IR 05000324/1997013

Insp Repts 50-324/97-13 & 50-325/97-13 on 971109-1227. Violations Noted.Major Areas Inspected:Operations, Engineering,Maint & Plant Support.Includes Results of Maint, Engineering & FP Insps

ML20199G750
Person / Time
Site:	Brunswick
Issue date:	01/23/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20199G672	List: IR 07100109/2012027 ML20199G667 ML20199G750
References
50-324-97-13, 50-325-97-13, NUDOCS 9802040338
Download: ML20199G750 (50)

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U. S. NUCLFAR REGULATORY COMMISSION REGION 11 Docket Nos: 50-325, 50 324 license Nos: DPR 71. DPR 62 Report No: 50-325/97-13. 50-324/97 13 Licensee: Carolina Power & Light (CP&L)

Facility: Brunswick Steam Electric Plant, Units 1 & 2 Location: 8470 River Road SE Southport, NC 28461 Dates: November 9 - December 27, 1997 Inspectors: C. Patterson Senior Resident Inspector E. Brown Resident inspector G. Guthrie, inspector in Training J. Coley Reactor inspector (M1.3. M8.6)

J. Lendhan. Reactor Inspector (E1.1. E1.4. E5.1. E E8.4. E8.5)

C. Doutt. Senior Instrumentation and Controls Engineer. Office of Nuclear Reactor Regulation (E1.1. E1.2. El.3)

G. Wiseman. Reactor Inspection (F2.1. F2.2. F2.3, F3.1. F5.1 F6.1. F7.1)

Approved by: M. Shymlock. Chief. Projects Branch 4 Division of Reactor Projects 9802040330 900123 PDR G ADOCK 05000324 PDR Enclosure 2

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EXECUTIVE SUMMARY Brunswick Steam Electric Plant. Units 1 & 2 NRC Inspection Report 50 325/97 13. 50-324/97-13 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 resu'ts of maintenance, engineering, and fire protection ir,pections by regional and headquarters inspector Operations e The inspector concluded that u.e cold weather program has been satisfactorily implemented. Adequate contingency plans and operator checks for proper operation of the systems were noted in the procedure Section 01.1).

• The inspector concluded. from a safety system walkdown, that the Containment Atmospheric Dilution system was being maintained as designed (Section 02.1).

• The clearance reviewed was prepared. authorized, and implemented in accordance with procedure (Section 02.2),

e The inspector concluded that the Plant Nuclear Safety Committee meeting provided an effective review of Unit I readiness for restart (Section 07.1).

e Inspe.; tor review determined that clearance records were not retained in accorcance with Technical Specifications (TS). The failure to maintain clearance records in accordance with TS was a violation (Section 07.2).

• The control of a short duration mid-cycle o:tage was excellent (Section 07.3).

• Licensee investigation determined that removal of the IB Reactor feedwater Pump at too high a power level caused larger than expected level transients. These transients combined with the improper functioning of the level contacts in the Reactor Recirculation Run back logic circuitry, resulted in the November 5-6. 199/ run backs (Section 08.3).

• The inspector concluded that the licensee's control of the 2C and 20 electrical bus maintenance was weak because they did not recognize DG in oberabilityconditionsduringtheimplementationoftt.eirclearance ( ection 08.4).

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2 Maintenance e Movement of the spent fuel shi) ping cask was perforrxo in accordance with methodology approved by t1e NRC in a letter dated December 2, 199 Adequate supervisory oversight was present during movement of the cask (Section M1.1).

• The inspector observed performance of calibration of two Reactor Core Isolation Cooling (RCIC) pressure switches. The work activities were completed without any identified questions or concerns (Section M1.2).

• Maintenance activities observed relating to equipmert qualification of electrical equipment were found to be conducted in a thorough and effective manner (Section M1.3).

. A violation was identified for a preventive maintenance procedure not indicating specific E0 requirements. This omission resulted in deficient Nelson flame seals in motor control centers not being detected during scheduled preventive maintenance activities (Section M1.3).

• The licensee continues to struggle with proper dispositioning of abnormal indications. The failure to maintain the Daily Surveillance Report in accordance with procedure was a violation. Abnormal values observed fer the Steam Jet Air Ejector radiation monitor and subsequent test indicated potential fuel failure for Unit 1 (Section M3.1),

• The licensee identified that the Unit 2 Core Spiay sparger differential alarm setpoints were outside of the TS allowable range. The cauce was attributed to voiding of the sparger nozzles similar to the phenomenon identified previously on Unit 1. The alarm setpoints were adjusted and the associated documentation was updated (Section M8.5).

-Engineerino >

+ An additional example of a violation was identified for an inadequate procedure for the conduct of E0 maintenance (Section E1.4). Two inspector followup items were identified to review revisions to instrument setpoint procedures and to review terminal block leakage current evaluations (Section El.1 and Section E1.4).

• A weakness was identified regarding a procedure reference to a drawing for accident temperature data which was not available for use and wording inconsistencies in the procedure (Section E1.1).

• The licensee was making progress in resolution of the technical issues and closure of CRs and JCOs (Section E1.4). The licensee training and qualification for E0 personnel meets NRC requirements (Section E5.1).

Instrument setpoint calculations were technica ly adequate and complied with NRC requirements (Section E1.2).

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Plant Support

. The ins)ector determined that each of the locked high radiation area doors w11ch were checked were locke lhe ins)ector concluded that the licensee is satisfactorily controlling locked ligh radiation areas in the plant (Section Rl.1).

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The inspector determined that several poor radiological work practices existed in a radioactive material storage area (Section Rl.2).

The inspector found the status and condition of the protected area fence i to be satisfactory (Section S2.1).

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Corrective maintenance on degraded fire protection systems was i

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accomplished in a timely manner. The maintenance and material condition of the fire protection equipment and features were satisfactory (Section F1.1).

. The inspector concluded that silicone foam penetration seal field verificction documentation was maintained by the licensee. The inst 311ation and repair procedures for penetration seals provided adequate guidance to ensure that materials were installed per design requirements. However, the designs were not supported by seal testing documentation, vendor data and inspection criteria, installer qualification and training records, and engineering evaluations that satisfy the guidance of Generic Letter 86-10 for deviations from the fire barrier configuration qualified by tests (Section F2.2).

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The inspector concluded that fire door surveillance procedures and acceptance criteria for verification of fire door clearances were in accordance with National Fire Protection Association (NFPA) guidanc However, an updated Final Safety Analysis Report (UFSAR) discrepancy associated documentation of fire door and frame evaluations was identified (Section F2.3).

. General housekeeping was satisfactory. Fire retardant plast.ic sheating and film materials were being used. Lubricants and oils were properly stored in approved safety containers. Controls for combustible gas bulk storage and cutting and welding operations were being enforce Controls were being properly maintained for limiting transient combustibles in designated separation zones and other restricted plant areas (Section F3.1).

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The fire brigade organization and qualification training .act the requirements of the site Procedures. Fire brigade turnout gear and fire fighting equipment were being properly maintained (Section F5.1).

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The coordination and oversight of the tacility's fire protection program had been reassigned from the previous Loss Prevention Unit organization to shift. Operations. The new organizat.onal structure met NRC guidelines and the licensee's fire protection program requirements (Section F6.1). .

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. The 1997 Nuclear Assessment Section assessment of the facility's fire protection program was comprehensive and was effective in identifying fire protection program performance deficiencies to management. Planned corrective actions in response tc the audit issues were substantial and included a fire p.'otection reorganization (Section F7.1).

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ReDort Details

~ Summary of Plant Status Unit I returned to power o)eration on November 14. 1997, following a mid-cycle outage that began on Novem)er 5. 1997, to remove leaking fuel assemblies. Two leaking fsel assemblies were identified and removed during the mid cycle outag However, indications of a potential fuel leaker remained after the unit returned to full power operation. At the end of the report period the unit had been on-line 42 day Unit 2 operated continuously during this report period. At the end of the report period the unit had been on-line continuously for 59 day Due to concerns about the control room dose, the licensee imposed an administrative limit on lodine until a Technical Specification (TS) amendment submitted was a) prove The licensee made a orocedure change to Administrative procedure 0Al-81. Water Chemistry Guidelines, setting the limit at 0.1 microcurie per gram dose equivalent L 'ine 131 compared to the TS value of 0.2 microcurie per gram. Also, the licet ;e has been providing weekly water chemistry data to NRR and the Resident Inspector for review. None of the data reviewed has exceeded the administrative limi Due to a reconstitution of the Environmental Qualification (EO) program and items identified, there are 12 of 24 Justification for Continued Operation (JCO) that remain open for both units. The following provides the status of the EQ JCOs and associated Engineering Service Requests (ESRs):

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1) ESR 97-00087. E0-Type JC0 for Improperly Configured Conduit Sea ) ESR 97-00574 Greyboot Connector ) ESR 97-00329 (old ESR 96-00625). E0 Type JC0 for EQ Fuses Without a Qualification Data Package (00P).

4) ESR 97-00289. Post A cident Sampling System (PASS) Valve Limit Switch Panel Wirin ) ESR 97-00238. JC0 for Standby Gas Treatment Motor Operated Valve (MOV) Position Indicator Rheosta ) ESR-97-00534. GE c' Type Terminal Strip ) ESR 97-00513. In-b Drywell Electrical Penetration ) ESR 97-00535. Target Rock Solenoids TB Spra ) ESR 97-00449, Degraded Junction Boxe ) ESR 97-00250. Conduit Union in EQ Boundar ) ESR 96-00425. Evaluation of E0 sealant ) ESR 97-00523. High Pressure Coolant Injection Auxiliary Oil Pump Motor Unit P10 13) ESR 97-00446. GE Radiation Detectors. closure date to be determined (TBD).

14) ESR 96-00503. Associated Circuit E0. closure date TB . _ _ _ _

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15) ESR 97-00330 (ola ESR 96 00501). Motor Control Center (MCC) E0 was closed by the licensee, but was reopened - closure date TB ) ESR 96-00426. Evaluation Quality class and E0 classification of PASS valves was scheduled for completion June 6, 1997. but closure date is TB ) ESR 97-00529. Failure of Unit 1 Drywell Motor, closure date TB ) ESR 96 00587 PASS Valves, closure date TB ) ESR 96 00627 ODP for Marathon 300 Terminal Blocks was scheduled for completion December 31, 1937 but revised to August 1. 1997, but closure date is now TB ) ESR 97-00229. JC0 for GE Condition Report (CR) 151 B Terminal Blocks was scheduled to be completed September 1, 1997, but closure date is now TB ) ESR 97-00256. Main Steam Insulation Valve Hiller Aci . tor JCO. was -

scheduled for completion September 2, 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 97-00435. MCC Fittings, closure date TB ) ESR 97-00602. Solenoid Valve Field Wiring, closure date TBD.

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In summary Unit I returned to power operation following completion of a mid-cycle outag Unit 2 o)erated continuously; however there were 12 outstanding JCOs in the E0 area for both unit I. Ooerations 01 Conduct of Operaticns 01.1 Cold Weather Preparation Insoection Scone (71714)

The inspector reviewed the licensee's cold weather program to determine whether it had been effectively implemente Observations and Findinas The inspector reviewed the licensee's cold weather 3rogram for adequacy and implementation by reviewing their Cold Weather 3111 and Freeze Protection Procedure. Operating Instruction 001-01.02: Fire Protection Procedure 0FPP-024. Freeze Protection of Fire Suppression System; and Preventive Maintenance Procedure OPM-HT001. Preventive Maintenance on Plant Freeze Protection and Heat Tracing. The inspector determined that the procedures were adequately implemented. Additionally, the procedures were adequately employed on multiple cold weather days. as observed by the inspecto The inspector conducted a walkdown of plant syn. , which were exposed to cold weather. Systems which were heat traced were observed for adequacy. The inspector looked for systems that did not have cold

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weather. heat trace installed. The inspector determined that the operation of the Makeup Water Tank system heat trace was not controlled by any procedure. The licensee stated that this heat trace system was being controlled b," operator knowledge only. The licensee initiated a procedure change request to place this heat trace system into their cold weather procedures. The inspector noted on the Unit 2 Condensate Storage Tank. High Pressure Coolant Injection (HPCI)/ Reactor Core Isolation ~ Cooling (RCIC) level switch vent line that a six inch portion of the lagging was missing at the top of the vent line and that the tin shielding was missing around the lagging at an elbow on the vent lin The lagging was wetted and degraded at the elbow. The inspector discussed these two items with the licensee. The licensee did not warrant these deficiencies as requiring corrective action. The inspector did not find other systems requiring heat trace that were not heat traced based on present system conditions and projected use of the systems observe Conclusions The inspector concluded that the cold weather program has been satisfactorily implemented. Adequate contingency plans and operator checks for proper operation of the systems were noted in the procedure Operational Status of Facilities and Equipment 02.1 Containment Atmosoheric Dilution (CAD) System Walkdown

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' Insoection Scope (71707)

On December 10. 1997, the inspector performed a walkdown of the CAD system in the Nitrogen and Off-Gas Services Buildin Observations and Findinos The CAD system is described in Updated Final Safety Analysis Report (UFSAR) Section 6.2.5. Combustible Gas Control in Containment. The CAD system provides long-term nitrogen makeup after a Loss of Coolant Accident (LOCA). This function is accomplished by vaporizing liquid nitrogen and feeding it into containment as required to maintain an oxygen concentration at or below five percent. The system is designed to Engineered Safety Feature (ESF) standards, all equipment for CAD service is designed with suitable redundancy and interconnections such that no single failure of an active component will render the system inoperable. This equipment includes one liquid nitrogen storage vesse two electric vaporizers, two flow-regulating stations. flow and temperature indicators. and appropriate redundant valves and interconnecting pipin The inspector traced the system piping in the Nitrogen and Off-Gas Services Building. The configuration was compared to plant drawing 0 02560. Containment Atmospheric Control System. The configuration was found to be like the plant drawing. The inspector observed an inch of

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frost on the outside of the piping insulation on both sides of valve HV-1 This valve is a manual isolation between the nitrogen tank and an 85 pound pressure regulating valv The inspector questioned why the frost was on the line. The licensee stated that the 90 pound relief valve setpoint was near the controlling pressure of the 85 pound regulator and some nitrogen was venting of The redundant pressure regulating valve was isolated and it's isolation valve (HV-12) was closed. The inspector questioned by keeping HV-12 closed, if the system was single failure proof. The licensee initiated CR 97-04128. CAD Tank Isolation Valve, to address this issue, The licensee concluded that no automatic action was required to address a LOC Manual alignment of the pressure regulator was acceptable since this was a long term post-LOCA actio Conclusions The inspector concluded, from a safety system walkdown, that the CAD system was being maintained as designe .2 Clearance Verification l Insoection Scoce (71707)

The inspector reviewed the tagout for the Unit 2 Residual Heat Removal

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(RHR) system to verify proper clearance preparation, authori7 n. and implementation, b. Observations and Findinas On December 10. 1997, the inspector performed verification of the proper alignment and tagging of clearance 2-97-1781 on the Unit 2 RHR Syste All accessible components were verified to-be in the proper position with the appropriate tags in place. The inspector reviewed Nuclear Generation Group Standard Procedure OPS-NGGC-1301. Equipment Clearanc The clearance package was adequately prepared, authorized, aad implemer.ted. The inspector subsequently verified proper clearance removal for those accessible component Conclusions The clearance reviewed was prepared, authorized, and implemented in accordance with procedure, i

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07 Quality Assurarm in Operations 07.1 Restart Plant Nuclear Safety Committee (PNSC) Insoection Scone (71707)

On November 11 and 12. 1997, the inspector attended the Unit 1 PNSC restart assessment following a mid-cycle outage to replace two leaking fuel assemblies, Observations and Findinos On November 11, 1997. PNSC was convened to review Unit I readiness for restart. The committee reviewed the fuel sipping results and core reloa Other maintenance activities during the outage were also reviewe The meeting was conducted in accordance with TS with attendance by all primary members, with no alternates. The meeting provided a thorough discussion of all agenda items. The PNSC Chairman concluded that the discussion of recirculation pump runbacks that occurred on November , during removal of the reactor feed pumps during the planned shutdown was not complet This item was statused as a restart constraint requiring another PNSC review prior to restart. Noteworthy in the review was the risk assessment review conducted for a failed Control Rod Drive (CRD) pump. During the mid-cycle outage one of the two CRD pump motors failed. The Probabilistic Safety Analysis (PSA)

person attended the comnittee meeting and presented the results from running the risk assessment model considering failure of both CRD Jump This risk was determined acceptable based on other TS required higi pressure injection sources such as HPCI and RCI On November 12. 1997, the inspector attended a second meeting. In this meeting discussion was held regarding the problem with run backs and it was concluded that this was due to a design deficiency that was already corrected and installed on Unit 2 and scheduled for Unit 1 at the time of the next refueling outage, Conclusions The inspector concluded that the PNSC meeting provided an effective-review of Unit I readiness for restar .2 Retention of Clearance Records Insoection Scope (71707)

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The inspector reviewed whether configuration management documents, specifically ciearances, were retained in accordance with TS 6.10. This specification requires that facility records be retained in accordance with the American National Standards Institute (ANSI) N45.2.9-1974 Collection. Storage, and Maintenance of Quality Assurance Record _

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6 Observations and Findinas During ins)ector review of clearance errors which resulted in damage to the Unit 23 recirculation pump seals, the licensee was unable to locate a clearance hung to facilitate repairs on the recirculation motor oil- -

cooler. Tha clearance. 2-97-1531. was hung _resulting in a configuration change for the B recirculation pump, but no maintenance on the system was performed. The clearance was removed from the field, thus restoring the system, and " rolled back" to allow use at a later dat Subsequently, a scheduler requested the clearance be deleted due to the repair activities being complete and approved without need for the clearance boundary. As a result of the deletion of the clearance, no record of the change in plant configuration was retaine The inspectoi : viewed TS 6.10. UFSAR Section 1.8. Regulatory Guide 1,88, and ANS1 N45.2.9-197 fhe inspector questioned the correctness of not retaining the clearance. Since a configuration change did occur despite the recirculation motor cooler activities not needing the cooler isolated. Nuclear Records Management Procedure ORMP-001. Indexing of Plant Records. defined those records required to be retained to satisfy the 0A requirements stated in ANSI N45.2.9-1974. Discussion with the licensee revealed that the records required to be retained did not include clearances. The inspector reviewed the Nuclear Generation Grou)

Standard Procedure OPS-NGGC-1301. Equipment Clearance, and the Brunswicc Required Records Lis Neither document required that clearances be retaine TS 6.10 requires facility records shall be maintained in accordance with ANSI N45.2.9-1974. ANSI N45.2.9-1974, in Section 3.2.7. Retention of Records. states that Appendix A to the standard defined the types of 0A records and the recommended retention periods. The failure to maintain data sheets or logs on equipment alignment consistent with ANSI N45.2.9-1974 is a violation. This violation is identified as VIO 50-325 (324)/97-13-01. Failure to Retain TS Required-0A Recor Conclusion Inspector review determined that clearance records were not retained in accordance with TS. The failure to maintain clearance records in accordance with TS was a violation.

07.3 Mid-Cycle Outaae (71707) Insoection Scope The inspector reviewed the mid-cycle outage activities to remove the leaking fuel assemblie Observations and Findinas Unit 1 was returned to power operation on November 14. 1997. This completed a mid-cycle outage in eight days. The unit was shutdow _-

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leaking fuel assemblies identified, removed, fuel reloaded and returned to power o)eration. This short duration outage was the quickest on record. T11s was accomplished with plant personnel without any major problems. This outage was planned and controlled similar to a regular refueling outag c. Conclusions The control of a short duration mid-cycle outage was excellen Miscellaneous Operations Issues (92700, 92901)

08.1 (Closed) Unresolved Item (URI) 50-325/96-15-01: Vessel Disassembly Without Secondary Containmen During a refueling outage, the reactor vessel head and steam dryer /separatorr assemblies were removed from the reactor vessel without secondary containment integrity (SCI) established. This issue was reviewed by the NRC Office of Nuclear Reactor Regulation. It was determined that the removal of the nead and assemblies without SCI established were not activities prohibited by TS 3.6.5.1. The potential

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by the NRC. However, maintenance of SCI curing vessel disassably was a logical extension of the defense-in-depth ap3 roach used in addressing the heavy loads issue and encouraged by the 4RC. The licensee's action in proceeding with vessel disassembly was not conservative. The licensee implemented controls during the Unit 2 refueling outage to maintain secondary containment operable during vessel disassembly. This issue was thoroughly evaluated as part of the licensee's Safe Shutdown Risk Management Assessmen .2 (Closed) Violation V10 50-325(324)/97-02-01: Locked Valve Out of Position The licensee's response to this violation was dated May 5, 1997, and was accepted by the NRC in a letter dated May 23. 1997. The corrective actions described in the response letter were verified as complete by the inspector. This violation is close .3 (Closed) URI 50-325/97-12-03: Recirculation Pumo Run backs On November 5. 1997, the licensee began a c0ntrolled shutdown for the Unit 1 forced outage in order to replace leaning fuel bundles. During the shutdown. Unit I received two recirculation pump run backs to the 45 percent limiter. During the second run back the five percent buffer region was entered and exited in accordance with procedures.

) Subsequently. no other transients or run backs were ercountered while removing the Reactor Feedwater Pumps (RFPs) from service. The licensee preliminarily attributed the first run back to a malfunction of the 1B discharge check valve causing diversion of the 1A RFP through the 1B discharge valve to the main condenser. The final analysis was provided in the root cause analysis for CR 97-3917. Unit 1 Plant Transients While

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Removing a Reactor Feed Pump from Service. The inspector reviewed the analysis and noted that the root cause attributed the run backs to the removal of the RFPs at too high of a power level and a design problem in the a) plication of the Metal-On-Silicon Field Effect Transistor (MOSFET)

switcl. The MOSFET was used in the 45 percent recirculation pump run back logic to indicate the below 182 inches reacter water level contact which is one of two contacts required to initiate the run bac Reactor water level perturbations are expected during the removal of the RFPs from service: however the magnitude of these perturbations seen for these events were outside of the operators expectations. The root cause analysis stated that removal of the RFP at 65 percent power was inappropriate in that 65 percent during this evolution has changed since power uprate. Before power uprate. RFPs were removed from service 3er 10P-32, Condensate and Feedwater System Operating Procedure, at or )elow 65 percent. Under current conditions 65 percent is approximately equivalent to 68 percent power pre-uprated power. The analysis attributed the magnitude of the perturbations to removal at too high of a power level. In addition, the licensee determined that when the first RFP was taken out of service, the less than 20 percent RFP flow contact for the 18 pump was made up and with the MOSFET improperly indicating below 182 inches water level the run backs were received. The design of the MOSFET causes the contact to not be able to properly position itself u'aon loss of the constant voltage supply. Therefore interruptions in tle voltage will cause the MOSFET contact to not function as designe The second Run back was also attributed to the MOSFE The licensee intends to replace the MOSFETs in the next Unit 1 outage, The inspector noted that the MOSFETs had already been replaced in Unit The licensee is reviewing plant operation to determine the appropriate power level for removal of the RFPs from service. Based on licensee satisfactory comaletion of the investigation into the cause for the multiple run bac(s on November 5-6, 1997 this item is close .4 (Closed) URI 50-325(324)/97-12-04: Diesel Doeration Low Voltace Auto Start Defeated The inspector reviewed the licensee's root cause investigation CR 97-03683, 4KV Bus 2C/2D Clearances. The licensee's investigation determined that the number 3 diesel generator (DG) undervoltage relay had been disabled in the same manrer as the number 4 DG during similar maintenance activities on different day The inspector verified that the licensee did not exceed TS action, limiting condition for operation, or time requirements for both electrical bus maintenance activities. The inspector found that, on October 9. 1997, the plant was under a TS action statement requirement per TS 3.8.2.1. to restore the inoperable bus to operable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The electrical

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bus was not restored, in this case, for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 58 minutes. This plant condition was not recognized as a problem until the root cause investigation was performed. The root cause investigation was found to

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be adequat The ins)ector concluded that the licensee *s control of the 2C and 2D electrical aus maintenance was weak because they did not recognize that the DG would be inoperable during the implementation of their clearance. This item is close II. Maintenance M1 Conduct of Maintenance M1.1 Spent Fuel Cask Movement a. Inspection Scooe (62707)

The inspector observed transfer of the spent fuel shipaing cask from th foot elevation to the transport v'hicle and from t1e transfer vehicle to the 117 foot elevation of the Unit 1 Reactor Buildin s b. Observations and Findinas On December 8. 1997, the inspector observed the removal of the spent fuel shipping cask, with fuel in the cask from the 117 foot to the 20 foot elevation in the Unit 1 Reactor Building. On December 15, 1997, the inspector observed shipping cask movement, without fuel in the cask, from the 20 foot elevation to the 117 foot elevation in the Unit 1 Reactor Building. During both evolutions the cask was transferred with the valve box covers removed while being moved by the non-single failure proof yoke. Approval for use of a non-single failure proof yoke for movement of the cask with the valve covers removed was granted to the-l licensee by the NRC in a letter dated December 2, 1997. Upon reaching the transfer vehicle on December 8. 1997. the cask was wiped down to reduce contami.1ation levels. During both movements the inspector noted that the area was adequately posted for the radiological conditions I present and i ealth pnysics personnel were present. The inspector noted that adequate maintenance supervisory oversight was present for both cask movement Subsequent surveys of the cask after removal from the Reactor Building revealed that the shipment exceeded required limits. This event was captured in CR 97-4161. S)ent Fuel Cask (IF-300). The cask was returned to the Reactor Building w1ere additional decontamination was conducte The licensee attributed the contamination levels seen to leaching of the contamination due to changing temperatures and weather condition c. Conclusions Movement of the spent fuel shi) ping cask was performed in accordance with methodology approved by t1e NRC in a letter dated December 2. 199 Adequate supervisory oversight was present during movement of the cas _ _a

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M1.2 RCIC Turbine Exhaust Diaphraam High Pressure Instrument Calibration Insoection Scoce (61726)

The inspector observed the performance of Maintenance Surveillance Test 2MST-RCIC230. RCIC Turbine Fxhaust Giaphragm High Pressure Instrument Channel Calibration, for the pressure switches 2-E51-PSH-N012A and 2-E51-PSH-N012 Observations and Findinas On December 24. 1997, with Unit 2 at 100 percent power the inspector observed the channel calibration for RCIC pressure switches 2 E51-PSH-N012A and 2-E51-PSH-N012 The inspector verified that duriug the performance of this channel calibration that HPCI and Automatic Depressurization System (ADS) were o)erable and that no othar work activities were being conducted whic1 could cause an inadvertent isolation. This test verified that, upon sensing of a high pressure condition between the t'arbine exhaust dia)hragms, an isolation signal is sent in accordance with TS 4.3.2.1 and Ta)les 3.3.2-2(4.b.6) and 4.3.2-1(4.b.6)

The inspector reviewed the work request / job order (WR/J0) AKNU 19 and the governing procedure 2MST-RCIC230. The procedure in use was verified to be the correct revision and the test instrumentation in use was within the allowable calibration duration. The inspector observed the

' procedure in use at all work locations and adequate communication was maintained throughout the test. The work observed was completed satisfactorily with no observed concern Conclusions The inspector observed performance of cal:uration of two RCIC pressure switches. The work activities were completed without any identified questions or concern M1.3 General Comments Insoection Scone (62700)

The inspector examined the following work activities involving EQ electrical equipment to verify maintenance implementation of EQ requirement *

WR/JO 97-ALVT-002 Verified Calibration of Unit 1 Loop B Residual Heat Removal (RHR) Service Water Pressure Switches Tag N SW-PS-1176 B and 1-SW-PS-11760

WR/JO 97-AGDR-002 Verified Calibration of Unit 1. Loop A. RHR Flow Transmitter (1-E11-FT-N015A). Converter (1-E11-FY-5119A). Square Root Converter (1-E11-FY-K600A)

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WR/JO 97-AAAS-002 Unit 2. Loop B. RHR Breaker Test in compartment DM 5 of GE IC 7700 Series MCC 2XB-2 Division Il Observations and findinos The above work was ,m cformed with the work packages present and in active use. Technicians were skillful, experienced, and knowledgeable of their assigned tasks. However, on December 10, 1997, while observing Instrumentation and Control (I&C) maintenance personnel perform work activities in accordance with WR/JO 97-AAAS-002, the inspector noted that one of the multiple cable electrical penetrations in the top of MCC 2-2XB-2 did not have Nelson flame guard putty on the inside surface as required by Maintenance Procedure OMMM 016. Environmental Qualification Maintenance Program. Revision 4. to properly seal the penetration. The inspector examined the putty installation on the top of the MCC cabinet for each of the penetrations and found the putty seal severely damaged on a second multiple cable penetration. In addition, cables were loose in both of the multiple cable penetrations. The applicable Environmental Otalification Data Package (ODP). ODP 67, requires missing or disturbed Nelson putty seals to be repaired or replaced. However, the PM procedure used to maintain and inspect the MCC's (PM Procedure OPM-MCC002. Revision 7. PM of GE Motor Control Centers and Switchboards)

did not have inspection requirements or acceptance criteria to ensure that putty seals were properiy sealing the cabinets. On September 17.

f 1997, a three-year PM conducted on MCC 2-2XB-2 would have identified l this discrepancy had procedure OPM-MCC002 included the acceptance criteria for the Nelson flame seal putty. A subsequent inspection performed on December 11. 1997 by the licensee, of 22 MCCs found an additional three MCC cabinet penetrations with damaged Nelson putty seals. In addition. 15 3ercent of the cables inspected in cabinet penetrations had putty w1ich appeared not to fully adhere to the cable in some areas. Failure of the procedure to implement E0 requirements for Nelson autty seals is identified as VIO 50-325(324)/97-13-0 Inadequate 3rocedure for the Conduct of E0 Preventive Maintenanc c. Conclusions Maintenance activities observed related to E0 of electrical equiament were found to be conducted in a thorough and effective manner, iowever, a violation was identified for a PM procedure not indicating specific E0 requirements. This omission resulted in deficient Nelson flame seals in MCCs not being dettcted during scheduled PM activitie M3 Maintenance Procedures and Documentation M3.1 Steam Jet Air Eiector Off-Gas Radiation Monitor increase Inspection Scoce (61726)

The inspector reviewed selected sections of Operating Instruction 101-03.1. Control Operator Daily Surveillance Report to ensure that i

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appropriate and prompt actions were taken to address abnormal TS surveillance values, b. Observations and Findinos On December 2. 1997. Unit 1 was in mode 1 at 100 percent powe The inspector reviewed the daily surveillance report as contained in Attachment 1 to 101-03.1 for November 30 through December 1. 199 The inspector noted that the values for the Steam Jet Air Ejector (SJAE)

i off-9as radiation monitors on aage 26 were between 1570 and 1780 millirem per hour (mR/hr) whici was greater than the T3/ Operating Limit

value of 1000 mR/hr. The SJAE off-gas radiation monitors provide for the detection of fuel element failures. The radiation levels are recorded in 101-03.1 to provide an indication whether SJAE off-gas radiation levels are approaching the alarm setpoint, which serves to ensure that dose rates for gaseous effluents do not exceed the limits l

prescribed in TS 3.11.2.1. Dose Rate.

l The inspector reviewed the associated procedures, work tickets, and discussed the abnormal values with the licensee. Step 4.2 c 'f 001-0 required the control operator to red circle all values wt are not within required limits. The inspector noted no indication on the attachment or in the operator logs that action had been taken or was expected to be performed to address the out-of-range values. Subsequent reviews of the daily log entries by the inspector indicated continual abnormal values and no red circle These failures were recorded in CR 97-4136. Daily Surveillance Repor The failure to red circle values not within required limits is a violatio This violation is identified as VIO 50-325/97 13-03. Failure to Note Abnormal TS Surveillance Value CR 97-4100. Questioned OG Data / Fuel Leak indicated that on December 3, 1997, a step increase of approximately 200 mR/hr was seen on the radiation monitor Subsequent sample results have shown an increase in the Sum of Six value ano changes in the fuel reliability index which are signs of potential fuel failure. In addition, the inspector noted that incorrect sensitivities were used during the November 25, 199 adjustment of the SJAE radiation monitor alarm setpoilts. This was documented by the licensee in CR 97-4046. SJAE Rad Mci. sensitivitie CR 97-4180 SJAE rad monitor setpoints, addressed coordination problems ',

between the Operations procedure used to request new radiation monitor setpoints, the Environmental and Radiological Control (E&RC) proced ce that calculates the new setpoint, and the Maintenance procedure that installs the new setpoints. By the time the radiation monitor setpoints were ready to be installed the new values needed to be recalculate The inspector determined as a result of the cited failure and the three additional CRs mentioned previously, that control and monitor'.ng of the alarm setpoint was poor. Previous instances of failing to properly disposition abnormal values were recorded by the NRC in Inspection Re) ort (IR) 50-325(324)/97-12, when inadequate corrective action was tacen for abnormally high drywell temperature. Tne abnormal temperature resulted in exceeding the calculated environmental limits for ten snubbers in the drywel ~

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c. Conclusions The licensee continues to struggle with proper dispositioning of abnormal indications. The failure to maintain the Daily Surveillance Report in accordance with procedure was a violation. Abnormal values observed for the Steam Jet Air Ejector radiation monitor and subsequent test indicate potential fuel failure for Unit M8 Miscellaneous Maintenance Issues (92902)

M (Closed) Licensee Event Reoort (LER) 50-325(324)/96-017-00: Invalid Loss of Coolant Accident Locic Actuation The invalid LOCA. initiation signal occurred during installation of test equipment to support surveillance testin P16nt systems responded as designed. The initiation signal resulted in the following actuation:

Automatic start of emergency DGs 1.2.3. and Automatic start of Unit 1 Core Spray (CS) pump 1 Automatic start of Unit 2 Nuclear Service Water (NSW) pump 2 Unit 1 Grou) 10 division 1 actuatio Closure of Jnit 1 Reactor Building Closed Cooling Water heat exchanger Service Water isolation valve.1-SW-V10 )ening of NSW header to vital header isolation valve. 1-SW-V117.

, Slutdown of 1A and 10 Unit 1 drywell coolers  ;

Corrective actions, described in the LER. were reviewed and verified by the inspector. -These included: appropriate administrative action with the involved technician; briefing of maintenance 1&C technicians on this event; providing maintenance I&C personnel managements expectations ft the restart of surveillance tests after problems have been encountered; restricting the use of Simpson Model 260 Voltage Ohm Meters (V0Ms) for circuit checks specified in maintenance surveillance tests: developing training to enhance technician knowledge of the effects of test equipment misalignment: and revising maintenance procedures to preclude similar event This event did not violate TS. This LER is close M8.2 (Closed) LER 50-325/97-009-00: Missed Increased Frecuency Inservice Testino Recuirement The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.Section XI, 1980 Edition through Winter 1981. Addenda Section IWV-3414(a), requires an increase in test frequency in the event an increase in stroke time of 25 percent or more from the previous test is observed. Contrary to this requirement, the test frequency was not increased as required. The required testing was missed by about two weeks. Upon discovery. the valve was tested and the stroke time was within the previous value and the test met the ASME Section XI requirement !

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The corrective actions to prevent recurrence of this event. described in the LER. were reviewed and verified by the inspector. Administrative controls have been revised to ensure completed test results are reviewed-in a timely manner and changes in test frequency are promptly initiate This event did not violate TSs. This event had minimal safety significance from a-valve operability viewpoint since the retest of the valve showed it was operable, ASME Section XI provides an intermediate condition that allows continued operation without need for immediate corrective action. From an administrative view, trending valve stroke times is an imaortant indication of valve performance. Corrective action taken s1ould improve this situation. This LER is close M8.3 FClosed) LER 50-325/97-001-00: Rod Block Monitor Surveillance

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nadeauacy

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A discovery that the surveillance procedure fer testing the rod block monitor (RBM). did not contain the pro 3er s 4 Ncessary to ensure ;

testing of the RBM instrument channel 3 int '

tion, This condition has existed since November 1996 for Unit 1, ma December 1996 for

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Unit 2. Upon discovery, the correct tests were performed on both units which indicated that the equipment was in calibration and capable of performing its safety functio The error was attributed to an inadequate administrative review of  ;

reformatting changes made in September 1996. The surveillance procedure changes were being upgraded in accordance with the generic procedure writers guide. However, these changes did not insert the proper steps to test the RBM inop instrument channel ' Corrective actions, described in the LER. were reviewed and verified by-the inspector. The inspector determined that this event did not violate TS since only the test for channel B was missed. The situation was corrected within the allowable time specified by TS 3/4. The-results of the RBM inop functional tests performed on toth units upon discovery, indicated that the equipment was in calibration and capable of performing its intended safety function. This LER is close M8.4 (Closed) LER 50-325(324)/95-022-00: HPCI System Discharae Flow Element Gasket Leak During performance of a post maintenance test on the HPCI system. the discharge flow element flanged gasket developed a 5 to 10 gallons per minute (gpm) leak. Several other problems were also observed with system operatio Investigation revealed that undersized flange studs had been originally installed on the flow element flange, allowing the Flexitallic gasket to be installed off center. The off centered gasket degraded during the post maintenance test. This condition existed on both units and prompted declaring a potential failure of the HPCI system to ]erform its intended safety functio With the HPCI system inoperable tie TS U

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oermitted continued reactor operation provide 1 the ADS. CS system, and RCIC were operable. This event was withir, Me TS requiremen Corrective measures as described in the LER were reviewed and verified by the inspector. This LER is close M8.5 (Closed) Ins)ection Follow-un item (IFI) 50-325/97-05-02: Abnormal CS Soarcer Brea t Detector Indication (Closed) VIC 50-325/97-06-03: Inadeauate CS Surveillance Procedure

.(Closed) LER 50-325/97 02: Core Soray Header Differential Pressure Instrumentation InoDerable On March 9.1997, en auxiliary o)erator (AO) was verifying instrumentation indications in tie Unit 1 Reactor Buildin The A0 observed.that the reading displayed for 1-E21-PDS-N004A. Core Spray Line Break Indicator, was not within TS 4.5.3.1.2.c.2 requirements. This

)ressure switch functioned to detect a break in the CS piping located l 3etween the vessel and the shroud. The differential pressure (dP)

sensor measures the pressure across the core. Due to the addition of

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indicated pressure drop to increase which would cause a more positive indicated dP. The out of tolerance condition had existed since November 1996 as stated in LER 50-325/97-02. During review of the associated surveillance procedures, the inspector determined that actual verification of the CS sparger alarm setpoint in relation to the

" normal" indicated instrument pressure was not being performed.

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Themfore. the licensee could not evaluate whether the alarm setpoint was within the " normal" TS range. This nonconformance resulted in VIO 50 325/97-05-02. Inadequate CS Surveillance Procedur The licensee performed reviews of data collected nonroutinely during 1995-1996 and in ESR 97-181 calculated a " normal" value for setpoint verification in the related surveillance procedures. The licensee subsecuently changed the alarm setpoints and updated the affected procec ures. Additionally. the licensee performed a review of the TS and determined that appropriate logging of required TS values was being accomplished. During the refueling outage for Unit 2 from Se]tember to October 1997 the licensee, with prior NRC approval, uprated t1e 100 percent _ rated thermal power 5 percent. The licensee included verification of CS sparger dP " normal" values as part of the uprate test program performed in accordance with S)ecial Procedure 2SP-97-20 Unit 2 Power Jprate Data Collection. The cleck served to record the CS sparger shutdown value The inspector reviewed ESR 97-634. ESP-97-204. CR 97-3870. LER 50-325/97-02, and other related documentation. The inspector verified that routine recording. upon entering mode 1. of the CS sparger dP was incorporated into 0)erating Instruction 001-03.3. Auxiliary Operator Daily Surveillance Report for both units. CR 97-3870. Core Spray Leak Detection, documented the discovery on October 29, 1997 by an AD, that

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the 2-E21-PDS-N004A. CS A Loop Leak Detection, was outside of its specified range. The instrument was declared inoperable and an LC0 was entered. The licensee determined the new CS dP range in ESR 97-634 Core Spr 3y Loop Line Breuk Detectio , Allowable Range Change. The new alarm setpoints were implemented and integrated into the affected surveillances. 3rocedures, and design documents. Based on completion of the review of t1e TS for other " normal" values not properly trended, adjustment of the dP alarm setpoints*to bring the setpoints into rvpliance with TS. and the institution of routine monitcring of the CS

.qarger " normal" values these items are close M8.6 (Clos (d) VIO 50-325(324)/97-02-04: Failure to Imolement the Renuirements of (a)(1) and (a)(2) of 10 CFR 50.65. The Maintenance Rule This violation reported that all historical data since July 10. 199 had not been obtained to establish baseline system / structure / component (SSC) performance, validate scoping, and set initial condition (a)(1)

and condition (a)(2) in the case of the reactor protection system (RPS),

Only corrective work. requests / job orders had been used for initial determination of functional failures. Therefore, instrument out-of-calibration data had not been reviewed for the period of July 10. 1993 through October 30. 1995. As an action related to Maintenance Rule implementation. Procedure OMMM-004. PM. was revised on October 30. 1995, to require that out-of-calibration data be evaluated for Maintenance Rule functional failure applicability. However, this requirement only collected subsequent instrument out-of-calibration dat As corrective action for this violation, the licensee reviewed all available instrument out-of-calibration data for the RPS and other components / systems which support the Maintenance Rule function Functional failures identified were evaluated against performance criteria to determine whether (a)(1) status should be assigne Although six condition reports were issued to evaluate additional functional failures, no system was required to be classified (a)(1)

based on this review. The inspector reviewed the licensee's corrective actions and held discussions with a)plicable management and engineering personnel concerning this issue. T1e inspector concluded that the licensee had taken the necessary corrective action to correct the deficient condition and had taken appropriate corrective action to prevent its recurrenc This item is close III. Enaineerina El Conduct of Engineering El.1 Review of Enaineerina Procedures Insoection Scoce (37550)

The inspectors reviewed the licensee's procedures which control the environmental qualification progra . _ _ _ _ _ _ _ _ _ _ _ _ -

4 17 Observations and Findinas

The inspectors reviewed the procedures listed below which control various activities related to the environmental qualification 3rogram to determine if the procedures implement the requirements of 10 C:R 5 Appendix B. and 10 CFR 50.49. The following procedures were reviewed:

EGR-NGGC 0005. Engineering Service Requests. Rev 6. dated Septembe" 5. 1997 EGR-NGGC-0007. Maintenance of Design Documents, Rev. 2. dated August 22, 1997

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EGR-NGGC 0153. Engineering Instrument Setpoints. Rev. 3. dated

August 22. 1997 l EGR-NGGC-0156. Environmental Qualification of mlectrical Equipment l Important to Safety. Rev. 4. dated October 8.1997 ENP-13.6 Equipment Data Base System. Control and Revision Rev. 12. dated June 25. 1997 MCP-NGGC-401.. Material Acquisition (Procurement Receiving, and Shipping). Rev. 3. dated August 26, 1997 The inspectors verified that the procedures provided adequate instructions for establishing, maintaining and implementing the requirements of'10 CFR 00.49 except for the issues discussed belo Section 9.6 of procedure EGR-NGGC-0156 provided the guidance for maintaining E0 qualification data packages (ODPs). The procedure specified that changes to ODPs are to be captured using the ESR process. The procedure required that ODPs were to be periodically updated as necessary to maintain auditability, to incorporate new

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requirements, to meet plant specific requirements, ard to keep the number of outstanding-changes at a reasonable level. However

procedure EGR-NGGC-0156 did not specify a clear time requirement for updating the CDPs. The inspectors also determined that procedure EGR-NGGC-0007 did not provide any requirements for updating ODP The failure to s]ecify specific criteria in procedures could result in the 0)Ps becoming unauditable which is contrary to the requirements of 10 CFR 50.49. The failure to maintain and u]date the ODPs was one of the causes of the violation whic1 resulted in the civil penalty identified in NRC Inspection Report (IR) 50-325(324)/96-14. The failure to establish clear, definite requirements for updating ODPs was identified as a violation example at the Shearon Harris Nuclear Plant in NRC IR 50-400/97-12. Since all Brunswick 00Ps are being revised and updated at the current time, a violation was not identified for this issue during the current inspection. The licensee's corrective actions for the Harris plant will resolve

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this problem since the Harris. Brunswick, n.d H. B. Robinson plants use the same corporate EGR-NGGC ?,ocedure Procedure EGR NG D 0153 provides the methodology to establish instrument setpoint margins sufficient to account for various instrument uncertainties and environmental effects including temperature, pressure, radiation, seismic, and insulation resistance errors Although procedure EGR-NGGC-0153 provided guidance on the treatment of environmental effects, the inspectors noted that in the discussion of temperature effects, the applicability of vendor 3 worst case performance specifications to plant specific conditions i was not clear. The inspectors also noted that requirements for seismic effects in procedure EGR-NGGC-0153 were not clear regarding t6 match / confirmation of vendor profiles to plant specific [ les or configuration, in addition, the inspectors noted that procedure EGR-NGGC-0153 referenced Drawing 0-03056. Service Environment Chart Normal &

Accident Conditions. Units 1 & 2. for information on accident temperature data to be used in instrument setpoint calculation The inspectors determined that-Drawing D-03056 was " frozen" on December 12. 1996, and was not available for use. The reason for removal of Drawing 0-03056 from use was documented in CR 96-04002 which identif9d the need to revise. and update Drawing D-03056-to incorporate f icironmental data from the Reactor Building Environmentai Renort (RBER), Revision 5. The inspectors noted in review of calculations initiated since December 1996, the RBER was referenced for temperature profiles in the re:ctor buildin The licensee indicated that a revision to EGR-NGGC-0153 will be initiated to resolve inconsistency in wording regarding the application of accident temperature / seismic effects to make it clear that vendor test results would fully envelope site specific profiles unless an evaluation has been aerformed to evaluate the-differences. Additional guidance will 3e included to characterize the requirements for engineering reviews of test-data to ensure seismic and environmental profiles are bounding for site specific conditions. The licensee indicated procedure EGR-NGGC-0153 will also be revised to either remove D-03056 as the reference for temperature data and replace it with the appropriate reference (the RBER) . or to correct the drawin The inspectors also identified that procedure EGR-NGGC-0153 unde-Section 9.5.1. Calibration Errors, was not clear regarding instrument calibration surveillance requirements for as-left, as-found or leave-alone zone tolerances. The licensee indicated that procedure EGR-NGGC-0153. Section 9.5.1. would be revised to clarify these requirements to indicate that calibration tolerances are the defined limits, above and below a desired value, within which an instrument loop signal may vary and not require a

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adjustmen Licensee engineers stated that calibration tolerances are understood to be "as-left" value The inspectors will review Procedure EGR-NGGC-0153 in a future inspection to followup on these issues. An ins)ector followup item (IFI). 50-325(324)/97-13 06. Revisions to )rocedure EGR-NGGC-0153, was identified to the licensee pending further review by liR Conclusions With the exception of the issues discussed above, the inspectors concluded that the licensee's procedures for implementation of the Environmental Qualification com) lied with the requirements of 10 CFR 50.49 and 10 CFR 50. Appendix 3. An IFI was identified to review procedure EGR-NGGC-0153 to verify that the licensee incorporates the above comments and clarifications. The reference to a " frozen" drawing to obtain accident temperature data and the wording inconsistencies discussed above were identificd to the licensee as a weaknes El.2 Review of Instrument Setooiit Calculations Insoection Stone (37550) ,

The inspectors reviewed randomly selected instrument setpoint calculations to deternine the adequacy of the licensee's calculation Observations and Finninos The inspectors reviewed the instrument setpoint calculations listed below and verified that the calculations were completed in accordance with NRC requirements. The inspectors verified that the calculations incorporated industry standards. Updated Final Safety Analysis Report commitments. Technical S)ecification requirements, and recommendations contained in iRC Regulatory Guide Calculations reviewed were as follows:

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-Calculation OE41-0036. Power Uprate HPCI Steamline Flow High Uncertainty and Scaling Calculatio Calculation ORWCU-0010. U1/U2 RWCU Flow Accuracy Calculation. Units 1 and 2 RWCU Differential Flow Leak Detection / BESS I& Calculation 0821-0068. Power Uprate Main Steam Line Flow High Setpoint Uncertainty and Scaling Calculatio Calculation 0-01534A-297. Insulation Resistance Degradation Calculatio From review of System Description SD-01.2. Reactor Vessel Instrumentation. and the Safety Evaluation by the Office of

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Nuclear Reactor Regulation. Conformance to Regulatory Guide 1.97 Revision 2. Brunswick Steam Electric Plant. Units 1 and 2. Dated May 14. 1985. the inspectors concluded that these calculations were. typical. The instrument setpoint calculations typically considered 140 F as the maximum temperature in the calculation From review of the calculations, the inspectors determined that instruments that perform a safety function are analyzed for a LOCA environment in the reactor building. The calculations showed that instrument uncertainties considered instrument temperature effects for a maximum temperature of 140' F which is bounding for the analyzed LOCA environmen The inspectors also determined that instruments relied upon to mitigate the effects of a high energy line break (HELB) were also evaluated by the licensee. For this instrumentation, environmental uncertainties-for a harsh environment were not required to be considered since the instrumentation function would occur before the reactor building temperature )rofiles listed in p the Reactor Building Environmental Report (REBR) Revision dated November 5. 1997, would reach 140 F and affect instrument performance. The ins)ectors noted that abnormal temperatures were not discussed in the-RBER. Discussions with licensee engineers disclosed that the design base accident event is based on an initial building environment airspace temperature of 104 F. The building temperatures ace measured and recorded daily by plant operators in accordance with procedure numbers 101-03.4.1 and 201-03.4,4. Unit 1 and 2 Control Operator Daily Check Sheets.. The

= operators are required to contact the duty engineer when the reactor building temperature exceeds 104 F so that engineering can perform an assessment of the effects of temperature on environmental qualificatio The inspectors noted that calculations for instrumentation which mitigates a HELB demonstrated that the instrument and associated equipment would not be exposed to a harsh environment before the instrumentation performed its safety function. In the instrument calculations reviewed by the inspectors instrument setpoints were based on a maximum temperature of 140 F (non-steam environment).

Although allowances were not made for a harsh environment. a seismic allowance was included in the calculation Review of the temperature profiles as shown in the Brunswick Reactor Building Environmental Report showed that the actuation isolation signal would occur before exceeding the temperature allowances assumed in the setpoint uncertainty calculations. An exce) tion was the High Pressure Coolant Injection (HPCI) line breat in the steam tunnel where the temperature profile showed that 140 F would be exceeded for ap3roximately 2.5 seconds before the isolation trip _ signal occurs. iowever this instrumentation would remain operable based on thermal delays. However, the HPCI isolation function would most likely be initiated by temperature

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sensors in the steam tunnel or HPCI room which would occur imediately with no time dela The inspectors concluded that the instrument setpoint calculations complied with NRC requirements and were technically adequat Review of the calculations showed that environmental effects, j- specifically accident temperature, were correctly evaluated in the calculations, Conclusions The inspectors concluded that the licensee's calculations were technically adequate and complied with NRC requirement The inspectors concurred with the licensee's conclusions that the setpoints for instruments relied upon to mitigate the effects of a KLB did not require inclusion of uncertainties for a harsh environment since the instruments perform their ft..iction before being effected by the harsh environment. Setpoints for instruments required for LOCA effects include the appropriate environmental uncertaintie El.3 Enaineerina Service Reaucst (ESR) 97-00426 Inspection Scoce (375501 '

The inspectors reviewed ESR 97-00426 which was prepared to address questions on instrument setpoint Observations and Findinas A review of procedures and various documents by an independent consultant resulted in questions involving environmental effects including uncertainties on instrument accuracy. These guestions .

were dccumented in an E-mail message dated June 20, 1997 Subject: '

E0 and Instrument Accuracy. The licensee addressed the referenced !

memo in Engineering Service Request ESR 97-00426. Revision dated September 18. 1997. ESR 97-00426 documents the evaluation completed by the licensee to address environmental effects on-instrumentation. The inspectors noted that the licensee response did not address the questions in the June 20, 1997 E-mail message point by point. but provided an evaluation that was more generic in nature. The inspectors noted that ESR 97-00426 was an engineering disposition (ED) type ESR. as defined in procedure EGR-NGGC-000 The use of this type ESR to respond to the E-mail cuestions was appropriate since the ESR only communicated existing cesign requirements, did not produce design output, and did not change existing engineering document The ESR concluded that instruments that aerform a safety function are analyzed for a LOCA environment in t1e reactor building. The instrument uncertainties consider -instrument temperature ef fects for a maximum temperature of 140"F which is the maximum bounding

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temperature for the analyzed LOCA environmen The inspectors noted that the word minimum had been incorrectly used in the fourth line, third paragraph in Section 2.0 of the ESR. The licensee stated that they will correct this error when the ESR is revised. as discussed belo ESR 97-00426 also concluded that harsh environmental effects have been appropriately accounted for in safety related uncertainty calculations. The ESR concluded that the isolaticr. aquence for a HELB due to main steam line break. reactor core isolation cooling l steam-line break, high pressure coolant injection steam line break, cr a piping failure in the reactor water cleanup system is such thtt the isolation function will occur before the instrumentation is exposed to harsh environmental effects. This conclusion was based on the instrumentation being able to perform its safety function prior to the temperature exceeding the temperature allowance assumed in the setpoint calculations. For area temperatures exceeding the setpoint temperature uncertainty allowance, the use of emergency operating procedures (EOPs),

operator action, and local temperature instrumentation would mitigate the event and provide the actions to determine and/or maintain. reactor level during a LOCA or HEL When temperatures exceed the temperatures (140 F) assumed in the setpoint calculations, plant operation is controlled through the

' COP A review'of E0P-03-SCCP Revision 5. Secondary Containment Control Procedure, and 2EOP-LPC Revision 1. Level / Power Control, shows that high area temperatures are an entry condition into secondary containment control procedure E0P when area temperatures exceed the maximum safe operating value requiring manual reactor sCrd E0P-03-SCCP Revision 5. refers the operators to Caution 1 to determine reactor level instrumentation operability. A review of Caution 1 disclosed that vessel level wide range instrumentation  ;

8B21 - LI - R604A/604B and C32 - PR - R609 are not to be used when secondary containment temperature exceeds 140 F. This exclusion was because the reference leg and associated instrumentation for these loops are in secondary containment. E0P Caution 1 then

)rovided compensation data for the remaining level instrumentation

]ased on drywell tem]erature, reactor saturation limit, and reactor pressur iowever, for secondary containment temperatures above 140 F. Caution 1 instrumentation may not be o)erable with instrumentation exposed to temperatures greater tlan 140*F during an event. In cases when vessel level can not adequately be determined, the E0Ps direct the operators to depressurize by initiating ADS and flood the vessel using low pressure emergency core cooling system .

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23 Conclusions The inspectors concluded that the licensee adequately addressed the questions in the June 20. 1997 E-mail message regarding instrument and E0 accuracy. However, the licensee stated that

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they will revise F.SR 97 00426 to address each question and recommendation ir. the E-mail message point by point to further clarify their response to the concerns / issues raised in the June 20, 1997 E mail messag El.4 Environmental Qualificat%1

, Insnection Scooe (37550.92903)

The inspectors reviewed the licensee's corrective actions for the

Environmental Qualification (FO) program, in response to findings l identified during Self-Assessment numbers 95-0041 and 96-0271 and the violations identified in NRC IR 50-325(324)/96-1 Observations and Findinas 1) Review of E0 Equipment Data Base The licensee's corrective actions to resolve the discrepancies in the E0 program identified by NRC (See IR 50-325 324/96-14)

include corrections to and updating of the Equipment Data Base System (EDBS). Numerous errors in EDBS had been identified and corrected by the licensee since the inspection findings were identified in IR 50-325(324)/96-14. The errors in EDBS were .

identified during E0 equipment walkdowns and review of various  !

data bases. In addition, numerous errors were identified in the EQ zones listed in EDBS for the location where various components were installed. These primarily occurred at. zone boundaries and were being resolved during review of walkdown dat The requirements for. recording and correcting E0 data in EDBS was s)ecified in- CP&L procedures EGR-NGGC-0156 and ENP-33, The-c1anges to EDBS to correct errors were processed using Form 100 of ENP-33.6. The Form 100 was design verified in the E0 unit and was then forwarded to appropriate personnel for entry into EDBS. All EDBS data entries made were independently verified by personnel in the Configuration Management group in the Design Control Uni The independent verification was performed to minimize o-eliminate data entry errors. Additional corrections to EDBS were ongoing to incorporate E0 walkdown ins)ection results and the revisions to EQ qualification data paccage The inspectors reviewed some randomly selected revisions to EDBS identified as a result of the E0 corrective actions and verified the EDBS data had been corrected. The inspectors also discussed the program for control of changes to EDBS with various licensee personnel who perform the day to day system revisions. These

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discussions disclosed that these individuals were cognizant of the requirements for controlling and making corrections to EDB ) Review of Qualification Data Packages The inspectors reviewed a draft cop.f of Revision 4 of ODP No. 4 titled. " Qualification Data Package For NAMCO EA180 Series Limit Switches" to determine if it adequately demonstrated environmental qualification for the safety related NAMCO switches for use inside the drywell in accordance with 10 CFR 50.49 and appropriate licensee E0 Prccedures. The package addressed the following:

qualification level (0588 Cat. I); tag numbers of equipment covered in the QDP: test report aaplicability; similarity of test specimens to installed equipment: E0 parameters. temperature, pressure, relative humidity, radiation, chemical spray, submergence; cualified life: E0 maintenance requirements; test anomalies; anc operating experience item During review 3f the Draft ODP. the inspectors identified the following questions / comments:

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The text in the CDP indicates that there were five anomalies in

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Qualification Test Report (OTR) 130 but only four anomalies were discussed in the ODP.

l .

Attachment 2 to the ODP included a calculation for qualified life l of the limit switches which was not signed as reviewe * Differences were noted in the system component evaluation worksheets (SCEW) for the same limit switches in the different unit * Data was missing from some of the SCEW sheets. That is, there were blanks on the data sheet For example, data on accuracy was left blan *

Some components were specified with Anaconda flex and others just stainless steel flex conduit. Additionally, only certain components were specified for weep hole * Page 49 section 4.1 Installation requirements indicates that the conduit seal may not be necessary for those limit switches installed in the Reactor Building. This requirement should be clear and should specifically list those limit switches which require conduit selling to ensure qualificatio * Page 13 lists the 16 Namco EA180 limit switches which had been installed. However only 14 were considered qualifieo by this OD Unit I limit switch tag numbers 1821-ZS-5373 and 1B21-ZS-5374 were excluded from the E0 requirements by ESR-97-00431. The Unit 2 equivalent switches were not discussed in the OD . _ _ _ _ .

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• In Section 2 of the 00P it was stated that it was a good maintenance practice to lubricate the NAMCO limit switche however. lubrication was not specified in Section 4 of the ODP which lists recommended maintenance practices.

In Section 4.2 of the ODP it was stated that the switches can be refurbished. However, a statement was made on page 21 that qualified replacement part kits were no longer available.

• A reference was made to abnormal temperatures on page 38 of the OD However, abnormal temperatures were not included in DR 227.

• The inspectors questioned apparent inconsistencies between activation energies and aging methods discussed in referenced qualification test reports (OTRs).

The licensee indicated that these comments would be evaluated by the E0 group and if appropriate, addressed in Revision 4 of the QDP when it is completed.

The inspectors reviewed a draft copy of Revision 7 of ODP-67 General Electric Company IC 7700 Series Motor Control Centers for BN The GE MCCs. located or, the 20, 50, and 80 foot elevations of the Units 1 and 2 Reactor Buildings, are subject to harsh environments resulting from postulated design basis accidents and ;

have a safety function to mitigate the consequences of these F accident The MCCs were qualified in ODP-67.

A series of similarity analysis were performed to demonstrate similarity between the tested configuration and supplied. The inspectors reviewed portions of DR 232. "Nutherm Report No. CPL-7806R. Qualification Test Results Applicable to Brunswick Nuclear Power Plant Safety-Related GE 7700 MCCs." Revision 0, dated June 30, 1997 which dccumented the similarity analysis. Section 2 of DR-232 contains a discussion on the similarity analysis between the components tested by NUTHERM and those installed in the Brunswick MCCs.

The similarity discussion covers fuses, stab assemblies, control transformers control and power wiring, overload heaters. overload relays, terminal boards, starters and contactors, molded case circuit breakers. circuit protectors. disconnect switches. potentiometers, and indicating lights. The similarity analyses were based on the similarity analyses contained in DR 1.1. GE Company NEDC-30696-P. May 1985. MCC Oualification Test Report Phase Il for CP&L Brunswick Plant, or were devices which could be directly linked to a test specimen and did not require a similarity analysis. Based on review of DR-232 NRC concluded that NOTHERM was able to establish that the com3onents they tested were in the same family as those provided by GE in t1e MCCs. This review was also dccumented in IR 50-325(324)/97-0 A draft copy of Revision 0 of ODP 99. R. G. Laurence Series 500 and 600 Solenoid Valves was reviewed. The inspectors verified that similarity analysis was included in the ODP >

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3) Review of EO Walkdown Data The inspectors reviewed E0 walkdown data which document inspection of E0 equipment-in the Unit 2 MSIV pit and drywell, and Unit 2 reactor building. Tha E0 walkdowns were performed in accordance with CP&L Special Procedure OSP-96-014. EQ Equipment Field Verification. The pyrpose of the walkdowns was to verify the accuracy of the manulacturer/model number listed in the licensee's data bases and to verify the equipment installed orientation and configuration were in accordance with the E0 qualification documentatio The ins)ectors reviewed walkdown records for scram '

pilot' solenoid valves, 1AMC0 limit switches, temperature elements, excess flow check valves, and pressure switches. The walkdown data was recorded on field inspection data sheets which were'then converted into an electronic data base. The inspectors verified that discrepancies identified during the walkdowns were documented either on a work request (WR/J0) for repair, or in a condition re) ort (CR). The ins)ectors reviewed completed WR/JO numbers 97-AF JR1, 97-AFUR2, 97- A UR3, and 97-AFUR4. These WR/J0s document drilling of weepholes in junction boxes in the Unit 2 MSIV pit to resolve a moisture intrusion issue. These boxes are associated

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with limit switches for the Unit 2 main steam isolation valves.

L The completed WR/J0s showed that the weepholes were drilled to resolve the concerns. The inspectors did not identify any

discrepancies in the records reviewe ) Review of Environmental Qualification Condition Reports The inspectors reviewed the licensee's corrective c.,ctions to L disposition the CRs listed below. These CRs were initiated by the licensee to-document and disposition nonconforming items whicn were identified during the ongoing E0 reconstitution project. The nonconforming items were identified as a result of E0 equipment walk h ns, review cnd updating of E0 equipment ODPs, omissions from the original program, or changes to the operating environment. The CRs reviewed were as follows

CR 97-02015 The licensee initiated CR 97-02015 on June 6. 1997 to document and disposition deficiencies that had been identified by the licensee's training staff during observation of simulator training when the fire protection system had not been isolated within the 15 minute time period after initiation of a HELB specified in 31 ant o)erating 3rocedures. The 15 minute time period is the

) asis w1ich esta)lished flood '.evels for E0 e and north and south RHR and core spray rooms.quipment in the HPCI Review of closure for CR 97-02015 disclosed that the licensee concluded that the issue has been adequately addressed by operator training, primarily through critiques which were held following the completion of the simulator training to discuss deficiencies noted during the training. In response to the CR. Action Items were Y

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assigned to the Operator Training group to incorporate the basis

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for the need to isolate the fire protection system into training materials. However, review of the training records on June 12, 1997, by personnel from the E0 group resulted in additional questions regarding the licensee s corrective actions. The records reviewed by the E0 personnel indicated that during

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simulator training, approximately 10 to 20 percent of the operators were failing to enter AOP-05,0, Radioactive Spills. High

. Radiation, and Airborne Activity, or were entering the AOP late

(after 15 minutes). The inspectors made an indepen6nt review of

, the training records reviewed by the E0 personnel. This review disclosed that the records the E0 personnel reviewed on June 12, 1997 were for the six month 02015 (January - June 1997) The .

period prior to reviewed inspectors initiation of CR 97-training records for July - September, 1997 and noted significant improvement in this area, although the HELB scenario was not included as part of the simulator training exercises in this time period. The training scenario did include a torus leak which required entry into A0P-05.0.

I The inspectors noted that the concern regarding flooding of

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instruments could also be caused by other accidents such as pipe L breaks in the service water or Reactor Building Closed Cooling l Water (RBCCW). Operator actions in these cases would be directed e by E0P-03 SCCP Secondary Containment Control Proccdure (SCCP),

based on high water leve'is in the HPCI and north and south RHR and core spray rooms. An uttry into E0P-03-SCCP would also result from flooding in these same rooms caused by activation of the fire protection system. As aaditional followup on this issue, the inspectors observed simulator training scenarios performed on December 3 and 17, 1997. Included in the scenario was a RCIC steam line break (HELB) and activitation of the fire protection system. Both crews participating in the training scenario isolated the fire protection system within the 15 minute time period. The inspectors also questioned some randomly selected reactor operators regarding the need for entry into A0P-0 following a HELB. The operators were cognizant of the basis of the actions in A0P-05.0 (need and reason for isolating the ' ire protection s CR 97-02015.ystem) and were familiar with the problem addrc ses by The inspectors verified the action items associated with the CR were complete CR 97-02015 was closed on December 11. 199 CR 97-01841. 97 02025. & 97-02408 These CRs documented various issues regarding possible effects of moisture on E0 equipment. CR 97-0184) was initiated to document the effect of spray from the fire protection system on E0 equipment in the reactor buildin The licensee has resolved all the issues associated with this CR except for drilling of weepholes in junction boxes whicn may be affected by the water spray. Licensee engineers are currently

- preparing instructions and procedures for completing this wor _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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The problem documented in CR 97-02025 concerned an issue which had been the subject of IE Circular 79-05. Moisture Leakage in Stranded Wire Conductors, which was issued by NRC on March 2 . This affects Patel seals which were used to seal some stranded wire conductors in instrument circuits. CR 97-0?408 documents several other moisture intrusion issues. The immediate corrective action taken to resolve these issues, as documented in CR 97-02408 was to hire an outside consultant to address the issues. The consultant has reviewed many of the issues documented in CR numbers 97-01841, 97-02025. and 97-02408 and made recommendations, some of which have been implemen.ed. The consultant also addressed another issue in the CRs involving current leakage in control circuit and the possible impact on ODPs and E0 of equipment. This concern was the effect of moisture intrusion through stranded wire conductors, sealed with Patel seals, which could result in leakage currents in instrument circuits. ESR 97 00440 was issued for the 120 volt AC circuits and ESR 97-00441 for DC circuits. These ESRs are currently being reviewed by licensee engineers. The current leakage issue was also applicable to questions raised regarding the NAMCO limit switches. The inspectors will review the licensee's evaluation of current leakage and its ap311 cation to evaluation of E0 equipment in a future inspection. T11s was identified to the licensee as IFl 50 325(324)/97-13-07. Review Technical Evaluation of Current

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Leakage and the Effect on EQ Equipment. pending further review by NR The licensee also aerformed an evaluation of the potential for moisture wicking t1 rough Patel seals. This evaluation was i documented in ESR 97-00423. 03erability Evaluation - Wickin Review of the ESR disclosed t.at the licensee performed a detailed evaluation of the Patel seals by comparison of the installations at Brunswick with the configurations tested by NRC at Sandia Laboratorics (NUREG/CR 0699. Jublished March.1979). The licensee's conclusions were t1at the design function of the instellea equipment will not be effected by moisture intrusion through the stranded wire. The ESR was based on a review of the duration of the design accidents and the resulting leakage currents caused by moisture intrusion into limit switche Further review of this ESR will be performed as part of IFI 50-325 (324)/97-13-07, discussed abov CR 97-02016 & 97-02074 CR numbers 97 02016 & 97-02074 were initiated to document issues involving NAMCO limit switches. The following issues were identified in the CRs:

• Inability to identify the date of manufacture of the switches since the codes for date of manufacture were painted ove __ ________ ____

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• Potential for paint to impair the operability of the switche The concern was that paint on the roller arms would impair mechanical function of the switches.

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• Chemical reaction between paint and internal switch components would cause corrosion of switches, leading to failure of the switche * Use of incorrect qualification test reports (0TRs) in the qualification test reports which qualified the switche * Effect of current leakage on switch operabilit A total of 14 NAMCO limit switches were covered under the E0 program. These switches were installed during modifications completed in 1983 and 1984 The licensee has determined that none of the switches were purchased or manufactured prior to 198 Therefore, the concern raised by IE Bulletin 79-28. Possible Malfunction of NAMC0 Model EA 180 Limit Switches at Elevated Temperatures, would not apply to the switches installed at Brunswick, Review of the licensee's response to IEB 79-28 disclosed that none of the potentially defective switches had been purchased by the Brunswick sit Review of the i1censee's corrective actions completed to date disclosed that the following actions have been completed:

The licensee has identified the date of manufacture for most of the NAMCO limit switche Additional manufacture dates may be identified when the Unit 1 walkdowns are completed during the Spring 1998 refueling outage. However, the licensee has conclusively determined that none of the switches would be affected by the defects identified in IEB 79-2 .

The switches were stroked in accordance with frequencies per the Technical Specifications which demonstrates that the mechanical function of the switches had not been impaired by the pain * The paint has been teste The test results show the not cause corrosion or deterioration of the switches paint would

. The ODP. has been revised to incorporate the correct OTRs. The ODP. ODP 49, was still in draf . The current leakage issue has been evaluated " ESR numbers 97-00440 and 97-00441, which are currently being reviewed by licensee engineer The licensee subsequently has determined that the switches were still within their qualified ;'fe. No equiament operability issues related to tv.e NAMCO ilmit switches lave been identifie _

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[R 97 02367 This CR was initiated on July 3, 1997 to document the failure to initiate CRs for nonconforming items, specifically, MCC door gaskets and non standard Raychem splices identified as a violation by NRC during an inspection documented in NRC 1R 50-325 (324)/97 08.- The licensee's corrective actions included completion of a review of all the E0 walkdown data sheets to identify any nonconforming equipment. Additional corrective actions included training of personnel in the E0 group regarding the corrective action program and assessment of the effectiveness of the corrective actions. These correcthe actions were also associated with other similar corrective action CRs. such as CR 97 01972 and CR 97-02465. The inspectors reviewed the completed corrective actions and concurred with closure of CR 97 02367. The CR was closed on December 14. 199 CR 97-02465 and 97-02672 This CR wac initiated on July 15, 1997, to document concerns on EQ operability determination This CR referenced CR numbers 97-01841, 97 02025. and 97 02408. discussed above, which involve moisture intrusion issues. As a result of the concerns raised in CR 97 02465, the E0 group presented an action plan to resolve the moisture intrusion issues (CR 97 02465) to the plant nuclear safety committee. Although, further review showed the operability determinations for the three CRs were correct, the root cause analysis concluded that there were other problems which resulted in CR 97-0246 The root cause of CR 97-02465 was attributed-to weak E0 project management. The root cause/ event review for the CR listed the causal-factors indicative of weak E0 3roject management to be poor communications within the E0 group, tie site position that E0 problems were primarily docunitation problems, and a poor corrective action culture within the E0 group. The poor corrective action culture was evidenced by corrective action items which were routinely extended, overdue, or completed late: failure to prepare JCOs: numerous CRs written against the E0 grou) for improper corrective actions: and closing CR action items )y other action items without completing the corrective actions. A violation of NRC requirements was identified in IR 50 325, 324/97-12 for failure of the licensee to implement their corrective action progra The licensee's corrective actions to address the issues raised in CR 97-02465 included increased management oversight aerforming a review of the E0 project schedule to complete the higlest priority work activities first, conducting more frequent E0 group meetings to improve communications within the E0 group, transferring some E0 group functions from the Design Control l%1t to a site organization. and performance of an effer' ve. dss review of the

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completed corrective actions. The CR was closed on December 17, 1997. The inspectors reviewed the completed corrective actions

.and concurred with closure of the CR. The ins)ectors concurred with the licensee's conclusions that the opera]ility determinations for the three referenced CRs were appropriate. NRC will perform review of the liccasee's actions to correct the l violation in future inspections, CR 97-02672, which was inniated on August 5. 1997, indicated that the Supervisor comments listed in CR 97 02465 were a misstatement of the consensus of opinion of individuals which met to discuss CR 97-02465. Review of CR 97-02672 disclosed that the CR did not raise any new issues or conceriis which had not been addressed by CR 97 02465. CR 97-02672 was closed on December 17, 199 NRC concurs with the licensee's conclusions and closure of the C CR 97 4059 This CR was initiated on December 2, 1997, to document concerns and questions on ESR 97-00426. The questions involved appropriateness of E0P actions, the need to include evaluation of drywell instrumentation in tic ESR, and various questions on instrument setpoints. The 1 Lensee completed a review of the questions raised in the CR and concluded that the ESR had addressed these issues, or the issues were beyond the scope of the ESR, For exam)le, appropriateness of E0P actions were approved by NRC for all BW1s and do not involve instrument setpoints. There are no instruments in the drywell which provide signals for automatic actuation. The inspectors reviewed the licensee's responses to the questions in the CR and concurred with the licensee's conclusions that no new corrective actions were required to resolve the concerns / questions raised in CR 97-04059-which had not been previously resolve ) Review of Environmental Qualification Requirements in Procurement Practices Th'e inspectors reviewed CP&L procedure MCP-NGGC-0401, Material Acquisition (Procurement. Receiving, and Shipping). Revision 4, dated August 26, 1997. This procedure specifies the instructions for procurement of safety related materials for use in CP&L nuclear plant. The inspectors noted that the requirements for obtaining reviews by E0 engineers is specified in the procedur Discussions with licensee engineers and review of previous revisions of the procedure disclosed that the procedure had been revised to strengthen the need for the E0 review in Revision 2 of

.MCP-NGGC 0401, effective April 15. 1997. Revision 2 added

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requirements that components that require environmental qualification:be reviewed by the E0 grou During review of CRs. the inspectors identified several examples of acceptance of materials / equipment by procurement engineering

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for use in E0 installations which were based on test reports which ;

had not been reviewed by the E0 group -These were documented in

'CR numbers 97-01970 and 97-03036, Several additional examples of discrepancies in documents prepared by procurement engineering which affected E0 equipment were also identified during review of procurement specifications and other documents su * as material evaluations. These discrepancies were documented in CR 97-04035 which tas initiated on November 25, 1997. The review of procurement documents was being performed as part of the corrective actions to address the E0 program discrepancies identified in IR 50-325(324)/96 14. This was listed as Commitment *

1. 4 in the licensee's December 19, 1996 Reply to Notice of ,

Violation, 6) Equipment Lubrication Requirements The inspectors reviewed CP&L procedure MMM-053. Equipment Lubrication Application Guidance and Lubricant Listing, Revision 6 dated November 11, 1997. This procedure provides a listing of plant equipment with recommended lubricants to be used, guidelines for lubrication of plant equipment, and lubricant sampling methods. The inspectors identified the following issues after reviewing the procedure:

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ODPs 26, 68, and 88 were not referenced in procedure MMM-053. These ODPs cover environmental qualification of Reliance electric motor Document References corresponding to above ODPs were not reference The types of lubricant specified fo, the Reliance motors in procedure MMM 053 differ from those listed in the ODPs 26 and 6 Procedure MMM 053 permits maintenance to change the lubricant without obtaining engineering review or approva Discussions with licensee engineers disclosed the CR 97-04015 was initiated on November 20, 1997, to document the fact that the procedure permits changes to lubricants without performance. of an engineering review. Action Item 40 to CR 97-02627 was issued to document a similar issue. This action item was closed by CR 97-0401 The inspectors determined that the licensee had not evaluated that the type of lubricants (Mobil) specified-in procedure MMM 053 for Reliance electric motors differed from those listed in ODP 26 and 68, Review of ODP 26. Revision 1. Joy Fan /Peliance Electric Company, Class 1E Continuous Duty, 20 HP, and ODP 68, Revision Standby Gas Treatment System - Fair Company Filter Unit and Control, showed that the electric motors were both qualification

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tested using Chevron SRI 2 greas The impact of using a dif ferent type of grease to lubricate the motors on the environmental qualification testing of the motors had not been documented by the licensee. The licensee initiated CR 97 04064 to document the fact that substitution of alternate lubricants had not been evaluated by E0 engineers. The failure to establish maintenance procedures appropriate to the circumstances for performing maintenance was identified to the licensee as another example of violation item 50 325(324)/97-13-02. Inadequate Procedure for the Conduct of E0 Preventive Maintenanc c. Conclusions

One violation example was identified regarding an inadequate E0 maintenance procedure for lubrication of E0 electric motors. Two inspector followup items were identified to followu) on revisions to instrument setpoint procedures and to review leacage current calculations. The licensee was making progress in resolving and closing CRt identified by the E0 group. As of the inspection dates, no 0DPs had been issue E5 Engineering staff Knowledge and Qualification E5.1 Trainino and Qualification of E0 Personnel Insnection Scone (37550)

The inspector reviewed the licensee's program for training and qualification of personnel in the E0 task force. including both CP&L and contract engineers, b. Observations and Findinos The requirements for performance of E0 equipment walkdowns are specified in CP&L Special Procedure OSP-96-014. E0 Equipment Field Verification. The prerequisite in procedure OSP 96-014 for individuals performing the walkdowns was to read the procedur The licensee qualified a number of individuals to perform the field walkdowns through a training program conducted in accordance with CP&L procedure TI-100. Conduct of Training. These individuals included Instrumentation and Control technician contract engineers, and personnel assigned to the E0 group who were qualified E0 engineers. The training for the qualified E0 engineers consisted of reading the procedure. orientation and on-the-job training to become familiar with the walkdown and data s gathering process. For other personnel, the training included reading of the procedures, formal classroom lecture demonstrations, performance of practical exercises, and on-the-job training. The walkdown group supervisor performed a detailed review of the result < of practical exercises and data gathered <

during initial walke is prior to signifying the individuals were s. _ __

qualified to perform walkdowns. The training provided for the walkdcwn personnel exceeded the procedural requirements. The E0 walkdown grou) supervisor stated that the level of training provided to t1e walkdown personnel war to assure that the walkdown results were very accurate and to preclude the need for repeat work. The inspectors revieweJ the training records for the walkdown personnel and verified that they had been trained in accordance with the licensee's program. The inspectors noted that the experience level for the walkdown personnel varied from a recent graduate engineer to individuals with more than 20 years of experience. The inspectors reviewed the walkdown inspection records prepared by various individuals in the walkdown group and noted that the original walkdown records were complete and accurate, with some exceptions. Discussions with the walkdown group supervisor disclosed that corrections noted on the records were the result of reviews perfnrmed to resolve discrepancies in the records. The changes were made as a result of additional walkdown inspections which were doc'mented in the records. In one case, an individual was terminated for failure to perform the walkdowns and complete the walkdown records properl This individual's work was reviewed by the licensee and corrected where necessar The inspectors also reviewed the training and qualification records for E0 technical personnel. These records included previous work experience, education and training, and CP&L specific training applicable to the E0 project. This training included E0 technical reviewer, E0 design verifier E0 calculations, and E0 ESR originato The inspectors also questioned the manager of the E0 group concerning work assignments within the E0 grou). That is, assignment of specific activities to individuals wit 1 previous experience in a particular area of specialization, such as review of requirements for qualification of motors or specific types of instrumentation. The E0 group manager has recently are)ared a directory of all engineers working within the E0 group w11c1 lists each engineer's experience and what work activities they have completed for the E0 project at Brunswick. The purpose of this directory was for the engineers within the group to know who has worked on various problems and issues so they could obtain assistance from these individuals when they become involved with similar technical issues. The directory was distributed'to personnel in the E0 group. The E0 group manager provided a copy of the directory to the inspectors and discussed the basis for the various work assignments within the group.which were based on the past work experience of the E0 technical personnel, c. Conclusions The inspector concluded that the licensee's program for training and qualification of E0 engineers meets NRC requirement ,

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E8 Hiscellaneous Engineering Issues (37551, 92903)

E (Closed) URI 50-325(324)/97-08-04: Control of Ecuioment Data Base System (EDBS) Information The licensee issued CR 97-02400. Non Validated EDBS Information, concerning rc, tine use of non-validated EDBS information. This wes associated with VIO 50 325(324)/97-08 03. Safety Relay Setting Change Made as Pen and Ink Changes to Procedure. The licensee replied to this violation on September 2. 1997. The reply discussed licensee corrective

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action regarding the use of EDBS. Likewise. the licensee responded on November 26. 1997 to VIO 50-325/97-11-01. Failure to Initiate Alternate Safe Shutdown Impairment. addressed corrective action fcr use of an EDBS non validated field for determination of an Alternate Safe Shutdown impairmen Plant procedure OENP-33.6. Equipment Data Base System Control and Revision, provides instructions for control of EDBS information. Color coding of fields in the electronic database represent the various types of data present. This procedure provides direction that certain types of data are not to be used until verifie Accordingly two previous violations address the use of non-verified EDBS information. The licensee corrective actions for these violations are being completed. The requirements for the control of information are in procedure OENP-33.6. Previous items address the concern of this URI. therefore this item is close E8.2 (Closed) LER 50-325(324)/97-04: Soent Fuel Shionina Cask Handlina Activities This report documented the discovery by the licensee that the heavy load analysis as described in tne UFSAR did not completely bound movement of the shiroing cask from the primary lift to the secondary lift with the valve box covers remove It was determined that movement of the cask with a non single failure proof yoke and less than full cask integr'ty constituted an unreviewed safety question (US0) in accordance with the requirements specified in 10 CFR Part 50.55 The failure to obtain prior approval for a previously unanalyzed condition was determined in IR 50-325(324)/97-12 to be a violation. In a letter to the NRC dated August 6. 1997, the licensee requested a license amendment for review of the US" The licensee re evaluated findings relative to the 30 foot dro: ~cident and qualified the transfer yoke using guidance provided in NUR b 0612. Control of Heavy Loads at Nuclear Power Plants. This evaluation contended that a fuel shipping cask drop event was not credible. therefore operation with less than full cask integrity was no longer a problem due to acceptable redundancy in the lifting yoke. In a letter to the licensee dated December 2. 1997, the NRC accepted the licensee determination that operation with the valve covers removed would not compromise the health and safety of the public due to acceotable redundancy of the lift devices. Based on the acceptance by the NRC of the licensee's evaluation and issuance of the enforcement action as described in IR 50-325(324)/97-12 this item is close .- . ._

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36 E8.3 (Closed) Inspector Followun item 50-325(324)/96-14-05.'Effect of EO Accuracy on Instrument Setooint Calculations.

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Review of procedures and various documents by an independent *

contJ1 tant had resulted in a number of questions regarding the

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effect of environmental effects (uncertainties) on instrument accuracy The questions / concerns were documented in an E mail *

message dated June 20, 1997. subjert E0 and Instrument Accuracy, in order to address the issues raised in the June 20 E mail message, a review of instrument setpoint calculations was performed by licensee instrumentation and controls (l&C)

engineers. . The review was documented in ESR 97-000426, which was

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discussed in paragraph El.3. above. The inspectors also reviewed various instrument setpoint calculations (documented in paragraph

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E1.2. above) and determined that E0 accuracy has been aroperly considered in the instrument setpoint calculations. T1e ins)ectors had no further questions regarding instrument setpoint metloaology or accuracy at this tim E JClosed) Violation item 50-325(324)/97-02-08. Failure to Imolement ao nsoection Procram for Safety-Related Miscellaneous Structural Steel The licensee responded to this violation in letters dated

April 30. 1997, and June 26. 1997 Subject: Reply to Notice of Violation. The licensee's corrective actions included revision of Specification 248-107 and review of other specifications to assure OC inspection criteria required by applicable codes and standards referenced in the UFSAR had been included in the specification Specifications reviewed included the following: 248-117 - Installation of Piping Systems: 048 012 - Installation of Electrical Cables: 006 001

- Design. Testing & Inspection of Concrete Mixes. Concrete Materials and High-Strength Bolts: 005-005 - Design. Testing, & Inspection of Concrete Mixes. Concrete Materials: 013 001 - Concrete Work: and 018-00 Miscellaneous Stee Additional corrective actions included inspection of a sample of safety related high strength bolts installed using Specification 248-107. The inspectors reviewed the results of the structural steel inspectior.s which were documented in ESR 97-00085.

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hiscellaneous Structural Steel Connection Inspections. The licensee also revised procedure MMP-013. to incorporate the specification 248-107 changes and trained OC. engineering and planning personnel on the changes to specification 248-107 which now require additional QC inspections. The inspectors reviewed records which documented inspections performed for selected USl A-46 modifications completcd on Unit 1 during the Fall.1997 refueling outage and verified the structural steel inspections were completed in accordance with the revised procedure Ee.5 (Closed) Violation item 50-325(324)/97 08-07. Failure to initiate Condition Reports to [,0cument Nonconformina E0 Items I

. The licensee September 2. 199 reshonded Subject: Reply to this violation to Notice of Violation. in a letter The dated

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licensee's corrective actions included training of E0 personnel on the corrective action program, a review of che E0 walkdown data sheets to identify any potential nonconforming conditions which had not been previously identified and dispositioned, and organizational changes to improve management o"ersight in the E0 group. CR 97-02367 was initiated by the licensee on July 3. 1997 to document and disposition the two s)ecific examples of failure to initiate CRs identified by NRC. Tie inspectors ceviewed the CR closecut records (CR was closed on December 14, 1997) and the licensee's corrective actions and verified that the actions were completed in accordance with the licensee's violation respons IV. Plant SuppEt R1 Radiological Protection and Chemistry Controls RI.1 Use of locks to Control Access a. Insnection Stone (71750)

The inspector verified a selected sampling of doors required to be locked, by plant TSs and procedures, fc r the purpose of radiation protection, b. Observations and Findinas The inspector reviewed Environmental & Radiological Control 0E&RC-004 Control of Locked High Radiation and Very High Radiation Areas, to determine the controls used to lock high radiation area doors and barriers. The inspector located a sampling of the locked high radiation area doors specified in OE&RC-0040 and tested them to ensure that they were locked. The ins)ector found that all the locked high radiation doors tested were locced, c. Conclusions The ins)ector determined that each of the locked high radiation area dcors w11ch were checked were locke The ins)ector concluded that the licensee is satisfactorily controll1ng locked ligh radiation areas in the plant.

R1.2 Radioactive Material Controls a. insoection Scqoe (71750)

The inspector conducted a housekeeping tour of radioactive material storage areas located in outside areas within the protected area, b, Observations and Findinas The inspector found several poor radiological work practices in the radiological material (RAM) storage area located aojacent to the

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Radiological Maintenance Service Building in the northwest corner of the p.*otected area. A bucket containing scaffolding brackets was half filled with water and was labeled as radioactive material. The label identified the brackets as contaminated. This practice had the possibility of allowing the potentially contaminated water to cause a spread of contamination in an RAM storage area. There was also scaffolding identified as radioactive lying unprotected on a wooden pa l l e'. .

The ~icensee conducted a walkdown of this area and the radiological service building, and identified multiple conditions requiring actio These items were identified in CR 97-04122. Nonconforming Material Condition, Conclusions The inspector determined that several poor radiological work practices existed in a radioactive material storage are S2 Status of Security Facilities and Equipment c2.1 Plant Access Control and Physical Barriers Inspection Scone (71750)

The inspector verified the status and condition of the protected area fencing, Qbser"ations Jnd Findinas The inspector performed a walkdown of the protected area fence. The fence was inspected for integrity such as corrosion on the posts, gaps in the fence, and general adequacy. The inepector noted no deficiencies, Conclusions The inspector found the status and condition of the protected area fence to be satisfactor F1 Control of Fire Protection Activities F1.1 Operability of Fire Protection Facilities and Eauioment Ipsoection Scone (64704)

The inspector reviewed the operation's fire protection daily impairment reports on the facility's fire protection systems and features, and inspected these items to determine the performance trends and the material conditions of this equipmen .__ _ _ __.- _ _ _ _ _ _._ _ _ _ _ . . _ - -

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b. - Observations end Findinas  !

A review of the Loss Prevention Unit daily Impairment Reports for-December 8 - 11, 1997.- indicated that the following fire-protection

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components or systems for safety related areas were out of service: ,

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fire Protection-System ~ Number of Imoairments Thermo-Lag Fire Barriers 2 Fire Doors 6

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Fire: Detection System - 3 1 Fire Suppression System 3 The inspector noted that a number of- fire doors were out of servic This high number was attributed to the current DG building fire door corrective action (door replacement and repairs) that was in process for discrepancies identified during a June 1997 licensee self assessment of the fire protection program.- Appropriate compensatory measures had been

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-1mplemented for the fire protection features which were out of servic The impairment status report provided the licensee with a good means of identifying out-of-service fire protection equipment and provided status

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for compensatory measures that were implemented. The corrective maintenance on degraded fire protection systems was accomplished in a

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timely manner,

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During the plant tours the inspector noted that the maintenance and material condition of the fire protection equipment were satisfactor Conclusions

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Correstive maintenance on degraded fire protection systems was accomplished in a. timely manner.>The maintenance and material condition

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of the fire protection equipment and features were satisfactor ,

F2 Status of Fire Protection Facilities and Equipment F2.1 E3ssive Fire Barriers

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Fire barriers ~ include penetration seals. wraps, walls. structural member--

fire resistanticoatings.. doors, dampers. etc. Fire barriers are used to-prevent the spread of fire and to protect redundant safe shutdown equipmen Laboratory testing of fire barrier materials is done only on a-limited range of test assemblies. In-)lant-installations can vary

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from the tested configurations. -Under tie provisions of Generic Letter (GL) 86-10. Implementation of Fire Protection Requirements, licensees are permitted to develop engineering evaluations justifying such deviation ;

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40 Silicone foam Penetration Seals a. Inspection Stone (64704)

The inspector reviewed the fire barrier ,,ilicone foam penetration seal design end testing. The inspector compared as-built fire barrier silicone foam penetratioh seals to fire endurance test configurations to verify that the as-built penetration seals reviewed were qualified by appropriate fire endurance tests, representative of, and bounded by, the design and construction of the fire endurance test specimens. During plant walkdowns the inspector observed the installation configurations of selected fire barrier silicone foam 3enetration seals to unfirm that the licensee had established an accepta)le design basis for those fire barriers used to separate safe shutdown function b. Observations and Findinas The inspector reviewed the fire barrier seal design and testing for six

of ten fire barrier silicone foam seal penetrations, Additional reviews I are documented in NRC 1Rs 50-325(324)/92-31, 93 08. and 93-3 The inspector reviewed Brunswick Specification No. 118 003, Revision Selection and Installation of Fire Barrier Penetration Seals

Corrective Maintenance Procedure OCMP-010, Revision 2, Installation of Fire Barrier, Pressure Boundary Penetration and Water / Moisture Seals: Fire Protection Procedure FFP-015. Revision 23, Fire Barrier Penetration Seal Work Control: Periodic Test OPT-34.6.7.12. Revision 3. Fire Barrier Penetration Seals: and the Fire Hazards Analysis (FHA) for the location and description of fire areas: and assessed the licensee's supporting technical justification and any available engineering evaluations for the sampled silicone foam type oenetration seals, The inspector's review focused on verifying that the following design and installation paramaters for the as-built configurations were adequately bounded and justified by the licensee's engineering evaluations:

. penetration opening sizes e thermal mass of penetrating items e clearances of penetrating items e unexposed surface temperatures The insoector found that penetration seal field verification documentation was maintained by the license However, the seal installers * qualification and training records were not readily available for review. Although the installation and repair procedures for penetration seals provided adequate guidance to ensure materials were installed per design requirements, the inspector could not verify that the established surveillance recuirements included vendor recommendations for inspection and icentification of silicone foam seal aging and shrinkag _ _ _ - - - _ - - . - - - . - - - - - - - . . - -

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The licensee was unable to locate the penetration seal testing documentation and the vtador data for the tested prototype configurations or GL 8610 engineering evaluation documentation that evahated the adequacy of the deviations from a tested fire barrier contiguration. This does not satisfy the guidar.ce of GL 8610. The i

licensee stated that industry documentation is available to support silicone foam penetration seal installations at Brunswick but the

.tiformation was maintained at other Carolina Power and Light (CP&L)

site The penetration seal testing documentation, vendor data and inspection criteria, installer qualification and training records, and evaluations of deviations from tested fire barrier configurations will be reviewed during a subsequent NRC inspection. This is identified as IFl 50 325 (324)/97-13 04. Review of Licensee Records and Engineering Evaluations to Establish the Fire Resistant Capabilities of Fire Rated Silicone foam Penetration Seals, Conclusions The inspector concluded that silicone foam penetration seal field verification documentation was maintained by the licensee. The installation and repair procedures for penetration seals provided adequate guidance to ensure that materials were installed per design requirements. However, the designs were not supported by seal testing documentation, vendor data and inspection criteria, installer qualification and training records, and engineering evaluations that satisfy the guidance of GL 8610 for deviations from the fire barrier configuration qualified by test F2.3 Fire Doors Insnection Scone (64704)

The inspector reviewed UFSAR Section 9.5.1.4.3.4.b. Fire Doors, and performed plant walkdowns to verify that the UFSAR wording was consistent with the observed plant installation configurations for selected fire doors installed in fire barriers used to separate safe shutdown function Observations and Findinas The UFSAR St.ction 9.5.1.4.3.4.b. Fire Doors, states that doors and frames are either listed by a national testing laboratory or are constructed similar to listed doors and frames. All doors and frames have been evaluated to assure satisfactory ratings. Results are documented in the FHA. During the review of the FHA the inspector identified that, while evaluations of fire doors and frames existed. the-licensee failed to document their results in the FHA. which is section 9.5.1.5 of the UFSA , r ,,- ~ , - , , - , , - , - - - - - - - , -v ,

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After discussions with the licensee. CR 97-04103 was issued to track the l failure to provide the results of fire door evaluations in the FH This UFSAR discrepancy was identified by the inspector and is discussed in Section F A review of the surveillance ins)ection and testing procedures for fire doors was performed to confirm tlat the licensee specified fire door clearance acce)tance criteria was in accordance with the guidance of National Fire )rotection Association (NFPA) 80. Standard for Fire Doors and Fire Windows. On December 10. 1997. the inspector observed ongoing door replacement and repair activities for fire doors in the DG building. No discrepancies were identified, Conclusions I

The inspector concluded that fire door surveillance prc:edures and acceptance criteria for verification o' fire daor clearances were in accordance with NFPA quidanc Howevr a UFSAR discrepancy associated documentation of fire door and frame eu.uations was identifie F2.4 UFSAR Review 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 inspector reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspector verified that the UFSAR wording was consistent with the observed plant practices, procedures, and/or parameter The inspector reviewed UFSAR Section 9.5.1.4.3.4.b, Fire Doors, as part of the fire protection program review activiti u , An inconsistency was noted in that the licensee failed to document the results of evaluations of fire doors and frames in the FHA which is section 9.5.1.5 of the UFSA This issue is discussed in Section F2.3. This item will be identified as part of URI 50-325(324)/97-13-05. UFSAR Discrepancy Fire Door F3 Fire Protection Procedures and Documentation F3.1 Fire Protection Procedures Insoection Scone (64704)

The inspector evaluated the adequacy and implementation of the licensee s Eire Protection Program described in the UFSAR and in Plant Operating Manual Fire Protection Procedure OPLP 01. Revision 6. Fire Protection Program Document. In addition a comparison was made of the program to selected NRC Safety Evaluation Reports which ap3 roved the station fire protection program. The inspector reviewed t7e following

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procedures for compliance with the NRC requirements and guidelines:

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OPLP-01. Revision 6. Fire Protection Program Document

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0FLP-01.1. Revision 12. Fire Protection Commitment Document

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OPLP-01.2 Revision 10. Fire Protection System Operabilit Action, and Surveillance Requirements

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FPP 005. Revision 15. Fire Watch Program

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FPP-008. Revision 24. Fire Protection Weekly inspection

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FPP 013. Revision 25. Transient Fire Load Evaluation

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FPP 014. Revision 17. Control of Combustible. Transient Fire loads

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and Ignition Sources Plant tours were also performed to assess procedure complian Obji.ervations and Findinas The listed procedures were issued to implement the facility's fire protection progra These procedures contained requirements for program administration, controls over combust 1 oles arid ignition sources, fire watch duties and training, and operability requirements for fire i

protection systems and features. The 3rocedures were well written and met the licensee's commitments to the 1R General plant walkdown inspections were perfoimed by the inspector to verify: acceptable housekeeping; compliance with the ]lant's fire prevention procedures such as control of transient com)ustibles:

operability of the fire detection and suppression systems: emergency '

lighting: and installation and operability of fire barriers, fire stop and penetration seals (fire doors, dampers, electrical penetration seals, etc.), Conclusions General housekeeping was satisfactory. Fire retardant plastic sheeting and film materials were being used. Lubricants and oils were properly stored in approved safety containers. Controls for combustible gas bulk storage and cutting and welding operations were being enforce Controls were being properly maintained for limiting t' alsient combustibles in designated separation zones and oth' restricted plant

. area F5 Fire Protection Staff Training and Qualification F5.1 EireBrioade Insoection Stone (64704)

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44 The inspector reviewed the fire brigade organization and training program for compliance with the NRC guidelines and program requirement ' Observations and Findinos

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The organization and training requirements for the plant fire brigade were established by Fire Protection Procedure 0FPP-051. Loss Prevention Emergency Response 0ualification/ Training and Drill Program. The fire brigade for each of five shifts was composed of an operations support fire protection technician shift incident commander (fire brigade leader) and at least four additional brigade members consisting of Auxiliary Operators. Chemistry Technicians and Maintenance personne Each operations shift also had a Senior Reactor Operator / Reactor  :

Operator Fire Brigade Advisor assigned to respond tr ' ires with the fire brigad As of the date of this inspection, there were a total of 48 fire brigade members 26 from operations and 22 from E&RC and Maintenance on the pic t fire brigade. The inspector verified that sufficient shift personal were available to staff each shift's fire brigade with at least five qualified fire brigade member A review of the training records for the fire brigade members indicated that the training, drill, respiratory and physical examination requirements for each active member were up to date and met the established site training requirement Fire Briaade Ecuioment:

The fire brigade turnout gear and a fire response vehicle and trailer with fire brigade equi) ment was stored in the Operations / Fire Protection equipment building. T1e_ inspector's inventory of the fire brigade equipment indicated that a sufficient number of turnout gear, consisting of coats, pants, boots, helmets, etc. , was provided to equip the fire brigade members expected to respond in the event of a fire or other emergency. The fire brigade turnout i., ear and fire fighting equipment were being properly maintained, Conclusions The fire brigade organization and qualification training met the-requirements of the site proced . Fire brigade turnout gear and fire fighting eouipment were being properly maintaine __ .__ -

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l e l 45 j F6 Fire Protection Organization and Administration F6.1 Fire Protection Mananement and OraanizatioD a. Inspection Scope (64704),

The licensee's management and administration of the facility's fire protection program were reviewed for compliance with the commitments to the NRC and to current NRC guideline b. Observations and Findinos During this report period the licensee reassigned the responsibility ior the administration and implementation of the fire protection program from the previous Loss Prevention Unit (LPU) to the Operations Shift and Support organizations. The LPU organization was dissolve The designated onsite manager responsible for the administration and implementation of the fire protection program was the Operations Manager, This responsibility had been delegated to the Operations Support Superintendent. The Operations Support Superintendent was responsible for the station fire protection program, fire protection surveillance testing of fire protection systems and equipment, and ensuring that the aopropriate fire prevention procedures and fire b:'igade programs were implemented. A Fire Protection Program C0ordinator reported to the Operations Support Superintenden Maintenarice of the 31 ant fire protection equipment was performed by the Maintenance Unit. cire protection related training was planned and conducted by the Brunswick Training Se: tion. Coordination of the station's fire protection program commitments and engineering functions was provided by a fire protection system engineer in the Brunswick Engineering Support Section, c. Conclusions The coordination and oversight of the facility's fire protection program had been reassigned from the previous LPU organization to Shift Operations. The new organizational structure met NRC guidelines and the licensee's fire protection program requirement F7 Quality Assurance in Fire Protection Activities F7.1 Fire Protection Audits a. Insoection Scope (64704)

The following audit report and the plant response to the issues were reviewed:

- Nuclear Assessment Section (NAS) Report B-FP-97-01. Brunswick Fire Protection and Loss Prevention Unit Assessment, dated August 1. 199 i

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b. Observations and Findinas The licensee's Nuclear Assessment Section performed an assessment of the fire protection program and LPU on June 16-27. 1997. The report for this assessment was Re) ort No. B FP-97 01. The assessment team determined that the LPJ fire prevention and fire response activities were adequate; however, its implementation of the fire protection

)rogram was ineffective based on a number of program elements found to

)e below acceptable standards. Findings from these assessments were categorized as strengths, issues, or weaknesses. The assessment report identified six program element issues and one weaknes The inspector reviewed the final audit report, the licensee's response to the identified issues. the planned corrective actions, and the NAS evaluation of the response adequac This NAS assessment of the facility's fire protection program was comprehensive and effective in identifying fire protection program performance deficiencies to management. The audit team identified deficiencies in LPU'c management oversight of fire protection procedures, training, problem identification, procedure performance standards, corrective actions, and personriel safety. Corrective actions in response to the identified issues were substantial and included a fire protection reorganization to integrate the former LPU organization into the shift Operations and Operations Sup) ort organizations under direct management of the Operations Support Manager, c. Conclusions The 1997 Nuclear Assessment Section assessment of tite facility's fire protection program was comprehensive and was effective in identifying fire protection program performance deficiencies to management. Planned corrective actions in response to the audit issues were substantial and included a fire protection reorganizatio Manaaetment Meetinas XI Exit Meeting Summary The inspector presented the inspection results to members of licensee management at tN conclusion of the ins)ection on January 8,1998. Post inspection briefings were conducted on )ecember 12, 1997. The licensee acknowledged the findings presented. The licensee stated that they had not determined if clearance records were required QA records.

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PARTIAL LIST OF PERSONS CONTACTED Licensee A. Brittain. Manager Security M. Christinziano, Manager Environmental and Radit lon Control

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W. Dorman. Supervisor Licensing and Regulatory Programs N. Gannon. Manager Maintenance J. Gawron. Manager Nuclear Assessment Section S. Hinnant. Vice President. Brunswick Steam Electric Plant K. Jury. Manager Regulatory Affairs R. Krich, Chief Engineer. Nuclear Engineering Department B. Lindgren. Manager Site Su) port Services J. Lyash. Manager Brunswick Engineering Support Section R. Mullis. Manager Operations Other licensee employees or contractors included office, operation, maintenance. chemistry, radiation, and corporate personne _

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INSPECTION PROCEDURES USED IP 37550: Engineering IP 37551: Onsite Eng11eering IP 6172 Surveillance Observations IP 62700: Maintenance Program implementation IP 62707: Maintenance Observations IP 64704: Fire Protection IP 71707: Plant 0)erations IP 71714: Freeze )rotection IP 71750: Plant Support Activities IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92901: Followup - Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-325(324)/97-13-01 VIO Failure to Retain TS Required QA Record (Section 07.2)

50 325(324)/97 13-02 VIO Inadequate Procedure for the Conduct of E0 Preventive Maintenance (Section M1.3, El.4.b.6)

50 325/97-13-03 VIO Failure to Note Abnormal TS Surveillance Values (Section M3.1)

50 325(324)/97-13-04 IFl Review of Licensee Records and Engineering Evaluations to Establish the Fire Resistant Capabilities of Fire Rated Silicone foam Penetration Seals (Section F2.2)

50-325(324)/97-13-05 URI UFSAR Discrepancy Fire Doors (Section F2.4)

50 325(324)/97-13 06 IFl Revisions to Procedure EGR-NGGC-0153 (Section El.1)

50-325(324)/97-13-07 IFl Review Technical Evaluation of Terminal Block Current Leakayc and the Effect on EQ Equipmen (Section El.4.b.4)

Closed 50-325/96-15-01 URI Vessel Disassembly Without Secondary Containment (Section 08.1)

50-325(324)/97 02-01 V10 Locked Valve Out of Position (Section 08.2)

50-325/97 12 03 URI Recirculation Pump Run back (Section 08.3)

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50-325(324)97-12-04 URI Diesel Generator Low Voltage Auto Start Defeated (Section 08.4)

50 325(324)/96-017-00 LER Invalid Loss of Coolant Accident (Section M8.1)

50_-325/97_009-00 LER Missed Increased Frequency inservice Testing Requirement (Section M8.2)

50-325/97-001-00 LER Rod Block Monitor Surveillance inadequacy (Section M8.3)

50-325(324)/95-022 00 LER High Pressure Coolant injection System Discharge Flow Element Gasket Leak (Section M8.4)

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50 325/97-05-02 IFl Abnormal CS Sp)arger Break Detector Indication (Section M /97-05-03 VIO Inadequate CS Surveillance Procedure (Section M8.5)

50 325/97-02 LER Core Spray Header Differential Pressure Instrumentation Inoperable (Section M8.5)

50-325(324)/97-02-04 VIO Failure to implement Requirements of the Maintenance Rule (Section M8.6)

50-325(324)/97-08-04 URI Control of EDBS Information (Section E8.1)

50-325(324)/97-04 LER Spent Fuel Shipping Cask Handling Activities (Section E8.2)

50-325(324)/96-14-05 IFI Effect of EQ Accuracy on Instrument Setpoint Calculations (Section E8.3)

50-325(324)/97-02-08 VIO Failure to Implement an Inspection Program for Safety-Related Miscellaneous Structural Steel (Section E8.4)

50-325(324)/97-08 07 VIO Failure to Initiate Condition Reports to Document Nonconforming EQ ltems (Section E8.5)

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