IR 05000456/1999012
| ML20211P236 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 09/03/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20211P195 | List: |
| References | |
| 50-456-99-12, 50-457-99-12, NUDOCS 9909130168 | |
| Download: ML20211P236 (15) | |
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t U.S. NUCLEAR REGULATORY COMMISSION REGIONlli
Docket Nos: 50-456, 50-457 License Nos: NPF-72, NPF-77 Report No: 50-456/99012(DRP); 50-457/99012(DRP)
Licensee: Commonwealth Edison Company Facility: Braidwood Nuclear Plant, Units 1 and 2 Location: RR #1, Box 84 Braceville,IL 60407 Dates: July 7 through August 16,1999 Inspectors: C. Phillips, Senior Resident inspector J. Adams, Resident inspector D. Pelton, Resident inspector J. Roman, Illinois Department of Nuclear Safety Approved by: Michael J. Jordan, Chief Reactor Projects Branch 3 Division of Reactor Projects l
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9909130168 990903 PDR ADOCK 05000456 Q PDR
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EXECUTIVE SUMMARY Braidwood Nuclear Plant, Units 1 and 2 NRC Inspection Report 50-456/99012(DRP); 50-457/99012(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period of resident inspection from July 7 through August 16,199 Ooerations
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The Inspectors observed control room operators throughout the inspection period and concluded that operators routinely performed good turnover briefings, control board operations, response to alarms, and three-way communications.' The inspectors particularly noted that operators closely monitored the ultimate heat sink temperature which was approaching the Technical Specification limit due to an exter.ded period of hot weather. The unit supervisors demonstrated good performance in the minimization of control room distractions, in the direction of personnel, in the conduct of briefings, and in the control of evolutions. (Section O1.1)
Maintenance
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The inspectors observed the performance of five surveillance tests. The inspectors concluded that the surveillance tests adequately tested the system, the operators followed the procedures, and that the procedures included the required testing discussed in the Technical Specifications. (Section M1.1)
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The inspectors observed portions of maintenance activities associated with the 2B residual heat removal pump motor bearing oil change out, and the inspection of limitorque valve actuator for the closed cooling water outlet from the 2B residual heat -
removal heat exchanger outlet valve. The inspectors concluded that activities were performed in accordance with the applicable procedures, the procedures provided the necessary information to perform the work, and that maintenance personnel were knowledgeable of the associated limiting conditions for operations (LCO). The inspectors concluded that the entry into and exit from LCO were properly entered into operating logs. (Section M1.2)
Enaineerina
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The inspectors concluded that the licensee properly performed a 10 CFR 50.59 screening for a temporary modification to the Unit 1 annunciator power supplies and properly performed a 10 CFR 50.59 safety evaluation for the upgrading of three safety j
injection relief valves. The inspectors concluded that the licensee's justifications were technically correct and referenced applicable vendor analyses, Updated Final Safety Analysis Report, Technical Specifications, and American Society of Mechanical Engineers Boiler and Pressure Vessel Code. (Section E1.1)
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The inspectors concluded that the operability evaluation conceming the seismic qualification of relays and large break loss of coolant accident computer code error reflected sound engineering judgement and safety focus, and were performed in accordance with the appropriate procedure. The inspectors concluded that corrective
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F actions were entered and 'were being tracked in the licensee's action tracking system,-
and the required compensatory action was properly implemented and the Technical Specification required report was submitted. (Section E1.2)
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Plant Suocort
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. The inspectors' concluded that the protected area fence and isolation zone were properly -
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l maintained. Although the licensee identified a trend conceming individuals leaving -
security doors unsecured after use, the inspectors concluded that immediate actions taken by security minimized the risk of unauthorized personnel gaining access to j secured areas within the plant.- (Section S2.1)
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Report Details Summarv of Plant Status
. Units 1 'and 2 entered the period at full power and remained at or near full power throughout the
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i-1. Operations
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01 Conduct of Operations 01.1 Control Room Observations Inspection Scope (71707)
The inspectors observed the conduct of operations during. normal operating conditions, during maintenance activities, and during the performance of surveillance tests. The inspectors performed extended observations in the main control room and the radioactive waste control room. The inspectors interviewed nuclear station operators, unit supervisors, and shift managers with regard to ongoing activitie Observations and Findinas ,
The inspectors observed main control room operators throughout the inspection perio '
The inspectors noted that the nuclear station operators were attentive, used operating procedures, used self-checks when manipulating equipment, and used three-way communications. The operators promptly addressed alarms, referred to the annunciator response procedures, and informed supervisors of alarms. The inspectors noted that unit supervisors minimized control room distractions, clearly directed personnel, clearly
. communicated personnel assignments and plant status during shift briefings, and effectively controlled evolutions. The inspectors particularly noted that operators closely monitored the ultimate heat sink temperature which was approaching the Technical
. Specification limit due to an extended period of hot weathe The inspectors observed three annunciators illuminated in the radioactive waste control room. These annunciators were the combined high and low level alarms for the auxiliary building equipment drain tank, the auxiliary building floor drain tank, and the turbine building floor drain tank. The inspectors determined that the level in each tank was below the low level alarm setpoint. Discussion with operations personnel (including the shift manager, field supervisor, and radioactive waste control room operators)
indicated that the tanks were maintained in this condition when pumped down and not in use per operations management direction. The low level alarms normally provided an indication to operators of inadvertent tank drainage. Although no loss of control of tank level had occurred, the inspectors were concemed about the potential for a lack of
- operator sensitivity to alarms as a result of operating with the low level alarms continuously illuminated. Operations management acknowledged the inspectors concems and agreed to revisit their position conceming operating with panel annunciators actuate c ,
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, Conclusions The inspectors observed control room operators throughout the inspection period and concluded that operators routinely performed good tumover briefings, control board operations, response to alarms, and three-way communications. The inspectors particularly noted that operators closely monitored the ultimate heat sink temperature
. which was approaching the Technical Specification limit due to an extended period of hot weather. The unit supervisors demonstrated good performance in the minimization of control room distractions, in the direction of personnel, in the conduct of briefings, and in the control of evolution Miscellaneous Operations issues (92901)
08.1 ' (Closed) Licensee Event Report (LER) 50-456/97001-00 "Both Main Control Room Chillers Inoperable due to Equipment Failure and Surveillance Testing Resulting in Entry into 3.0.3." in April 1997, the licensee placed the OA main control room chiller in
" equipment test" in support of diesel generator testing. Placing the OA main control room chiller in equipment test rendered it inoperable requiring entry into Technical Specification limiting condition for operations (LCO) 7.6-1 A. Once diesel generator
' testing was completed, the licensee attempted to restore the OA main control room chiller to operable status by securing the OB main control room chiller and starting the OA main control room chiller. While securing the OB main control room chiller, essential service water (SX) system valve OSX-063B (SX system supply to the OB chiller) started to close as expected but then position indication was lost. An operator was dispatched to investigate the problem with OSX-0638. The operator discovered that the valve's breaker had tripped on thermal overload. The OB main control room cooler was declared inoperable and LCO's 7.6-1 A and 3.0.3 were entered. Within approximately 7 minutes of entering LCO 3.0.3, the OA main control room chiller was started, declared operable, and LCO 3.0.3 was exited. The licensee determined that valve OSX-063B's
. gear box had been leaking oil onto the valve's position indication limit switch contact The leaking oil covered contacts which temporarily prevented the valve's motor operator from de-energizing once the valve had shut. This resulted in an additional current draw and a thermal overload condition which tripped the breaker. The inspectors observed valves OSX-063A and OSX-063B, reviewed operability determination 97-100, and discussed valve modification plans with system engineering personnel. The inspectors !
agreed with the licensee's determination of operability. Plans to modify these valves to l prevent recurrence of thermal overload trips were being tracked via engineering request i ER9802178. This item is close .2 (Closed) LER 50-457/97002-00 " Manually Opened Reactor Trip Breakers Due to j Decoder and Encoder Control Card Failure." in September,1997, while in Mode 3, operators were performing rod control system testing which required the reactor trip i
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breakers to be in the closed position. While control rod bank "C" was being withdrawn with demand group step counters at 48 steps, operators observed the digital rod ;
position indication for control rod B-8 jump to 96 steps. The surveillance procedure requires opening the reactor trip breakers if, at any time during control rod movement,
. the digital rod position indication versus demand indication disagrees by greater than or equal to 24 steps. Operators immediately opened the reactor trip breakers, entered the LCO for Technical Specification 3.3.1.3, and made a 4-hour Event Notification due to manually opening the reactor trip breakers. The licensee's subsequent investigation determined that a circuit board on the digital rod position indication Train B decoder and
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encoder control had failed. The licensee's corrective actions included entering the event into their corrective actions program, replacing the defective card, and re-performing the surveillance prior to unit stad-up. The inspector reviewed the operator's actions and determined that they were appropriate. The inspectors determined that the licensee repor1ed the event within the required time frame, and took appropriate corrective actions. This issue is close .3 (Closed) LER 50-456/97003-00. " Manually Opened Reactor trip Breakers Due To a Control Rod Drive Bank Overlap Malfunction." In May 1997, the licensee was performing a start-up of Unit 1. While manually withdrawing control banks "C" and "D,"
the nuclear station operator noted that control bank "C" was.not moving. Bank "C" stopped at 218 steps verses the expected withdrawal point of 228 steps. The unit supervisor directed the nuclear station operator to manually trip the Unit 1 reactor. The licensee determined that an improper rod bank overlap condition (overlap point S6)
existed due to oxidation on switch contacts in the associated thumbwheel switch for overlap point S6 circuitry. The effected switch was cycled to clean the oxidation from the contacts. The rod control system was_ subsequently retested, in accordance with Braidwood Engineering Surveillance Procedure (BwVS) 500-2, " Checkout of the Bank Overlap Unit," with no recurrence of the overlap problem. The licensee modified BwVS 500-2 to include a section which exercises all six of the bcA overlap circuitry thumbwheel switches prior to testing. The inspectors reviewed the latest revision of BwVS 500-2 and determined that the procedure did include a section for exercising the bank overlap thumbwheel switches. In January 1998, a similar thumbwheel switch oxidation problem occurred on Unit 2. The licensee determined that the oxidation of the contacts, being an age-related degradation, warranted periodic switch replacemen The licensee entered this periodic switch replacement into their "predefine" program which ensured that the thumbwheel switches would be periodically replaced (once every six refueling outages). This item is close D8.4 (Closed) Unresolved item 50 456/99007-01(DRP) " Notice of Enforcement Discretion For Gas Pockets identified in The Unit 1 Emergency Core Cooling System (ECCS)
- Piping." The inspectors reviewed the root cause evaluation, problem identification forms (PlFs), and the corrective actions regarding the ECCS discharge piping gas pocket formation identified on May 13,1999. On May 13, the licensee discovered that gas
. pockets had formed in an area common to both trains of ECCS system discharge piping. This resulted in the failure of Technical Specification surveillance procedure
' 1BwOSR 3.5.2.2-2, "ECCS Venting and Valve Alignment Surveillance," Revision O. The licensee entered Technical Specification 3.0.3 and requested a Notice of Enforcement Discretion from the NRC to allow continued operation of Unit 1. The licensee performed a root cause evaluation. The licensee determined that the gas was nitrogen that had leaked past check valves downstream of a safety injection system accumulator. The root cause evaluation and corrective actions were thorough and well documented. The licensee identified gas bubbles coming out of solution during surveillance testing (1BwOSR 3.5.2.2-2, "ECCS Venting and Valve Alignment Surveillance, Revision 0) on
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February 19,1999, and again during the same surveillance on April 19,1999. The qualitative acceptance criteria of 1BwOSR3.5.2.2-2 stated that if a significant amount of
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gas was found while venting then the ECCS discharge piping needed to be ultrasonically tested. When gas bubbles were identified during the venting surveillance test in February the ECCS discharge piping was ultrasonically tested. However, when
' the gas bubbles were identified in April operations management did not consider the presence of gas bubbles to be a degraded condition and a FIF was not writte k
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Consequently no root cause or apparent cause evaluation was performed. Also, there J
was no information (i.e.', remarks) in the surveillance test procedure that indicated that gas bubbles had been observed during the venting of the ECCS system. The inspectors determined that operators had identified the presence of gas bubbles in the ECCS system through interviews with operators involved with the surveillance testin ) The inspectors concluded that the qualitative acceptance criteria had not been l consistently addressed. The inspectors discussed this issue with operations j management who agreed to review the acceptance criteria specified in l- 1BwOSR3.5.2.2-2 to ensure it was sufficient to ensure consistent actions were taken in
. the event that the presence of gas bubbles are identified during ECCS system ventin This item is closed.
l 11. Maintenance M1 Conduct of Mainte' nance l
~ M1.1 Observation of Miscellaneous Surveillance Activities
' a. ' Inspection Scope (61726)
.The inspectors observed all or portions of the following surveillance activities:
- 2BwOSR 3.5.2.5, "ECCS Subsystem Automatic Valve Actuation 18-Month Surveillance," Revision 1;
2BwVSR 5.5.8.RH.2, "ASME (American Society of Mechanical Engineers)
Surveillance Requirements For Residual Heat Removal Pump 2RH01PB," i Revision OE1;
.- Braidwood Electrical Surveillance (BwHS) 4002-066, " Periodic Protective Relay Calibration," Revision 3; i
+ BwHS 4009-061, " Unit Two Diesel Driven Auxiliary Feedwater Pump Room and Day Tank Room Low Pressure CO 2[ Carbon Dioxide) System Actuation Surveillance," Revision 3; and . l
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- - 1BwOSR 3.3.2.7-6098, " Unit One ESFAS [ Engineered Safety Function Actuation System] Instrumentation Slave Relay Surveillance (Train B Automatic Safety injection - K609)," Revision ~ Observations and Findinas The inspectors observed the performance of the above listed surveillance tests. For each surveillance test, the inspectors obse'vedr the establishment of initial conditions required for the surveillance test, the operation of equipment, the communications
- between the licensed operators in the control room and non-licensed operators in the
!' auxiliary building, and the restoration of affected equipment. The inspectors determined I that these activities were performed in accordance with the applicable procedure. The inspectors reviewed the data obtained during the surveillance tests and noted that it met the required acceptance criteria specified in the surveillance test procedures. The 7 ,
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t inspectors also reviewed the associated portions of the Updated Final Safety Analysis Report and the Technical Specifications and determined that the surveillance test l procedures demonstrated the systems performed as designe Conclusions The inspectors observed the performance of five surveillance tests. The inspectors concluded that the surveillance tests adequately tested the system, the operators followed the procedures, and that the procedures included the required testing i- discussed in the Technical Specification M1.2 - Maintenance Activity Observations I Inspection Scope (62707)
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.. The inspectors observed all or portions of the following maintenance activities:
+- Work Request 970084831-01, "2B Residual Heat Removal Pump Motor Bearing Oil Change Out"; and
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Work Request 970097408-01, " Inspection of Limitorque Valve Actuator for Closed Cooling Water Outlet from the 2B Residual Heat Removal Heat Exchanger Outlet Valve 2CC94128." Observations and Findinas t
The inspectors attended the heightened-level-of-awareness meetings; reviewed the above work packages; walked down the work areas; questioned personnel conceming the scope of the work;' observed the establishment of required system conditions; and observed the use of foreign material exclusion controls. Based on the inspectors .
observation, there were no concems. The inspectors also reviewed the associated LCO and reviewed the control room operating logs for LCO entry and exit log entries. The inspectors noted no problems with the logbook documentatio Conclusions The inspectors observed portions of maintenance activities associated with the 2B residual heat removal pump motor bearing oil change out, and the inspection of ,
limitorque valve actuator for the closed cooling water outlet from the 2B residual heat i removal heat exchanger outlet valve. The inspectors concluded that activities were i performed in accordance with the applicable procedures, the procedures provided the necessary information to perform the work, and that maintenance personnel were knowledgeable of the associated LCO. The inspectors concluded that the entry into and exit from LCO were property entered into operating log ~ M8 . Miscellaneous Maintenance issues (92902) i
.M8.1 ~ (Closed) Unresolved item 50-456/457/98008-01(DRS) " Untested Safeguards Actuation Relay Contacts." In May of 1998, the licensee identified that contacts associated with the safeguards actuation relays were not tested. The purpose of the relays was to simultaneously (block) start the emergency core cooling pumps in the
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' event of a safety injection signalif offsite power is available. The licensee's position
. paper stated that the sequencing of the loads was credited.for in the Updated Final .
' Safety Analysis Report and that the block start capability was redundant and not required to be tested by Technical Specification 4.3.1,4.3.2, or 4.8.1. The inspectors reviewed the licensee's position paper and discuss the appropriateness of not testing .
the safeguards actuation relay contacts with the office of Nuclear Reactor Regulatio The inspectors determined that Technical Specification testing of the safeguards actuation relay was not required and that the safeguards actuation relay would not prevent the proper operation of the parallel sequencing relays. As a good engineering practice, the licensee implemented changes to the appropriate procedures for the testing of the safeguards actuation relay contacts. The inspectors reviewed these changes and noted the procedures contained steps to test the safeguards actuation relay contacts. No violations of Nuclear Regulatory Commission requirements were identified. This issue is close . Engineering E1 Conduct of Engineering E Review of Comoleted 10 CFR 50.59 Screenina and Safety Evaluation Inspection Scope (37551)
The inspectors reviewed one 10 CFR 50.59 screening and one safety evaluation:
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BRW-FCS-1999-832, " Installation of Temporary Modification 99-1-002, j Revision 3"; and i i
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93W-SE-1999-395, " Upgrade of Safety injection Relief Valves." )
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The inspectors reviewed Nuclear Station Procedure (NSP)-CC-3005, "10 CFR 50.59 Safety Evaluations Process," Revision Observations and Findings in BRW-FCS-1999-832, the licensee addressed the installation of a temporary
- modification to enhance the reliability of the Unit 1 annunciator power supplies. The licensee determined that the issue did not warrant a safety evaluation. Based on a review of the screening documentation, the inspectors determined that NSP-CC-3005 screening process was followed, applicable sections of the design basis were considered, and justifications were technically correct. In BRW-SE-1999-395, the licensee addressed the upgrading of three safety injection relief valves. The licensee concluded that no unreviewed safety question existed. The inspectors reviewed the safety evaluation and agreed with the licensee's conclusions. The inspectors determined that the licensee's justifications were technically correct and referenced the applicable sections of the Updated Final Safety Analysis Report, Technical
' Specifications, and ASME Boiler and Pressure Vessel Code.' The inspectors noted that the safety evaluation was completed in accordance with the NSP-CC-3005 safety evaluation process requirement _
- Conclusions
- The inspectors concluded that the_ licensee properly performed a 10 CFR 50.59
' screening for a temporary modification to the Unit 1 annunciator power supplies and
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properly performed a 10 CFR 50.59 safety evaluation for the upgrading of three safety injection relief valvesJ The inspectors concluded that the licensee's justifications were technically correct and referenced applicable vendor analyses, Updated Final Safet Analysis Report, Technical Specifications, and American Society of Mechanical Engineers Boiler and Pressure Vessel Cod E1.2 Operability Evaluation Reviews InsrMion Scooe (37551)
The inspectors reviewed the following documents:
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Operability Evaluation 99-018, " Seismic Qualification of Model AR and ARD Relays,".
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Operability Evaluation 99-019, "Large Break Loss of Coolant Accident Computer Code Error," and
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NSP CC-3001, " Operability Determination Process," Revision The inspectors discussed the operability evaluations with operations, site engineering, and system engineering personne b. . Observations and Findings The inspectors reviewed the documentation for each operability evaluations and initially were concemed that operability evaluation 99-018 lacked the necessary detail to provide
. reasonable assurance that the affected relays and their associated equipment would remain operable following a seismic event. tiowever, after discussing the issue with site engineering personnel and reviewing circuit and seismic analysis documentation, the l inspectors determined that the operability determination process requirements had been j me j l
The inspectors discussed the operability evaluations with operations personnel and were 1 initially concemed that two shift managers were not aware of the contents of the large j break loss of coolant accident computer code error operability evaluation (99-019). ,
However, the inspectors -letermined that the lack of awareness with operability i evaluation contents was of minor concem and was of little safety significance due to the
. fact that there were no operational contingency actions required to support continued operability. In addition the shift manager would be informed immediately upon discovery if the total peaking factor failed to meet surveillance test procedure acceptance criteria and would be able to take appropriate corrective action immediately as required by ,
-Technical Specification Action Statement 3.2.1 A. " Heat Flux Hot Channel Factor....Not l Within Limits" which requires a power reduction within 15 minute !
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The inspectors reviewed compensatory and corrective actions with engineering and
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_ operations personnel. The inspectors noted that corrective actions were identified and 10 l
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.. wore being tracked by the licensee's action tracking system. The inspectors verife' d
. that a compensatory action required by operability evaluation 99-019 to reduce the total
= peaking factor limit in the Core Operating Limits Report from 2.6 to 2.45 had been completed and had been reported to the Nuclear Regulatory Commission in accordance
. with Technical Specification 5.6. Conclusions The inspectors concluded that the operability evaluation concerning the seismic qualification of relays and large break loss of coolant accident computer code error reflected sound engineering judgement and safety focus, and were performed in accordance with the appropriate procedure. The inspectors concluded that corrective actions were entered and were being tracked in the licensee's action tracking system, and the required compensatory action was properly implemented and the Technical Specification required report was submitte E8 Miscellansous Engineering issues (92903) .
E (Closed) Violation 50 456/457/98014-03 " Loose Bolts on Emergency Diesel Generator Lubricating Oil Heat Exchangers." On August 18,1998, the inspectors observed that 10 of the 16 bolts on the 2A and one bolt on the 2B emergency diesel generator lubricating oil heat exchanger erid bells were loose. The licensee tightened
. the loose bolts, checked the other emergency diesel generator lubricating oil heat exchangers for loose bolts, performed an operability assessment, conducted a prompt investigation, entered the problem into their corrective action program, and performed a root cause analysis. The licensee identified two root causes for this event. The first was not establishing and using specific tightening values for the emergency diesel generator lube oil cooler end bell fasteners. The second was not performing periodic hot re-tightening of the emergency diesel generator lube oil cooler end bell fastener The licensee completed the implementation of corrective actions to prevent recurrenc The inspectors verified that the corrective actions have been effective in preventing recurrence of loose bolts on the diesel generator heat exchangers. This issue is close IV, Plant Suonort P8 Miscellaneous EP lasues (92904)
P (Closed) Inspection Followun item (IFI) 50-456/457/98007.ni " Staffing of the Operational Support Center (OSC) was Slow Following the Alert Declaration." In May 1998, the inspectors identified that staffing of the OSC was slow following the 1 declaration of a simulated Alert. Radiation protection personnel (notified via pagers) l
. arrived approximately 20 minutes after the Alert declaration. The remainder of the OSC personnel did not arrive until approximately 45 minutes after the Alert declaration. At that time, the inspectors determined that the people arriving late did not carry pagers 2 and had not heard a plant announcement concoming the declaration of the Alert. The licensee subsequently upgraded the stations public address system and provided additional pagers to all OSC management responders. The licensee also made OSC management responders responsible for ensuring back-up personnel were brought to the OSC upon receipt of a page of an emergency declaration. On August 4,1999, the inspectors observed the performance of a " table-top" training session during which an
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unannounced emergency declaration was made, including a group page. The
. Inspectors observed the staffing of the OSC and determined that within approximately 20 minutes of the emergency declaration, the OSC was staffed by 10 management
. Individuals and approximately 40 non-management personnel from radiation protection, chemistry, engineering, mechanical maintenance, electrical maintenance, and instrument maintenance departments.: Actions taken by the licensee to improve OSC staffing timeliness appeared to have been effective. This item is close S2- ' Status of Security Facilities and Equipment S Security Controls Insoection Scope (71750)
The inspectors reviewed Braidwood Station Memorandum Number 61, " Security Violations," observed vital area controls, verified the integrity of the protected area boundary, and the maintenance of the isolation zon Observations and Findinos The inspectors walked down the protected area fence and determined that the fence had no uncontrolled openings, was not damaged or degraded, and showed no signs of erosion at the base. The inspectors observed that the established isolation zone was
- free of objects. The inspectors observed personnel entering and exiting vital areas and determined that proper entry and exit requirements were followed and that security equipment in place functioned as intended. The station identified a trend with the control of security doors. The licensee identified numerous occurrences of doors being left unsecured after people pass through them. In accordance with Braidwood Station Memorandum Number 61, personnel were required to assure security doors were closed after use. The licensee began tracking this trend via a " trend" PlF in an attempt to determine the cause and corrective actions necessary to prevent recurrence. Despite the noted trend, the inspectors concluded that immediate actions taken by security
- minimized the risk of unauthorized personnel gaining access to secured areas within the plant, Conclusions The inspectors concluded that the protected area fence and isolation zone were properly maintained. Although the licensee identified a trend conceming individuals leaving security doors unsecured after use, the inspectors concluded that immediate actions taks;, by security minimized the risk of unauthorized personnel gaining access to secured arees'within the plan V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on August 16,1999. The licensee acknowledged the findings j presented. The inspectors asked the licensee whether any materials examined during the !
inspection should be considered proprietary. No proprietary information was identifie l
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
- J. Bailey, Nuclear Operator
- M. Cassidy, Regulatory Assurance - NRC Coordinator R. Graham, Work Control Manager
- L. Guthrie, Maintenance Manager ,
- A. Haeger, Radiation Protection Manager
- C. Herzog, Services Manager
- F. Lentine, Design Engineering Manager
- T. Luke, Engineering Manager
- K. Schwartz, Station Manager -
- T. Simpkin, Regulatory Assurance Manager T. Tulon, Site Vice President -
- R. Wegner, Operations Manager NBR
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- J Adams, Resident inspector
- M. Jordan, Chief, Reactor Projects Branch 3
- C. Phillips, Senior Resident inspector
.*D. Pelton, Resident inspector T. Tongue, Project Engineer ADBE J. Roman
- Denotes those who attended the exit interview conducted on August 16,199 l l
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INSPECTION PROCEDURES USED-IP 37551: Onsite Engineering
. IP 61726:- Surveillance Observations IP 62707: - Maintenance Observation -
IP 71707: Plant Operations IP 71750: Plant Support Activities .
IP 92901: Followup - Plant Operations -
IP 92902: Fo'lowup - Plant Maintenance -
IP 92903: Followup- Engineering IP 92904: Followup - Plant Support -
ITEMS OPENED, CLOSED, AND DISCUSSED Opened None-( Closed 50-456/97001-00 LER main control room chiller inoperable 50-457/97002-00 LER failure of circuit board 50-456/97003-00 LER thermal overloads trip 50-456/457/98007-01 IFl staffing of OSC slow 50-456/457/98008-01- URI failure to test relays 50-456/457/98014-03 VIO loose bolts on emergency diesel generator .50-456/99007-01- URI - NOED for gas pockets in 1ECCS piping Discussed lNone
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LIST OF ACRONYMS USED ASME American Society of Mechanical Engineers BwHS Braidwood Electrical Surveillance BwOSR Braidwood Operations Surveillance Requirement BwVSR Braidwood Engineering Surveillance Procedure Requirement CFR- Code of Federal Regulations ECCS Emergency Core Cooling System EP Emergency Preparedness IFl inspection Followup item LER Licensee Event Report-LCO- Limiting Condition for Operation NSP Nuclear Station Procedure NRC Nuclear Regulatory Commission OSC Operational Support Center PIF Problem Identification Form RP Radiation Protection RP&C Radiological Protection & Chemistry SX Essential Service Water URI Unresolved item VIO Violation 15 ,
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