IR 05000324/1998007

Insp Repts 50-324/98-07 & 50-325/98-07 on 980607-0718. Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support

ML20237B923
Person / Time
Site:	Brunswick
Issue date:	08/11/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20237B917	List: IR 05000324/1998007 ML20237B914 ML20237B921
References
50-324-98-07, 50-324-98-7, 50-325-98-07, 50-325-98-7, NUDOCS 9808200086
Download: ML20237B923 (31)

Text

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U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos: 50-325, 50-324 License Nos: DPR-71. DPR-62 Report No: 50-325/98-07. 50-324/98-07 Licensee: Carolina Power & Light (CP&L)

Facility: Brunswick Steam Electric Plant. Units 1 & 2 Location: 8470 River Road SE Southport. NC 28461 l

Dates: June '7 - July 18, 1998 Inspectors: C. Patterson. Senior Resident Inspector E. Brown Resident Inspector E. Guthrie Resident Inspector E. Testa. Health Physics Inspector (Sections R1-R5. R8.2)

W. Smith Inspector In Training N. Stinson. Inspector In Training Approved by: M. Ernstes. Acting Chief. Projects Branch 4 Division of Reactor Projects l

l Enclosure 2

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9808200006 980811 PDR ADOCK 05000324 G PDR I

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EXECUTIVE SUMMARY Brunswick Steam Electric Plant. Units 1 & 2 I NRC Inspection Report 50-325/98-07, 50-324/98-07 l I

This integrated inspection included aspects of licensee operation i maintenance, engineering, and plant support. The report covers a 6-week i period of resident ins)ection and includes the results of a radiation protection inspection ]y regional inspector Doerations

. High temperatures were indicated around the reactor recirculation pump I motors and reactor building closed loop cooling water system's supply line. The temperatures observed were in excess of the limits prescribed in the Updated Final Safety Analysis Report (UFSAR). The licensee intended to perform evaluations to allow an increase in these systems required temperature. An unresolved item was issued to track the resolution of the elevated temperatures for these systems (Section 02.1). i e Accessible valves and electrical components for the Reactor Core l Isolation Cooling System were verified to be properly aligne I Housekeeping was determined to be satisfactory (Section 02.2).

• Shift staffing and task allocation were adequate and potential changes l to the STA/SRO and R0/EC positions would meet regulatory requirements (Section 06.1).

. The licensee conducted thorough reviews of Improved Technical Specifications implementation readiness during several Plant Nuclear Safety Committee meetings (Section 07.1).

. The same level of control and risk assessment was a] plied to a short maintenance outage as a normal refueling outage. T11s was identified as a strength (Section 07.2).

Maintenance

. The licensee identified a 63-inch crack in a fuel rod from the Unit 1 core. Debris was determined to be the initiator of the failure. Good planning was observed for work in full protective clothing with an area temperature greater than 85 degrees Fahrenheit. Adequate health physics and supervisory oversight were present (Section M2.1).

. During the Spring 1998 Unit 1 outage an increase in errors was noted in maintenance and test activities. Several examples of multiple barrier l breakdowns, such as independent verification, were identified as a

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violation of plant procedures. The licensee initiated a human performance improvement initiative to address the declining trend (Section M4.1).

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E-Enaineerina

. The inspector identified a corrective action violation in identification  ;

and untimely initiation of corrective actions regarding an adverse '

condition involving the Run Control Relay (RCR) and Jet Assist Time Relay (JATR) DG air bellows time delay relays (Section E2.1).

. The inspector found that the licensee had maintained Special Nuclear Material (SNM) accountability for a Local Power Range Monitor (LPRM) in question. The inspector found that the licensee's warehouse inventory had an error in the number of LPRM units on hand due to not having well defined procedural guidance on SNM in their material control procedures (Section E3.1).

. The licensee acknowledged that design control activities warranted additional attention. These activities were accuracy of Equipment Data Base System (EDBS). staffing of design control organization. and UFSAR updates (Section E7.1).

Plant Sucoort

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. The licensee's radiation control practices for radiological posting and labeling, personnel dosimetry, radiological surveys, radiation ,

controlled area ingress and egress control, and whole body counting were consistent with the licensee's Radiation Control and Protection Manual and relevant sections of 10 CFR 20 (Section R1.1).

. The licensee's practices for calibrating radiological survey instrumentation were consistent with the licensee's Calibration and Use of Eberline RM-14 Radiation Monitor 3rocedure and relevant sections of 10 CFR 20. The licensee maintained 3reathing air bottle calibrations and breathing air quality in accordance with the licensee's Radiation Control and Protection Manual and relevant sections of 10 CFR 20 (Section R2.1).

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. The licensee performed procedural change reviews in accordance with its Performance of Nuclear Safety Reviews procedure and relevant sections of 10 CFR 50 (Section R3.1).

. The licensee effectively implemented training and qualification for Environmental and Radiation Control management and technicians in accordance with the training and qualification program requirements as required in Technical Specifications 6.3.1 and 6.4 (Section R5.1). 1

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. The triennial fire protection audit was determined to be a thorough l

review of the program. Audit findings reflected extensive auditor knowledge of fire protection and the associated requirement Additional assessments were scheduled to independently review corrective

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actions for the fire protection program (Section F7.1). I

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i Report Details )

Summary of Plant Status I

Unit 1 operated continuously during this report period. At the end of the report period the unit had been on-line continuously for 51 day Unit 2 was shutdown on June 17, 1998, for a several day maintenance outag This ended 238 days of continuous operation. At the end of the report period the unit had been on-line continuously for 27 days. The unit operated with two control rods inserted to suppress power around a leaking fuel assembl I. Operations l 02 Operational Status of Facilities and Equipment 02.1 Elevated Temperatures Insoection Scooe (71707)

l The inspector monitored plant conditions due to the high ambient I temperature during June 1998. Plant operating temperatures were compared to design limits as stated in the updated final safety analysis ;

report (UFSAR) and vendor documentatio ' Observations and findinas On June 9.1998, the inspector observed Control Room indications of drywell (DW) temperature. The 23 foot (ft.) elevation for the Unit 2 DW indicated 135 degrees Fahrenheit ( F). UFSAR Section 9.4.6.2. Primary Containment Cooling System stated "the system contains multiple fan-coil cooling units' and was " designed to maintain an average tem 3erature of 135 F. with 128 F around the recirculation pump motors." T1e inspector determined that the recirculation pump motors extended into the 23 f elevation which indicated 135 F. The inspector discussed this issue with the on-shift perscnnel. The licensee indicated that as a result of failures of multiple DW cooling fans a forced outage was planned for June 17, 1998 to repair the cooler On June 29, 1998. after completion of the forced outage the inspector observed an elevated temperature for the 23 ft. elevation of the Unit 2 DW. The control room recorder indicated a temperature of 145.6 F. The inspector questioned the on-shift personnel regarding the temperature requirements contained in UFSAR section 9.4. After further !

t discussion, the licensee initiated Condition Report (CR) 98-1625, 128 l

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Deg F.UFSAR Temperature. The CR indicated that emergency temperature i assumptions were 150 F at a winding temperature of 266 F max. The j inspector had not observed temperatures in excess of 150 F around the DW j recirculation pump motors. Inspector review of the vendor manual i indicated that the rating of the motor was based on the cooling air not exceeding 134.6 F. Based on this information the inspector questicn e l engineering personnel regarding maintaining the recirculation motrr i ratin ;

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l 2 l On June 19, 1998, the inspector attended a licensee meeting which  !

discussed plant status. The licensee indicated that they were having  ;

problems maintaining Reactor Building Closed Cooling Water (RBCCW)

system supply temperatures within specifications. The licensee indicated that this was not a safety related system and that the

1. emperature s) edification of concern was not a Technical Specification  !

(TS) limit, t1erefore they were going to raise the specification to some  :

number greater than 100 degrees. The inspector informed the licensee l that UFSAR. Section 9.2.2.2, RBCCW System Description, specified that

"An adequate supply of service water is available during all modes of o)eration to maintain the RBCCW supply at less than 100 F." Upon review t1e licensee identified that RBCCW supply temperature had exceeded the UFSAR temperature of 100 F for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, when the temperature rose to 100.3 F, on June 18, 1998. The licensee informed the inspector that they were going to change the UFSAR temperature to a higher value. During this inspection period this temperature on Unit 2 has been as high as 104 F for periods of several hours at a tim Based on questions regarding the impact of altering the allowed i temperatures, this item is unresolved. This unresolved item (URI) is ,

identified as URI 50-325(324)/98-07-01, Elevated Temperature ! Conclusions High temperatures were indicated around the reactor recirculation pump motors and reactor building closed loop cooling water system's supply line. The temperatures observed were in excess of the limits prescribed in the UFSAR. The licensee intends to perform evaluations to allow an increase in these systems required temperature. An unresolved item was  !

issued to track the resolution of the elevated temperatures for these  ;

system .2 Reactor Core Isolation Coolina (RCIC) Walkdown Insoection Scooe (71707)

The ins)ector performed a verification of the operating line-up for the  !

Unit 1 3CIC syste i l Observations and findinas '

l"- Accessible valves and electrical components were verified to be l correctly aligned in accordance with Operating Procedure 10P-16. Reactor

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Core Isolation Cooling System, Revision (Rev.) 43 and drawings D-25029 sheets 1 and 2. Valves in the system were appropriately labeled and no gross packing leakage or missing actuator handwheels were identifie The inspector observed that no equipment was staged which might threaten system performance. No transient combustibles were identified in the area, nor were any ignition sources observed. The ins)ector observed rust on the condensate storage tank. No leakage was o) served from the rusted areas identified. The licensee generated CR 98-1641. CST corrosion near CO-V152, to document the finding. The licen:'e verified

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adequate tank thickness and structural integrity of the associated high-pressure coolant injection (HPCI)/RCIC instrumentation lines, c. Conclusions

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Accessible valves and electrical components for the Reactor Core

! Isolation Cooling System were verified to be properly aligned.

i Housekeeping was determined to be satisfactor Operations Organization and Administration I

06.1 Shift Staffino and Task Allocation a. Insoection Scope (71707)

The inspector reviewed operations shift staffing and task allocation during this inspection perio b. Observations and Findinas The inspector' questioned the licensee's overall shift staffing and task allocation following recent problems the licensee incurred with meeting 1 Alternate Safe Shutdown (ASSD) shift manning requirements. ASSD shift  !

manning problems were discussed in Inspection Report (IR) 50-325(324)/ l 98-06 paragraph F The inspector reviewed the licensee's records and procedures, and conducted observations and discussions with the licensee regarding shift staffing and task allocation changes and challenges presented to the licensee. The inspector was informed by the licensee that they had been considering two changes regarding shift staffing and task allocations.

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The first change was to have the Shift Technical Advi.sor (STA) stand the Senior Control Operator (SCO) watch position concurrently. Up until the end of this review they had not implemented this dual-role positio The second change was to have Reactor Operators (RO)s serve as Emergency Communicators (EC)s. The licensee wanted the flexibility to use the R0 as EC only during non-taxing emergency response scenarios for the R Normally the licensee expressed that they wanted the Auxiliary Unit 0)erator (AUD) to fulfill the function of EC during emergency respons W1 ether the RO or AUD was used as EC was left to the discretion of the Site Emergency Coordinator, who was typically the Shift Superintenden The ins)ector reviewed several NRC Generic Letters. Information Notices, and a NJREG which discussed this matte The inspection program was continuing to monitor any effects resulting from changes in shift i staffing and task allocations for events which were called into question.

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c. Conclusions The inspector concluded that the shift staffing and task allocation were adequate and that the potential changes to the STA/SR0 and R0/EC positions would meet regulatory requirement Quality Assurance in Operations 07.1 Lmoroved Standard Technical Specifications Implementation a. Insoection Scoce (71707. 40500)

The inspector attended several Plant Nuclear Safety Committee (PNSC) I meetings that discussed the licensee's preparation for implementation of Improved Technical Specifications (ITS).

b. Observations and Findinas On July 2. 1998, the inspector attended a PNSC meeting convened to review the licensee's readiness for implementation of the ITS. The I

inspector noted that the proper persons were present to meet the quorum TS requirements. The PNSC agenda comprised of receiving affirmations of readiness from the appropriate organizations to implement the ITS with a projected implementation date of July 11, 1998. The inspector noted that all affirmations of readiness were received by the PNSC Chairman but several issues were not resolved to allow implementation on July ll, i 1998. The most significant item holding up implementation of ITS was '

the Thermal Hydraulic Instability modificatio This was due to the necessary time needed to complete procedure revisions and the final installation of the modification, which needed to be completed concurrently with ITS implementation. The licensee assigned action items and I rmmitted to reconvene PNSC on July 16, 1998, to complete the readiness review for ITS implementation. PNSC appropriately established a later date of July 25, 1998 to implement IT The inspector attended a portion of a PNSC meeting conducted July 16, 1998. There was a thorough discussion of ITS implementation issue The PNSC Chairman provided a good safety focus for the meeting. An additional review of open items was scheduled prior to July 29, 199 Conclusions The licensee conducted thorough reviews of ITS impleraentation readiness during several PNSC meeting .2 Maintenance Outaae Control Inspection Scope (71707)

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b. Observations and Findinas I This maintenance outage was conducted to make permanent repairs to the main circulating water system debris filter, re)lacement of the failed DW cooling fan motors, com)letion of remaining E0 work and repair of DW floor drain sump pumps. T1e inspector noted that the licensee applied the same controls to the maintenance outage used during a refueling outag A safe shutdown risk management assessment was conducte The outage center was staffed during the outage. A startup PNSC was

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conducted on June 19, 1998, to review all items in detail. The outage was completed on schedule without difficult c. Conclusions The same level of control and risk assessment was a) plied to a short I maintenance outage as a normal refueling outage. T11s was identified as a strengt II. Maintenance M2 Maintenance and Material Condition of Facilitie.s and Equipment  ;

M2.1 Failed Fuel Inspections a. Inspection Scone (62707)

The inspector observed the inspection of a failed fuel bundle removed from Unit b. Observations and Findinos On June 24. 1998, the inspector observed visual inspection activities conducted on the 117 ft. elevation of the Unit 1 Reactor Building. The licensee identified a 63-inch crack along a single fuel rod. There was a discernable fret mark located on the fuel rod. Ultrasonic and eddy current testing was also aerformed. The licensee in conjunction with the fuel vendor contend tlat debris was the initiator of the failur Additional inspection of another bundle removed in February 1998 was conducted. However, no obvious fret marks were identifie Inspection activities were conducted satisfactorily. The inspector noted adequate health physics presence as well as adequate supervisory oversight. Good planning was demonstrated for the safety of the workers, while working with an area temperature greater than 85 F in full protective clothing around the spent fuel pool. Activities observed were performed consistent with the guidance provided in Administrative Instruction 0AI-106. Foreign Material Exclusion for the Refueling Floor. Rev.10. such as use of capture devices in the foreign material exclusion area around the spent fuel pool.

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l c. Conclusion The licensee identified a 63-inch crack in a fuel rod from the Unit 1

, core. Debris was determined to be the initiator of the failur Adequate health physics and supervisory oversight were present. Good planning was observed for work in full protective clothing with an area temperature greater than 85 M4 Maintenance Staff Knowledge and Performance M4.1 Maintenance Configuration Control Insoection Scooe (62707)

The inspector reviewed the circumstances surrounding an identified negative trend in the control of test and maintenance activitie This trend was made evident by several events, recorded in CRs and their associated root cause analysis:

. 98-1077. MCC 1XC breaker flash; e 98-1280. Unexpected Rod Bloc . 98-0789. Inadvertent y scram:

• 98-1104 Valve Out of Position (1-E41-V159. HPCI Pump Discharge

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Line Check Valve):

• 98-1220. Torus Master Clearance; l

l * 98-1219. 1-E51-F029 Clearance problem.

' Observations and Findinos URI 50-325(324)/98-06-03 identified a negative trend during the outage in the licensee's control of test and maintenance activities. These events dealt with the failure of multiple processes including procedural adherence, communication, and independent verification. Several events were caused by i.nstrument and control (I&C) personnel. Maintenance Management Manual 0MMM-001 Maintenance: Conduct of Operations. Rev. 32, required that procedures be performed as written unless permission to deviate has been given by a supervisor.

l Several events were related to clearances issues during the refueling l outage. Conduct of Operations Manual. Operating Instruction 001-1.09.

l Equipment Tagging. Rev. 2. required that equipment tagging provide a high degree of personnel and equipment safety as well as maintain the statas and integrity of important plant components and systems. As a result of these events and others, on June 12, 1998, the licensee initiated an industry " Excellence in Human Performance" improvement plan to promote excellence in human performance to reduce human error _ _ _ _ - _ _ - . _ _ _ _ -_-

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l CR 98-1077. MCC 1XC breaker flash

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On May 1,1998, around 2:45 p.m. the fire brigade was called to respond

to a brief " fire ball" in a MCC compartment on the 20 ft elevation of l the Unit 1 Reactor Building. The " fire ball" was observed while an I&C technician was reinstalling the 1-G31-F001 control breaker after completion of preventive maintenance (PM) activities. The fire alarm '

was sounded and the fire brigade responded promptly. As a result of  !

proper implementation of site safety precautions, no personnel injuries j occurred. The cause was attributed to a jum)er being left on the '

breaker after testing was performed during t1e P The inspector reviewed the following procedures: l l

. Preventive Maintenance OPM-BKR008. PM-Functional Testing of Molded Case Circuit Breakers Rev. 1 . Administrative Instruction 0AI-59, Jumpering and Wire Removal, l Rev. 25, l

. Maintenance Management Manual 0MMM-001, Maintenance: Conduct of Operations. Rev. 32,

. Plant Program Procedure OPLP-1.2. Independent Verification, j Rev. 11,

. Administrative Procedure 0AP-10. Procedure Use and Adherence, )<

Rev. )

i The root cause analysis associated with CR 98-1077 indicated that inadequate self-checking was the root cause. Based on review of the above procedures and a demonstration of the test technique used, the inspector does not concur that inadequate self-checking was the root cause. While adequate self-checking may have prevented this event, the inspector concluded that performance of the procedure as written did not discuss use of the jumpers. Jumpers were controlled by 0AI-59 but were not addressed in the procedur CFR 50 Appendix B Criteria V, Instructions. Procedures, and Drawings, requires that activities affecting quality shall be described in documented instructions and shall be accomplished in accordance with these instructions. OMMM-001 required procedures to be used in accordance with the approved level of use as provided in 0AP-1 AP-10 allowed performance of reference-use procedures, such as OPM-BKR008, from memory, however adherence to the procedure was still require The failure to follow OPM-BKR008 is a violation. This violation is identified as the first example of VIO 50-325(324)/98-07-02, l Configuration Control Problem . UnexDected Rod Block On May 15, 1998, an unexpected rod block was identified after completing a weekly average power range monitor (APRM) surveillance test on Unit The unit was in refueling and core alterations were in progress. After the performance of Maintenance Surveillance Test 1MST-APRM21W. APRM 12 percent Rod Block, 15 percent RPS Trip, and Inop Chan Funct Test / Cal,

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I l 8 Rev. 9. the licensee determined that a lifted lead had not been reterminated during restoration of the APRM to service. CR 98-1280.

Unexpected Rod Block, was initiated. This CR indicated that during testing, a technician erroneously signed off the determination step as not applicable (N/A).

The failure to determinate wire 1-C11-A3-2 in accordance with IMST-

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APRM21W is a violation. This violation is identified as the second i example of VIO 50-325(324)/98-07-02. Configuration Control Problems.

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As part of the inspector's review of this event a cosy of the completed maintenance surveillance test (MST) was recuested. )uring licensing review of the requested paperwork, it was cetermined that contrary to the CR the ste) which was indicated as N/A. was signed off as complete in the MST. C1 98-1449. Documentation Practice, was initiated to document this finding. The licensee indicated ttat upon returning to the steps for the wire lift. the N/A'd ste)s had been lined out and the steps signed off as performed. However, tais could not be verified by the inspector due to the field copy of the procedure being destroye The licensee indicated that the technicians took a " clean" co)y of the procedure and transferred the test performance information. iowever, the steps initially N/A'd and reperformed were not reproduced during the transfer, all steps in the procedure indicated that the steps had been performed sequentially without any deviations. The inspector questioned the loss of the record of the activities performed and of the change in plant configuration. The licensee indicated that the only documentation error was the failure to note on the " master" copy that a l'

CR had been initiated as a result of the failure to reland the wire lif . Inadvertent k scram <

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On April 1.1998, the licensee was performing testing in accordance with Maintenance Surveillance Tect 2MST-APRM290. APRM Flow Bias Flow Units i C & D Channel Calibration. Rev. 21 Technicians placed the mode switch i for APRM C in an unlabeled third posl u n. This action resulted in a half scram. The procedure indicated that flow unit C mode switc instead of the mode switch for APRM C. was to be moved to an unlabeled third position. Licensee review determined that the unfamiliarity of the technicians with the instrumentation was a factor. In addition, the technician providing concurrent verification was unfamiliar with the instrumentation. During the testing, the technician questioned the correctness of the switch that was about to be positioned. Despite questions with positioning of the APRM C mode switch, both technicians agreed to Josition the wrong switch without stopping and contacting a knowledgea)le individual. The inspector determined that configuration control was not maintained as a result of inadequate verificatio The failure to properly align the flow unit C mode switch in accordance with 2MST-APRM290 is a violation. This violation is identified as the third example of VIO 50-325(324)/98-07-02. Configuration Control l Problem _ _ - _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _

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l CR 98-1104 salve Out of Position (1-E41-V159. HPCI Pumo Discharae Line

! Check Valve)

l On May 1. 1998, the licensee was hanging clearance 1-98-00060 to establish a boundary for local leak rate testing (LLRT) feedwater (FW)

flow element inspection, eddy current testing on a FW heater, and repair work on the 1-FW-V8. HPCI Pump Discharge Line Check Valv 'Alve 1-E41-V159..was required to be gagged open during the performance of LLRT for l the 1-E41-F006. HPCI. Injection Valve. Testing of the 1-E41-F006 was L determined to be unsatisfactor Additional review by the licensee determined tlat the 1-E41-V159 had )

been gagged in the closed position instead of open. The associated

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root cause analysis indicated that the maintenance crew installing the gag assumed that the valve was ope Subsequent, independent verification by Operations personnel, also assumed that the valve was ,

gagged o)en. The Operations personnel remarked that the indicator was left in Jetween the open and closed positions, due to the assumption that the maintenance personnel who originally positioned the valve were l

correct. The erroneous assumption by the Operations personnel and unfamiliarity with the 1-E41-V159 were identified as contributors to this event. The inspector noted that despite unfamiliarity being a contributor to this event, the licensee determined that no training on this uniquely configured valve was necessary. Inspector review of this event determined that the unfamiliarity of the maintenance personnel was attributable to the crew being contractor The failure to properly align the 1-E41-V159 in accordance with clearance 1-98-00060 is a violation. This violation is identified as the fourth example of VIO 50-325(324)/98-07-02. Configuration Control Problem . Torus Master Clearance On May 11. 1998, during a walkdown of the reactor turbine generator board (RTGB), operators questioned the absence of the fluid boundary- 1 between the torus and work activities being conducted on the HPCI RCIC,  !

and Core Spray (CS) systems. Additional review determined that the torus master clearance. 1-98-0001 had been canceled and removed. The fluid boundary was established by the master clearance and three sub-tier clearances (1-98-00061. 1-98-00062, and 1-98-00391) for the out-of- '

service systems. Removal of the master clearance removed portions of the barrier providing protection to personnel performing work on the i other system The licensee determined that to save time the torus implementation manager requested a contract individual serving as the project planner to sign the implementation manager off of the torus master clearanc The project planner updated the status of the torus work as satisfactory, signed the implementation manager off, and proceeded to sign off the sub-tier clearances as satisfactory despite the other l clearances referencing the mechanical maintenance supervisor and the

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L L 10 Unit 1 Senior Reactor Operator (SRO) as supervisors for the uncompleted

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removal of the torus master clearance. personnel . consequently authorized

( The licensee indicated that the error was due to improper assumptions on the part of the project planner and poor communication. The inspector, through review of this event and discussion with the licensee, concluded that _ inadequate oversight of the clearance process by operations was a contributing cause. The licensee indicated that the clearance holder

• was respcasible for assuring that all work activities were complete '

l before signing off.the clearance, which prompts restoration of plant

configuration. The inspector questioned whether the SR0 was expected to

, independently verify that work activities were complete when canceling a ('

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clearance. The licensee indicated that the SRO only verified that the clearance. documentation was complete. The inspector noted that this process, in effect. allowed contract personnel to affect plant configuration without verification or validation of proper plant conditions by a member of the license The failure to properly maintain the clearance boundary established under clearances 1-98-00001 1-98-00061. 1-98-00062, and 1-98-00391

! while work activities were still being performed is a violation. This violation is identified as the fifth example of VIO 50-325(324)/98-07-

, 02. Configuration Control Problem . 1-E51-F029 Clearance Problem On May 12. 1998, during removal of clearance 1-98-0001, the licensee L discovered that motor-operated valve 1-E51-F029. RCIC Suppression Pool Suction Valve, was found missositione The governing clearance 1-98-00062 required the valve to 3e open. The valve was found with a-clearance tag in the closed position. Licensee investigation revealed

,. that the motor had not been electrically disconnected. As a result the licensee contends that the valve was initially positioned to the open position. Plant conditions, specifically low steam sup)ly pressure for this system may have caused an isolation which closed t1e valv Inspector review of this event and discussion with the licensee revealed that the clearan:e was not @ quately established. The failure to properly maintain the clearance boundary establis' aJ under clearances 1-98-00001. 1-98-00061, 1-98-0062, and 1-98-00391 while work activities were still being ,)erformed is a violation. This violation is identified E as the sixth example of VIO 50-325(324)/98-07-02. Configuration Control-Problem Conclusions During the Spring 1998 Unit 1 outage an increase in errors was noted in maintenance and test activities. Several examples of multiple barrier breakdowns, such as independent verification, were identified as a I- violation of plant procedures. The licensee initiated a human performance improvement initiative to address the declining tren _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ ___ ______ _ _-

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M8- Miscellaneous Maintenance Issues-(92902)

M (Closed) Violation VIO 50-325(324)/97-11-03: Pen and Ink Changes to I&C Procedure 26. 1997, admitting The licensee reshonded the violatio he cause of theto violation this violation onthe was that November individual responsible for a) proving the changes to the procedure inappropriately determined that tie scope of the changes constituted a clerical' chang AP-010. Procedure Use and Adherence, was changed to eliminate those sections of the procedure allowing minor editorial enhancements and i

field corrections. The inspector verified that a procedure revision to 0AP-010 was mad This violation was close M8.2- (Closed) Violation 50-324/97-11-06: Rigging to Safety Equipment Without a. Safety Evaluation

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.The licensee responded to this violation in a letter dated November 2 ~199 The licensee had completed all the. corrective actions assigned

from the root cause analysis except performing an effectiveness review.

I Additional treining was conducted for all individuals involved, with rigging particularly to safety related components. Rigging to safety

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related components was determined by the licensee to be a knowledge weakness. A procedure revision.was comploted for Maintenance Management Manual 0MMM-022. Instructions for Placement of Temporary Loads (e. Rigging. Scaffolding. Ladders. Personnel) in an effort to enhance the procedure by specifying the requirements for rigging off safety related equipment and the requirements for rigging releases. The inspector found that the corrective actions were appropriate, therefore this item 1.s close M8.3- (Closed) Deviation DEV 50-325(324)/96-08-03: Failure to Complete Maintenance Procedure Backlog This item was closed in IR 50-325(324)/98-06 as 50-325(324)/96-08-01, but should have been closed as 50-325(324)/96-08-0 M8.4 (Closed) Unresolved item URI 50-325(324)/98-06-03: Configuration Control of Maintenance and Test Activities l Based on the issuance of violation 50-325(324)/98-07-02. Configuration

Control.. this item is closed.

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II Enaineerina E2 Engineering Support of Facilities and Equipment E2.1 Diesel Generator (DG) Air Bellows Time Delay Relav Failures Insoection Scope (37551. 61726. 71707)

The inspector reviewed air bellows time delay relay failures used in two applications on each of the DGs: one application was the Run Control Relay (RCR) and the second. application was the Jet Assist Time Relay (JATR). Observations and Findinas I On June 24, 1998, the inspector reviewed CR 98-01567. DG1 RCR Relay Calibration Failure. The CR stated that during the performance of calibration procedure OPIC-TMR006. Calibration of Alfen Bradley Model 700DC Time Delay Relays. Rev. 2. that DG1 RCR relay time delay contacts failed to actuate. This condition was identified on the CR as an operability concern or a potential reportable event. The immediate action taken was to replace the relay, which after replacement calibrated satisfactoril The DG1 RCR relay. failure CR identified corrective actions as 1)

determine past operability of DGl. 2) perform a root cause analysis of the RCR relay failure, and 3) replace the DG2 and DG3 RCR relays during the next available opportunity, which had a due date of December 3 (The DG4 RCR relay was replaced in 1996 after failing to l consistently actuate within the required calibration time). The same CR stated that a similar time delay relay, used in the JATR relay application, had racently failed on DG2 during the performance of the same relay calibration procedure. The JATR failure had occurred on February 2. 1998 and was discussed in CR 98-00556. DG2 JATR Failure.

l The inspector reviewed the JATR relay failure CR and noted that this relay was replaced upon failure of the relay contacts to change state during calibration. The inspector also noted that the CR performed a past operability determination, which was not finished until May 1 , even though the CR did not identify this as being an operability concer Following review of both of the CRs for the RCR and JATR relay failures the inspector expressed concern for possible common mode failure of the air bellows time delay relays and questioned the licensee why the other RCR and JATR relays on the other DGs were not tested following the !

second identified failure. The licensee believed the failure mechanism to be associated with the calibration procedure which used a mechanical actuating lever to actuate the time-delay and the contacts. The licensee b lieved that the relay functioned normally under a normal electrical solenoid actuation. The inspector noted that the licensee had conducted operability determinations, but made no conclusions on the I

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failure mechanism yet concluded that no further testing of the other DGs was necessar This was discussed with the licensee after which they reevaluated their course of action and determined that they would

perform an electrical actuation of each relay, both the RCR and JATR l relays, that were installed in each DG. This testing began on June 26.

l 1998 and finished on June 29, 1998. All of the relays showed a change

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of state when the electrical solenoid actuated the relays. The inspector observed the relay testing on DG3 and DG4, on June 26, 199 and noted no problems with the testin The electrical actuation of the relays verified operability of the DGs.

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Prior to this test on June 26, 1998, the licensee was using age testing I on the relays, which had been removed upon the calibration failures, to i deter.nine the failure mechanism. The licensee was leaning toward

hardened lubrication as the failure mechanism but had not come to a l conclusion after about five months. The failed JATR relay had been l tested on March 10, 1998, which was 36 days after the failure, and then again on May 12, 1998, which was 63 days after failure. Each time the

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I relay was tested the first mechanical actuation failed and subsequent mechanical actuation passe The inspector noted that CR 98-00556 was not generated until Narch even though the event oate was identified as February 2.199 The CR stated that it was generated to track the failure as a  ;

Maintenance Rule Functional Failure. The inspector also noted that l CR 98-01567 was not generated until June 23, 1998 with an event date of l June 15. 1998. The DGs were considered high safety significant systems and they had the highest risk contribution in Brunswick's Probabilistic Risk Assessment. Additionally, the impact of common mode failure of the relays could have been significant since the JATR relay provides an air boost to the DG turbocharger on emergency loadN of the diesel and the RCR relay allows for 2 roper load sequencing of the aiesel in certain emergency scenarios w1en the diesel is already running in local manua Based on this, the CR 3rocess was not performed in a timely manner, since response should 3e commensurate with the safety significance of the component and the system. Additionally, the inspector noted that the root cause analysis for CR 98-00556 had not been completed as of June 24, 199 Plant Program Procedure OPLP-04, Corrective Action Management. Section 6.9 required that level I and level II root cause  !

analysis root cause analysis be performed in 28 days. However, this root cause analysis had been extended two times. The extensions were performed in accordance with the procedure. The root cause analysis was extended because age testing and arialysis had not conclusively determined the root cause. The inspector noted that the root cause l analysis may not have conclusively determined a failure cause and that the electrical testing of the relays immediately determined the o>erability condition of the diesels. The root cause analysis method close an analysis method that was slow to conclusively determine j operability and common mode failure condition CFR 50. Appendix B. Criterion XVI. Corrective Action, requires that measures shall be established to assure that conditions adverse to

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l l 14 quality, such as failures and malfunctions are promptly identified and corrected. In the case of significant conditions adverse to cualit the measures shall assure that the cause of the condition is cetermined l and corrective action taken to preclude repetition. The identification l of the significant condition adverse to quality. the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management. The failure to initiate corrective actions in a timely manner for both CR 98-00556 and CR 98-l 01567 are two examples that constitute a violation of 10 CFR 5 )

Appendix B. Criterion XVI. Corrective Action. This violation is being i tracked as 50-325(324)/98-07-03. Diesel Generator Relay Failures.

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c. Conclusions t

l The inspector identified a corrective action violation in identification and untimely initiation of corrective actions regarding an adverse

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condition involving the RCR and JATR DG air bellows time delay relay E3 Engineering Procedures and Documentation E3.1 Nonfuel Soecial Nuclear Material (SNM) Accountability Insoection Scooe (37551) l The inspector reviewed an inventory discrepancy associated with a Local Power Range Monitor (LPRM) for the potential loss of accountability for SNM and material handling quality control issue b. Observations and Findinas On June 30, 1998, the inspector reviewed the licensees nonfuel records I of accountability for SNM associated with an inventory discrepancy which was identified in a CR 98-01330. Inventory Discrepancy-LPRM. The SNM records reviewed by the inspector indicated that the SNM had been accounted for since the LPRM was received in 1990. The CR indicated that the inventory error existed in the warehouse inventory records and that they had one more on hand than they actually did. The inspector verified through discussions with the licensee that the loss of inventory in the warehouse records for this LPRM affected the number of components on hand and did not affect the quality control certifications for that component. When the inspector reviewed the nonfuel SNM records for the LPRM in question. the inspector found that the warehouse inventory discrepancy had existed since 199 The inspector noted that the warehouse and engineering responsibilities

, for accountability of the nonfuel items such as LPRMs. were not clearly l defined. This was believed to be the cause of the incorrect inventor The warehouse was not issuing these type of items from their inventory when they left the warehouse as they normally would for most other items. The inspector determined this to be for two reasons. One reason was that the warehouse relied on engineering's role in SNM accountability to track the material. The second reason was for

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( 15 i monetary issues associated with warehouse inventor During

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discussions with the licensee, they expressed that they were going to propose a procedure change to the Material Control Procedures which L would enhance and provide guidance in the procedure specifying Material and Control Services responsibility and accountability for SNM. The licensee had changed their warehouse policy to issue all items when leaving the warehouse as part of the corrective actions for CR 98-0133 c. Conclusions The inspector found that the licensee had maintained SNM accountability for an LPRM in question. The inspector found that the licensee's warehouse inventory had an error in the number of LPRM units on hand due to not having well defined procedural guidance on SNM in their material control procedur E7 Quality Assurance in Engineering Activities E7.1 Desian Control a. Insoection Scone (37551)

The inspector reviewed the status and condition of the equipment database system (EDBS) to evaluate the extent of database errors, problems, and corrective actions being taken by the licensee to correct and improve the system. The inspector also reviewed the status of the design control group and recent Nuclear Assessment Section (NAS) finding concerning UFSAR update l b. Observations and Findinos EQ3BS On June 11. 1998. the inspector initiated an inspection of the EDBS program based on two recent CRs that identified two specific problem One of the problems was that the EDBS program was not providing the correct physical locations of electrical panels in the plant. This type of information was considered as data stored in type B data fields specified by Nuclear Generation Group Standard Procedure. EGR-NGGC-001 Equipment Database System Program. Rev. O. Section 3.7. Type B Data !

Fields, stated that this field contained " data utilized in making key '

decisions regarding plant operation or program requirements (configuration information)." The inspector notec that Section Purpose Statement. stated that this procedure " satisfies requirements of 10 CFR 50. Appendix B for type A and type B data fields." The second problem identified in the CRs was that the EDBS was being used in many plant operating practices and that incorrect data continued to affect important aspects of safe plant operations. This problem was identified in CR 98-01466. Incorrect EDBS Information, subsequent to a fire dril The operators attempted to use EDBS to obtain the locations of several

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lighting distribution panels to assist in fire fighting, but the

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l The ir 'ector reviewed the Corrective Action Program (CAP) database for l all c 1997 and 1998 to determine the extent of the errors and problems that had been identified. The inspector found, through a query on EDBS in the CAP database, that 92 CRs were generated. The inspector reviewed each of the CRs and found that 18 of them represented the general problems with the databas The inspector found that each CR corrective action addressed only the J particular problem. The inspector did not find a CR which determined a negative trend or intended to review the database to determine the extent of condition. Concerns with the corrective actions for the EBDS 4 program were identified as an Inspector Followup Item (IFI). This is l identified as IFI 50-325(324)/98-07-04 EDBS Program Corrective Action The inspector questioned the if any adverse trends had been identified by the licensee with the EDBS database and what if any im]rovements were in process to upgrade the program based on the problems tlat were indicated. The licensee indicated that extensive auditing and corrections had been made to the type A data field but that nothing was planned nor were the resources available to do the same with the type B data field. CR 98-00087 EDBS Field Verification, specifically addressed the type A data field for verification of accuracy of the information, action item 3 only assigned the type A data fiel The inspector reviewed the licensee's response to NRC 10 CFR 50.5 A letter concerning the adequacy and availability of design basis information. The inspector noted that the response discusses the use of EDBS as an item used to govern the control of structuring system, and component configuratio Desian Control The inspector reviewed the engineering organization chart of July i 1998. The design control superintendent, design review supervisor, and five engineer positions were vacant. The establishment of the design control unit under the corporate office was done to address issues such as maintaining the UFSAR up to date and other issues discussed in the NRC 10 CFR CO.54.f lette However, several key positions have been vacant for .nth UFSAR Uodates The inspector read NAS assessment re> ort 98-07 One of the issues was that the administration of pending clanges and interface between the UFSAR and other station processes was not adequate to ensure the accuracy of the UFSAR. The inspector noted that the licensee had recently completed a comprehensive review of the UFSAR for correction of i errors. This review identified over 200 discrepancies in the UFSA However, this NAS finding indicates that the effort to keep the UFSAR

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accurate since the audit needs attention.

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17 i Conclusions

! The licensee acknowledged that design control activities warranted L additional attention. These activities were accuracy of EDBS. staffing l of design control organization, and UFSAR update E8 Miscellaneous Engineering Issues (92903)

!- E8.1 (Closed) Violation VIO 50-325(324)96-15-06: Repeat Failure to take y Adequate Corrective Actions for Chlorine Detector Failures i

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This violation was reviewed in irs 50-325(324)/97-08 and 97-09. The

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t licensee responded to this violation on December.20. 1996. A l supplemental reply was received on February 14. 1997. An update to the I supplemental reply was received August 1.1997, indicating that the system reliability had been improved. Three previous Licensee Event Reports (LERs) and a violation were closed in IR 97-08. This item

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remained open pending improvement in reliability of the detectors. In i

, IR 50-325(324)/97-09, a modification that relocated the detector out of a high air velocity region to behind a filter was discussed. The inspector reviewed the detector performance with the system enginee Over. the past year, during monthly tests, no failures were found. This item is closed.

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.E8.2 LClosed) Violation VIO 50-325(324)/97-02-07: Failure to Initiate CR for HPCI Valve -Time Discrepancy This violation was for failure to initiate a CR once a discrepancy was

. identified with the UFSAR. The licensee responded to this violation on L June 26, 1997 admitting the violation. The Unit 1 HPCI minimum flow bypass to the suppression pool allowable stroke time was 10 second The licensee had identified a change to make the UFSAR and maximum inservice test program allowable time of 20 seconds. The UFSAR discrepancy was identified in April 1996. No CR was initiated at the time of discovery as required by plant procedure. On February 13, 1997, when'the. valve stroke time was found to be 10.04 seconds. HPCI was declared inoperable and the valve motor replaced. This would not have been necessary had the change been processed to allow a nominal 10 seconds with 20 seconds maximum valve stroke time. The inspector reviewed the licensee's corrective action regarding counseling of

. individuals and training of personnel. This item is close E8.3 (Closed) Licensee Event Reoort LER 50-324/97-04-00. 01: Safety Relief Valves Exceeded Technical Specification Setpoints Limits l This LER documented the test results of 11 safety relief valves (SRVs)

removed from Unit 2 during the fall of 1997 refueling outage. The testing at a laboratory indicated five of 11 valves were found to lift L outside the + or - 1 percent TS 3.4.2 set)oint limit that existed at the time the valves were installed. On Novem3er 1. 1996, the TS limit was

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amended to 3 percent. Two of the valves exceeded the setpoint limit l of 1. percent but not 3 percent. Of the three valves below the l l

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! set)oint only one exceeded the three percent at -3.2 percent. The most pro)able cause of the setpoint drift was oxide accumulation on the pilot disc seat which resulted in a small increase in the effective pilot disc seating area.

i Historically. SRV lift setpoints drift high primarily due to the

! formation of an oxide bond between the pilot seat and disc. Past LERs l were 1-95-007. 2-96-002, and 1-96-01 To resolve this issue. a l

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modification of the Unit 1 and 2 pilot disc surface was implemente The modification involved platinum particle deposition on the SRV pilot disc surface by means of ion beam implementatio The 11 SRVs removed l during the fall 1997 refueling outage were equipped with this new l material . This modification with the increase in setpoint drift to + or

!~ - 3 percent was expected to resolve this issue. This appeared to be effective except for the one valve at -3.2 percent.

! However, during the spring of 1998. refueling outage for Unit 1. 4 of the 11 valves removed tested outside the + or - 3 percent TS limit These facts were reported in LER 50-325/98-003. An evaluation of the test results is expected to be submitted by August 14. 1998. Although the TS limits were exceeded, the condition is bounded by a vendor safety l analysis. Accordingly LER 50-324/97-04-00. 01 is closed. Conti1ued l evaluation of this generic industry problem continues under LER f>0-325/

98-003.

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E8.4 (Closed) Unresolved Item URI 50-325(324)/97-11-05: Use of Technical l Support Memorandums i

This item concerned the use of Technical Sup) ort Memorand ms (TSMs)

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although the process was deleted in 1992. T1e licensee issued CR 97-03353. Procedures Referencing TSMs. to document and resolve this issu This problem was noted because an auxiliary operator daily check sheet referenced a TSM for control of temperatures in reactor building. The inspector questioned the n eference to a TSM that was no longer active.

, The licensee conducted a word search of procedures and identified

! approximately 90 references. Procedure revisions were initiated to delete these references. Four items were questioned by the inspector

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, concerning configuration changes. The licensee added a specific action

! to CR 97-03353 to review these four items This was completed June 18.

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1998. The licensee determined that none of the TSMs involved changes in engineering records or design information. The inspector reviewed this action item with the licensee. Based on the actions taken to resolve the reference and use of TSMs this item was close E8.5 (Closed) Licensee Event Report LERs 50-324/96-03-00. 01. 02: Operation in Excess of Maximum Power Level Specified in Operating Licens (Closed) Licensee Event Reoort LERs 50-325/97-05-00. 01: Feedwater Flow Indication Discrepancy (Closed) Violation VIO 50-324/96-442-1013: Failure to Operate the Facility Within License Thermal Power Limit L__________________ _ _ _ _ _ - - _ _ _ _ _ . _ _ _ _ _ J

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.LClosed) Violation VIO 50-324/96-442-2013: Failure to Maintain Average Planar Linear Heat Generation Rate Within Approved Limits The licensee responded to this violation on January 13. 1997, admitting the violation. The licensee corrected the process computer database for feedwater temperature compensation that resulted in the error for calculation of thermal power for Unit 2. To provide a more accurate determination of indicated thermal power, the licensee inspected the feedwater flow elements in each unit and documented these results in LERs. This data was analyzed in detail along with past calibration methods. The details were provided in the LERs and supporting licensee root cause analysis and ESRs. The final results indicated each unit had exceeded the thermal power limit, but the errors were less than 2 percent. Although violations, the margin of error was bounded by the uncertainty used in the transient analysis. The inspector reviewed the licensee's closure Jackage for these items. The actions taken were thorough and compre1ensiv These items are close I Plant Sucoort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Radiological Control and Exposure a. Insoection Scoce (83750)

The inspectors observed radiological posting of areas, labeling of material and radiological surveys in clean areas. The inspectors also had discussions with workers who were entering posted areas and collecting radioactive material and reviewed lesson plans for radiation worker training and licensee procedures. Requirements and guidance were provided in:

. Title 10 CFR 20 Subpart F - Surveys and Monitoring

• 10 CFR 20 Subpart J - Precautionary Procedures

. NRC Circular No. 81-07 Control of Radioactively Contaminated Material

• Updated Final Safety Analysis Report Chapter 12. Radiation Protection

. Lesson Plans for General Employees and Radiation Control Technicians Training Observations and Findinas Radiological Postina. Labelina and Control of Radioactive Materials Radioactive Material Areas. Contamination Areas. Radiation Areas, High Radiation Areas and Locked High Radiation Areas were appropriately posted and consistent with the training presented to licensee workers that enter those areas and regulatory requirements. During a plant tour, the inspectors observed that radioactive material had been appropriately stored in the Reactor and Radwaste Building. Radwaste t

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Container Interim Storage Facility, and outside storage areas. The ins)ectors verified a sample of posted Locked High Radiation doors were locce The licensee had previously identified deficiencies in control of radioactive material and labeling by documenting 20 CR as of July . The inspectors observed radiation control technicians surveying and labeling radioactive material collected from posted Contaminated l Areas by workers. During tours the inspectors verified that radioactive i material was properly labeled consistent with NRC requirements and licensee procedures and that workers were aware of hazard communicated by the " Caution Radioactive Material" label The inspectors toured break areas outside of the radiologically I controlled area in the Clean Maintenance Sho). Service Buildin l Operations and Maintenance Building and Warelouse with a radiation  !

control technician and requested smears be taken at sclocted locations to assess the effectiveness of contamination control measures. All smears were determined to be within established plant limits. During the l tours the inspectors observed personnel modesty clothing (scrubs), on a chair located in the corner of the Operations and Maintenance Lunch ,

Roo Scrubs may be used as part of the protective clothing for i entering contamination area The scrubs were removed to the radiological control exit and counted in a Small Articles Monitor (SAM). The SAM was capable of detecting gamma contamination and was set to alarm high if 5.000 disintegrations per minute (dpm) were detected. The scrubs (two tops with a vendor's marking) resulted in a reading of about 176.000 dpm on two different SAM monitors. The scrubs were frisked with a CM7A and a RM-14 contamination monitors, and measured less than 100 net counts per minute. A qualitative gamma spectroscopy analysis identified Mn-54. Co-58. Co-60 and Cs-137 isotopes and a total activity of approximately 3.12 E-02 microcurie per cubic centimeter (uCi/cc). The radiation control technician followed up with a thorough smear survey of the lunch room where the scrubs were discovered and determined all smears were less than 200 dpm/100 cm3 and within established plant limits for clean areas outside of the radiological control are The licensee initiated CR BNP 98-01779 and included this as another example of radioactive material found outside the radiological control area that was originally documented in CR 98-0038 CR 98 00383 was not closed at the time of this inspectio The inspectors reviewed the Contaminated Square Footage Data. At the l time of the inspection, the licensee was tracking approximately 41.943

! square feet (sq. ft.), or approximately 6 percent of the accessible area (699.669 sq. ft.) as contaminated. The licensee had established the contaminated square footage target of approximately 7 percent (48,976 sq. ft.).

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Radiation Work Permit and Radiation Worker Practices

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The inspectors reviewed selected Radiation Work Permits (RWPs) for adequacy of the radiation protection requirements based on work scope, location, and condition For the RWPs reviewed. the inspectors noted that appropriate )rotective clothing, and dosimetry were indicate During tours of t1e plant, the inspectors observed the adherence of plant workers to the RWP requirements. The inspector observed that personal dosimetry was being worn in the appropriate locatio Personnel Dosimetry. Exoosure Controls. and the ALARA Procram The inspectors toured the health physics facilities. Reactor Building (including the refueling floor). Turbine Building, outside radioactive material storage areas (RMSAs), radwaste processing area, and the Unit 1 DW health physics control point. From record review, the inspectors determined that the licensee was tracking and trending personnel contamination events (PCEs). As of July 15. 1998, the licensee had tracked approximately 145 PCEs which included buth skin and clothing contaminations. In addition, the licensee was also tracking 78 evaluated risk personnel contamination events, which were contamination events where dose was assigned. Of the tracked PCEs. approximately 30 occurred in' designated clean area The inspectors took independent smears to assess contamination control in the Reactor Building. DW control point. Turbine Building. Radwaste processing area radioactive material storage areas, storage areas, and on the refueling floor. All smears were counted and determined to be  !

less than 1000 dpm/100 cm l The inspectors observed workers properly using friskers at the exit location from the controlled areas. The inspectors also observed workers properly exiting the protected area through the exit portal The inspectors selectively reviewed the whole body counting program 3rocedure. Radiological Control Instruction (RCI-8)- Bioassay Progra Rev.13. DOS-NGGC-0020 - Whole Body Counter (WBC) System Calibration, i Rev. 3 and the Whole Body Count Calibration record The daily calibration checks were performed as required. The WBC system met Minimum Detectable Activity values. The ins]ectors determined that the licensee was following the requirements of tie reviewed procedures.

c. Conclusions l The licensee met requirements for area posting and labeling of l radioactive material. Contamination control was effective in containing I

levels of contamination to appropriate areas. Workers were complying with established radiological work practices. The Whole Body Counter had been properly calibrated and personnel were implementing the program l in accordance with established procedure L_-____________ __

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l 22 R Status of RP&C Facility and Equipment

R2.1 Radiological Survey Instrumentation and Breathina Air Bottles Insoection Scoce (83750)

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The inspectors reviewed radiological survey instrumentation calibration practices and records', and breathing air bottles. Requirements were in 10 CFR 20.1501. 20.1602. 20.2103, and Pa'ct 20 Subpart Observations and Findinas

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The inspectors toured the radiological instrumentation calibration area and observed that all radioactive sources were maintained in a locked y room, which was posted as a very high radiation area in accordance with l 10 CFR 20.1602. The key used to gain entry to the room was under the l control of the health physics technician at the radiation controlled l area (RCA) access control point.

! The inspectors reviewed procedure OE&RC-0320. Calibration and Use of Eberline RM-14 Radiation Monitor, Rev. 19. dated September 27, 1998, and observed a radiation control (RC) technician calibrate an Eberline RM-14 Radiation Monitor. The RC technician calibrated the instrument in l accordance with procedure OE&RC-0320 and the Vendor Technical Manual.

l- The inspectors reviewed the training requirements and training records i

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of the RC technicians assigned to calibrate instruments and determined '

i that the technicians met the training requirements to perform the calibration. The inspectors also reviewed a sample of the calibration

and source check records for the RM-14 Eberline Radiation Monitors used f .at the site'and found the instruments to be properly calibrated and l source checked in accordance with the procedure OE&RC-0320 and the

! requirements set forth in 10 CFR 20.1501. The inspectors also observed L

that calibration records were retained for a period of at least 3 years l as required by 10 CFR 20.2103.

i The inspectors reviewed the licensee's records for qualifying breathing l air as Grade D. The inspectors examined breathing air bottles ready for i

use for physical integrity, current calibration, and gauge indication The breathing air bottles observed were being maintained in satisfactory condition.

I Conclusions

! The licensee maintained adequate control of radioactive sources. The L

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licensee's procedure and practices for calibrating the Eberline RM-14 l Radiation Monitor met regulatory requirements. Breathing air met Grade !

D or better quality requirements anc breathing air bottles were 1 maintained in a satisfactory condition.

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R3 RP&C Procedures and Documentation R3.1 Radiation Control Procedure Revision a-. Insoection'Scoce'(83750)

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The inspectors examined selected procedural changes and the licensee's procedural-~ review proces Requirements were in-10 CFR 50.59 and 10 CFR- -

50.71(e). j

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, Observations and Findinas The inspectors reviewed the last five revisions (Revisions 26, 27, . 29. and 30) of procedure OE&RC-0215. Release of Materials From the L Radiation Control Area, and found each revision adequately met the l requirements, as committed to in the UFSAR section 12.5.2.2.

l Instrumentation. The licensee performed either an Editorial Review or a

Technical Review of each revision of.the procedure. The inspectors ( determined that the last five revisions of procedure OE&RC-0215 did'not

! involve a change to the Technical-Specifications or an unreviewed safety f

question. The licensee had updated the FSAR to include the procedure changes.as required by 10 CFR 50.7)(e).

I l s Conclusions

, The inspectors determined that procedure OE&RC-0215 adequately met the l-

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requirements in the UFSAR section 12.5.2.2 Instrumentation. Also, the '

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licensee's review of procedural changes was in accordance with 10 CFR ,

50.5 i R5 . Staff Training and Qualification in RP&C R5.1 Qualification Requirements L Insoection Scooe (83750) l The inspectors evaluated the qualifications of the Environmental and 1

' Radiation Control (E&RC) management staff and selected technicians to '

determine if all qualification requirements were met in accordance with  !

, Technical Specifications 6.3.1. Also evaluated was the training of the i

RC technicians to. determine'if they met the training requirements in

' Technical Specifications 6.4 Trainin : Observations and Findinas Technical Specifications 6.3.1 states "Each member of the facility staff shall meet or. exceed the minimum qualification of. ANSI N18.1-1971 for comparable positions, except for (1) the Manager - Environmental &

Radiation Control who .shall meet or exceed the qualifications of  :

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Regulatory Guide 1.8 September 1975 and..." The inspectors observed ~

through a review of the qualification records that the-E&RC Manager (Radiation Protection. Manager), Radiation Protection Superintenden l

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l other E&RC management staff, and the RC technicians met the qualification requirements in Technical Specifications 6.3.1. The inspectors also cetermined through a review of the training requirements and training records'that the RC technicians had been properly trained

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to )erform their assigned duties as committed to in the licensee's Tec1nical Specifications . Conclusions I

i The inspectors concluded that the E&RC Manager, Radiation Protection

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Superintendent, other E&RC management staff, and RC technicians met the qualification requirements of Techrical Specification 6.3.1. Also, the l RC technicians had been pro)erly trained for the duties they had been L assigned as required by Tec1nical Specifications 6.4.

i- R8 Miscellaneous RP&C Issues (92904)

i R (Closed) Violation 50-325(324)/97-11-08: Contract Employee in Radiation Controlled Area Without Proper Dosimetry The licensee responded to this violation in a letter dated November 26.

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' 1997 which stated that measures would be evaluated to reduce the probability that an individual would enter the RCA without a dosimeter.

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The inspector found that this evaluation implemented one corrective measure. The licensee placed large signs at RCA entry points asking if they remembered their electronic dosimeters. A root cause investigation was performed and all action items complete. The inspector found that

! no similar events have occurred since this event. The inspector I

concluded that the licensees corrective actions were appropriate, therefore this item is close R8.2 (Closed) LER 50-325(324)/96-009: Solid Radwaste UFSAR Discrepancies The wet solid radioactive waste processing system had operated without a r documented safety evaluation (10 CFR 50.59). The inspectors verified

that the licensee UFSAR ( Rev.15) included changes to reflect the actual plant configuration for use of the Radwaste Processing System and that a safety evaluation (ESR 96-00500 Rev 0) had been performed which included a structural / seismic evaluation of the containers enclosing the High Integrity Containers. The inspectors walk down verified completion of upgrades enclosing the processing area, which provided a high efficiency particulate absorber unit and that sufficient curbing to contain spills was in place. This LER is close F7 Quality Assurance in Fire Protection F7.1 Nuclear Assessment Oversiaht Insoection Scooe (71707. 71750. 40500)

The inspector attended meetings and reviewed findings from several NAS audits, reviews and. CRs

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b. Observations and Findinos The inspector reviewed the findings present during several NAS audits, reviews and CRs. The inspector found that the audits, in general. were

" adec uat The findings for the triennial fire protection audit B-FP-98-01 demonstrated a thorough program review. The findings were technical I in-nature and indicated extensive subject matter and regulatory knowledge by the auditors.

L The audit determined the fire protection program to be ineffectiv !

! however the audit indicated that the fire brigade was effective with a s -

declining trend. The audit identified seven potential issues and four weaknesses. Three issues not previously identified or observed by the inspector were identified. These issues included deficiencies in the i fire brigade training program, conduct of fire drills, and the failure l to implement changes to the fire protection program as indicated in a '

l recent UFSAR change. Weaknesses were identified to still exist in the I I

control of transient combustible material, procedure adequacy and

performance, and documentation of deviations from National Fire j

! Protection Association code The inspector discussed the fire protection audit with the NAS manager.

l The inspector indicated that many of the findings including th determination that the program was ineffective appeared to be repetitive i

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of the 1997 assessment (B-FP-97-01), and questioned the adequacy of the l corrective actions completed for the 1997 NAS audit findings. The l licensee initiated CR 98-1631, NAS Fire Protection Follow-up. As a '

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result, additional follow-up assessments were planned to independently review progress in restoring the effectiveness of the fire program, c. Conclusions I The. triennial fire protection audit was determined to be a thorough review of the program. Audit findings reflected extensive auditor knowledge of. fire protection and the associated requirement Additional assessments were scheduled to independently review corrective actions for the fire protection progra Manaaement Meetinas XI: Exit Meeting Summary The inspector presented the inspection results to members of licensee management at the conclusion of the inspection on July 24, 1998. The l

licensee acknowledged the findings presented.

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. ' PARTIAL LIST OF PERSONS CONTACTED License '

A. Brittain. Manager Security

~ Mi Christinziano. ManagerL Environmental and Radiation Control

- W. Dorman. Supervisor Licensing'and Regulatory Programs l N. Gannon. Manager-Maintenance l J. Gawron Manager Nuclear Assessment l

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'K. Jury.1 Manager Regulatory Affairs

"J.'Keenan. Vice President. Brunswick Steam Electric Plant i .B. Lindgren' Manager Site-Support Services .

L ?J. Lyash. Brunswick Engineering Support Section.

L G. Miller. Manager Brunswick Engineering Support Section R. Mullis. Manager Operations-L

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Other licensee employees or contracts included office, operation, maintenanc chemistry, radiation. and corporate personne .

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'NRC -

.

K. Barr l

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.

ja

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n ,

s l

I e

n

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o ll. ..

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INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 40500: Effectiveness of Licensee Controls in Identifying. Resolving, and Preventing Problems IP 61726: Surveillance Observations IP 62707: Maintenance Observations

, IP 71707: Plant Operations Plant Support Activities

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IP 71750:

IP 83750: Occupational Radiation Exposure IP 93902: Followup - Maintenance IP 92903: Followup - Engineering IP 92904: Followup - Plant Support ITEMS OPENED, CLOSED, AND DISCUSSED ,

l Ooened 1 50-325(324)/98-07-01 URI Elevated Temperatures (paragraph 02.1)

50-325(324)/98-07-02 VIO Configuration Control Problems (paragraph M4.1)

50-325(324)/98-07-03 VIO Diesel Generator Relay Failures (paragraph E2.1)

50-325(324)/98-07-04 IFI Equipment Database System Program Corrective Actions (paragraph E7.1)

Closed

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50-325(324)/97-11-03 VIO Pen and Ink Changes to I&C Procedure (paragraph M8.1)

50-324/97-11-06 VIO Rigging to Safety Equipment Without a Safety Evaluation (paragraph M8.2)

50-325(324)/96-08-03 DEV Failure to Complete Maintenance Procedure Backlog (paragraph M8.3)

50-325(324)/98-06-03 URI Configuration Control of Maintenance and Test Activities (paragraph M8.4)

50-325(324)/96-15-06 VIO Repeat Failure to Take Adequate Corrective Actions for Chlorine Detector Failures (paragraph E8.1)

50-325(324)/97-02-07 VIO Failure to Initiate CR for HPCI Valve Time Discrepancy (paragraph E8.2)

l 50-324/97-004-00 LER Safety Relief Valves Exceeded Technical Specification Setpoint Limits (paragraph E8.3)

e

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50-324/97-004-01 LER Safety Relief Valves Exceeded Technical Specification Setpoint Limits (paragraph E8.3)

50-325(324)/97-11-05 URI Use of Technical Support Memorandums (paragraph E8.4)

50-324/96-003-00 LER Operation in Excess of Maximum Power Level Specified in Operating License (paragraph E8.5)

50-324/96-003-01 LER Operation in Excess of Maximum Power Level Specified in Operating License (paragraph E8.5)

50-324/96-003-02 LER Operation in Excess of Maximum Power Level Specified in Operating License (paragraph E8.5)

50-325/97-005-00 LER Feedwater Flow Indication Discrepancy (paragraph E8.5)

50-325/97-005-01 LER Feedwater Flow Indication Discrepancy (paragraph E8.5)

50-324/96-442-1013 VIO Failure to Operate the Facility Within License Thermal Power Limit (paragraph E8.5)

50-324/96-442-2013 VIO Failure to Maintain Average Planar Linear Heat Generation Rate Within Approved Limits (paragraph E8.5)

50-325(324)/97-11-08 VIO Contract Employee in Radiation Controlled Area Without Proper Dcsimetry (paragraph R8.1)

50-325(324)/96-009-00 LER Solid Radwaste UFSAR Discrepancies (Paragraph R8.2)

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