ML20311A593

1Q/2000 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Last modified : April 01, 2002

2Q/2000 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

2Q/2000 Inspection Findings - Crystal River 3 Page 2 of 4 Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

(Section 4OA3)

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

2Q/2000 Inspection Findings - Crystal River 3 Page 3 of 4 Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

2Q/2000 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Last modified : April 01, 2002

3Q/2000 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

(Section 4OA3)

Inspection Report# : 2000003(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown

3Q/2000 Inspection Findings - Crystal River 3 Page 2 of 4 and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining

3Q/2000 Inspection Findings - Crystal River 3 Page 3 of 4 surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

3Q/2000 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Last modified : March 29, 2002

4Q/2000 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

(Section 4OA3)

Inspection Report# : 2000003(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request

4Q/2000 Inspection Findings - Crystal River 3 Page 2 of 4 had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Barrier Integrity Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

4Q/2000 Inspection Findings - Crystal River 3 Page 3 of 4 Inspection Report# : 2000007(pdf)

Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

4Q/2000 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Last modified : March 28, 2002

1Q/2001 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

1Q/2001 Inspection Findings - Crystal River 3 Page 2 of 4 (Section 4OA3)

Inspection Report# : 2000003(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Barrier Integrity Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

1Q/2001 Inspection Findings - Crystal River 3 Page 3 of 4 Inspection Report# : 2000007(pdf)

Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

1Q/2001 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Last modified : March 28, 2002

2Q/2001 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

2Q/2001 Inspection Findings - Crystal River 3 Page 2 of 4 (Section 4OA3)

Inspection Report# : 2000003(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Barrier Integrity Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

2Q/2001 Inspection Findings - Crystal River 3 Page 3 of 4 Inspection Report# : 2000007(pdf)

Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

2Q/2001 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Last modified : March 27, 2002

3Q/2001 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

3Q/2001 Inspection Findings - Crystal River 3 Page 2 of 4 Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

(Section 4OA3)

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining

3Q/2001 Inspection Findings - Crystal River 3 Page 3 of 4 surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

3Q/2001 Inspection Findings - Crystal River 3 Page 4 of 4 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Last modified : March 26, 2002

4Q/2001 Inspection Findings - Crystal River 3 Page 1 of 4 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1.

If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

4Q/2001 Inspection Findings - Crystal River 3 Page 2 of 4 Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol.

For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training.

(Section 4OA3)

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold.

The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee

4Q/2001 Inspection Findings - Crystal River 3 Page 3 of 4 replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance.

(Section 2OS2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ),

and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Public Radiation Safety Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

4Q/2001 Inspection Findings - Crystal River 3 Page 4 of 4 Miscellaneous Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Last modified : March 01, 2002

1Q/2002 Inspection Findings - Crystal River 3 Page 1 of 5 Crystal River 3 Initiating Events Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1. If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional.

1Q/2002 Inspection Findings - Crystal River 3 Page 2 of 5 (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure. These methods are proceduralized and addressed in operator training. (Section 4OA3)

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve

1Q/2002 Inspection Findings - Crystal River 3 Page 3 of 5 rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold. The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance. (Section 2OS2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

1Q/2002 Inspection Findings - Crystal River 3 Page 4 of 5 Public Radiation Safety Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance.

Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough.

However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions.

Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective

1Q/2002 Inspection Findings - Crystal River 3 Page 5 of 5 actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Last modified : July 22, 2002

2Q/2002 Inspection Findings - Crystal River 3 Page 1 of 7 Crystal River 3 Initiating Events Significance: Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Corrective actions to address a feedwater transient which occurred on December 15, 2001, had not been implemented.

The inspectors identified that corrective actions to address a feedwater transient which occurred on December 15, 2001, had not been implemented. This issue was more than minor because the feedwater transient required operator intervention in order to stabilize the plant and resulted in cavitation of a feedwater booster pump, which if it had tripped or become damaged, could have resulted in more severe consequences. Therefore, it was important that corrective actions should have been implemented. This finding was determined to be of very low safety significance (Green) by the significance determination process because the impact was limited to a slightly increased likelihood of a plant transient. (Section 4OA2.c)

Inspection Report# : 2002006(pdf)

Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Jun 29, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly position a component during clearance (tagging) activities Technical Specification 5.6.1, Procedures, states that written procedures shall be implemented covering the activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. The regulatory guide list in Appendix A, includes procedures for equipment control (tagging). Licensee procedure OPS-NGGC-1301, Equipment Clearance implements this requirement and states, in step 4.10, that the "Tag Hanger positions components as specified on the Clearance Checklist." Contrary to the above, the tag hanger for Clearance Checklist 35216, on April 4, 2002, failed to position the emergency feedwater pump (EFP-3) fuel rack, in the "tripped" position prior to placing a "Diesel Engine Tripped" tag on the fuel rack. The rack was in the Normal, not tripped position and this was not identified by file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

2Q/2002 Inspection Findings - Crystal River 3 Page 2 of 7 the independent second-checker. This is being treated as a Non-Cited Violation. The violation is in the licensee corrective action program as Nuclear Condition Report 58819.

Inspection Report# : 2002002(pdf)

Significance: Jun 29, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to complete an accurate risk assessment per 10 CFR 50.65(a)(4) 10 CFR 50.65 (a)(4) requires, in part, that before performing maintenance activities (including but not limited to surveillances, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, the licensee failed to assess the maintenance risk for all plant maintenance to be performed during the week of February 11, 2002. Specifically, the completed assessment failed to account for raw water pumps within licensee-established risk assessment scope that were concurrently out of service. This is being treated as a Non-Cited Violation. The violation is in the licensee corrective action program as Nuclear Condition Report 58911.

Inspection Report# : 2002002(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1. If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures.

Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

2Q/2002 Inspection Findings - Crystal River 3 Page 3 of 7 Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

2Q/2002 Inspection Findings - Crystal River 3 Page 4 of 7 emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure.

These methods are proceduralized and addressed in operator training. (Section 4OA3)

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold. The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

2Q/2002 Inspection Findings - Crystal River 3 Page 5 of 7 repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance. (Section 2OS2).

Inspection Report# : 2000002(pdf)

Public Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

2Q/2002 Inspection Findings - Crystal River 3 Page 6 of 7 Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems Based on the results of the inspection, one finding and several negative observations were identified. The licensee was effective at identifying problems Based on the results of the inspection, one finding and several negative observations were identified. The licensee was effective at identifying problems at a low threshold and putting them into the corrective action program. Although two issues were identified that the licensee had not entered into the corrective action program, these were considered isolated instances and not indicative of a weakness in this area. Generally, the licensee properly evaluated issues and implemented effective and timely corrective action. Formal root causes for issues classified as significant conditions adverse to quality were especially thorough and detailed. The inspectors identified several examples in which condition reporting evaluations lacked thoroughness or were too narrowly focused, and some corrective actions were not comprehensive or were not implemented as intended. One finding of very low safety significance was identified. The inspectors identified that corrective actions to address a feedwater transient had not been implemented. Licensee audits and self-assessments were effective in identifying deficiencies in the corrective action programs. In addition, audit and assessment findings were consistent with the inspectors' observations. Based on interviews of plant personnel from various departments, personnel indicated that they felt free to input safety issues and conditions adverse to quality into the corrective action and employee concerns programs. A safety conscious work environment was evident at Crystal River.

Inspection Report# : 2002006(pdf)

Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

2Q/2002 Inspection Findings - Crystal River 3 Page 7 of 7 Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/03/2003

3Q/2002 Inspection Findings - Crystal River 3 Page 1 of 7 Crystal River 3 Initiating Events Significance: Sep 28, 2002 Identified By: Self Disclosing Item Type: FIN Finding Corrective Actions Green. The licensee's corrective actions for a failed power cable were insufficient to prevent recurrence of a partial loss of offsite power event. The finding was more than minor because it increased the likelihood of a loss of offsite power.

The finding was determined to be of very low safety significance by the safety determination process because it did not involve a total loss of offsite power and power remained available for safety equipment. (Section 4OA2)

Inspection Report# : 2002003(pdf)

Significance: Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Corrective Actions to Address a Feedwater Transient Green. The inspectors identified that corrective actions to address a feedwater transient which occurred on December 15, 2001, had not been implemented. This issue was more than minor because the feedwater transient required operator intervention in order to stabilize the plant and resulted in cavitation of a feedwater booster pump, which if it had tripped or become damaged, could have resulted in more severe consequences. Therefore, it was important that corrective actions should have been implemented. This finding was determined to be of very low safety significance (Green) by the significance determination process because the impact was limited to a slightly increased likelihood of a plant transient. (Section 4OA2.c)

Inspection Report# : 2002006(pdf)

Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Worker Fatigue Green. The inspectors identified that the licensee did not consider worker fatigue in the licensee investigation of a potential loss of 4160 volt bus that involved worker performance issues. This finding is more than a minor because it was viewed as a precursor to a significant event (loss of decay heat removal). The finding is considered to be of very low safety significance because no actual loss of equipment occurred. (Section 1R20)

Inspection Report# : 2001004(pdf)

Mitigating Systems Significance: Jun 29, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to complete an accurate risk assessment per 10 CFR 50.65(a)(4) 10 CFR 50.65 (a)(4) requires, in part, that before performing maintenance activities (including but not limited to

3Q/2002 Inspection Findings - Crystal River 3 Page 2 of 7 surveillances, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, the licensee failed to assess the maintenance risk for all plant maintenance to be performed during the week of February 11, 2002. Specifically, the completed assessment failed to account for raw water pumps within licensee-established risk assessment scope that were concurrently out of service. This is being treated as a Non-Cited Violation. The violation is in the licensee corrective action program as Nuclear Condition Report 58911.

Inspection Report# : 2002002(pdf)

Significance: Jun 29, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly position a component during clearance (tagging) activities Technical Specification 5.6.1, Procedures, states that written procedures shall be implemented covering the activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. The regulatory guide list in Appendix A, includes procedures for equipment control (tagging). Licensee procedure OPS-NGGC-1301, Equipment Clearance implements this requirement and states, in step 4.10, that the "Tag Hanger positions components as specified on the Clearance Checklist." Contrary to the above, the tag hanger for Clearance Checklist 35216, on April 4, 2002, failed to position the emergency feedwater pump (EFP-3) fuel rack, in the "tripped" position prior to placing a "Diesel Engine Tripped" tag on the fuel rack. The rack was in the Normal, not tripped position and this was not identified by the independent second-checker. This is being treated as a Non-Cited Violation. The violation is in the licensee corrective action program as Nuclear Condition Report 58819.

Inspection Report# : 2002002(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Follow Procedures During Electrical Maintenance on the Engineered Safeguards Bus 10 CFR Part 50, Appendix B, Criterion V, states that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures.

Contrary to this requirement, on October 9, 2001, during maintenance on 4160 volt engineered safeguards (ES) bus 3A using Work Request 365187: 1) Although the work request stated "Check Bus for Voltage before starting work", dead bus checks were not done; 2) Although licensee Administrative Instruction AI-610, Electrical Safety, required a maintenance risk assessment be performed on all work on energized equipment with the work assessed as medium or high risk, the work request had not been risk assessed and was classified low risk; and 3) Although licensee Administrative Instruction AI-504, Guidelines for Cold Shutdown and Refueling, stated "Power supplies (for operating safety equipment shall be) controlled by physical barriers with signs" an energized power supply for the operating decay heat removal equipment accessed by a worker was not controlled by a physical barrier with a sign. This was identified in the licensee's corrective action program as CR-42306 (Green).

Inspection Report# : 2001004(pdf)

Significance: Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Main Steam Safety Valve Setpoints Outside Required Tolerance Longer than Allowed by Technical Specifications Technical Specification 3.7.1.A requires that all main steam line code safety valves shall be operable with lift settings as specified in Table 3.7.1-1. If one or more main steam line code safety valves are inoperable, actions must be taken in accordance with Action A of Technical Specification 3.7.1. Contrary to the above, as of September 26, 2001 and September 29, 1999, main steam line safety valves did not have lift settings in accordance with Table 3.7.1-1, and the requirements of Action A of Technical Specification 3.7.1 were not met. This condition was identified by the licensee and documented in Nuclear Condition Report NCR 48648. This condition was reported in LER 50-332/01-002 (Green).

3Q/2002 Inspection Findings - Crystal River 3 Page 3 of 7 Inspection Report# : 2001004(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Installation Error Results in Containment Isolation Valve Inoperable Longer than Allowed by Technical Specifications Technical Specification 3.6.3 Condition C requires that a containment penetration flow path be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if the associated isolation valve is not operable. Contrary to this from May 13 to 14, a feedwater check valve (FWV-46) was not operable, and the penetration was not isolated. This was identified in the licensee's corrective action program as CR-42306. This finding is only of very low significance because it only affects the barrier integrity cornerstone and all other mitigating systems were functional. (Green)

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Requirements for Inoperable Fire Damper.

Crystal River 3 Operating License Requirement 2.C.(9) requires that FPC shall implement and maintain in effect all provisions of the approved fire protection program. Table 6.7a of the Fire Protection Plan, requires that when a fire barrier penetration is not functional, either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. Contrary to the above, for various times prior to January 3, 2001, the air return fire barrier damper for the B engineered safeguards 4160 volt switchgear room was not functional and neither fire watch provision was met. This issue was described in the licensee corrective action program as PC 01-0012 and is being treated as a Non-cited violation.

Inspection Report# : 2000005(pdf)

Significance: Dec 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Fire Protection Plan Requirements When Two Cable Spreading Room Fire Dampers Were Not Operable.

Crystal River 3 Operating License Requirement 2.C.(9) requires that all provisions of the approved fire protection program be implemented. Table 6.7a of the Fire Protection Plan requires that when a fire barrier penetration is not functional, the licensee shall either establish a continuous fire watch on at least one side of the barrier, or verify the operability of fire detectors on one side of the barrier and establish an hourly fire watch patrol. For various periods of time from February 1999 to October 10, 2000, both exhaust fire barrier dampers (AHFD-47 and 83) for the cable spreading room were not functional and the fire watch provisions of the Fire Protection Plan were not met. The violation is in the licensee's corrective action program as Precursor Card 00-2918. (Section 4OA7)

Inspection Report# : 2000004(pdf)

Significance: Sep 30, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Meet Technical Specification Requirements for Pressurizer Heaters Green. A Non-Cited Violation of Technical Specification 3.4.8, Condition B was identified because requirements for electrical power supplies to the pressurizer heaters were not met in one instance. Breakers supplying one train of emergency power to the pressurizer heaters were removed from service for greater than the allowed period. The finding was determined to be of very low safety significance. Florida Power Corporation recently identified that procedures did

3Q/2002 Inspection Findings - Crystal River 3 Page 4 of 7 not incorporate the power supply technical specification requirements and subsequently identified this single instance of noncompliance through a detailed review. The intent of the requirements is to ensure that the reactor coolant system is maintained subcooled with natural circulation flow under specific plant conditions. Although the pressurizer heaters support pressure control during natural circulation, there were other alternative methods available to maintain pressure.

These methods are proceduralized and addressed in operator training. (Section 4OA3)

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Aug 11, 2001 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed by the NRC to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Process White performance indicator threshold in May, 2001. The White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector determined that the licensee's root cause evaluation for the reactor coolant leakage was acceptable and that the licensee had taken appropriate corrective actions. The source of the leakage was seal ring leakage from a valve located between the reactor and the decay heat removal system. The licensee stated that the most likely cause of the valve leakage, initially identified in April 2000, was uneven seating forces around the pressure retaining valve seal ring. The uneven forces had likely resulted from either a 1999 valve rebuild or reactor coolant system pressure cycles associated with plant shutdown and subsequent heatup. A plant shutdown and restart in May 2001 was associated with a distinct increase in the leakage that caused the performance indicator to cross the White threshold. The leakage exceeded the White performance indicator threshold for about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during which time, the reactor remained shutdown. Prior to restarting the reactor following the leakage increase, the licensee installed a canopy over the top of the valve, moving the pressure retaining surface from the seal ring to the canopy. The modification stopped the leakage and the performance indicator returned to the Green performance level.

Inspection Report# : 2001007(pdf)

Significance: N/A Nov 01, 2000 Identified By: NRC Item Type: FIN Finding Reactor Coolant System Leakage This supplemental inspection was performed to assess Florida Power Corporation's activities associated with increased reactor coolant system leakage. The leakage exceeded the NRC's Reactor Oversight Program White performance indicator threshold in August, 2000. The leakage remained above the threshold until September 9, 2000, when the plant was shutdown for repairs. This White indicator threshold is set at 50 percent of the leakage allowed by Technical Specifications and corresponds to a performance level that may result in increased NRC oversight. Using Inspection Procedure 95001, the inspector found that the licensee's root cause analysis for the leakage was acceptable and that the licensee had taken or planned appropriate corrective actions. The primary contributor to the leakage was identified as seat leakage thorough a pressurizer safety valve. Leakage from the pressure seal on a decay heat system valve also contributed to the overall reactor coolant system leakage. The licensee replaced the safety valve and conducted leak sealant repairs on the decay heat valve. These actions significantly reduced the reactor coolant system leakage and returned the performance indicator to the Green performance level. The licensee subsequently identified the root cause of the safety valve seat leakage and developed appropriate corrective actions. The licensee plans to complete permanent repairs to the decay heat valve at the next available opportunity.

Inspection Report# : 2000007(pdf)

Emergency Preparedness

3Q/2002 Inspection Findings - Crystal River 3 Page 5 of 7 Occupational Radiation Safety Significance: Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Radiological Collective Dose Expenditure Estimates Not Revised.

Collective dose expenditures for three high dose rate/dose evolutions conducted during the October 1999 Refueling Outage exceeded their original dose expenditure estimates by more than 50 percent. For steam generator tube maintenance activities, actual dose expenditures exceeded both the original and revised dose projections by more than 50 percent. For eddy current testing and scaffolding activities, revisions to the dose estimates were not conducted and documented until after the original dose expenditure estimates were exceeded. Differences between the original and revised estimates resulted from elevated dose rates, expanded job scope, and/or worker performance. Since the tasks did not result in any individual doses exceeding 10 CFR Part 20, Subpart C, Occupational Dose Limits, this finding was determined to be of very low safety significance. (Section 2OS2).

Inspection Report# : 2000002(pdf)

Significance: Jul 01, 2000 Identified By: Licensee Item Type: NCV NonCited Violation High Radiation Area Controls Not Fully Effective.

On October 23, 1999, Health Physics (HP) technicians providing high radiation area job coverage failed to provide positive controls in accordance with Improved Technical Specification 5.8.1.c, for two contract workers performing leadscrew cleaning and inspection activities under Radiation Work Permit 99-0146. The two workers received cumulative doses of 330 and 550 millirem which exceeded the 250 millirem (mrem) cumulative dose expected for the task. Since there was no substantial potential for overexposure to occur based on the expected job duration (1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), and the maximum general area dose rates (300 mrem per hour), this finding was determined to be of very low safety significance. This finding was identified as a Non-Cited Violation (NCV) for failure to provide continuous health physics coverage required by Improved Technical Specification 5.8.1.c for work conducted in a High Radiation Area (Section 2OS1.2).

Inspection Report# : 2000002(pdf)

Public Radiation Safety Physical Protection Significance: Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure that individuals provide identification prior to being granted unescorted access.

Green. A Non-cited violation of a license condition and procedural requirements was identified when two NRC inspectors were granted unescorted access to Crystal River 3 without being required to produce a valid picture identification. Provisions in the Crystal River Physical Security Plan and the requirements of Security Procedure SEC-NGGC-2101 for obtaining identification information prior to granting access were not met. The finding was of very

3Q/2002 Inspection Findings - Crystal River 3 Page 6 of 7 low safety significance because, although the identification information was not verified as required prior to access, the individuals granted access met all requirements for authorization for unescorted access. (Section 3PP2)

Inspection Report# : 2000005(pdf)

Miscellaneous Significance: N/A Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems Based on the results of the inspection, one finding and several negative observations were identified. The licensee was effective at identifying problems at a low threshold and putting them into the corrective action program. Although two issues were identified that the licensee had not entered into the corrective action program, these were considered isolated instances and not indicative of a weakness in this area. Generally, the licensee properly evaluated issues and implemented effective and timely corrective action. Formal root causes for issues classified as significant conditions adverse to quality were especially thorough and detailed. The inspectors identified several examples in which condition reporting evaluations lacked thoroughness or were too narrowly focused, and some corrective actions were not comprehensive or were not implemented as intended. One finding of very low safety significance was identified. The inspectors identified that corrective actions to address a feedwater transient had not been implemented. Licensee audits and self-assessments were effective in identifying deficiencies in the corrective action programs. In addition, audit and assessment findings were consistent with the inspectors' observations. Based on interviews of plant personnel from various departments, personnel indicated that they felt free to input safety issues and conditions adverse to quality into the corrective action and employee concerns programs. A safety conscious work environment was evident at Crystal River.

Inspection Report# : 2002006(pdf)

Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Report a Condition Prohibited by Technical Specifications 10 CFR 50.73 (a)(2)(B), requires that any condition prohibited by plant technical specifications shall be reported by the licensee in a Licensee Event Report within 60 days after discovery of the event. Contrary to the above, the licensee determined that two main steam safety valves had setpoints outside of the Technical Specification Table 3.7.1-1 required tolerance, the actions of the technical specification were not taken, and the condition was not reported within 60 days after discovery (September 1999). This condition was identified by the licensee and documented in Nuclear Condition Report NCR 51139 (No Color).

Inspection Report# : 2001004(pdf)

Significance: N/A Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Corrective Action Program Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. The priority grading system ensured timely resolution and corrective actions commensurate with safety significance. Corrective action backlog and precursor card evaluation timeliness were well managed. Root cause analyses were thorough. However, issues addressed in NRC inspection findings were not specifically reviewed to ensure adequate corrective actions. Licensee self-assessments and audits were effective in identifying deficiencies in the corrective action program. These deficiencies were entered into the corrective action program and, for the most part, resulted in the implementation of corrective actions. However, numerous Health Physics peer assessment recommendations, although entered in the corrective action program, were closed with inadequate documentation of disposition and corrective actions. A safety conscious work environment was present

3Q/2002 Inspection Findings - Crystal River 3 Page 7 of 7 where employees felt free to raise safety concerns.

Inspection Report# : 2000006(pdf)

Significance: Oct 06, 2000 Identified By: NRC Item Type: FIN Finding Cross Cutting Issue - Corrective Action Effectiveness Green. A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for precursor card 99-4142. This precursor card addressed deficiencies involved with rigging of the reactor vessel plenum during reactor assembly. NRC Non-cited Violation 50-302/99-07-01, Reactor Plenum Rigged Improperly, also addressed this issue. The licensee did not fully assess the nature and extent of the issue. Consequently, important causal factors were not identified and corrective actions to prevent recurrence were not thorough. This issue was determined to have very low safety significance because the licensee adequately addressed the potential adverse impact on equipment prior to reactor startup. The licensee's examination did not identify any damage to the reactor vessel or plenum. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2.2).

Inspection Report# : 2000006(pdf)

Last modified : December 02, 2002

4Q/2002 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 Initiating Events Significance: Sep 28, 2002 Identified By: Self Disclosing Item Type: FIN Finding Corrective Actions Green. The licensee's corrective actions for a failed power cable were insufficient to prevent recurrence of a partial loss of offsite power event.

The finding was more than minor because it increased the likelihood of a loss of offsite power. The finding was determined to be of very low safety significance by the safety determination process because it did not involve a total loss of offsite power and power remained available for safety equipment. (Section 4OA2)

Inspection Report# : 2002003(pdf)

Significance: Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Corrective Actions to Address a Feedwater Transient Green. The inspectors identified that corrective actions to address a feedwater transient which occurred on December 15, 2001, had not been implemented. This issue was more than minor because the feedwater transient required operator intervention in order to stabilize the plant and resulted in cavitation of a feedwater booster pump, which if it had tripped or become damaged, could have resulted in more severe consequences. Therefore, it was important that corrective actions should have been implemented. This finding was determined to be of very low safety significance (Green) by the significance determination process because the impact was limited to a slightly increased likelihood of a plant transient. (Section 4OA2.c)

Inspection Report# : 2002006(pdf)

Mitigating Systems Significance: Jun 29, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to complete an accurate risk assessment per 10 CFR 50.65(a)(4) 10 CFR 50.65 (a)(4) requires, in part, that before performing maintenance activities (including but not limited to surveillances, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, the licensee failed to assess the maintenance risk for all plant maintenance to be performed during the week of February 11, 2002. Specifically, the completed assessment failed to account for raw water pumps within licensee-established risk assessment scope that were concurrently out of service. This is being treated as a Non-Cited Violation. The violation is in the licensee corrective action program as Nuclear Condition Report 58911.

Inspection Report# : 2002002(pdf)

Significance: Jun 29, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly position a component during clearance (tagging) activities Technical Specification 5.6.1, Procedures, states that written procedures shall be implemented covering the activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. The regulatory guide list in Appendix A, includes procedures for equipment control (tagging). Licensee procedure OPS-NGGC-1301, Equipment Clearance implements this requirement and states, in step 4.10, that the "Tag Hanger positions components as specified on the Clearance Checklist." Contrary to the above, the tag hanger for Clearance Checklist 35216, on April 4, 2002, failed to position the emergency feedwater pump (EFP-3) fuel rack, in the "tripped" position prior to placing a "Diesel Engine Tripped" tag on the fuel rack. The rack was in the Normal, not tripped position and this was not identified by the independent second-checker. This is being treated as a Non-Cited Violation. The violation is in the licensee corrective action program as Nuclear Condition Report

4Q/2002 Inspection Findings - Crystal River 3 Page 2 of 2 58819.

Inspection Report# : 2002002(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems Based on the results of the inspection, one finding and several negative observations were identified. The licensee was effective at identifying problems at a low threshold and putting them into the corrective action program. Although two issues were identified that the licensee had not entered into the corrective action program, these were considered isolated instances and not indicative of a weakness in this area. Generally, the licensee properly evaluated issues and implemented effective and timely corrective action. Formal root causes for issues classified as significant conditions adverse to quality were especially thorough and detailed. The inspectors identified several examples in which condition reporting evaluations lacked thoroughness or were too narrowly focused, and some corrective actions were not comprehensive or were not implemented as intended. One finding of very low safety significance was identified. The inspectors identified that corrective actions to address a feedwater transient had not been implemented. Licensee audits and self-assessments were effective in identifying deficiencies in the corrective action programs. In addition, audit and assessment findings were consistent with the inspectors' observations. Based on interviews of plant personnel from various departments, personnel indicated that they felt free to input safety issues and conditions adverse to quality into the corrective action and employee concerns programs. A safety conscious work environment was evident at Crystal River.

Inspection Report# : 2002006(pdf)

Last modified : March 25, 2003

1Q/2003 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 1Q/2003 Plant Inspection Findings Initiating Events Significance: Sep 28, 2002 Identified By: Self Disclosing Item Type: FIN Finding Corrective Actions Green. The licensee's corrective actions for a failed power cable were insufficient to prevent recurrence of a partial loss of offsite power event. The finding was more than minor because it increased the likelihood of a loss of offsite power.

The finding was determined to be of very low safety significance by the safety determination process because it did not involve a total loss of offsite power and power remained available for safety equipment. (Section 4OA2)

Inspection Report# : 2002003(pdf)

Significance: Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Corrective Actions to Address a Feedwater Transient Green. The inspectors identified that corrective actions to address a feedwater transient which occurred on December 15, 2001, had not been implemented. This issue was more than minor because the feedwater transient required operator intervention in order to stabilize the plant and resulted in cavitation of a feedwater booster pump, which if it had tripped or become damaged, could have resulted in more severe consequences. Therefore, it was important that corrective actions should have been implemented. This finding was determined to be of very low safety significance (Green) by the significance determination process because the impact was limited to a slightly increased likelihood of a plant transient. (Section 4OA2.c)

Inspection Report# : 2002006(pdf)

Mitigating Systems Barrier Integrity Emergency Preparedness file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/22/2003

1Q/2003 Inspection Findings - Crystal River 3 Page 2 of 2 Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Jun 21, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems Based on the results of the inspection, one finding and several negative observations were identified. The licensee was effective at identifying problems at a low threshold and putting them into the corrective action program. Although two issues were identified that the licensee had not entered into the corrective action program, these were considered isolated instances and not indicative of a weakness in this area. Generally, the licensee properly evaluated issues and implemented effective and timely corrective action. Formal root causes for issues classified as significant conditions adverse to quality were especially thorough and detailed. The inspectors identified several examples in which condition reporting evaluations lacked thoroughness or were too narrowly focused, and some corrective actions were not comprehensive or were not implemented as intended. One finding of very low safety significance was identified. The inspectors identified that corrective actions to address a feedwater transient had not been implemented. Licensee audits and self-assessments were effective in identifying deficiencies in the corrective action programs. In addition, audit and assessment findings were consistent with the inspectors' observations. Based on interviews of plant personnel from various departments, personnel indicated that they felt free to input safety issues and conditions adverse to quality into the corrective action and employee concerns programs. A safety conscious work environment was evident at Crystal River.

Inspection Report# : 2002006(pdf)

Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 07/22/2003

2Q/2003 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 2Q/2003 Plant Inspection Findings Initiating Events Significance: Sep 28, 2002 Identified By: Self Disclosing Item Type: FIN Finding Corrective Actions Green. The licensee's corrective actions for a failed power cable were insufficient to prevent recurrence of a partial loss of offsite power event. The finding was more than minor because it increased the likelihood of a loss of offsite power.

The finding was determined to be of very low safety significance by the safety determination process because it did not involve a total loss of offsite power and power remained available for safety equipment. (Section 4OA2)

Inspection Report# : 2002003(pdf)

Mitigating Systems Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure To Implement Inservice Testing Program Requirements (Section 1R22).

A finding was identified for failure to implement increased frequency testing of a safety-related pump, after the pump differential pressure was found in the Alert range of the ASME Code,Section XI test on December 2, 2002. When tested on May 22, 2003, the pump was found in the Action range and was declared inoperable. A non-cited violation of Technical Specification 5.6.2 was identified. The finding is greater than minor because an engineering evaluation was required to assure that accident analysis requirements were met during the subsequent period of operation with differential pressure below the design minimum value. If the finding had not been corrected, pump performance could have resulted in the safety system not being capable of performing its design function to remove residual heat following an accident. The finding is of very low safety significance because the maximum period of operation below the design minimum differential pressure was of short duration and redundancy existed that assured the safety function remained available. (Section 1R22)

Inspection Report# : 2003004(pdf)

Barrier Integrity Emergency Preparedness file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 10/08/2003

2Q/2003 Inspection Findings - Crystal River 3 Page 2 of 2 Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 10/08/2003

3Q/2003 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 3Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Protect One Train of Safe Shutdown Equipment From Fire Damage The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix R, Section III.G.2, Fire Protection of Safe Shutdown Capability, for failure to protect certain electrical cables for safe shutdown equipment from fire damage in three fire areas. The licensee has corrected related identified procedural deficiencies and plans to resolve the noncompliance with cable protection through licensing correspondence with the NRC.

This finding is greater than minor safety significance because it involved a lack of required fire barriers for equipment relied upon for safe shutdown following a fire and because it affected the objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the availability and reliability of systems that mitigate initiating events to prevent undesirable consequences. The finding is of very low safety significance because licensee's proceduralized manual actions are reasonably accomplishable and training would have enabled operators to maintain the makeup function sufficiently to maintain reactor coolant system process variables within acceptable ranges. Therefore, the inspectors identified this issue as a Green finding as described in Inspection Procedure 71111.05, Fire Protection.

(Section 4OA5)

Inspection Report# : 2003005(pdf)

Significance: Sep 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Control Complex Cooling Trains A self-revealing non-cited violation of Crystal River 3 Technical Specification 3.7.18 was identified. Following Train B chiller maintenance on December 19, 2002, and Train A chiller maintenance on February 25, 2003, neither train of control complex cooling was operable because control complex chiller motor overload relays had been improperly set below their design values. The problem was identified on June 11, 2003, when both chiller motors tripped on overload current, when an overload current condition had not occurred.

The self-revealing finding is greater than minor safety significance because it resulted in a loss of the control complex cooling safety function and affected the availability and reliability of the Mitigating Systems Cornerstone of Reactor Safety that is used to mitigate events. The finding is of very low safety significance because the alternate non-safety Appendix R cooling system and feedwater pump (FWP-7) were available to mitigate transients involving systems that could be affected by the loss of cooling. (Section 4OA3) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 01/12/2004

3Q/2003 Inspection Findings - Crystal River 3 Page 2 of 2 Inspection Report# : 2003005(pdf)

Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure To Implement Inservice Testing Program Requirements (Section 1R22).

A finding was identified for failure to implement increased frequency testing of a safety-related pump, after the pump differential pressure was found in the Alert range of the ASME Code,Section XI test on December 2, 2002. When tested on May 22, 2003, the pump was found in the Action range and was declared inoperable.

A non-cited violation of Technical Specification 5.6.2 was identified. The finding is greater than minor because an engineering evaluation was required to assure that accident analysis requirements were met during the subsequent period of operation with differential pressure below the design minimum value. If the finding had not been corrected, pump performance could have resulted in the safety system not being capable of performing its design function to remove residual heat following an accident. The finding is of very low safety significance because the maximum period of operation below the design minimum differential pressure was of short duration and redundancy existed that assured the safety function remained available. (Section 1R22)

Inspection Report# : 2003004(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 01/12/2004

4Q/2003 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 4Q/2003 Plant Inspection Findings Initiating Events Significance: Dec 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Perform the Magnetic Particle Calibration The inspector identified a non-cited violation of Technical Specification 5.6.1.1 for failure to follow procedural requirements involving incorrect calibration of a magnetic particle testing (MT) yoke. This finding could have inhibited the identification of indications or flaws on American Society of Mechanical Engineers (ASME) Class 2 Safety-Related Feed Water to Once Through Steam Generator (OTSG) "A" piping.

This finding is more than minor because if left uncorrected, it could result in a more significant safety concern. Failure to correctly perform the calibration could reduce the ability to discover indications or flaws which could lead to pipe breaks. The issue was determined to be of very low safety significance because the likelihood of a loss of coolant accident (LOCA) initiator was not affected, the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not have available, and the finding did not increase the likelihood of a fire or flood. (Section 1R08)

Inspection Report# : 2003006(pdf)

Mitigating Systems Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Protect One Train of Safe Shutdown Equipment From Fire Damage The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix R, Section III.G.2, Fire Protection of Safe Shutdown Capability, for failure to protect certain electrical cables for safe shutdown equipment from fire damage in three fire areas. The licensee has corrected related identified procedural deficiencies and plans to resolve the noncompliance with cable protection through licensing correspondence with the NRC.

This finding is greater than minor safety significance because it involved a lack of required fire barriers for equipment relied upon for safe shutdown following a fire and because it affected the objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the availability and reliability of systems that mitigate initiating events to prevent undesirable consequences. The finding is of very low safety significance because licensee's proceduralized manual actions are reasonably accomplishable and training would have enabled operators to maintain the makeup function sufficiently to maintain reactor coolant system process variables within acceptable ranges. Therefore, the inspectors identified this issue as a Green finding as described in Inspection Procedure 71111.05, Fire Protection.

(Section 4OA5)

Inspection Report# : 2003005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 04/22/2004

4Q/2003 Inspection Findings - Crystal River 3 Page 2 of 3 Significance: Sep 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Control Complex Cooling Trains A self-revealing non-cited violation of Crystal River 3 Technical Specification 3.7.18 was identified. Following Train B chiller maintenance on December 19, 2002, and Train A chiller maintenance on February 25, 2003, neither train of control complex cooling was operable because control complex chiller motor overload relays had been improperly set below their design values. The problem was identified on June 11, 2003, when both chiller motors tripped on overload current, when an overload current condition had not occurred.

The self-revealing finding is greater than minor safety significance because it resulted in a loss of the control complex cooling safety function and affected the availability and reliability of the Mitigating Systems Cornerstone of Reactor Safety that is used to mitigate events. The finding is of very low safety significance because the alternate non-safety Appendix R cooling system and feedwater pump (FWP-7) were available to mitigate transients involving systems that could be affected by the loss of cooling. (Section 4OA3)

Inspection Report# : 2003005(pdf)

Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure To Implement Inservice Testing Program Requirements (Section 1R22).

A finding was identified for failure to implement increased frequency testing of a safety-related pump, after the pump differential pressure was found in the Alert range of the ASME Code,Section XI test on December 2, 2002. When tested on May 22, 2003, the pump was found in the Action range and was declared inoperable.

A non-cited violation of Technical Specification 5.6.2 was identified. The finding is greater than minor because an engineering evaluation was required to assure that accident analysis requirements were met during the subsequent period of operation with differential pressure below the design minimum value. If the finding had not been corrected, pump performance could have resulted in the safety system not being capable of performing its design function to remove residual heat following an accident. The finding is of very low safety significance because the maximum period of operation below the design minimum differential pressure was of short duration and redundancy existed that assured the safety function remained available. (Section 1R22)

Inspection Report# : 2003004(pdf)

Barrier Integrity Significance: Dec 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Identify and Correct a Small Pressure Boundary Leak in The Pressurizer Upper Level Instrument Tap Nozzles A self-revealing non-cited violation of Technical Specification 3.4.12.a was identified. Small cracks in the pressurizer upper level instrument tap nozzles resulted in pressure boundary leakage since late 2000.

file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 04/22/2004

4Q/2003 Inspection Findings - Crystal River 3 Page 3 of 3 The finding was greater than minor because the breach in the reactor coolant system (RCS) affected the RCS barrier performance attribute of the Barrier Integrity Cornerstone objective. However, the cracks were very small, were axial in direction, and therefore, were not expected to grow large enough to challenge the structural stability of the nozzle. A Phase 3 analysis was performed and because the likelihood of a LOCA initiator was not affected, the finding was determined to be of very low safety significance. (Section 4OA3.3)

Inspection Report# : 2003006(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CR3\cr3_pim.html 04/22/2004

1Q/2004 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 1Q/2004 Plant Inspection Findings Initiating Events Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: FIN Finding Loss of Design Control When an Improper Circuit Card Placed in the Integrated Control System Caused a Reactor Trip A self-revealing Green finding was identified for a loss of design control during maintenance on the integrated control system which later resulted in a reactor trip.

This finding is more than minor because it affected the configuration control attribute of the initiating event cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. The issue was of very low safety significance because although there was a reactor trip, mitigating systems remained available and were not affected. Because no safety systems were affected, the finding did not involve a violation of regulatory requirements. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3.1)

Inspection Report# : 2004003(pdf)

Significance: Dec 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Perform the Magnetic Particle Calibration The inspector identified a non-cited violation of Technical Specification 5.6.1.1 for failure to follow procedural requirements involving incorrect calibration of a magnetic particle testing (MT) yoke. This finding could have inhibited the identification of indications or flaws on American Society of Mechanical Engineers (ASME) Class 2 Safety-Related Feed Water to Once Through Steam Generator (OTSG) "A" piping.

This finding is more than minor because if left uncorrected, it could result in a more significant safety concern. Failure to correctly perform the calibration could reduce the ability to discover indications or flaws which could lead to pipe breaks. The issue was determined to be of very low safety significance because the likelihood of a loss of coolant accident (LOCA) initiator was not affected, the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not have available, and the finding did not increase the likelihood of a fire or flood. (Section 1R08)

Inspection Report# : 2003006(pdf)

Mitigating Systems Significance: Sep 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Control Complex Cooling Trains A self-revealing non-cited violation of Crystal River 3 Technical Specification 3.7.18 was identified. Following Train B chiller maintenance on December 19, 2002, and Train A chiller maintenance on February 25, 2003, neither train of control complex cooling was operable because control complex chiller motor overload relays had been improperly set below their design values. The problem was identified on June 11, 2003, when both chiller motors tripped on overload current, when an overload current condition had not occurred.

The self-revealing finding is greater than minor safety significance because it resulted in a loss of the control complex cooling safety function and affected the availability and reliability of the Mitigating Systems Cornerstone of Reactor Safety that is used to mitigate events. The finding is of very low safety significance because the alternate non-safety Appendix R cooling system and feedwater pump (FWP-7) were available to mitigate transients involving systems that could be affected by the loss of cooling. (Section 4OA3)

Inspection Report# : 2003005(pdf) 07/14/2004

1Q/2004 Inspection Findings - Crystal River 3 Page 2 of 3 Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Protect One Train of Safe Shutdown Equipment From Fire Damage The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix R, Section III.G.2, Fire Protection of Safe Shutdown Capability, for failure to protect certain electrical cables for safe shutdown equipment from fire damage in three fire areas. The licensee has corrected related identified procedural deficiencies and plans to resolve the noncompliance with cable protection through licensing correspondence with the NRC.

This finding is greater than minor safety significance because it involved a lack of required fire barriers for equipment relied upon for safe shutdown following a fire and because it affected the objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the availability and reliability of systems that mitigate initiating events to prevent undesirable consequences. The finding is of very low safety significance because licensee's proceduralized manual actions are reasonably accomplishable and training would have enabled operators to maintain the makeup function sufficiently to maintain reactor coolant system process variables within acceptable ranges. Therefore, the inspectors identified this issue as a Green finding as described in Inspection Procedure 71111.05, Fire Protection. (Section 4OA5)

Inspection Report# : 2003005(pdf)

Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure To Implement Inservice Testing Program Requirements (Section 1R22).

A finding was identified for failure to implement increased frequency testing of a safety-related pump, after the pump differential pressure was found in the Alert range of the ASME Code,Section XI test on December 2, 2002. When tested on May 22, 2003, the pump was found in the Action range and was declared inoperable.

A non-cited violation of Technical Specification 5.6.2 was identified. The finding is greater than minor because an engineering evaluation was required to assure that accident analysis requirements were met during the subsequent period of operation with differential pressure below the design minimum value. If the finding had not been corrected, pump performance could have resulted in the safety system not being capable of performing its design function to remove residual heat following an accident. The finding is of very low safety significance because the maximum period of operation below the design minimum differential pressure was of short duration and redundancy existed that assured the safety function remained available. (Section 1R22)

Inspection Report# : 2003004(pdf)

Barrier Integrity Significance: Dec 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Identify and Correct a Small Pressure Boundary Leak in The Pressurizer Upper Level Instrument Tap Nozzles A self-revealing non-cited violation of Technical Specification 3.4.12.a was identified. Small cracks in the pressurizer upper level instrument tap nozzles resulted in pressure boundary leakage since late 2000.

The finding was greater than minor because the breach in the reactor coolant system (RCS) affected the RCS barrier performance attribute of the Barrier Integrity Cornerstone objective. However, the cracks were very small, were axial in direction, and therefore, were not expected to grow large enough to challenge the structural stability of the nozzle. A Phase 3 analysis was performed and because the likelihood of a LOCA initiator was not affected, the finding was determined to be of very low safety significance. (Section 4OA3.3)

Inspection Report# : 2003006(pdf)

Emergency Preparedness Occupational Radiation Safety 07/14/2004

1Q/2004 Inspection Findings - Crystal River 3 Page 3 of 3 Public Radiation Safety Physical Protection Miscellaneous Last modified : May 05, 2004 07/14/2004

2Q/2004 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 2Q/2004 Plant Inspection Findings Initiating Events Significance: Jun 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures in 10 CFR 50.59 Screening An NRC identified, Non-Cited Violation (NCV) of Technical Specification 5.6.1.1 was identified for failure to fully implement a procedure which required a 10 CFR 50.59 evaluation to be completed for a one-time test of the power operated relief valve (PORV). Because the required evaluation was not completed, the licensee was unaware that the test would result in opening the PORV. As a result, the PORV unexpectedly opened for a very short period while the plant was operating and caused a reactor pressure transient.

This finding is more than minor because it affected the Primary System Loss of Coolant Accident (LOCA) Initiator attribute of the Initiating Events Cornerstone. The issue was of very low safety significance because although PORV opened for a short period of time with the reactor operating at power, mitigating systems, including the PORV block valve, were available had the valve failed to shut. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3)

Inspection Report# : 2004004(pdf)

Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: FIN Finding Loss of Design Control When an Improper Circuit Card Placed in the Integrated Control System Caused a Reactor Trip A self-revealing Green finding was identified for a loss of design control during maintenance on the integrated control system which later resulted in a reactor trip.

This finding is more than minor because it affected the configuration control attribute of the initiating event cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. The issue was of very low safety significance because although there was a reactor trip, mitigating systems remained available and were not affected. Because no safety systems were affected, the finding did not involve a violation of regulatory requirements. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3.1)

Inspection Report# : 2004003(pdf)

Significance: Dec 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Perform the Magnetic Particle Calibration The inspector identified a non-cited violation of Technical Specification 5.6.1.1 for failure to follow procedural requirements involving incorrect calibration of a magnetic particle testing (MT) yoke. This finding could have inhibited the identification of indications or flaws on American Society of Mechanical Engineers (ASME) Class 2 Safety-Related Feed Water to Once Through Steam Generator (OTSG) "A" piping.

This finding is more than minor because if left uncorrected, it could result in a more significant safety concern. Failure to correctly perform the calibration could reduce the ability to discover indications or flaws which could lead to pipe breaks. The issue was determined to be of very low safety significance because the likelihood of a loss of coolant accident (LOCA) initiator was not affected, the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not have available, and the finding did not increase the likelihood of a fire or flood. (Section 1R08)

Inspection Report# : 2003006(pdf)

Mitigating Systems Significance: Sep 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Control Complex Cooling Trains

2Q/2004 Inspection Findings - Crystal River 3 Page 2 of 3 A self-revealing non-cited violation of Crystal River 3 Technical Specification 3.7.18 was identified. Following Train B chiller maintenance on December 19, 2002, and Train A chiller maintenance on February 25, 2003, neither train of control complex cooling was operable because control complex chiller motor overload relays had been improperly set below their design values. The problem was identified on June 11, 2003, when both chiller motors tripped on overload current, when an overload current condition had not occurred.

The self-revealing finding is greater than minor safety significance because it resulted in a loss of the control complex cooling safety function and affected the availability and reliability of the Mitigating Systems Cornerstone of Reactor Safety that is used to mitigate events. The finding is of very low safety significance because the alternate non-safety Appendix R cooling system and feedwater pump (FWP-7) were available to mitigate transients involving systems that could be affected by the loss of cooling. (Section 4OA3)

Inspection Report# : 2003005(pdf)

Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Protect One Train of Safe Shutdown Equipment From Fire Damage The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix R, Section III.G.2, Fire Protection of Safe Shutdown Capability, for failure to protect certain electrical cables for safe shutdown equipment from fire damage in three fire areas. The licensee has corrected related identified procedural deficiencies and plans to resolve the noncompliance with cable protection through licensing correspondence with the NRC.

This finding is greater than minor safety significance because it involved a lack of required fire barriers for equipment relied upon for safe shutdown following a fire and because it affected the objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the availability and reliability of systems that mitigate initiating events to prevent undesirable consequences. The finding is of very low safety significance because licensee's proceduralized manual actions are reasonably accomplishable and training would have enabled operators to maintain the makeup function sufficiently to maintain reactor coolant system process variables within acceptable ranges. Therefore, the inspectors identified this issue as a Green finding as described in Inspection Procedure 71111.05, Fire Protection. (Section 4OA5)

Inspection Report# : 2003005(pdf)

Barrier Integrity Significance: Dec 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Identify and Correct a Small Pressure Boundary Leak in The Pressurizer Upper Level Instrument Tap Nozzles A self-revealing non-cited violation of Technical Specification 3.4.12.a was identified. Small cracks in the pressurizer upper level instrument tap nozzles resulted in pressure boundary leakage since late 2000.

The finding was greater than minor because the breach in the reactor coolant system (RCS) affected the RCS barrier performance attribute of the Barrier Integrity Cornerstone objective. However, the cracks were very small, were axial in direction, and therefore, were not expected to grow large enough to challenge the structural stability of the nozzle. A Phase 3 analysis was performed and because the likelihood of a LOCA initiator was not affected, the finding was determined to be of very low safety significance. (Section 4OA3.3)

Inspection Report# : 2003006(pdf)

2Q/2004 Inspection Findings - Crystal River 3 Page 3 of 3 Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : September 08, 2004

3Q/2004 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 3Q/2004 Plant Inspection Findings Initiating Events Significance: Jun 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures in 10 CFR 50.59 Screening An NRC identified, Non-Cited Violation (NCV) of Technical Specification 5.6.1.1 was identified for failure to fully implement a procedure which required a 10 CFR 50.59 evaluation to be completed for a one-time test of the power operated relief valve (PORV). Because the required evaluation was not completed, the licensee was unaware that the test would result in opening the PORV. As a result, the PORV unexpectedly opened for a very short period while the plant was operating and caused a reactor pressure transient.

This finding is more than minor because it affected the Primary System Loss of Coolant Accident (LOCA) Initiator attribute of the Initiating Events Cornerstone. The issue was of very low safety significance because although PORV opened for a short period of time with the reactor operating at power, mitigating systems, including the PORV block valve, were available had the valve failed to shut. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3)

Inspection Report# : 2004004(pdf)

Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: FIN Finding Loss of Design Control When an Improper Circuit Card Placed in the Integrated Control System Caused a Reactor Trip A self-revealing Green finding was identified for a loss of design control during maintenance on the integrated control system which later resulted in a reactor trip.

This finding is more than minor because it affected the configuration control attribute of the initiating event cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. The issue was of very low safety significance because although there was a reactor trip, mitigating systems remained available and were not affected. Because no safety systems were affected, the finding did not involve a violation of regulatory requirements. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3.1)

Inspection Report# : 2004003(pdf)

Significance: Dec 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Perform the Magnetic Particle Calibration The inspector identified a non-cited violation of Technical Specification 5.6.1.1 for failure to follow procedural requirements involving incorrect calibration of a magnetic particle testing (MT) yoke. This finding could have inhibited the identification of indications or flaws on American Society of Mechanical Engineers (ASME) Class 2 Safety-Related Feed Water to Once Through Steam Generator (OTSG) "A" piping.

This finding is more than minor because if left uncorrected, it could result in a more significant safety concern. Failure to correctly perform the calibration could reduce the ability to discover indications or flaws which could lead to pipe breaks. The issue was determined to be of very low safety significance because the likelihood of a loss of coolant accident (LOCA) initiator was not affected, the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not have available, and the finding did not increase the likelihood of a fire or flood. (Section 1R08)

Inspection Report# : 2003006(pdf)

Mitigating Systems Significance: Sep 25, 2004

3Q/2004 Inspection Findings - Crystal River 3 Page 2 of 3 Identified By: NRC Item Type: NCV NonCited Violation Failure To Investigate Deficient Condition of Boric Acid Leakage Affecting The Low Pressure Injection System As Required By Boric Acid Corrosion Control Procedure The inspectors identified a Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion V, for failure to follow boric acid corrosion control program procedures that required an investigation of boric acid leakage identified on decay heat pump DHP-1B.

This finding is more than minor because if left uncorrected it could become a more significant concern, that being loss of integrity of components in the low pressure injection system. The finding was of very low safety significance because only minimal corrosion was observed when inspected. (Section 1RO4)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure To Establish Adequate Corrective Actions For Fire Brigade Response Results In A Recurrent Problem The inspectors identified a Non-Cited Violation (NCV) of Crystal River 3 Operating License Condition 2.C.(9) when prompt corrective measures were not taken to ensure the availability of a fire brigade member to respond to a fire emergency.

This finding is more than minor because if left uncorrected, adequate fire response capability would be challenged which would be a more significant safety concern. A significance determination process review assumed fire confinement was affected with a low degradation rating which resulted in the finding being screened as having very low safety significance. The finding involved the cross-cutting element of problem and identification of resolution, in that interim corrective actions were narrowly focused and ineffective to prevent recurrence. (Section 1RO5)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Redundant Channels of A Post-Accident Monitoring Function Not Operable Due To Reversed Power Supplies Redundant channels of a post-accident monitoring function not operable due to reversed power supp A self-revealing Non-Cited Violation (NCV) of Technical Specification 3.3.17 D was identified when both channels of the Degrees of Subcooling Monitor were found to have their respective power supplies crossed.

The finding was more than minor because the failure of degrees of subcooling monitor indication during certain LOCA scenarios could challenge the control room operators in taking timely action to establish the plant conditions (trip reactor coolant pumps within one minute) needed to assure safety. The finding was of very low safety significance because operators retained the ability to diagnose a loss of subcooling margin using emergency operating procedures had a loss of subcooling margin occurred. (Section 4OA3)

Inspection Report# : 2004005(pdf)

Barrier Integrity Significance: Dec 27, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Identify and Correct a Small Pressure Boundary Leak in The Pressurizer Upper Level Instrument Tap Nozzles A self-revealing non-cited violation of Technical Specification 3.4.12.a was identified. Small cracks in the pressurizer upper level instrument tap nozzles resulted in pressure boundary leakage since late 2000.

The finding was greater than minor because the breach in the reactor coolant system (RCS) affected the RCS barrier performance attribute of the Barrier Integrity Cornerstone objective. However, the cracks were very small, were axial in direction, and therefore, were not expected to grow large enough to challenge the structural stability of the nozzle. A Phase 3 analysis was performed and because the likelihood of a LOCA initiator was not affected, the finding was determined to be of very low safety significance. (Section 4OA3.3)

Inspection Report# : 2003006(pdf)

3Q/2004 Inspection Findings - Crystal River 3 Page 3 of 3 Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jul 02, 2004 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The licensee's corrective action program was generally effective at identifying problems at an appropriate threshold level and entering them into the corrective action program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluation for issues classified as significant conditions adverse to quality were especially comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. The inspectors generally found that the scope and depth of corrective actions implemented by the licensee were appropriate for the severity and risk significance of the problem identified. Industry operating experience items were effectively evaluated for applicability and entered into the corrective action program (CAP). Nuclear Assessment Section (NAS) audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self- assessments were appropriately entered into the corrective action program and addressed. Based on discussions conducted with plant employees from various departments the inspectors did not identify any reluctance to report safety concerns. Further, the inspectors concluded that the licensee was aggressive in addressing potential chilling effect issues. However, the inspectors observed from the more recent data reviewed that several lower threshold issues had not been entered into the CAP. In addition, several examples were identified where problem evaluations lacked thoroughness or were narrowly focused.

Inspection Report# : 2004007(pdf)

Last modified : December 29, 2004

4Q/2004 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 4Q/2004 Plant Inspection Findings Initiating Events Significance: Jun 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures in 10 CFR 50.59 Screening An NRC identified, Non-Cited Violation (NCV) of Technical Specification 5.6.1.1 was identified for failure to fully implement a procedure which required a 10 CFR 50.59 evaluation to be completed for a one-time test of the power operated relief valve (PORV). Because the required evaluation was not completed, the licensee was unaware that the test would result in opening the PORV. As a result, the PORV unexpectedly opened for a very short period while the plant was operating and caused a reactor pressure transient.

This finding is more than minor because it affected the Primary System Loss of Coolant Accident (LOCA) Initiator attribute of the Initiating Events Cornerstone. The issue was of very low safety significance because although PORV opened for a short period of time with the reactor operating at power, mitigating systems, including the PORV block valve, were available had the valve failed to shut. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3)

Inspection Report# : 2004004(pdf)

Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: FIN Finding Loss of Design Control When an Improper Circuit Card Placed in the Integrated Control System Caused a Reactor Trip A self-revealing Green finding was identified for a loss of design control during maintenance on the integrated control system which later resulted in a reactor trip.

This finding is more than minor because it affected the configuration control attribute of the initiating event cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. The issue was of very low safety significance because although there was a reactor trip, mitigating systems remained available and were not affected. Because no safety systems were affected, the finding did not involve a violation of regulatory requirements. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3.1)

Inspection Report# : 2004003(pdf)

Mitigating Systems Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure To Investigate Deficient Condition of Boric Acid Leakage Affecting The Low Pressure Injection System As Required By Boric Acid Corrosion Control Procedure The inspectors identified a Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion V, for failure to follow boric acid corrosion control program procedures that required an investigation of boric acid leakage identified on decay heat pump DHP-1B.

This finding is more than minor because if left uncorrected it could become a more significant concern, that being loss of integrity of components in the low pressure injection system. The finding was of very low safety significance because only minimal corrosion was observed when inspected. (Section 1RO4)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation

4Q/2004 Inspection Findings - Crystal River 3 Page 2 of 3 Redundant Channels of A Post-Accident Monitoring Function Not Operable Due To Reversed Power Supplies Redundant channels of a post-accident monitoring function not operable due to reversed power supp A self-revealing Non-Cited Violation (NCV) of Technical Specification 3.3.17 D was identified when both channels of the Degrees of Subcooling Monitor were found to have their respective power supplies crossed.

The finding was more than minor because the failure of degrees of subcooling monitor indication during certain LOCA scenarios could challenge the control room operators in taking timely action to establish the plant conditions (trip reactor coolant pumps within one minute) needed to assure safety. The finding was of very low safety significance because operators retained the ability to diagnose a loss of subcooling margin using emergency operating procedures had a loss of subcooling margin occurred. (Section 4OA3)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure To Establish Adequate Corrective Actions For Fire Brigade Response Results In A Recurrent Problem The inspectors identified a Non-Cited Violation (NCV) of Crystal River 3 Operating License Condition 2.C.(9) when prompt corrective measures were not taken to ensure the availability of a fire brigade member to respond to a fire emergency.

This finding is more than minor because if left uncorrected, adequate fire response capability would be challenged which would be a more significant safety concern. A significance determination process review assumed fire confinement was affected with a low degradation rating which resulted in the finding being screened as having very low safety significance. The finding involved the cross-cutting element of problem and identification of resolution, in that interim corrective actions were narrowly focused and ineffective to prevent recurrence. (Section 1RO5)

Inspection Report# : 2004005(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jul 02, 2004 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The licensee's corrective action program was generally effective at identifying problems at an appropriate threshold level and entering them into the corrective action program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluation

4Q/2004 Inspection Findings - Crystal River 3 Page 3 of 3 for issues classified as significant conditions adverse to quality were especially comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. The inspectors generally found that the scope and depth of corrective actions implemented by the licensee were appropriate for the severity and risk significance of the problem identified. Industry operating experience items were effectively evaluated for applicability and entered into the corrective action program (CAP). Nuclear Assessment Section (NAS) audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self- assessments were appropriately entered into the corrective action program and addressed. Based on discussions conducted with plant employees from various departments the inspectors did not identify any reluctance to report safety concerns. Further, the inspectors concluded that the licensee was aggressive in addressing potential chilling effect issues. However, the inspectors observed from the more recent data reviewed that several lower threshold issues had not been entered into the CAP. In addition, several examples were identified where problem evaluations lacked thoroughness or were narrowly focused.

Inspection Report# : 2004007(pdf)

Last modified : March 09, 2005

1Q/2005 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 1Q/2005 Plant Inspection Findings Initiating Events Significance: Jun 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures in 10 CFR 50.59 Screening An NRC identified, Non-Cited Violation (NCV) of Technical Specification 5.6.1.1 was identified for failure to fully implement a procedure which required a 10 CFR 50.59 evaluation to be completed for a one-time test of the power operated relief valve (PORV). Because the required evaluation was not completed, the licensee was unaware that the test would result in opening the PORV. As a result, the PORV unexpectedly opened for a very short period while the plant was operating and caused a reactor pressure transient.

This finding is more than minor because it affected the Primary System Loss of Coolant Accident (LOCA) Initiator attribute of the Initiating Events Cornerstone. The issue was of very low safety significance because although PORV opened for a short period of time with the reactor operating at power, mitigating systems, including the PORV block valve, were available had the valve failed to shut. The cause of the finding involved the cross-cutting element of human performance. (Section 4OA3)

Inspection Report# : 2004004(pdf)

Mitigating Systems Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure To Investigate Deficient Condition of Boric Acid Leakage Affecting The Low Pressure Injection System As Required By Boric Acid Corrosion Control Procedure The inspectors identified a Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion V, for failure to follow boric acid corrosion control program procedures that required an investigation of boric acid leakage identified on decay heat pump DHP-1B.

This finding is more than minor because if left uncorrected it could become a more significant concern, that being loss of integrity of components in the low pressure injection system. The finding was of very low safety significance because only minimal corrosion was observed when inspected. (Section 1RO4)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Redundant Channels of A Post-Accident Monitoring Function Not Operable Due To Reversed Power Supplies Redundant channels of a post-accident monitoring function not operable due to reversed power supp A self-revealing Non-Cited Violation (NCV) of Technical Specification 3.3.17 D was identified when both channels of the Degrees of Subcooling Monitor were found to have their respective power supplies crossed.

The finding was more than minor because the failure of degrees of subcooling monitor indication during certain LOCA scenarios could challenge the control room operators in taking timely action to establish the plant conditions (trip reactor coolant pumps within one minute) needed to assure safety. The finding was of very low safety significance because operators retained the ability to diagnose a loss of subcooling margin using emergency operating procedures had a loss of subcooling margin occurred. (Section 4OA3)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: NRC

1Q/2005 Inspection Findings - Crystal River 3 Page 2 of 2 Item Type: NCV NonCited Violation Failure To Establish Adequate Corrective Actions For Fire Brigade Response Results In A Recurrent Problem The inspectors identified a Non-Cited Violation (NCV) of Crystal River 3 Operating License Condition 2.C.(9) when prompt corrective measures were not taken to ensure the availability of a fire brigade member to respond to a fire emergency.

This finding is more than minor because if left uncorrected, adequate fire response capability would be challenged which would be a more significant safety concern. A significance determination process review assumed fire confinement was affected with a low degradation rating which resulted in the finding being screened as having very low safety significance. The finding involved the cross-cutting element of problem and identification of resolution, in that interim corrective actions were narrowly focused and ineffective to prevent recurrence. (Section 1RO5)

Inspection Report# : 2004005(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jul 02, 2004 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The licensee's corrective action program was generally effective at identifying problems at an appropriate threshold level and entering them into the corrective action program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluation for issues classified as significant conditions adverse to quality were especially comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. The inspectors generally found that the scope and depth of corrective actions implemented by the licensee were appropriate for the severity and risk significance of the problem identified. Industry operating experience items were effectively evaluated for applicability and entered into the corrective action program (CAP). Nuclear Assessment Section (NAS) audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self- assessments were appropriately entered into the corrective action program and addressed. Based on discussions conducted with plant employees from various departments the inspectors did not identify any reluctance to report safety concerns. Further, the inspectors concluded that the licensee was aggressive in addressing potential chilling effect issues. However, the inspectors observed from the more recent data reviewed that several lower threshold issues had not been entered into the CAP. In addition, several examples were identified where problem evaluations lacked thoroughness or were narrowly focused.

Inspection Report# : 2004007(pdf)

Last modified : June 17, 2005

2Q/2005 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 2Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish appropriate quantitative acceptance criteria to assure Crystal River 3 Technical Specification 3.8.1 operability of the offsite power supply The inspectors identified a non-cited violation when the licensee failed to establish appropriate quantitative acceptance criteria to assure offsite power operability for compliance with Crystal River 3 Technical Specification 3.8.1.

This finding is more than minor because if left uncorrected, a more significant safety concern could occur if a low voltage condition of the offsite power supply was not immediately recognized and corrected. The finding was of very low safety significance according to the SDP Phase 1 worksheet since none of the functions identified in phase 1 were degraded as a result of this deficiency. (Section 1R22)

Inspection Report# : 2005003(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly evaluate and correct emergency diesel generator loss of fuel oil header prime condition caused by leakage past the fuel header check valves A self revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly evaluate and correct a long standing emergency diesel generator (EGDG) loss of fuel oil header prime condition caused by leakage past the fuel header check valves.

As a result, two separate slow fast start failures occurred in the A' EGDG during fast start surveillance tests, conducted on April 23, 2004 (NCR 125149) and March 23, 2005 (NCR 154522). A third related failure was identified in July 5, 2001, when the B' EGDG failed to start during a monthly surveillance test (NCR 44603).

This finding is more than minor because it directly affected the mitigating system cornerstone objective of ensuring the reliability and operability of a mitigating system. This issue was of very low safety significance because the condition did not affect the capability of the A' EGDG to perform its design safety function. In addition, the slower fast start time was bounded by the accident analysis calculations.

Corrective actions included, replacing the fuel oil check valves with a higher closing spring force valve, priming the system three times per month, and initiating actions to modify the fuel oil system. (Section 4OA3)

Inspection Report# : 2005003(pdf)

Significance: TBD Jun 16, 2005 Identified By: NRC Item Type: AV Apparent Violation Unprotected Post-Fire Safe Shutdown Cables and Related Non-feasible Local Manual Operator Action Preliminary Greater than Green. An apparent violation of 10 CFR 50, Appendix R, Section III.G.2, for failure to physically protect or separate cables from fire damage and instead relying on an unapproved local manual operator action. The unprotected cables were associated with a common electrical protection and metering circuit which was installed such that fire damage to a cable in or just above the 3A 4160V engineered safeguards (ES) switchgear could result in tripping and locking out all feeder breakers to both 4160V ES busses, resulting in a loss of all safety-related alternating current power.

In addition, the local manual operator action to reset the 3B emergency diesel generator breaker lockout on the 3B 4160V ES switchgear was determined to be non-feasible. During a severe fire in the adjacent 3A 4160V Switchgear Room the fire response activities would cause the location for the operator action (the 3B 4160V Switchgear Room) to be exposed to hot smoke, water mist, and water on the floor.

This finding is greater than minor because it degraded the defense in depth for fire protection and also because it is associated with the protection against external factors attribute and degraded the reactor safety mitigating systems cornerstone objective. The finding adversely

2Q/2005 Inspection Findings - Crystal River 3 Page 2 of 3 affected the reliability and capability of equipment required to achieve and maintain a safe shutdown condition following a severe fire in the 3A 4160V ES Switchgear Room. (Section 4OA5.01)

Inspection Report# : 2005007(pdf)

Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure To Investigate Deficient Condition of Boric Acid Leakage Affecting The Low Pressure Injection System As Required By Boric Acid Corrosion Control Procedure The inspectors identified a Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion V, for failure to follow boric acid corrosion control program procedures that required an investigation of boric acid leakage identified on decay heat pump DHP-1B.

This finding is more than minor because if left uncorrected it could become a more significant concern, that being loss of integrity of components in the low pressure injection system. The finding was of very low safety significance because only minimal corrosion was observed when inspected. (Section 1RO4)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Redundant Channels of A Post-Accident Monitoring Function Not Operable Due To Reversed Power Supplies Redundant channels of a post-accident monitoring function not operable due to reversed power supp A self-revealing Non-Cited Violation (NCV) of Technical Specification 3.3.17 D was identified when both channels of the Degrees of Subcooling Monitor were found to have their respective power supplies crossed.

The finding was more than minor because the failure of degrees of subcooling monitor indication during certain LOCA scenarios could challenge the control room operators in taking timely action to establish the plant conditions (trip reactor coolant pumps within one minute) needed to assure safety. The finding was of very low safety significance because operators retained the ability to diagnose a loss of subcooling margin using emergency operating procedures had a loss of subcooling margin occurred. (Section 4OA3)

Inspection Report# : 2004005(pdf)

Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure To Establish Adequate Corrective Actions For Fire Brigade Response Results In A Recurrent Problem The inspectors identified a Non-Cited Violation (NCV) of Crystal River 3 Operating License Condition 2.C.(9) when prompt corrective measures were not taken to ensure the availability of a fire brigade member to respond to a fire emergency.

This finding is more than minor because if left uncorrected, adequate fire response capability would be challenged which would be a more significant safety concern. A significance determination process review assumed fire confinement was affected with a low degradation rating which resulted in the finding being screened as having very low safety significance. The finding involved the cross-cutting element of problem and identification of resolution, in that interim corrective actions were narrowly focused and ineffective to prevent recurrence. (Section 1RO5)

Inspection Report# : 2004005(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety

2Q/2005 Inspection Findings - Crystal River 3 Page 3 of 3 Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jul 02, 2004 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The licensee's corrective action program was generally effective at identifying problems at an appropriate threshold level and entering them into the corrective action program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluation for issues classified as significant conditions adverse to quality were especially comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. The inspectors generally found that the scope and depth of corrective actions implemented by the licensee were appropriate for the severity and risk significance of the problem identified. Industry operating experience items were effectively evaluated for applicability and entered into the corrective action program (CAP). Nuclear Assessment Section (NAS) audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self- assessments were appropriately entered into the corrective action program and addressed. Based on discussions conducted with plant employees from various departments the inspectors did not identify any reluctance to report safety concerns. Further, the inspectors concluded that the licensee was aggressive in addressing potential chilling effect issues. However, the inspectors observed from the more recent data reviewed that several lower threshold issues had not been entered into the CAP. In addition, several examples were identified where problem evaluations lacked thoroughness or were narrowly focused.

Inspection Report# : 2004007(pdf)

Last modified : August 24, 2005

3Q/2005 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 3Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly assess and correct condition of water in the 1A diesel fuel tank An NRC identified, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly assess and correct a long standing issue associated with minor amounts of water intrusion into the 1A Emergency Diesel Generator (EGDG) diesel fuel tank (DFT). As a result, the DFT remained susceptible to water intrusion during a postulated peak high tide associated with a probable maximum hurricane which could have affected the operability of the 1A EGDG. The licensee entered the issue into the corrective action program and is performing a root cause evaluation to determine short and long term corrective actions.

This finding is more than minor because it affected the protection against external factors attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring availability of a mitigating system. During a design basis flood event, enough water could have entered the DFT through a loose cap adapter connection to render the 1A EGDG inoperable. Using NRC Manual Chapter 0609, "Significance Determination Process," Appendix A, Phase 1, this finding was determined to require a Phase 3 analysis since the finding screened as potentially risk significant due to a flooding initiating event. A Regional Senior Reactor Analyst performed the Phase 3 evaluation and determined that the finding was of very low safety significance. This finding also involved a cross cutting aspect of Problem Identification and Resolution, because station personnel missed several opportunities to properly assess and correct this degraded condition.(Section 1R06)

Inspection Report# : 2005004(pdf)

Significance: Sep 07, 2005 Identified By: NRC Item Type: VIO Violation Unprotected Post-Fire Safe Shutdown Cables and Related Non-feasible Local Manual Operator Action 10 CFR 50.48(b)(1) requires, in part, that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G, Fire Protection of Safe Shutdown Capability.

Section III.G.2 states that, except as provided for in Section III.G.3, where cables or equipment (including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground) of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:

a. separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating (Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier.);

b. separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.); or

c. enclosure of cable, equipment, and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.)

Contrary to the above, on January 26, 2005, the licensee failed to ensure that one of the redundant trains of systems necessary to achieve and maintain hot shutdown conditions would be free of fire damage via one of the three means specified in 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, cables for the electrical protection and metering circuit located in the 3A 4160-V engineered safeguards (ES) switchgear room were vulnerable to fire damage that could disable both the 3A 4160-V ES switchgear and the redundant train 3B 4160-V ES switchgear resulting in a loss of all safety-related alternating current power.

This violation is associated with a White Significance Determination Process finding for Unit 3 in the mitigating systems cornerstone.

3Q/2005 Inspection Findings - Crystal River 3 Page 2 of 3 Inspection Report# : 2005011(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish appropriate quantitative acceptance criteria to assure Crystal River 3 Technical Specification 3.8.1 operability of the offsite power supply The inspectors identified a non-cited violation when the licensee failed to establish appropriate quantitative acceptance criteria to assure offsite power operability for compliance with Crystal River 3 Technical Specification 3.8.1.

This finding is more than minor because if left uncorrected, a more significant safety concern could occur if a low voltage condition of the offsite power supply was not immediately recognized and corrected. The finding was of very low safety significance according to the SDP Phase 1 worksheet since none of the functions identified in phase 1 were degraded as a result of this deficiency. (Section 1R22)

Inspection Report# : 2005003(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly evaluate and correct emergency diesel generator loss of fuel oil header prime condition caused by leakage past the fuel header check valves A self revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly evaluate and correct a long standing emergency diesel generator (EGDG) loss of fuel oil header prime condition caused by leakage past the fuel header check valves.

As a result, two separate slow fast start failures occurred in the A' EGDG during fast start surveillance tests, conducted on April 23, 2004 (NCR 125149) and March 23, 2005 (NCR 154522). A third related failure was identified in July 5, 2001, when the B' EGDG failed to start during a monthly surveillance test (NCR 44603).

This finding is more than minor because it directly affected the mitigating system cornerstone objective of ensuring the reliability and operability of a mitigating system. This issue was of very low safety significance because the condition did not affect the capability of the A' EGDG to perform its design safety function. In addition, the slower fast start time was bounded by the accident analysis calculations.

Corrective actions included, replacing the fuel oil check valves with a higher closing spring force valve, priming the system three times per month, and initiating actions to modify the fuel oil system. (Section 4OA3)

Inspection Report# : 2005003(pdf)

Barrier Integrity Significance: Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections The inspectors identified a Non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for inadequate corrective action during review of the results of the refueling outage 12 steam generator Tube End Crack inspections. As a result, Crystal River 3 operated with the calculated leakage exceeding the Technical Specification leakage limit. The licensee entered this condition into their corrective action program.

The finding is more than minor because it was associated with steam generator tube integrity and affected the barrier integrity cornerstone, and if left uncorrected, a more significant safety concern could occur if appropriate corrective actions were not applied to unexpected results found during steam generator inspection activities. This finding represents a cross cutting aspect of problem identification and resolution. The finding was of very low safety significance because it did not result in loss of structural integrity of the steam generators, the small increase in estimated leak rate under main steam line break accident scenarios would not have any significant effect on core damage frequency or large early release frequency, and the contained location of flaws in the tubes makes it impossible to cause spontaneous tube ruptures.

Inspection Report# : 2005009(pdf)

Significance: SL-IV Sep 26, 2005 Identified By: NRC

3Q/2005 Inspection Findings - Crystal River 3 Page 3 of 3 Item Type: NCV NonCited Violation Completeness and Accuracy of Information Provided to the NRC Concerning Steam Generator Inspection Results The inspectors identified a Non-cited violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information, for several examples of inaccuracies and incomplete information in required reports and correspondence. The licensee entered this condition into their corrective action program.

This violation was assessed using traditional enforcement because it impacted the regulatory process. The issue is more than minor because the NRC relies on complete and accurate information to reach conclusions concerning the allowable time between steam generator inspections. It was determined to be a Severity Level IV violation because it was not willful, the technical issue associated with the incomplete and inaccurate information was of very low safety significance, and the NRC had not yet made a regulatory decision based on the information.

Inspection Report# : 2005009(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : November 30, 2005

4Q/2005 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 4Q/2005 Plant Inspection Findings Initiating Events Significance: Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate procedure guidance resulted a Loss of Condensate flow and a Manual Reactor Trip A self-revealing finding was identified for failure to provide adequate condensate system procedural guidance to preclude condensate pump operation at critical speed. As a result, prolonged operation at critical speed caused the condensate pump to fail and subsequently, the reactor was manually tripped in anticipation of a loss of the normal heat sink. The licensee entered this issue into the licensee's corrective action program as nuclear condition reports (NCRs) 174440 and 174442.

This finding is more than minor because it affected the procedure quality attribute of the Initiating Events cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. This finding also affected the equipment reliability attribute of the Mitigating Systems Cornerstone objective and resulted in a loss of the normal heat sink. Because two Cornerstones were affected, a Phase 2 analysis was required. The consequences of the finding were assessed through the Significance Determination Process (SDP) Phase 2, and although the likelihood of a unit trip was increased and resulted in a loss of the normal heat sink, the exposure time for this condition was less than 3 days and all other mitigation capabilities described on the Phase 2, SDP worksheet for transient (reactor trip) core damage sequences were maintained. Consequently, the finding was determined to be of very low safety significance (Green). The finding was associated with non-safety related equipment and therefore, no violation of regulatory requirements occurred (Section 4OA3.2).

Inspection Report# : 2005005(pdf)

Mitigating Systems Significance: Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly assess and correct condition of water in the 1A diesel fuel tank An NRC identified, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly assess and correct a long standing issue associated with minor amounts of water intrusion into the 1A Emergency Diesel Generator (EGDG) diesel fuel tank (DFT). As a result, the DFT remained susceptible to water intrusion during a postulated peak high tide associated with a probable maximum hurricane which could have affected the operability of the 1A EGDG. The licensee entered the issue into the corrective action program and is performing a root cause evaluation to determine short and long term corrective actions.

This finding is more than minor because it affected the protection against external factors attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring availability of a mitigating system. During a design basis flood event, enough water could have entered the DFT through a loose cap adapter connection to render the 1A EGDG inoperable. Using NRC Manual Chapter 0609, "Significance Determination Process," Appendix A, Phase 1, this finding was determined to require a Phase 3 analysis since the finding screened as potentially risk significant due to a flooding initiating event. A Regional Senior Reactor Analyst performed the Phase 3 evaluation and determined that the finding was of very low safety significance. This finding also involved a cross cutting aspect of Problem Identification and Resolution, because station personnel missed several opportunities to properly assess and correct this degraded condition.(Section 1R06)

Inspection Report# : 2005004(pdf)

Significance: Sep 07, 2005 Identified By: NRC Item Type: VIO Violation Unprotected Post-Fire Safe Shutdown Cables and Related Non-feasible Local Manual Operator Action 10 CFR 50.48(b)(1) requires, in part, that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G, Fire Protection of Safe Shutdown Capability.

Section III.G.2 states that, except as provided for in Section III.G.3, where cables or equipment (including associated non-safety circuits that

4Q/2005 Inspection Findings - Crystal River 3 Page 2 of 3 could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground) of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:

a. separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating (Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier.);

b. separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.); or

c. enclosure of cable, equipment, and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.)

Contrary to the above, on January 26, 2005, the licensee failed to ensure that one of the redundant trains of systems necessary to achieve and maintain hot shutdown conditions would be free of fire damage via one of the three means specified in 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, cables for the electrical protection and metering circuit located in the 3A 4160-V engineered safeguards (ES) switchgear room were vulnerable to fire damage that could disable both the 3A 4160-V ES switchgear and the redundant train 3B 4160-V ES switchgear resulting in a loss of all safety-related alternating current power.

This violation is associated with a White Significance Determination Process finding for Unit 3 in the mitigating systems cornerstone.

Inspection Report# : 2005011(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish appropriate quantitative acceptance criteria to assure Crystal River 3 Technical Specification 3.8.1 operability of the offsite power supply The inspectors identified a non-cited violation when the licensee failed to establish appropriate quantitative acceptance criteria to assure offsite power operability for compliance with Crystal River 3 Technical Specification 3.8.1.

This finding is more than minor because if left uncorrected, a more significant safety concern could occur if a low voltage condition of the offsite power supply was not immediately recognized and corrected. The finding was of very low safety significance according to the SDP Phase 1 worksheet since none of the functions identified in phase 1 were degraded as a result of this deficiency. (Section 1R22)

Inspection Report# : 2005003(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly evaluate and correct emergency diesel generator loss of fuel oil header prime condition caused by leakage past the fuel header check valves A self revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly evaluate and correct a long standing emergency diesel generator (EGDG) loss of fuel oil header prime condition caused by leakage past the fuel header check valves.

As a result, two separate slow fast start failures occurred in the A' EGDG during fast start surveillance tests, conducted on April 23, 2004 (NCR 125149) and March 23, 2005 (NCR 154522). A third related failure was identified in July 5, 2001, when the B' EGDG failed to start during a monthly surveillance test (NCR 44603).

This finding is more than minor because it directly affected the mitigating system cornerstone objective of ensuring the reliability and operability of a mitigating system. This issue was of very low safety significance because the condition did not affect the capability of the A' EGDG to perform its design safety function. In addition, the slower fast start time was bounded by the accident analysis calculations.

Corrective actions included, replacing the fuel oil check valves with a higher closing spring force valve, priming the system three times per month, and initiating actions to modify the fuel oil system. (Section 4OA3)

Inspection Report# : 2005003(pdf)

4Q/2005 Inspection Findings - Crystal River 3 Page 3 of 3 Significance: Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections The inspectors identified a Non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for inadequate corrective action during review of the results of the refueling outage 12 steam generator Tube End Crack inspections. As a result, Crystal River 3 operated with the calculated leakage exceeding the Technical Specification leakage limit. The licensee entered this condition into their corrective action program.

The finding is more than minor because it was associated with steam generator tube integrity and affected the barrier integrity cornerstone, and if left uncorrected, a more significant safety concern could occur if appropriate corrective actions were not applied to unexpected results found during steam generator inspection activities. This finding represents a cross cutting aspect of problem identification and resolution. The finding was of very low safety significance because it did not result in loss of structural integrity of the steam generators, the small increase in estimated leak rate under main steam line break accident scenarios would not have any significant effect on core damage frequency or large early release frequency, and the contained location of flaws in the tubes makes it impossible to cause spontaneous tube ruptures.

Inspection Report# : 2005009(pdf)

Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Completeness and Accuracy of Information Provided to the NRC Concerning Steam Generator Inspection Results The inspectors identified a Non-cited violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information, for several examples of inaccuracies and incomplete information in required reports and correspondence. The licensee entered this condition into their corrective action program.

This violation was assessed using traditional enforcement because it impacted the regulatory process. The issue is more than minor because the NRC relies on complete and accurate information to reach conclusions concerning the allowable time between steam generator inspections. It was determined to be a Severity Level IV violation because it was not willful, the technical issue associated with the incomplete and inaccurate information was of very low safety significance, and the NRC had not yet made a regulatory decision based on the information.

Inspection Report# : 2005009(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : March 03, 2006

1Q/2006 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 1Q/2006 Plant Inspection Findings Initiating Events Significance: Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate procedure guidance resulted a Loss of Condensate flow and a Manual Reactor Trip A self-revealing finding was identified for failure to provide adequate condensate system procedural guidance to preclude condensate pump operation at critical speed. As a result, prolonged operation at critical speed caused the condensate pump to fail and subsequently, the reactor was manually tripped in anticipation of a loss of the normal heat sink. The licensee entered this issue into the licensee's corrective action program as nuclear condition reports (NCRs) 174440 and 174442.

This finding is more than minor because it affected the procedure quality attribute of the Initiating Events cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. This finding also affected the equipment reliability attribute of the Mitigating Systems Cornerstone objective and resulted in a loss of the normal heat sink. Because two Cornerstones were affected, a Phase 2 analysis was required. The consequences of the finding were assessed through the Significance Determination Process (SDP) Phase 2, and although the likelihood of a unit trip was increased and resulted in a loss of the normal heat sink, the exposure time for this condition was less than 3 days and all other mitigation capabilities described on the Phase 2, SDP worksheet for transient (reactor trip) core damage sequences were maintained. Consequently, the finding was determined to be of very low safety significance (Green). The finding was associated with non-safety related equipment and therefore, no violation of regulatory requirements occurred (Section 4OA3.2).

Inspection Report# : 2005005(pdf)

Mitigating Systems Significance: Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly assess and correct condition of water in the 1A diesel fuel tank An NRC identified, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly assess and correct a long standing issue associated with minor amounts of water intrusion into the 1A Emergency Diesel Generator (EGDG) diesel fuel tank (DFT). As a result, the DFT remained susceptible to water intrusion during a postulated peak high tide associated with a probable maximum hurricane which could have affected the operability of the 1A EGDG. The licensee entered the issue into the corrective action program and is performing a root cause evaluation to determine short and long term corrective actions.

This finding is more than minor because it affected the protection against external factors attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring availability of a mitigating system. During a design basis flood event, enough water could have entered the DFT through a loose cap adapter connection to render the 1A EGDG inoperable. Using NRC Manual Chapter 0609, "Significance Determination Process," Appendix A, Phase 1, this finding was determined to require a Phase 3 analysis since the finding screened as potentially risk significant due to a flooding initiating event. A Regional Senior Reactor Analyst performed the Phase 3 evaluation and determined that the finding was of very low safety significance. This finding also involved a cross cutting aspect of Problem Identification and Resolution, because station personnel missed several opportunities to properly assess and correct this degraded condition.(Section 1R06)

Inspection Report# : 2005004(pdf)

Significance: Sep 07, 2005 Identified By: NRC Item Type: VIO Violation Unprotected Post-Fire Safe Shutdown Cables and Related Non-feasible Local Manual Operator Action 10 CFR 50.48(b)(1) requires, in part, that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G, Fire Protection of Safe Shutdown Capability.

Section III.G.2 states that, except as provided for in Section III.G.3, where cables or equipment (including associated non-safety circuits that

1Q/2006 Inspection Findings - Crystal River 3 Page 2 of 3 could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground) of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:

a. separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating (Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier.);

b. separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.); or

c. enclosure of cable, equipment, and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.)

Contrary to the above, on January 26, 2005, the licensee failed to ensure that one of the redundant trains of systems necessary to achieve and maintain hot shutdown conditions would be free of fire damage via one of the three means specified in 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, cables for the electrical protection and metering circuit located in the 3A 4160-V engineered safeguards (ES) switchgear room were vulnerable to fire damage that could disable both the 3A 4160-V ES switchgear and the redundant train 3B 4160-V ES switchgear resulting in a loss of all safety-related alternating current power.

This violation is associated with a White Significance Determination Process finding for Unit 3 in the mitigating systems cornerstone.

Inspection Report# : 2005011(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish appropriate quantitative acceptance criteria to assure Crystal River 3 Technical Specification 3.8.1 operability of the offsite power supply The inspectors identified a non-cited violation when the licensee failed to establish appropriate quantitative acceptance criteria to assure offsite power operability for compliance with Crystal River 3 Technical Specification 3.8.1.

This finding is more than minor because if left uncorrected, a more significant safety concern could occur if a low voltage condition of the offsite power supply was not immediately recognized and corrected. The finding was of very low safety significance according to the SDP Phase 1 worksheet since none of the functions identified in phase 1 were degraded as a result of this deficiency. (Section 1R22)

Inspection Report# : 2005003(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly evaluate and correct emergency diesel generator loss of fuel oil header prime condition caused by leakage past the fuel header check valves A self revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly evaluate and correct a long standing emergency diesel generator (EGDG) loss of fuel oil header prime condition caused by leakage past the fuel header check valves.

As a result, two separate slow fast start failures occurred in the A' EGDG during fast start surveillance tests, conducted on April 23, 2004 (NCR 125149) and March 23, 2005 (NCR 154522). A third related failure was identified in July 5, 2001, when the B' EGDG failed to start during a monthly surveillance test (NCR 44603).

This finding is more than minor because it directly affected the mitigating system cornerstone objective of ensuring the reliability and operability of a mitigating system. This issue was of very low safety significance because the condition did not affect the capability of the A' EGDG to perform its design safety function. In addition, the slower fast start time was bounded by the accident analysis calculations.

Corrective actions included, replacing the fuel oil check valves with a higher closing spring force valve, priming the system three times per month, and initiating actions to modify the fuel oil system. (Section 4OA3)

Inspection Report# : 2005003(pdf)

1Q/2006 Inspection Findings - Crystal River 3 Page 3 of 3 Significance: Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections The inspectors identified a Non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for inadequate corrective action during review of the results of the refueling outage 12 steam generator Tube End Crack inspections. As a result, Crystal River 3 operated with the calculated leakage exceeding the Technical Specification leakage limit. The licensee entered this condition into their corrective action program.

The finding is more than minor because it was associated with steam generator tube integrity and affected the barrier integrity cornerstone, and if left uncorrected, a more significant safety concern could occur if appropriate corrective actions were not applied to unexpected results found during steam generator inspection activities. This finding represents a cross cutting aspect of problem identification and resolution. The finding was of very low safety significance because it did not result in loss of structural integrity of the steam generators, the small increase in estimated leak rate under main steam line break accident scenarios would not have any significant effect on core damage frequency or large early release frequency, and the contained location of flaws in the tubes makes it impossible to cause spontaneous tube ruptures.

Inspection Report# : 2005009(pdf)

Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Completeness and Accuracy of Information Provided to the NRC Concerning Steam Generator Inspection Results The inspectors identified a Non-cited violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information, for several examples of inaccuracies and incomplete information in required reports and correspondence. The licensee entered this condition into their corrective action program.

This violation was assessed using traditional enforcement because it impacted the regulatory process. The issue is more than minor because the NRC relies on complete and accurate information to reach conclusions concerning the allowable time between steam generator inspections. It was determined to be a Severity Level IV violation because it was not willful, the technical issue associated with the incomplete and inaccurate information was of very low safety significance, and the NRC had not yet made a regulatory decision based on the information.

Inspection Report# : 2005009(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : May 25, 2006

2Q/2006 Inspection Findings - Crystal River 3 Page 1 of 3 Crystal River 3 2Q/2006 Plant Inspection Findings Initiating Events Significance: Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate procedure guidance resulted a Loss of Condensate flow and a Manual Reactor Trip A self-revealing finding was identified for failure to provide adequate condensate system procedural guidance to preclude condensate pump operation at critical speed. As a result, prolonged operation at critical speed caused the condensate pump to fail and subsequently, the reactor was manually tripped in anticipation of a loss of the normal heat sink. The licensee entered this issue into the licensee's corrective action program as nuclear condition reports (NCRs) 174440 and 174442.

This finding is more than minor because it affected the procedure quality attribute of the Initiating Events cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. This finding also affected the equipment reliability attribute of the Mitigating Systems Cornerstone objective and resulted in a loss of the normal heat sink. Because two Cornerstones were affected, a Phase 2 analysis was required. The consequences of the finding were assessed through the Significance Determination Process (SDP) Phase 2, and although the likelihood of a unit trip was increased and resulted in a loss of the normal heat sink, the exposure time for this condition was less than 3 days and all other mitigation capabilities described on the Phase 2, SDP worksheet for transient (reactor trip) core damage sequences were maintained. Consequently, the finding was determined to be of very low safety significance (Green). The finding was associated with non-safety related equipment and therefore, no violation of regulatory requirements occurred (Section 4OA3.2).

Inspection Report# : 2005005(pdf)

Mitigating Systems Significance: Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Failure to Conduct an Extent of Condition Review after Three Motor Operated Valves Were Found with Their Pinion Gears Installed Incorrectly A Green finding was identified by the inspectors for failure to conduct an extent of condition evaluation when three motor operated valves (MOVs) which were thought to not be susceptible to incorrect pinion gear installation were found with their pinion gears installed backwards.

This finding is more than minor because it affected the equipment performance attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring reliability of a mitigating system. Using NRC Manual Chapter 0609, Significance Determination Process, Appendix A, Phase 1, this finding was determined to be of very low significance (Green), because the finding has not resulted in a loss of safety function and was not screened as potentially risk significant due to external events. The primary cause of the finding was related to the cross cutting area of Problem Identification and Resolution, in that station personnel failed to determine the need for additional MOV inspections when three MOVs which were initially thought to not be susceptible to incorrect pinion gear installation were found with reversed pinion gears, one of which was also discovered with an improperly staked pinion key.

Inspection Report# : 2006009(pdf)

Significance: Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly assess and correct condition of water in the 1A diesel fuel tank An NRC identified, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for failure to properly assess and correct a long standing issue associated with minor amounts of water intrusion into the 1A Emergency Diesel Generator (EGDG) diesel fuel tank (DFT). As a result, the DFT remained susceptible to water intrusion during a postulated peak high tide associated with a probable maximum hurricane which could have affected the operability of the 1A EGDG. The licensee entered the issue into the corrective action program and is performing a root cause evaluation to determine short and long term corrective actions.

2Q/2006 Inspection Findings - Crystal River 3 Page 2 of 3 This finding is more than minor because it affected the protection against external factors attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring availability of a mitigating system. During a design basis flood event, enough water could have entered the DFT through a loose cap adapter connection to render the 1A EGDG inoperable. Using NRC Manual Chapter 0609, "Significance Determination Process," Appendix A, Phase 1, this finding was determined to require a Phase 3 analysis since the finding screened as potentially risk significant due to a flooding initiating event. A Regional Senior Reactor Analyst performed the Phase 3 evaluation and determined that the finding was of very low safety significance. This finding also involved a cross cutting aspect of Problem Identification and Resolution, because station personnel missed several opportunities to properly assess and correct this degraded condition.(Section 1R06)

Inspection Report# : 2005004(pdf)

Significance: Sep 07, 2005 Identified By: NRC Item Type: VIO Violation Unprotected Post-Fire Safe Shutdown Cables and Related Non-feasible Local Manual Operator Action 10 CFR 50.48(b)(1) requires, in part, that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G, Fire Protection of Safe Shutdown Capability.

Section III.G.2 states that, except as provided for in Section III.G.3, where cables or equipment (including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground) of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:

a. separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating (Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier.);

b. separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.); or

c. enclosure of cable, equipment, and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. (In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.)

Contrary to the above, on January 26, 2005, the licensee failed to ensure that one of the redundant trains of systems necessary to achieve and maintain hot shutdown conditions would be free of fire damage via one of the three means specified in 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, cables for the electrical protection and metering circuit located in the 3A 4160-V engineered safeguards (ES) switchgear room were vulnerable to fire damage that could disable both the 3A 4160-V ES switchgear and the redundant train 3B 4160-V ES switchgear resulting in a loss of all safety-related alternating current power.

This violation is associated with a White Significance Determination Process finding for Unit 3 in the mitigating systems cornerstone.

Inspection Report# : 2005011(pdf)

Barrier Integrity Significance: Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections The inspectors identified a Non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for inadequate corrective action during review of the results of the refueling outage 12 steam generator Tube End Crack inspections. As a result, Crystal River 3 operated with the calculated leakage exceeding the Technical Specification leakage limit. The licensee entered this condition into their corrective action program.

The finding is more than minor because it was associated with steam generator tube integrity and affected the barrier integrity cornerstone, and if left uncorrected, a more significant safety concern could occur if appropriate corrective actions were not applied to unexpected results found during steam generator inspection activities. This finding represents a cross cutting aspect of problem identification and resolution. The finding was of very low safety significance because it did not result in loss of structural integrity of the steam generators, the small increase in estimated leak rate under main steam line break accident scenarios would not have any significant effect on core damage frequency or large early release frequency, and the contained location of flaws in the tubes makes it impossible to cause spontaneous tube ruptures.

Inspection Report# : 2005009(pdf)

2Q/2006 Inspection Findings - Crystal River 3 Page 3 of 3 Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Completeness and Accuracy of Information Provided to the NRC Concerning Steam Generator Inspection Results The inspectors identified a Non-cited violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information, for several examples of inaccuracies and incomplete information in required reports and correspondence. The licensee entered this condition into their corrective action program.

This violation was assessed using traditional enforcement because it impacted the regulatory process. The issue is more than minor because the NRC relies on complete and accurate information to reach conclusions concerning the allowable time between steam generator inspections. It was determined to be a Severity Level IV violation because it was not willful, the technical issue associated with the incomplete and inaccurate information was of very low safety significance, and the NRC had not yet made a regulatory decision based on the information.

Inspection Report# : 2005009(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that in general, problems were properly identified, evaluated, prioritized, and corrected within the licensees problem identification and resolution program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluations for issues classified as significant conditions adverse to quality were comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. One exception was noted concerning corrective action for identified deficiencies with three motor-operated valves.

The processes and procedures of the licensees corrective action program (CAP) were generally adequate; thresholds for identifying issues were appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. Nuclear Assessment Section audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self-assessments were appropriately entered into the CAP and addressed. However, the inspectors observed that several lower threshold issues had not been entered into the CAP.

Based on discussions and interviews conducted with plant employees from various departments, the inspectors did not identify any reluctance to report safety concerns.

Inspection Report# : 2006009(pdf)

Last modified : August 25, 2006

3Q/2006 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 3Q/2006 Plant Inspection Findings Initiating Events Significance: Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate procedure guidance resulted a Loss of Condensate flow and a Manual Reactor Trip A self-revealing finding was identified for failure to provide adequate condensate system procedural guidance to preclude condensate pump operation at critical speed. As a result, prolonged operation at critical speed caused the condensate pump to fail and subsequently, the reactor was manually tripped in anticipation of a loss of the normal heat sink. The licensee entered this issue into the licensee's corrective action program as nuclear condition reports (NCRs) 174440 and 174442.

This finding is more than minor because it affected the procedure quality attribute of the Initiating Events cornerstone and resulted in an event that upset plant stability and challenged critical safety functions. This finding also affected the equipment reliability attribute of the Mitigating Systems Cornerstone objective and resulted in a loss of the normal heat sink. Because two Cornerstones were affected, a Phase 2 analysis was required. The consequences of the finding were assessed through the Significance Determination Process (SDP) Phase 2, and although the likelihood of a unit trip was increased and resulted in a loss of the normal heat sink, the exposure time for this condition was less than 3 days and all other mitigation capabilities described on the Phase 2, SDP worksheet for transient (reactor trip) core damage sequences were maintained. Consequently, the finding was determined to be of very low safety significance (Green). The finding was associated with non-safety related equipment and therefore, no violation of regulatory requirements occurred (Section 4OA3.2).

Inspection Report# : 2005005(pdf)

Mitigating Systems Significance: Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Failure to Conduct an Extent of Condition Review after Three Motor Operated Valves Were Found with Their Pinion Gears Installed Incorrectly A Green finding was identified by the inspectors for failure to conduct an extent of condition evaluation when three motor operated valves (MOVs) which were thought to not be susceptible to incorrect pinion gear installation were found with their pinion gears installed backwards.

This finding is more than minor because it affected the equipment performance attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring reliability of a mitigating system. Using NRC Manual Chapter 0609, Significance Determination Process, Appendix A, Phase 1, this finding was determined to be of very low significance (Green), because the finding has not resulted in a loss of safety function and was not screened as potentially risk significant due to external events. The primary cause of the finding was related to the cross cutting area of Problem Identification and Resolution, in that station personnel failed to determine the need for additional MOV inspections when three MOVs which were initially thought to not be susceptible to incorrect pinion gear installation were found with reversed pinion gears, one of which was also discovered with an improperly staked pinion key.

Inspection Report# : 2006009(pdf)

3Q/2006 Inspection Findings - Crystal River 3 Page 2 of 2 Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that in general, problems were properly identified, evaluated, prioritized, and corrected within the licensees problem identification and resolution program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluations for issues classified as significant conditions adverse to quality were comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. One exception was noted concerning corrective action for identified deficiencies with three motor-operated valves.

The processes and procedures of the licensees corrective action program (CAP) were generally adequate; thresholds for identifying issues were appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. Nuclear Assessment Section audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self-assessments were appropriately entered into the CAP and addressed. However, the inspectors observed that several lower threshold issues had not been entered into the CAP.

Based on discussions and interviews conducted with plant employees from various departments, the inspectors did not identify any reluctance to report safety concerns.

Inspection Report# : 2006009(pdf)

Last modified : December 21, 2006

4Q/2006 Inspection Findings - Crystal River 3 Page 1 of 2 Crystal River 3 4Q/2006 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Failure to Conduct an Extent of Condition Review after Three Motor Operated Valves Were Found with Their Pinion Gears Installed Incorrectly A Green finding was identified by the inspectors for failure to conduct an extent of condition evaluation when three motor operated valves (MOVs) which were thought to not be susceptible to incorrect pinion gear installation were found with their pinion gears installed backwards.

This finding is more than minor because it affected the equipment performance attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring reliability of a mitigating system. Using NRC Manual Chapter 0609, Significance Determination Process, Appendix A, Phase 1, this finding was determined to be of very low significance (Green), because the finding has not resulted in a loss of safety function and was not screened as potentially risk significant due to external events. The primary cause of the finding was related to the cross cutting area of Problem Identification and Resolution, in that station personnel failed to determine the need for additional MOV inspections when three MOVs which were initially thought to not be susceptible to incorrect pinion gear installation were found with reversed pinion gears, one of which was also discovered with an improperly staked pinion key.

Inspection Report# : 2006009 (pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation

4Q/2006 Inspection Findings - Crystal River 3 Page 2 of 2 Failure to Conduct Adequate Surveys for Liquid Effluent Releases The inspectors identified a non-cited violation (NCV) of 10 CFR 20.1501(a) for failure to perform accurate surveys to demonstrate compliance with Technical Specification (TS) 5.6.2.3 Offsite Dose Calculation Manual (ODCM) controls used to maintain doses to members of the public from radioactive effluents as low as reasonably achievable (ALARA) in accordance with Appendix I to 10 CFR 50 design criteria as specified in 10 CFR 50.36a. Specifically, as of December 4, 2006, the licensee failed to conduct adequate dose evaluations to demonstrate compliance with TS 5.6.2.3 for radioactive liquid effluent releases made from the station discharge tank SDT-1 to a percolation pond located within the owner controlled area. The failure to conduct accurate dose evaluations for this liquid release pathway impaired the licensees ability to demonstrate compliance with ODCM ALARA limits for the liquid radioactive waste processing equipment and operations. The issue was entered into the licensees corrective action program for resolution.

The violation is more than minor because it adversely affects the program and process attribute of the Public Radiation Safety cornerstone, in that it involved an occurrence in the licensees radioactive effluent release program that is contrary to NRC regulations. The finding was determined to be of very low safety significance because preliminary calculations based on recently determined dilution factors for the settling pond demonstrated that resultant offsite dose values were small fractions of the ODCM limits (Appendix I to 10 CFR Part 50 design criteria). Further, evaluations of radionuclide concentrations in the effluent were conducted in accordance with 10 CFR 20.1302(b) (2)(i) to demonstrate compliance with 10 CFR 20.1301 limits. (Section 2PS1)

Inspection Report# : 2006005 (pdf)

Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that in general, problems were properly identified, evaluated, prioritized, and corrected within the licensees problem identification and resolution program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluations for issues classified as significant conditions adverse to quality were comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. One exception was noted concerning corrective action for identified deficiencies with three motor-operated valves.

The processes and procedures of the licensees corrective action program (CAP) were generally adequate; thresholds for identifying issues were appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. Nuclear Assessment Section audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self-assessments were appropriately entered into the CAP and addressed. However, the inspectors observed that several lower threshold issues had not been entered into the CAP.

Based on discussions and interviews conducted with plant employees from various departments, the inspectors did not identify any reluctance to report safety concerns.

Inspection Report# : 2006009 (pdf)

Last modified : March 01, 2007

Crystal River 3 1Q/2007 Plant Inspection Findings Initiating Events Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Address Marine Fouling Resulted in a Plant Transient A self-revealing finding was identified for the failure to address the marine fouling failure mode in the scope of the existing preventive maintenance on the intake screen wash auto start system. As a result, reactor power had to be decreased to 80 percent to maintain condenser operating temperature limits. The licensee entered the issue into the corrective action program. Corrective actions included cleaning both the low and high side differential level sensing tubes, replacing tubes as needed, and implementing preventive maintenance procedures to periodically clean the tubes.

The finding was more than minor since it affected the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically the Operating Experience (OE) Program, in that, the licensee did not adequately implement OE through changes to station procedures to provide instructions to clean the sensing tubes during preventive maintenance on the system. (Section 4OA2.2)

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Replace a Non-refurbished ICS Module Resulted in a Reactor Trip A self-revealing finding was identified for failure to replace a non-refurbished integrated control system (ICS) multiplier module that had been temporarily installed during the Fall 2005 refueling outage. As a result, an age-related failure of a multiplier module resulted in an automatic reactor trip. The licensee entered the issue into the corrective action program.

Corrective actions completed and/or proposed include: installation of a refurbished multiplier module; development of an engineering refueling outage turnover checklist to ensure formal followup actions are implemented whenever components not of desired quality are installed; and briefing of engineering personnel of this event.

The finding was more than minor because it affected the equipment reliability attribute of the Initiating Events Cornerstone and resulted in an automatic reactor trip that upset plant stability and challenged critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of the finding is related to the cross-cutting area of Human Performance, specifically Decision Making in that the licensee did not adequately communicate decisions and the basis for decisions to personnel who have a need to know the information. (Section 4OA3.1)

Inspection Report# : 2007002 (pdf)

Barrier Integrity Emergency Preparedness

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Identify and Correct Repetitive Raw Water System Flush Water Strainer Baskets Degradation A self-revealing, non-cited Violation of 10 CFR 50, Appendix B, Criterion XVI was identified for failure to identify and take appropriate corrective actions for repetitive failures of the raw water pumps bearing flush water strainer baskets. As a result, both raw water pumps, RWP-2B and RWP-3B, were inoperable for a period greater than that allowed by Improved Technical Specifications when shell debris passed through a corroded strainer and clogged the cyclone separator discharge piping. The licensee entered the issue into the corrective action program. New strainer baskets made of a material compatible with service conditions were installed. Additional corrective actions include: performing routine engineering review of degraded conditions found during preventative maintenance activities; revision to applicable surveillance procedures, and counseling of maintenance and engineering personnel on the need to identify and document adverse conditions in the corrective action program.

The finding was more than minor because it affected the equipment reliability attribute of the Mitigating System Cornerstone and resulted in a raw water train being inoperable for a period of time greater than allowed by Improved Technical Specifications. The finding was assessed through the Significance Determination Process (SDP) Phase 1 screening worksheet and determined to be of very low safety significance since the raw water pumps with a degraded flush water system had a very high likelihood of performing their safety function during a loss of offsite power event. A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically, the licensee did not document the adverse condition of degraded strainer baskets in the corrective action program after it was determined that the filtering ability of the cyclone separator was a required design function. (Section 4OA3.2)

Inspection Report# : 2007002 (pdf)

Significance: Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Failure to Conduct an Extent of Condition Review after Three Motor Operated Valves Were Found with Their Pinion Gears Installed Incorrectly A Green finding was identified by the inspectors for failure to conduct an extent of condition evaluation when three motor operated valves (MOVs) which were thought to not be susceptible to incorrect pinion gear installation were found with their pinion gears installed backwards.

This finding is more than minor because it affected the equipment performance attribute of the mitigating system cornerstone and affected the cornerstone objective of ensuring reliability of a mitigating system. Using NRC Manual Chapter 0609, Significance Determination Process, Appendix A, Phase 1, this finding was determined to be of very low significance (Green), because the finding has not resulted in a loss of safety function and was not screened as potentially risk significant due to external events. The primary cause of the finding was related to the cross cutting area of Problem Identification and Resolution, in that station personnel failed to determine the need for additional MOV inspections when three MOVs which were initially thought to not be susceptible to incorrect pinion gear installation were found with reversed pinion gears, one of which was also discovered with an improperly staked pinion key.

Inspection Report# : 2006009 (pdf)

Occupational Radiation Safety Public Radiation Safety Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Surveys for Liquid Effluent Releases The inspectors identified a non-cited violation (NCV) of 10 CFR 20.1501(a) for failure to perform accurate surveys to demonstrate compliance with Technical Specification (TS) 5.6.2.3 Offsite Dose Calculation Manual (ODCM) controls used to maintain doses to members of the public from radioactive effluents as low as reasonably achievable (ALARA) in accordance with Appendix I to 10 CFR 50 design criteria as specified in 10 CFR 50.36a. Specifically, as of December 4, 2006, the licensee failed to conduct adequate dose evaluations to demonstrate compliance with TS 5.6.2.3 for radioactive liquid effluent releases made from the station discharge tank SDT-1 to a percolation pond located within the owner controlled area. The failure to conduct accurate dose evaluations for this liquid release pathway impaired the licensees ability to demonstrate compliance with ODCM ALARA limits for the liquid radioactive waste processing equipment and operations. The issue was entered into the licensees corrective action program for resolution.

The violation is more than minor because it adversely affects the program and process attribute of the Public Radiation Safety cornerstone, in that it involved an occurrence in the licensees radioactive effluent release program that is contrary to NRC regulations. The finding was determined to be of very low safety significance because preliminary calculations based on recently determined dilution factors for the settling pond demonstrated that resultant offsite dose values were small fractions of the ODCM limits (Appendix I to 10 CFR Part 50 design criteria). Further, evaluations of radionuclide concentrations in the effluent were conducted in accordance with 10 CFR 20.1302(b) (2)(i) to demonstrate compliance with 10 CFR 20.1301 limits. (Section 2PS1)

Inspection Report# : 2006005 (pdf)

Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Jun 23, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that in general, problems were properly identified, evaluated, prioritized, and corrected within the licensees problem identification and resolution program. Evaluation of issues was generally comprehensive and technically adequate. Formal root cause evaluations for issues classified as significant conditions adverse to quality were comprehensive and detailed. Overall, corrective actions developed and implemented for issues were effective in correcting the problems. One exception was noted concerning corrective action for identified deficiencies with three motor-operated valves.

The processes and procedures of the licensees corrective action program (CAP) were generally adequate; thresholds for identifying issues were appropriately low, and in most cases, corrective actions were adequate to address conditions

adverse to quality. Nuclear Assessment Section audits and departmental self-assessments were effective in identifying issues and directing attention to areas that needed improvement. Licensee identified weaknesses and issues in self-assessments were appropriately entered into the CAP and addressed. However, the inspectors observed that several lower threshold issues had not been entered into the CAP.

Based on discussions and interviews conducted with plant employees from various departments, the inspectors did not identify any reluctance to report safety concerns.

Inspection Report# : 2006009 (pdf)

Last modified : June 01, 2007

Crystal River 3 2Q/2007 Plant Inspection Findings Initiating Events Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Address Marine Fouling Resulted in a Plant Transient A self-revealing finding was identified for the failure to address the marine fouling failure mode in the scope of the existing preventive maintenance on the intake screen wash auto start system. As a result, reactor power had to be decreased to 80 percent to maintain condenser operating temperature limits. The licensee entered the issue into the corrective action program. Corrective actions included cleaning both the low and high side differential level sensing tubes, replacing tubes as needed, and implementing preventive maintenance procedures to periodically clean the tubes.

The finding was more than minor since it affected the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically the Operating Experience (OE) Program, in that, the licensee did not adequately implement OE through changes to station procedures to provide instructions to clean the sensing tubes during preventive maintenance on the system. (Section 4OA2.2)

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Replace a Non-refurbished ICS Module Resulted in a Reactor Trip A self-revealing finding was identified for failure to replace a non-refurbished integrated control system (ICS) multiplier module that had been temporarily installed during the Fall 2005 refueling outage. As a result, an age-related failure of a multiplier module resulted in an automatic reactor trip. The licensee entered the issue into the corrective action program. Corrective actions completed and/or proposed include: installation of a refurbished multiplier module; development of an engineering refueling outage turnover checklist to ensure formal followup actions are implemented whenever components not of desired quality are installed; and briefing of engineering personnel of this event.

The finding was more than minor because it affected the equipment reliability attribute of the Initiating Events Cornerstone and resulted in an automatic reactor trip that upset plant stability and challenged critical safety functions.

Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of the finding is related to the cross-cutting area of Human Performance, specifically Decision Making in that the licensee did not adequately communicate decisions and the basis for decisions to personnel who have a need to know the information. (Section 4OA3.1)

Inspection Report# : 2007002 (pdf)

Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance Procedures for Hydrostatic Seals Necessary to Protect Safety-Related Equipment from Internal Flooding The inspectors identified a non-cited violation (NCV) of Improved Technical Specification 5.6.1.1.a, for failure to adequately establish and implement procedures required by Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance (PM). Specifically, no procedure, program or process existed to periodically inspect hydrostatic barriers to identify and repair any degradation of the seals which provide protection of safety-related equipment from internal flooding. Corrective actions completed or planned include: Repair and qualify applicable fire seals as hydrostatic barriers and establish a hydrostatic penetration seal preventative maintenance program.

The finding is more than minor because it affected the protection against external factors (i.e. flood hazard) attribute of the Mitigating System cornerstone and could have impacted the availability of mitigating equipment during an internal flood event if left uncorrected. The inspectors determined that several degraded fire barrier seals did not meet hydrostatic barrier acceptability requirements. The finding was assessed through the SDP Phase 1 screening and determined to be of very low safety significance since the as-found condition of the hydrostatic barriers would not have resulted in the loss or degradation of safety-related mitigating equipment in the event of an internal flood.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Identify and Correct Repetitive Raw Water System Flush Water Strainer Baskets Degradation A self-revealing, non-cited Violation of 10 CFR 50, Appendix B, Criterion XVI was identified for failure to identify and take appropriate corrective actions for repetitive failures of the raw water pumps bearing flush water strainer baskets. As a result, both raw water pumps, RWP-2B and RWP-3B, were inoperable for a period greater than that allowed by Improved Technical Specifications when shell debris passed through a corroded strainer and clogged the cyclone separator discharge piping. The licensee entered the issue into the corrective action program. New strainer baskets made of a material compatible with service conditions were installed. Additional corrective actions include:

performing routine engineering review of degraded conditions found during preventative maintenance activities; revision to applicable surveillance procedures, and counseling of maintenance and engineering personnel on the need to identify and document adverse conditions in the corrective action program.

The finding was more than minor because it affected the equipment reliability attribute of the Mitigating System Cornerstone and resulted in a raw water train being inoperable for a period of time greater than allowed by Improved Technical Specifications. The finding was assessed through the Significance Determination Process (SDP) Phase 1 screening worksheet and determined to be of very low safety significance since the raw water pumps with a degraded flush water system had a very high likelihood of performing their safety function during a loss of offsite power event.

A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically, the licensee did not document the adverse condition of degraded strainer baskets in the corrective action program after it was determined that the filtering ability of the cyclone separator was a required design function.

(Section 4OA3.2)

Inspection Report# : 2007002 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Surveys for Liquid Effluent Releases The inspectors identified a non-cited violation (NCV) of 10 CFR 20.1501(a) for failure to perform accurate surveys to demonstrate compliance with Technical Specification (TS) 5.6.2.3 Offsite Dose Calculation Manual (ODCM) controls used to maintain doses to members of the public from radioactive effluents as low as reasonably achievable (ALARA) in accordance with Appendix I to 10 CFR 50 design criteria as specified in 10 CFR 50.36a. Specifically, as of December 4, 2006, the licensee failed to conduct adequate dose evaluations to demonstrate compliance with TS 5.6.2.3 for radioactive liquid effluent releases made from the station discharge tank SDT-1 to a percolation pond located within the owner controlled area. The failure to conduct accurate dose evaluations for this liquid release pathway impaired the licensees ability to demonstrate compliance with ODCM ALARA limits for the liquid radioactive waste processing equipment and operations. The issue was entered into the licensees corrective action program for resolution.

The violation is more than minor because it adversely affects the program and process attribute of the Public Radiation Safety cornerstone, in that it involved an occurrence in the licensees radioactive effluent release program that is contrary to NRC regulations. The finding was determined to be of very low safety significance because preliminary calculations based on recently determined dilution factors for the settling pond demonstrated that resultant offsite dose values were small fractions of the ODCM limits (Appendix I to 10 CFR Part 50 design criteria). Further, evaluations of radionuclide concentrations in the effluent were conducted in accordance with 10 CFR 20.1302(b) (2)

(i) to demonstrate compliance with 10 CFR 20.1301 limits. (Section 2PS1)

Inspection Report# : 2006005 (pdf)

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : August 24, 2007

Crystal River 3 3Q/2007 Plant Inspection Findings Initiating Events Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Address Marine Fouling Resulted in a Plant Transient A self-revealing finding was identified for the failure to address the marine fouling failure mode in the scope of the existing preventive maintenance on the intake screen wash auto start system. As a result, reactor power had to be decreased to 80 percent to maintain condenser operating temperature limits. The licensee entered the issue into the corrective action program. Corrective actions included cleaning both the low and high side differential level sensing tubes, replacing tubes as needed, and implementing preventive maintenance procedures to periodically clean the tubes.

The finding was more than minor since it affected the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically the Operating Experience (OE) Program, in that, the licensee did not adequately implement OE through changes to station procedures to provide instructions to clean the sensing tubes during preventive maintenance on the system. (Section 4OA2.2)

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Replace a Non-refurbished ICS Module Resulted in a Reactor Trip A self-revealing finding was identified for failure to replace a non-refurbished integrated control system (ICS) multiplier module that had been temporarily installed during the Fall 2005 refueling outage. As a result, an age-related failure of a multiplier module resulted in an automatic reactor trip. The licensee entered the issue into the corrective action program. Corrective actions completed and/or proposed include: installation of a refurbished multiplier module; development of an engineering refueling outage turnover checklist to ensure formal followup actions are implemented whenever components not of desired quality are installed; and briefing of engineering personnel of this event.

The finding was more than minor because it affected the equipment reliability attribute of the Initiating Events Cornerstone and resulted in an automatic reactor trip that upset plant stability and challenged critical safety functions.

Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of the finding is related to the cross-cutting area of Human Performance, specifically Decision Making in that the licensee did not adequately communicate decisions and the basis for decisions to personnel who have a need to know the information. (Section 4OA3.1)

Inspection Report# : 2007002 (pdf)

Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance Procedures for Hydrostatic Seals Necessary to Protect Safety-Related Equipment from Internal Flooding The inspectors identified a non-cited violation (NCV) of Improved Technical Specification 5.6.1.1.a, for failure to adequately establish and implement procedures required by Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance (PM). Specifically, no procedure, program or process existed to periodically inspect hydrostatic barriers to identify and repair any degradation of the seals which provide protection of safety-related equipment from internal flooding. Corrective actions completed or planned include: Repair and qualify applicable fire seals as hydrostatic barriers and establish a hydrostatic penetration seal preventative maintenance program.

The finding is more than minor because it affected the protection against external factors (i.e. flood hazard) attribute of the Mitigating System cornerstone and could have impacted the availability of mitigating equipment during an internal flood event if left uncorrected. The inspectors determined that several degraded fire barrier seals did not meet hydrostatic barrier acceptability requirements. The finding was assessed through the SDP Phase 1 screening and determined to be of very low safety significance since the as-found condition of the hydrostatic barriers would not have resulted in the loss or degradation of safety-related mitigating equipment in the event of an internal flood.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Identify and Correct Repetitive Raw Water System Flush Water Strainer Baskets Degradation A self-revealing, non-cited Violation of 10 CFR 50, Appendix B, Criterion XVI was identified for failure to identify and take appropriate corrective actions for repetitive failures of the raw water pumps bearing flush water strainer baskets. As a result, both raw water pumps, RWP-2B and RWP-3B, were inoperable for a period greater than that allowed by Improved Technical Specifications when shell debris passed through a corroded strainer and clogged the cyclone separator discharge piping. The licensee entered the issue into the corrective action program. New strainer baskets made of a material compatible with service conditions were installed. Additional corrective actions include:

performing routine engineering review of degraded conditions found during preventative maintenance activities; revision to applicable surveillance procedures, and counseling of maintenance and engineering personnel on the need to identify and document adverse conditions in the corrective action program.

The finding was more than minor because it affected the equipment reliability attribute of the Mitigating System Cornerstone and resulted in a raw water train being inoperable for a period of time greater than allowed by Improved Technical Specifications. The finding was assessed through the Significance Determination Process (SDP) Phase 1 screening worksheet and determined to be of very low safety significance since the raw water pumps with a degraded flush water system had a very high likelihood of performing their safety function during a loss of offsite power event.

A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically, the licensee did not document the adverse condition of degraded strainer baskets in the corrective action program after it was determined that the filtering ability of the cyclone separator was a required design function.

(Section 4OA3.2)

Inspection Report# : 2007002 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct Adequate Surveys for Liquid Effluent Releases The inspectors identified a non-cited violation (NCV) of 10 CFR 20.1501(a) for failure to perform accurate surveys to demonstrate compliance with Technical Specification (TS) 5.6.2.3 Offsite Dose Calculation Manual (ODCM) controls used to maintain doses to members of the public from radioactive effluents as low as reasonably achievable (ALARA) in accordance with Appendix I to 10 CFR 50 design criteria as specified in 10 CFR 50.36a. Specifically, as of December 4, 2006, the licensee failed to conduct adequate dose evaluations to demonstrate compliance with TS 5.6.2.3 for radioactive liquid effluent releases made from the station discharge tank SDT-1 to a percolation pond located within the owner controlled area. The failure to conduct accurate dose evaluations for this liquid release pathway impaired the licensees ability to demonstrate compliance with ODCM ALARA limits for the liquid radioactive waste processing equipment and operations. The issue was entered into the licensees corrective action program for resolution.

The violation is more than minor because it adversely affects the program and process attribute of the Public Radiation Safety cornerstone, in that it involved an occurrence in the licensees radioactive effluent release program that is contrary to NRC regulations. The finding was determined to be of very low safety significance because preliminary calculations based on recently determined dilution factors for the settling pond demonstrated that resultant offsite dose values were small fractions of the ODCM limits (Appendix I to 10 CFR Part 50 design criteria). Further, evaluations of radionuclide concentrations in the effluent were conducted in accordance with 10 CFR 20.1302(b) (2)

(i) to demonstrate compliance with 10 CFR 20.1301 limits. (Section 2PS1)

Inspection Report# : 2006005 (pdf)

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : December 07, 2007

Crystal River 3 4Q/2007 Plant Inspection Findings Initiating Events Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Address Marine Fouling Resulted in a Plant Transient A self-revealing finding was identified for the failure to address the marine fouling failure mode in the scope of the existing preventive maintenance on the intake screen wash auto start system. As a result, reactor power had to be decreased to 80 percent to maintain condenser operating temperature limits. The licensee entered the issue into the corrective action program. Corrective actions included cleaning both the low and high side differential level sensing tubes, replacing tubes as needed, and implementing preventive maintenance procedures to periodically clean the tubes.

The finding was more than minor since it affected the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. A contributing cause of the finding is related to the cross-cutting area of Problem Identification and Resolution, specifically the Operating Experience (OE) Program, in that, the licensee did not adequately implement OE through changes to station procedures to provide instructions to clean the sensing tubes during preventive maintenance on the system. (Section 4OA2.2)

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Replace a Non-refurbished ICS Module Resulted in a Reactor Trip A self-revealing finding was identified for failure to replace a non-refurbished integrated control system (ICS) multiplier module that had been temporarily installed during the Fall 2005 refueling outage. As a result, an age-related failure of a multiplier module resulted in an automatic reactor trip. The licensee entered the issue into the corrective action program. Corrective actions completed and/or proposed include: installation of a refurbished multiplier module; development of an engineering refueling outage turnover checklist to ensure formal followup actions are implemented whenever components not of desired quality are installed; and briefing of engineering personnel of this event.

The finding was more than minor because it affected the equipment reliability attribute of the Initiating Events Cornerstone and resulted in an automatic reactor trip that upset plant stability and challenged critical safety functions.

Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was determined to be of very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of the finding is related to the cross-cutting area of Human Performance, specifically Decision Making in that the licensee did not adequately communicate decisions and the basis for decisions to personnel who have a need to know the information. (Section 4OA3.1)

Inspection Report# : 2007002 (pdf)