IR 05000261/1998011
| ML14181B036 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 12/23/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14181B035 | List: |
| References | |
| 50-261-98-11, NUDOCS 9901210236 | |
| Download: ML14181B036 (15) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No:
50-261 License No:
DPR-23 Report No:
50-261/98-11 Licensee:
Carolina Power & Light (CP&L)
Facility:
H. B. Robinson Unit 2 Location:
3581 West Entrance Road Hartsville, SC 29550 Dates:
October 25 - December 5, 1998 Inspectors:
B. Desai, Senior Resident Inspector A. Hutto, Resident Inspector Approved by:
Brian R. Bonser, Chief Reactor Projects Branch 4 Division of Reactor Projects 9901210236 981223 Enclosure PDR ADOCK 05000261 G
EXECUTIVE SUMMARY H. B. Robinson Power Plant, Unit 2 NRC Integrated Inspection Report 50-261/98-11 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a six-week period of resident inspectio Operations
The conduct of operations was professional, risk informed, and safety-conscious (Section 01.1).
- A detailed system walkdown found that the Safety Injection System was appropriately configured to perform its design function (Section 02.1).
- The licensee program for cold weather protection was adequately implemented. Freeze protection panels and circuits, as well as, temporary enclosures and heaters were appropriately installed and maintained per procedure (Section 02.2).
- A clearance associated with a charging pump was correctly implemented. A review of existing clearances identified that clearances older than three months had been evaluated in accordance with licensee procedures and did not impact plant safety (Section 02.3).
- The onsite review functions of the Plant Nuclear Safety Committee (PNSC) were conducted in accordance with Technical Specifications (TSs). During the PNSC meetings topics were thoroughly discussed and evaluated (Section 07.1).
Maintenance
Maintenance activities were conducted in accordance with applicable work documents and procedures (Section M1.1).
- Surveillance testing activities observed were adequately performed (Section M2.1).
- During an Emergency Diesel Generator (EDG) fuel oil system flow test the Reactor Operator performing the test exhibited a good questioning attitude by identifying a procedural step that would have rendered both EDGs inoperable. The procedural deficiency was not recognized by propedure writers and reviewers or by operators during a pre-surveillance briefing (Section M2.2).
- Licensee response and actions with regard to a Reactor Coolant System (RCS) flow transmitter tubing fitting leak were conservative and technically sound. Compensatory measures were instituted to monitor the leak until a shutdown repair can be mad There were no operability concerns identified and the RCS leak was within TS leak rate limits (Section M2.3).
- Engineering
The inspectors noted inconsistencies with the actual dedicated shutdown diesel generator governor speed droop setting and the setting described in the system description and vendor manual. The safety significance was minimal as an operator is stationed at the diesel control panel during station blackout to control frequency. The system engineer appropriately initiated a condition report to revise procedures to reflect vendor recommendations (Section E2.1).
Plant Support
Radiation control and security practices were properly conducted. Inspected areas in the Radiological Control Area (RCA) were noted to be appropriately posted and secured as necessary. The security plan was effectively implemented and compensatory actions were initiated when required (Section R1.1, S1.1).
Report Details Summary of Plant Status Robinson Unit 2 operated at full power throughout the report perio I. Operations
Conduct of Operations 0 General Comments (71707)
The inspectors conducted frequent control room tours to verify proper staffing, operator attentiveness and communications, and adherence to approved procedures. The inspectors routinely attended operations turnovers, management reviews, and plan-of the-day meetings to maintain awareness of overall plant operations. Operator logs, Condition Reports (CR), and instrumentation were routinely reviewed. Plant tours were conducted to verify operational safety and compliance with Technical Specifications (TS). In general, the inspectors concluded that the conduct of operations was risk informed, professional, and safety-consciou.2 World Association of Nuclear Operators (WANO) Assessment Review (71707)
The inspectors reviewed the interim report for the WANO assessment that was conducted in 1998. The inspectors did not note any significant nuclear safety issues requiring immediate NRC attentio Operational Status of Facilities and Equipment 0 Engineered Safety Feature System Walkdown Inspection Scope (71707)
The inspectors conducted a detailed system walkdown of the Safety Injection (SI)
syste Observations and Findings The inspectors conducted a detailed system walkdown of the SI system to assess the general condition of system components, including labeling, to verify that system valve positions matched the system drawings and station operating procedures, and to assess plant housekeeping and radiological conditions around system component The inspectors determined that the SI system was capable of performing its design function for both standby and accident conditions. No misaligned valves were identified, housekeeping and component labeling were adequate. A minor discrepancy
identified by the inspectors related to a caution tag was resolved promptly by the licensee. The inspectors also reviewed the applicable sections of the Updated Final Safety Analysis Report (UFSAR) and identified no discrepancie Conclusions A detailed system walkdown found that the SI System was appropriately configured to perform its design functio.2 Cold Weather Preparations Inspection Scope (71707, 71714)
The inspectors conducted an independent review of the licensee's process and preparations for protection of the plant from the effects of cold weathe Observations and Findings The inspectors reviewed the three procedures that provide instructions and necessary precautionary measures to protect the plant from the effects of cold weather. These procedures included Operating Procedure (OP)-925, "Cold Weather Operation,"
Revision 16, Administrative Procedure (AP)-008, "Cold Weather Preparations," Revision 3 and Electrical Distribution Procedure (EDP)-009, "Freeze Protection Panels," Revision 10. The inspectors performed a walkdown of the plant and periodically verified the status of the freeze protection panels, temporary enclosures, and portable heaters that were installed and maintained. No problems were noted. The inspectors also reviewed the open maintenance work orders and confirmed that the open work orders on the freeze protection system were adequately managed. Additionally, the inspectors reviewed the Nuclear Assessment Section (NAS) audit that was conducted related to cold weather preparations and noted that there were no significant finding The inspectors did note that OP-925 defined cold weather as being imminent with outside air temperature less than or equal to 420F. With cold weather imminent, Section 8.2.3 was applicable, which required certain actions to be performed by the outside auxiliary operators. The inspectors identified that the licensee had not always implemented OP-925, though the outside air temperature had reached below 420F. The shift supervisor had chosen to not implement OP-925 based on projected temperatures not expected to go significantly below 420F. The inspectors discussed this issue with the operations staff and determined that the safety significance of not implementing OP 925 under these conditions was minimal; however, following this discussion with the inspectors, the licensee decided to enhance OP-925 to give more flexibility to the operating staff in implementing OP-92 Conclusions The licensee program for cold weather protection was adequately implemented. Freeze protection panels and circuits, as well as, temporary enclosures and heaters were appropriately installed and maintained per procedur *
0 Clearance Walkdown (62707, 71707) Inspection Scope The inspectors performed a walkdown of a system clearance and performed a review of other active clearance Observations and Findinqs The inspectors verified proper implementation of clearance 98-01121 on November 2 This clearance was established to support scheduled maintenance on the A charging pump. The inspectors verified that valves as well as electrical breakers were aligned appropriately to provide an adequate boundary for the scheduled maintenance activiti No discrepancies were identified during verification of the clearanc The inspectors also reviewed the status of other currently hanging clearance Operations procedure OPS-NGGC-1301, "Equipment Clearance," Revision 3, states that a CR should be generated or verified to exist, for each clearance that has been in effect for greater than three months. The CR is to evaluate the impact of the long standing clearance and determine an appropriate solution to resolving the underlying cause. The inspectors noted that there were five clearances currently hanging that were greater than three months old. The inspectors verified that each of these clearances had a CR that evaluated the resolution associated with the equipment under clearanc The inspectors also discussed the clearance with the operations support supervisor and determined that none of the clearances adversely impacted plant safet Conclusions A clearance associated with a charging pump was correctly implemented. A review of existing clearances identified that clearances older than three months had been evaluated in accordance with licensee procedures and did not impact plant safet Quality Assurance In Operations 0 Plant Nuclear Safety Committee and Nuclear Assessment Section Oversight (40500, 71707)
The inspectors periodically attended Plant Nuclear Safety Committee (PNSC) meetings during the inspection report period. The presentations to the committee were thorough and the presenters readily responded to all questions. The committee members asked probing questions and were well prepared. The committee members displayed an understanding of the issues. The inspectors also reviewed Nuclear Assessment Section audits and concluded that they were appropriately focused to identify and enhance safety. The inspectors concluded that the onsite review functions of the PNSC were conducted in accordance with TSs. During the PNSC meetings topics were thoroughly discussed and evaluate Miscellaneous Operations Issues (92901)
0 (Closed) Licensee Event Report (LER) 50-261/97-07-01: Condition Outside Design Basis Due to Inoperable Over Power Delta Temperatures (OPDT) Channels. This issue was initially discussed in NRC Inspection Report (IR) 50-261/97-08 as Unresolved Item (URI) 50-261/97-08-01, Failure to Properly Calibrate the Low Trip Set point for OPDT. This URI was dispositioned as Violation 50-261/97-09-02, Failure to Properly Calibrate OPDT, which was closed in NRC IR 50-261/98-01 following verification of the adequacy of licensee corrective actions. Additionally, LER 50-261/97-07-00 associated with the condition was also closed in NRC IR 50-261/97-09. LER 50-261/97-07-01, was submitted by the licensee to supplement LER 50-261/97-07-00 to correct some minor typographical error.2 (Closed) LER 50-261/97-11-01: Reactor Trip Due to Condensate Pump B Shaft Failur LER 50-261/97-11-00 related to the event was closed in NRC IR 50-261/97-14. The details associated with the reactor trip were discussed in NRC IR 50-261/97-12. This LER was submitted by the licensee to supplement the initial LER by providing information related to the root cause of the condensate shaft failure. This failure was attributed to cyclic fatigue. Additionally, the A condensate pump shaft was non destructively tested during Refueling Outage 18. The testing concluded that there were no indications on the pump shaft or stub shaft ends and that continued usage was acceptable. The inspectors reviewed the supplement and determined that the corrective actions taken by the licensee were adequat II. Maintenance M1 Conduct of Maintenance M Maintenance Observations (62707)
The inspectors observed all or portions of the following maintenance activities:
WR/JO 98-AGEM1, "SG 'C' L/L Level Relay Replacement"
MMM-042, "Documentation of Temporary Lead Lifts and Jumpers, Valve Manipulations, and Switch Manipulations or Thermocouple Polarity Test",
Revision 3
WR/JO ALEL 001, " Service Water Pump 'C' Replacement"
CM-10, "Service Water Pump Overhaul", Revision 8 The inspectors found the work performed under these activities to be professional and thorough. All work observed was performed with the work package or the specific maintenance procedure present and in active use. Technicians were experienced and
- knowledgeable of their assigned tasks. The inspectors frequently observed supervisors and system engineers monitoring job progress. Peer checking and self checking techniques were being use M2 Maintenance and Material Condition of Facilities and Equipment M Surveillance Observations (61726)
The inspectors observed all or portions of the following surveillance tests:
OST-251-1, "RHR Pump 'A' and Components Test", Rev. 9
OST-402-2, "EDG 'B' Diesel Fuel Oil System Flow Test", Revision The inspectors found that the testing was performed in accordance with applicable procedure M Emergency Diesel Generator (EDG) Fuel Oil Operations Surveillance Test (OST) Inspection Scope (61726, 71707)
The inspectors reviewed and assessed the circumstances surrounding a procedural problem encountered during the performance OST-402-2, "EDG 'B' Diesel Fuel Oil System Flow Test," Revision Observations and Findings Procedure OST-402-2 is a quarterly surveillance of the fuel oil transfer system supplying the B EDG. On October 19, operations personnel were performing OST-402-2 when a procedural step was identified that would have rendered both EDGs inoperable. The Reactor Operator (RO) noted the discrepancy prior to performing the step. The procedure was halted, the system was restored to the pre-test configuration, and a CR, 98-02361, was written to document the deficienc The inspectors reviewed the evaluation for CR 98-02361. The inspectors also reviewed Revision 8, and Revision 7 to OST-402-2. OST-402-2 was revised (Revision 8) to address flow measurement problems with the controlotron flow measuring instrumen The instrument experienced drifting during a previous performance of the OST (CR 98 01397). Steps were added in Revision 8 to re-zero the controlotron at several steps during the test. As part of the re-zeroing process, the procedure was written to open the A fuel oil transfer pump breaker prior to running the A transfer pump. At this point in the procedure the B fuel oil transfer pump breaker would also be open. Completing this step would have simultaneously de-energized both fuel oil transfer pumps preventing the filling of the fuel oil day tanks from the storage tanks rendering both EDGs inoperable. This was the licensee's first attempt to perform the OST following the implementation of Revision The inspectors determined that opening the A fuel oil transfer pump breaker was not necessary to successfully complete the OST. The inspectors concluded that the procedure writer and reviewers exhibited a lack of attention to detail in developing a procedure revision that would have rendered both EDGs inoperable. Additionally, the pretest briefing conducted by operations did not identify this procedure deficiency prior to performing the OST. The RO exhibited a good questioning attitude during performance of the OST and prevented both EDGs from becoming inoperabl As part of the corrective actions implemented by the licensee Revision 9 to the procedure was written to delete the steps which opened the A fuel oil transfer pump breaker. OST-402-1 was also corrected as it had been revised in a similar way. All persons involved with the event were counseled. No additional corrective action was planned. The OST was subsequently performed without any problem Conclusions During an EDG fuel oil.system flow test the RO performing the test exhibited a good questioning attitude by identifying a procedural step that would have rendered both EDGs inoperable. The procedural deficiency was not recognized by procedure writers and reviewers or by operators during a pre-surveillance briefin M Fitting Leak on Reactor Coolant System (RCS) Flow Transmitter Sensing Line Inspection Scope (37551, 62707)
The inspectors reviewed and assessed licensee activities associated with a leaking fitting in the C RCS loop flow transmitter tubing. This fitting is located inside containmen Observations and Findings A fitting on the high pressure sensing line for RCS flow transmitter FT-435 was found to be leaking during a containment entry on November 8. This fitting was located on a common high pressure sensing line for the three flow instruments for the C reactor coolant loop. The flow instrument provides input to the reactor protection system. A containment entry was made on November 11 to quantify the leak rate and to scope the job site for repair planning and for development of a temporary modification to contain the leakag The inspectors reviewed Engineering Service Request (ESR) 96-00113, "Modification to Improve Problem Differential Pressure Transmitters." The tubing/fitting in question was installed during the last refueling outage as part of ESR 96-00113. The leak was apparently a result of improper tubing fit-up at a T type compression fitting in a sloping section of the tubing. The fitting was leaking on both sides of the T with a quantified leak rate of approximately eight gallons per da Following numerous discussions, as well as review of other industry data and experience, the licensee made a decision to defer repairs to the transmitter tubing until the next forced outage or Refueling Outage (RFO) 19. This decision was based on a
risk evaluation of the fitting leak degrading or possibly failing verses the risks of maneuvering the plant to a power level to make on-line repairs, and the large amount of radiation dose associated with isolating and repairing the tubing (approximately two person-rem). With the plant at full power a failure of the tubing would trip the reactor with all three loop flow transmitters failing low, and would result in an estimated RCS leak rate of 20 to 80 gallons per minute. It was engineering's judgement that the risk of failure was low based on normal thread engagement on the fittings and the lack of steam at the leak precluding thread erosion or cuttin Interim compensatory measures were instituted to make weekly containment entries to measure leak rate so that any adverse trends could be promptly identified. The licensee installed a temporary camera in containment for real time monitoring of the leak in the control room. The inspectors concluded that the licensee's course of action was reasonable and included compensatory measures to allow monitoring of any adverse trends associated with the leak. The inspectors identified no concerns with the operability of the loop flow instruments or TS leak rate limits. Additionally, entries into containment following the discovery have confirmed that the leak rate has not increase The inspectors verified that a post modification inspection of the fittings was performe The post modification instructions did not require but recommended that the system pressure be at Normal Operating Pressure (NOP) for the post modification inspectio A review of the Shift Supervisor Operator (SSO) logs indicated that the plant was in transition to but had not reached NOP the day the post modification test was performe Conclusions Licensee response and actions with regard to an RCS flow transmitter tubing fitting leak were conservative and technically sound. Compensatory measures were instituted to monitor the leak until a shutdown repair can be made. There were no operability concerns identified and the RCS leak was within TS leak rate limit M8 Miscellaneous Maintenance Issues (92902)
M (Closed) LER 50-261/97-09-00: Technical Specification Violation Due to Inadequate Surveillance Test Procedure. This issue related to periodic testing of containment isolation valve leakage was identified by the licensee on July 2, 1997. The licensee attributed the condition to personnel error and appropriately retested the isolation valves as well as revised the applicable surveillance test procedure to specify the proper vent path for leakage measurement. This issue was dispositioned as NCV 50-261/97-08-03 following verification of the adequacy of licensee corrective action Ill. Engineering E2 Engineering Support of Facilities and Equipment E Dedicated Shutdown (DS) Diesel Generator Speed Droop Settings Inspection Scope (37551)
The inspectors identified and assessed inconsistencies associated with the DS diesel generator speed droop setting and the system description and vendor manua Observations and Findings On November 10, during a general walk down of the DS diesel, the inspectors noticed a speed droop setting of 50 on the governor. The inspectors reviewed the system description for the DS diesel which indicated that for parallel operation, the speed droop setting should be 50 and for standby operation a setting of zero. The inspector informed the system engineer of this discrepancy. The system engineer confirmed that the vendor manual recommended a speed droop setting of 30 to 50 for parallel operation (condition during testing) and a setting of zero for single unit operation (condition during station blackout when the DS diesel is expected to be utilized). The DS diesel operations surveillance test leaves the governor speed droop setting at 5 During a station blackout, the DS diesel is designed to supply power to the DS bus for safe shutdown loads. With the speed droop set at fifty, frequency would be expected to drop as the diesel is loaded. Current procedural guidance contained in procedures EPP-22, Revision 14, "Energizing Plant Equipment Using the Dedicated Shutdown Diesel Generator" and DSP-002, Revision 20, "Hot Shutdown Using the Dedicated Alternate Shutdown System" did not specify a speed droop setting for emergency/ non parallel operation of the DS diesel generator. There is minimal safety significance however, as DS procedures have an operator stationed at the DS diesel control panel who can appropriately maintain frequency utilizing the manual speed control. A speed droop setting of zero requires less operator action. The system engineer initiated a CR, 98-02557, with recommended corrective actions to revise procedures to agree with the vendor recommendations. The inspectors checked the speed droop setting for both emergency diesels and they were both found to be set at zero in the standby condition. Conclusions The inspectors noted inconsistencies with the actual dedicated shutdown diesel generator governor speed droop setting and the setting described in the system description and vendor manual. The safety significance was minimal as an operator is stationed at the diesel control panel during station blackout to control frequency. The system engineer appropriately initiated a CR to revise procedures to reflect vendor recommendation E8 Miscellaneous Engineering Issues (37551, 92903)
E (Closed) LERs 50-261/97-05-00, 97-05-01, 97-05-02: Condition Outside Design Basis:
Spent Fuel Shipping Cask Unreviewed Safety Question (USQ). This issue was discussed in NRC IR 50-261/97-10 and tracked as EEl 50-261/97-10-04, Failure to Meet 10 CFR 50.59 Requirements for USQ Related to Spent Fuel Cask Movemen Subsequently, the licensee submitted a USQ to the NRC on August 28,1997. The NRC reviewed the submittal and granted enforcement discretion in a letter dated November 7, 1997. The enforcement discretion was based on this issue being an old design issue as described in the NRC Enforcement Policy. EEl 50-261/97-10-04 was closed in NRC IR 50-261/98-06. In addition to submitting the USQ, as corrective action, the licensee had placed procedure SFS-001, "IF-300 Shipping Cask Operation," on administrative hold thus suspending spent fuel cask handling operation E (Closed) LER 50-261/97-06-00: Condition Outside Design Basis Due to Design Installation Error Safety Injection Pump Control Power Cabling. This issue was identified on May 21, 1997 during the NRC Architectural Engineering (A/E) team inspection. The C Safety Injection (SI) pump control cables were routed in the same cable tray as the A and B SI pump cables. This configuration was identified as a condition outside the design basis. URI 50-261/97-201-14, SI cable separation deficiency was identified as a result. This item was dispositioned as violation 50-261/98 03-06, Failure to Provide Accurate Information to NRC on Cable Separation for the SI pumps. The licensee completed a modification to reroute the cables for the "C" SI pump. Violation 50-261/98-03-06 was closed in NRC IR 50-261/98-0 E (Closed) LERs 50-261/97-08-00, 97-08-01: Condition Outside Design Basis Inadequate Safety Injection Pump Net Positive Suction Head (NPSH). This item was originally discussed in NRC IR 50-261/97-201 as URI 50-261/97-201-09, SI, RHR, and CS Pump NPSH. The issue was dispositioned as an escalated enforcement item 50-261/98-03 04, resulting in a Severity Level Ill violation. Following verification of completion of licensee corrective action, which included a modification to the SI system, the violation was closed in NRC IR 50-261/98-09. LER 50-261/97-08-00 and LER 50-261/97-08-01 discussed the circumstances surrounding this issu IV. Plant Support R1 Radiological Protection and Chemistry Controls R General Comments (71750)
The inspectors periodically toured the Radiological Control Area (RCA) during the inspection period. Radiological control practices were observed and discussed with radiological control personnel including RCA entry and exit controls, survey postings, locked high radiation area controls, and radiological area material condition. The inspectors concluded that radiation control practices were being conducted in accordance with procedures. The inspectors also toured the radwaste building and found that radwaste storage containers and laundry bags were in good condition and
appropriately labeled. In addition, outside radwaste storage areas and structures were properly posted and exhibited correct labeling and effective housekeeping. The inspectors found that housekeeping throughout the plant was effective in maintaining areas free of unnecessary equipment and debris. Relatively few contaminated areas were noted and posted locked high radiation areas were properly secured against unauthorized entr S1 Conduct of Security and Safeguards Activities S General Comments (71750)
During the period, the inspectors toured the protected area and noted that the perimeter fence was intact and not compromised by erosion or disrepair. Isolation zones were maintained on both sides of the barrier and were free of objects which could shield or conceal an individual. The inspectors periodically observed personnel, packages, and vehicles entering the protected area and verified that necessary searches, visitor escorting, and special purpose detectors were used as applicable prior to entr Lighting of the perimeter and of the protected area was acceptable and met illumination requirement P8 Miscellaneous Security and Safeguards Issues (92904)
P (Closed) LER 50-261/97-SO4-00: Unauthorized Entry Into Protected Area As a Result of Personnel Error. This event was attributed to personnel error and occurred on April 8, 1997. As a result of the hand geometry bio-metric system being out-of-service, security had established compensatory measures which included manual verification of authorized individuals. During this time a licensee employee whose security badge was not currently active in the security computer was allowed to enter the Protected Are The security guard who was verifying current access failed to adequately review the authorized access list. This incident was reported to the NRC operations center on April 8. Licensee corrective actions included a review of the work performed by the individual as well as a review of the areas visited, disciplinary action towards the security guard involved, and review of the incident with other members of the security force. The inspectors reviewed the completion of the corrective actions and determined that they were adequate. The inspectors noted that no similar incidents have occurred since this inciden V. Management Meetings X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management a the conclusion of the inspection on December 4, 1998. The licensee acknowledged the findings presented at the exit meeting. No proprietary information was identifie PARTIAL LIST OF PERSONS CONTACTED Licensee T. Cleary, Manager, Operations H. Chernoff, Supervisor, Licensing/Regulatory Programs J. Clements, Manager, Site Support Services R. Duncan, Manager, Robinson Engineering Support Services J. Fletcher, Manager, Maintenance J. Moyer, Director, Site Operations R. Steele, Manager, Outage Management R. Warden, Manager, Manager, Regulatory Affairs D. Young, Vice President, Robinson Nuclear Plant NRC B. Desai, Senior Resident Inspector A. Hutto, Resident Inspector INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726:
Surveillance Observations IP 62707:
Maintenance Observation IP 71707:
Plant Operations IP 71714:
Cold Weather Preparations IP 71750:
Plant Support Activitie IP 92901:
Followup - Operations IP 92902:
Followup - Maintenance IP 92903:
Followup - Engineering IP 92904 Followup - Plant Support
ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Closed 50-261/97-07-01 LER Condition Outside Design Basis Due to Inoperable Over Power Delta Temperatures Channels (Section 08.1)
50-261/97-11-01 LER Reactor Trip Due to Condensate Pump "B" Shaft Failure (Section 08.2)
50-261/97-09-00 LER Technical Specification Violation Due to Inadequate Surveillance Test Procedure (Section M8.1)
50-261/97-05-00 LER Condition Outside Design Basis Spent Fuel Shipping Cask 50-261/97-05-01 Unreviewed Safety Question Determination (Section E8.1)
50-261/97-05-02 50-261/97-06-00 LER Condition Outside Design Basis Due to Design Installation Error Safety Injection Pump Control Power Cabling (Section E8.2)
50-261/97-08-00 LER Condition Outside Design Basis Inadequate Safety Injection 50-261/97-08-01 Pump Net Positive Suction Head (Section E8.3)
50-261/97-SO4-00 LER Unauthorized Entry Into Protected Area As a Result of Personnel Error (Section P8.1)