equivalent (TEDE): 1384 mrem: Committed Effective Dose Equivalent (CEDE): 49 mrem: and Shallow Dose Equivalent (SDE): 3192 '

mrem. The inspector determined the licensee had adequately monitored and tracked individual occupational radiation exposures in accordance with 10 CFR Part 20 requirements and that all doses reported were at a small percentage of applicable regulatory limit The inspector reviewed and discussed with licensee representatives the

)rogram for controlling access to high radiation areas (HRAs), locked ligh radiation 6reas (LHRAs), and very high radiation areas (VHRAs).

These areas were ins)ected during tours for proper posting and access controls. _ No !! ras, _HRAs, or VHRAs were identified where required posting was needed but not posted. Areas controlled as LHRAs were inspected and found locked in accordance with licensee procedur '

Key controls for entry into locked and very high radiation areas were evaluated against the requirements of the licensee's administrative control procedure and determined to be controlled in accordance with procedure. A sample of survey instruments and respirators available for issuance were inspected and determined to have current calibration dates. Radiation workers during Jeak traffic periods were observed exiting the RCA in accordance wit 1 procedures for frisking out of the RCA to include properly clearing small articles with the small articles monitor. Numerous smears were taken by the inspector throughout the RCA. counted in the licensee's counting lab, and determined to be less than the licensee's contamination control limit for areas required to be controlled as contaminated. During inspection of the tool issuance rooms. good controls for slightly contaminated tools inside the RCA and for clean tools outstde the RCA were note During an inspection walkdown conducted on the morning of March 26, 1997. (as documented in Brunswick Inspection Resort 97-05, dated May 1 , as Unresolved item 97-05-08, Movement of lighly Contaminated Valve From Work Area to Storage), the inspector toured the Radioactive Material Storage Container Building (RMSCB) and identified a standard  :

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five gallon bucket with no radioactive material label that alarmed the l small article monitor with a digital readout of approximately million dpm. The bucket contained hiahly contaminated components

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including a 2 inch diaphragm valve which surveyed at 474,000 dpm tygon tubing that surveyed at 129,000 dpm, and a bag of nuts and bolts that-surveyed at 681,000 dp In addition to a safety concern with !

radioactive material found not controlled in accordance with licensee

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labeling procedures the inspector requested information during the March 1997 inspection as to the origin and contributing circumstances related to the bucket and its contents. Licensee investigations conducted since the earlier inspection (as documented in Inspection Report 97-05) were determined by the inspector to be adequate but inconclusive as to who was responsible as well as to the precise circumstonces or Jrigins of the bucket. During facility walkdowns conducted during this inspection, on the morning of Se)tember 24. 1997, the inspector identified a standard 55 gallon drum wit 11n the Unit i Reactor Building. near the control rod drive rebuild room. that was not labeled. Upon survey the drum was found to contain low levels of contamination (6000 dpm/100 sq.cm.) and constituted another example of a radioactive material container that should have been labeled in accordance with the licensee's labeling procedure. The failure of the licensee to label the five gallon bucket of highly contaminated radioactive materials, as well as the additional example of f ailure to label a 55 gallon drum with low levels of contamination, with a clearly visible label bearing the radiation symbol and the words " Cautio Radioactive Material" constitutes a violation of licensee procedure HPS-NGGC-0003. Radiological Posting. Labeling and Surveys. Revision 3. dated August 27, 1997. Paragraph 9.2. Tagging cnd Labeling of Radioactive Material which requires each container holding radioactive material to be so labeled. This violation is identified as VIO 50-325(324)/

97-11-07. Failure to Label Containers of Radioactive Material in Accordance with Procedur Conclusions The radiological controls progren was effectively implemented with good occupational exposure controls demonstrated during outage condition Good radiological control performance was evident in the occupational exposure control activities observed by the inspectors. One violation was identified for failure to label containers of radioactive material in accordance with procedur R4 Staff Knowledge and Performance in RP&C R4.1 Contract Worker in RCA Without Proper Dosimetry Inspection Scone (71750)

The inspector observed radiation workers compliance with the licensee's procedures for radiation protection, Ohs _ervations and Findinas During inspection activities on August 18, 1997. the inspector questioned a contract worker to verify whether the individual within the RCA was knowledgeable of the items contained on the associated RWP. was wearing proper dosimetry and the dosimetry was properly located. The individual was not aware of his RWP information. Also, the inspector observed that the worker was not wearing an electronic dosimeter as

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required by Environmental and Radiation Control 0E&RC-0230, Issue and Use of Radiation Work Permit. In addition, the inspector noted that the individual was carrying cigarette The inspector promptly notified the licensee of the finding. The licensee restricted the worker from entering the RCA and reconstructed the worker's exposure for that day. A standdown was conducted to '

reemphasize RCA access and basic General Employee Training (GET)

requirement s. Further review of this individual's RCA access history detennined that this worker had entered the RCA without signing on to an <

RWP or obtaining electronic dosimetry on two other occasion ,

TS 6.81.a requires that written procedures shall be established, implemented, and maintained covering the activities in Appendix A of Regulatory Guide (RG) 1.33. November 1972. RG 1.33. Appendix A. Section G requires procedures for Personnel Monitoring and Special Work Permits. Failing to sign on to an RWP, and to acquire appropriate dosimetry, are violations of OE&RC-230. This violation is identified as VIO 50-325(324)/97-11-08. Contract Employee in Radiation Controlled Area Without Proper Dosimetr On August 27. 1997, another individual was discovered by the licensee to have entered the RCA. Jerformed work, and exited without an electronic dosimeter as required )y 0E&RC-230. The tidividual's access to the RCA was restricted and other disciplinary actions were taken. On September 9. 1997, a contract worker was escorting another contract worker to cttend hands-on training on the Unit 2 117 elevation. U)on preparing to enter the RCA, a health abysics technician observed tlat the escorted contract worker did not lave a thermoluscent dosimeter in accordance with OE&RC-23 The inspector determined through review of these events and an event described in IR 50-325(324)/97-09 where the RCA boundary was violated without a health ph sics technician, that contractor knowledge of  ;

general radiologica control requirements was poo c-. Conclusions A violation was identified for a contract worker's repeated failure to log on to an RWP and obtain the recuired monitorinC dosimetr '

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Contractor knowledge of general raciological control requirements was poor as evidenced by several radiological control noncomformances and noted as a weakness,

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R7 Quality Assurance in Radiological Protection and Chemistry Activities R7.1 Positive Whole Body Count

, Insnection Scoce (71750)

The inspector reviewed the licensee's root cause/ event review concerning a contract worker who had a positive whole body count upon termination of site dutie Observations and Findinas

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On July 17, 1997, a contract employee was exiting the station after completion of work. This work consisted of several tasks or tha refuel floor. The individual was found to have a positive whole body csLat which indicated internal radioactivity of 43 nanocurie The licensee conducted a thorough review of this event and documented their conclusions in root cause/ event review CR 97-02514. Positive Whole Body Count. This review concluded that the most likely source was d ,

random particle ingestion occurred durir.g protective clothing handlin The individual was assessed a committed effective dose equivalent of 0.049 rem. The licensee's review evaluated a number of other factors related to this event such as work conditions, whole body counter accuracy, et The inspector reviewed the licensee's CR and other information provided to the inspector. A quality check was also reviewed related to how some information was coamunicated to the inspector. The information provided was consistent with the results in the final root cause/ event revie Conclusions The inspector concluded that the licensee conducted a thorough review of i this event. The information provided to the NRC was consisten R8 Hiscellaneous Radiation Protection and Chemistry Issues R8.1 ALARA Program Effectiveness Inspection Stone (Q?.5 Part 20 of Title 10 to the Code of Federal Regulations requires that licensees use, to the extent practicable, procedures and engineering t controls based upon sound radiation protection principles to achieve

! occupational doses and doses to members of the public that are as low as i reasonably achievable. 'The ALARA area was evaluated to determine whether the licensee was establishing and tracking performance against ALARA goals, whether continuing ALARA initiatives are ongoing to reduce dose, and to evaluate the overall effectiveness of the ALARA program,

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45 Observations and Findinas j Through September 23. 1997, the licensee projected a B213R1 refueling

- outage dose of 109.8 person rem and actually achieved a dose of 10 rem which was approximately equal to the goal. The overall outage dose goal of 209 person rem contained 11 person rem for emergent work and the licensee was on track to achieve its outage goal as of the week of inspection. The licensee is also on track to achieve their annual dose goal based on low dose accrual during power operation periods during 1997. During the first half of 1997 (180 days). BNP operated and accumulated a total of 80 person rem or 222 person mrem / day per uni This compares favorably with the BNP 336 person-mrem / day aer unit in .i 1996 (a decrease of greater than 30%) but remains above tie average of 197 person mrem / day per unit accumulated among other comparable BWR The annual dose goal, if achieved. is still at a relatively high level but represents good dose performance for the site. A short un)lanned outage in planned for Unit 1 later in 1997 to replace fuel leaters and this emergent work will add significantly to the annual ex)osure. The inspector reviewed with the licensee current and planned A_ ARA initiative During 1997, the licensee has undertaken several dose reduction initiatives including expanded application of shieldi.1 additional advanced radiation worker training (including training on mock ups for high dose jobs during outage), hot spot reduction, expanded t remate monitoring system, and hydrogen water chemistry managemen Although substantial dose saving was potentially achievable the licensee did not, undertake a full system chemical decon during the B213R1 outage due to cost / benefit considerations. Other potential dose saving ALARA initiatives that have been canceled for 1997 due to cost / benefit considerations include installation of more permanent shielding and decommissioning of radwaste equipmen Overall, the inspector determined that collective dose is being adequately controlled and reduced, Conclusions

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Overall, based on an evaluation of ALARA initiatives and ALARA work plans for high dose work evolutions, the inspector concluded that the licensee's ALARA program was adequately controlling collective dose and that collective dose was on a favorable reducing tren R8.2 Review of Previously identified Inspection Findinos (92904)

(Closed) VIO 50-325(324)/96-16-02: Failure to implement a Radiological Control Procedure Consistent With 10 CFR 20.1502 (a)(2).

The licensee has revised their applicable dosimetry / monitoring procedures to be consistent with the requirements of 10 CFR Part 20. The licensee has completed comprehensive corrective action that included, in part, a revised dosimetry requirement that all site personnel (including administrative and support personnel outside the protected area) be monitored for radiation exposure by TLD at all times while onsite. This item is closed, y

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(Closed) IFI 50-325(324)/97-05-07: Actions to Reduce Tritium in the Stabilization Pond The licensee has revised their applicable groundwater monitoring program procedure, effective September 24, 1997, to reflect increased groundwater monitoring and additional test wells to monitor tritium levels in and around the stabilization pond. Samples taken from the new test points since the time of the earlier inspection (see IR 97-05) were determined to be less than lower limits of detection for tritiu Releases of tritium from the site are w thin regulatory limits. The licensee had not as yet completed other initiatives to reduce tritium gcneration. such as routing condensate from the turbine building cooling system containing t;itium to the radwaste treatment system, as of the date of this inspection. This item is close .(Q gsed) URI 50.325(324)/97-05 08: Unlabeled Container of Radioactive Material This URI is closed based on issuance of the violation in section R1.2.

S4 Security and Safeguards Staff Knowledge and Performance S4.1 Site Access Controls Inspection Scone (71'50)

The in>pector observed physical security staff performance during routine personnel access, Observations and Findinas On August 27. 1997, the inspector obs u ved worker access into the protected area. A large number of workers were attempting to access into the plant. Multiple alarms were sounding on multiple lanes. As one worke.' would alarm the explosive or metal detector, the next worker would proceed through the detector. As a result, several workers requiring a pat-down or a search with hand held detectors were accumulating around the x-ray viewing station. The security officer in charge halted personnel access to allow adecuate time for personnel search activities. The inspector determinec that the security officer in charge took appropriate measures to maintain proper control of personnel access activities, The inspector observed individuals placing personal items onto a small table prior to walking through the metal detector. Due to the large number of individuals and the positioning of the security personnel, the inspector detcrmined that the potential existed for items to proceed through the access area undetected. This oversight was seen as a weakness in the plant's access control activities. The inspector discussed this issue with the licensee. Upon notification the licensee directed that security personnel establish a position with an

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unobstructed view of the table. In addition. the licensee will review removing the tables. None of the items the inspector observed being place on the table included any prohibited item Conclusions Appro)riate actions were observed during morning peak access activitie A weatness was identified which had the potential to allow items to bypass normal monitoring method F2 Status of Fire Protection Facilities and Equipment F2.1 Fire Separation Zones Insoection Scone (71750)

The inspector checked the fire separation zones to be sure they were free of transient combustible material Observations and Findinas During a routine tour of the Unit 2 Reactor Building on August 21. 1997, the inspector identified that there was transient combustible material in a fire separation zone on the 50 foot elevation. The area was unattended and the materials were not in use. The material was in an area labeled by signs that indicated transient combustibles were not allowed. The inspector informed the control room supervisor of this proble Tne licensee removed the material from the area and accelerated its corrective action for a violation in NRC IR 50-325(324)/97-08 concerning control transient combustible material in separation zones. CR 97-02837. Combustibles in Se)aration Zones was written by the licensee to ente; this problem into tie corrective action progra The inspector toured the plant buildings on August 24. 1997, and noted the licensee had placed signs on stands in the fire separation zones to make the areas more visible to )lant workers. Also, painting on the floor in fire reparation zones lad been started to clearly identify these area Conclusions The inspector concluded that the licensee was still struggling with control of transient combustible materials. but had accelerated their corrective actions to clearly identify areas where transient combustibles were not allowed.

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F8 Hiscellaneous Fire Protection Issues (92904)

(Closed) Unresolved item (URI) 50-325(324)/97 08-14: Designation of fire Separation Zones LCJosed)VIO 50-325(324)/97-08-13: Failure to Follow Fire Protection Program Procedures During inspection activities documented in IR 50-325(324)/97-08, the ins)ector discovered transient combastibles stored within various areas wit 11n the plant. Subsequently. Malation 50-325(324)/97-08-13. Failure to Follow Fire Protection Program Procedures, was issued. On one occasion transient combustibles were discovered within the Unit 1 ECCS Mini-Steam Tunnel. The entrance to RB1-6 was posted with a sign forbidding the storage of transient combustibles. After review of the a)plicable documents the licensee determined that, despite the postin t7e area was not a separation zone. The licensee performed a review of the Appendix R licensing basis for the se)aration zones. A verification was performed by the licensee to ensure tlat all those zones so designated through commitments to the NRC were adequately designated throughout the sit The inspector reviewed the licensee corrective actions for the violatio The licensee removed separation zone postings from the ECCS Mini-Steam Tunnels and the 17 foot elevation HPCI rooms for both unit Floor markings and additional posting were added to existing zones on the 20 foot and 50 foot elevations of the Reactor Buildings and in zone DG-8 in the Diesel Generator Buildin New zones posted included areas outside the drywell equipment hatch for Unit 1 and above the ECCS Mini-Steam Tunnel in both units. Based on completion of the corrective actions for violation 50-325(324)/97-08 13, Failure to Follow Fire Protection Progre Procedures these items are close V. Manaaement Mee h XI Exit Meetino Summa The inspector presented the inspection results to members of licensee management at the conclusion of the inspection on October 1. 1997 and on October 17, 1997 to clarify some GL-89-10 inspection issues. Post inspection briefings were conducted on August 28. 29, and September 25, 26. 1997. The hcensee acknowledged the findings presented.

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.. . PARTIAL LIST OF PERSONS CONTACTED Licensee A. Brittain. Manager Security M. Christinziano. Manager Environmental and Radiation Control W. Dorman Supervisor Licensing and Regulatory Programs N. Gannon. Manager Maintenance J. Gawron. Manager Nuclear Assessment S. Hinnant. Vice President. Brunswick Steam Electric Plant K. Jury. Manager Regulatory Affairs W. Levis. Director Site Operations B. Lindgren. Manager Site Support Services R. Lopriore Manager Training J. Lyash. Plant General Manager G. Miller. Manager Brunswick Engineering Support Section R. Mullis. Manager Operations Other licensee employees or contractors included office, operation, maintenance, chemistry, radiation. and corporate personnel.

A. Aiello E. Brown J. Coley E. Girard E. Guthrie M. Holbrook F. Jape C. Patterson W. Rankin T. Scarbrough M. Shymlock

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INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 41500: Training and 0:ialification IP 61726: Surveillance Observations IP 62706: Maintenance IP 62707: Maintenance Observations IP 71001: Requalification Inspection IP 71707: Plant Operations IP 71750: Plant Support Activities IP 73753: Inservice Ins)ection IP 8375P: Occupational bdiation Exposure IP 92901: Followup - Operations IP 92902: Followup - Maintenance IP 92903: Followp - Engineering IP 92904: Followup - Plant Support Temporary Instruction 2515/109: Safety Related Motor-Cperated Valve Testing and Surveillance ITEMS OPENED, CLOSED, AND DISCUSSED Onened 50-325/97-11-01 VIO Failure to Initiate Alternate Safe Shutdown impairment (paragraph 03.1)

50-325(324)/97-11-02 V10 PNSC Ouorum Too Many Alternates (paragraph 07.2)

50-325(324)/97-11-03 V10 Pen and ink Changes to 1 & C Procedure (paragraph M3.1)

50-325(324)/97 11-04 VIO Inadequate Corrective Actions for MOVs (paragraphs E1.1.b.2. El.1.b.3 E1.1.b.5)

50-325(324)/97-11-05 URI Use of Technical Support Memorandums (paragraph E3.1)

50-324/97-11-06 VIO Rigging to Safety Equi > ment Without a Safety Evaluation (paragraph E4.1)

50-325(324)/97-11-07 V10 Failure to label Containers of Radioactive Material in Accordance with Procedure (paragraph R1.2)

50-325(324)/97-11-08 VIO Contract Employee in Radiation Controlled Area Without Proper Dosimetry (paragraph R4.1)

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Closed 50 325(324)/96-10-02 V10 Failure to Disposition Repeated 1.0R Failures (paragraph 08.1)

50-325(324)/94-20-01 IFl Electric Torque Calculations for DC MOV Motors (paragraph E8.1)

50-325(324)/94-20-02 IFl MOV 2-E51-F046 Thermal Overlocd Trip (paragraph E8.2)

50-325(324)/96-16-02 V10 Failure to implement Radiological Control Procedure (paragraph R8.2)

50-325(324)/97-05 07 IFl Actions to Reduce Tritium in Stabilization Pond (paragraph R8.2)

50-325(324)/97-05-08 URI Movement of Highly Contaminated Valve From Work Area to Storage (paragraph RI.2)

50-325(324)/97-08-14 URI Designation of Fire Separation Zones (paragraph F8)

50 325(324)/97-08-13 VIO Failure to follow Fire Protection Program Procedures (paragraph F8)

Discussed 50-325(324)/96-15-04 Ifl Material Condition of Remote Shutdown Panels (paragraph M8.1)

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