IR 05000341/2001017
| ML020160485 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/16/2002 |
| From: | Ring M Division Reactor Projects III |
| To: | O'Connor W Detroit Edison |
| References | |
| IR-01-017 | |
| Download: ML020160485 (38) | |
Text
ary 16, 2002
SUBJECT:
FERMI 2 NUCLEAR POWER STATION NRC INSPECTION REPORT 50-341/01-17(DRP)
Dear Mr. OConnor:
On December 29, 2001, the NRC completed an inspection at your Fermi 2 Nuclear Power Station.
The enclosed report documents inspection findings which were discussed on December 21, 2001, with Mr. Cobb, and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Specifically, this inspection focused on plant operations and radiation protection.
Based upon the results of this inspection, the inspectors identified two issues of very low safety significance (Green) which were determined to involve a violation of NRC requirements.
However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these issues as a Non-Cited Violation, in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you deny this Non-Cited Violation, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-001; and the NRC Resident Inspectors at the Fermi 2 Nuclear Power Station.
W. OConnor, Jr. -2-In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/NRC/ADAMS/index.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Mark A. Ring, Chief Branch 1 Division of Reactor Projects Docket No. 50-341 License No. NPF-43
Enclosure:
Inspection Report 50-341/01-17(DRP)
REGION III==
Docket No: 50-341 License No: DPR-43 Report No: 50-341/01-17(DRP)
Licensee: Detroit Edison Company Facility: Enrico Fermi, Unit 2 Location: 6400 N. Dixie Hwy.
Newport, MI 48166 Dates: November 17 through December 29, 2001 Inspectors: S. Campbell, Senior Resident Inspector J. Larizza, Resident Inspector R. Alexander, Radiation Specialist T. Kim, Project Manager Approved by: Mark A. Ring, Chief Branch 1 Division of Reactor Projects
SUMMARY OF FINDINGS IR 05000341-01-17(DRP), on 11/17-12/29/01, Detroit Edison Company, Fermi 2 Nuclear Power Station. Maintenance Risk Assessments and Emergent Work Evaluation.
The inspection was conducted by resident and specialist inspectors. The inspection identified two Green findings which were examples of a Non-Cited Violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply are indicated by No Color or by the severity level of the application violation. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at http://www.nrc.gov/NRR/OVERSIGHT/index.html.
Cornerstone: Initiating Events
- Green. The inspectors identified an example of a Non-Cited Violation of Technical Specification 5.4.1.a for using the incorrect procedure for restoring the north reactor feedwater pump following emergent work activities that involved inappropriate opening of the north reactor feedwater pump discharge valve. Control room operators used a system operating procedure that required plant conditions of 950 pounds per square inch reactor pressure and both north and south reactor feedwater pump turbines operating. However, actual conditions were about 650 pounds per square inch reactor pressure and only the south reactor feedwater pump turbine was operating.
The finding had an actual impact of: 1) discharging about 1.8 million pounds mass per hour of cold water moderator to the reactor vessel, 2) an unexpected power excursion from about 4 to 11 percent, causing a one-half scram signal from intermediate range monitor E, 3) an unexpected reactor water level increase to 225 inches, which was above the Level 8 trip setpoint, and 4) sending isolation signals to the high pressure coolant injection pump, reactor core isolation coolant pump and the only operating south reactor feedwater pump (stopping water to the reactor vessel). The finding was of very low safety significance because the event occurred during reactor startup and at low reactor power level and the power level excursion was not significant. Because the finding was of very low safety significance and the finding was captured in the licensees corrective action program, this finding is being treated as an example of a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (Section 1R13).
Cornerstone: Mitigating Systems
- Green. The inspectors identified an example of a Non-Cited Violation of Technical Specification 5.4.1.a for not completing the valve lineup while venting and draining the Division 2 residual heat removal system after completing heat exchanger relief valve testing. The operator failed to complete the instructions for venting and draining the Division 2 residual heat removal system before the system was refilled and caused an inadvertent discharge of approximately 400 gallons of contaminated water into the reactor building.
The finding was more than minor for the following reasons: 1) high contamination levels in the reactor building resulted from the spill, 2) the potential loss of residual heat removal cooling water from the system, and 3) the potential challenge to electrical equipment wetted from the spill. The finding was of very low safety significance because neither personnel contamination nor personnel overexposure occurred, electrical equipment was not damaged, and the residual heat removal system was not required at the time of the event. Because the finding was of very low safety significance and the finding was captured in the licensees corrective action program, this finding is being treated as an example of a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (Section 1R13).
Report Details 1. REACTOR SAFETY Cornerstone: Mitigating Systems Plant Status At the beginning of the inspection period, the plant had been shutdown to conduct the eighth refueling outage. On November 27, 2001, the refueling outage was completed and operators commenced raising reactor power. On November 28, 2001, power ascension stopped at 4 percent, when an unexpected power increase to 11 percent followed by an expected level increase to the Level 8 trip setpoint occurred while placing the north reactor feedwater pump in service. After resolution of the problems, restart of the unit occurred on November 29, 2001, and operators raised reactor power and synchronized the unit to the grid on November 30, 2001. The plant reached 100 percent on December 2, 2001. Power remained at 100 percent until December 6, 2001, when an operator broke a vent line on a stator cooling water heat exchanger causing the control room operators to scram the plant manually upon losing stator cooling water pressure. Following repairs, operators restarted the unit on December 6, 2001, and synchronized the main turbine to the electrical grid on December 8, 2001. Reactor power remained at 100 percent until operators decreased power to 65 percent to conduct a control rod shuffle on December 15, 2001. After completing the shuffle, reactor power was raised to 100 percent on December 16, 2001. Reactor power remained at 100 percent through the remainder of the inspection period.
1R01 Adverse Weather (71111.01)
a. Inspection Scope On December 4 and December 20, 2001, the inspectors used procedure 27.000.04, Freeze Protection Lineup Verification, to a conduct a partial walkdown of the residual heat removal (RHR) service water complex and reactor/auxiliary building to verify freeze protection readiness.
b. Findings No findings of significance were identified.
1R04 Equipment Alignments (71111.04S)
a. Inspection Scope The inspectors conducted a complete system alignment verification of the condensate and core spray systems. The verification included a review of documents to determine the correct system lineup, including abnormal and emergency operating procedures, drawings, the Updated Final Safety Analysis Report, and the vendors manuals. Also, the inspectors reviewed outstanding maintenance work requests on the system and any
deficiencies that affect the ability of the system to perform its function. Outstanding design issues including temporary modifications, operator workarounds, and items tracked by engineering department personnel were reviewed. The walkdown identified any discrepancies between the existing system equipment lineup and correct lineup.
b. Findings No findings of significance were identified.
1R05 Fire Protection (71111.05Q)
a. Inspection Scope The inspectors toured the following areas to determine whether combustible hazards were present, fire extinguishers were properly filled and tested, the CARDOX units were operable, hose stations were properly maintained, and if the fire hazard analysis drawings were correct:
+ Third Floor Reactor Building (Zone 7)
+ Fifth Floor Reactor Building, Refuel Floor (Zone 9)
+ Third Floor Auxiliary Building, Control Room (Zone 9)
+ Division 2 RHR Building (Zone 2)
+ First Floor Reactor Building (Zone 5)
+ Reactor Building South East Quadrant (Zone 2)
b. Findings Following a review of maintenance activities for fire detection equipment (fire indicating lights, fire detection bells, ionization detectors, carbon dioxide shutoff dampers, and smoke detectors), the inspectors noted that about 168 components were not tested per fire detection zone operability procedures. The licensee initiated Condition Assessment Resolution Document (CARD) 01-20330 in response to the inspectors concerns. This item will be an unresolved item, (URI 50-341/01-17-01) pending the inspectors review of the licensees evaluation of the CARD and a review of the criteria for testing fire detection equipment in various zones.
1R05 Fire Protection (71111.05A)
a. Inspection Scope On December 18, 2001, the inspectors observed the licensees fire brigade respond to an unannounced simulated fire on the first floor of the radwaste building in the chemical storage area. The inspectors observed proper use of protective clothing and self-contained breathing apparatus, the availability of sufficient fire fighting equipment, effective radio communications and effective fire brigade leader directions. The inspectors noted that the pre-planned drill scenario was followed and the drill objectives were met, and observed a drill critique at the termination of the drill.
b. Findings No findings of significance were identified.
1R07 Heat Sink Performance (71111.07)
a. Inspection Scope The inspectors reviewed surveillance procedure 47.205.01, RHR Division 1 (North)
Heat Exchanger Performance Test, and reviewed data collected during the test. The inspectors reviewed 1996, 1998 and 2000 data for the previous Division 1 heat exchanger tests and examined the performance trending.
b. Findings No findings of significance were identified.
1R12 Maintenance Rule Implementation (71111.12Q)
a. Inspection Scope The inspectors reviewed the system health reports, associated CARDs, white papers for probabilistic risk assessment on conditional probabilities, and the control room unit logs for the following systems to determine whether the maintenance rule program had been implemented appropriately by assessing the characterization of failed structures, systems, and components. The inspectors also determined whether goal setting and performance monitoring were adequate for the following systems:
+ Condensate Storage Tank (P1100)
+ Mechanical Draft Cooling Towers (E1156)
+ Turbine Building Closed Cooling Water System (P4300)
+ Safety Relief Valves (B2104)
+ RHR Service Water (E1151)
+ Condensate System (N2000)
b. Findings No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)
1. Restoration of Division 2 RHR System a. Inspection Scope On November 24, 2001, operators found Division 2 RHR keep fill isolation valve E1100F208 mispositioned open while restoring the system. The inspectors reviewed work packages, safety tagging records and Level 3 CARD 01-19330 and interviewed operators who were involved in the event.
b. Findings The inspectors identified one Green finding involving an example of a Non-Cited Violation of Technical Specification 5.4.1.a for inadequate implementation of the RHR system operating procedure, which resulted in the inadvertent discharge of about 400 gallons of water from the Division 2 RHR keep fill system into the reactor building.
On November 24, 2001, the licensee completed an emergent work item per work request 000Z991909 to test the RHR Division 2 heat exchanger B outlet line relief valve E1100F025B. During system restoration, about 400 gallons of water were discharged through the Division 2 RHR supply to thermal recombiner water spray cooler T4804001B vent valve E1100F255 into the reactor building. The operators were using system operating procedure 23.205, RHR System, to fill and vent the RHR system with water from the condensate storage tank. A non-licensed operator noticed the leak into the reactor building and notified the control room. A control room operator diverted the water flow from the open valve to the torus by opening Division 2 RHR test line valve E1150F028B and Division 2 RHR torus cooling line isolation valve E1150F027B. Radiological surveys indicated contamination levels above the 100,000 dpm/100 cm2 high contamination limits at 110,000 dpm/100 cm2. Water had migrated from the third floor to the basement and dripped onto electrical cable trays and equipment.
The licensees investigation determined the cause to be inadequate implementation of system operating procedure 23.205 to complete the valve lineup after system draining.
The operator, who performed Section 7.6, Draining Division 1(2) RHR to Torus, completed only the portion of the procedure to drain the system. The operator did not perform the section that instructed closing of valve E1100F255 after the system was drained and vented. An incorrect assumption was made that the associated safety tagging record (2001-006932) would restore all valves to the proper lineup after the work was completed. The safety tagging record did not list a restoration position for valve E1100F0255. Attachment 1B, Division 2 RHR Initial Valve Lineup, of procedure 23.205 required that valves E1100F208 and E1100F255 be closed.
Drawing 6M721-5706-1, RHR Division 2 Functional Operating Sketch, listed valve E1100F255 as closed.
The performance deficiency associated with this event was inadequate implementation of the valve lineup section in the RHR system operating procedure that led to the unexpected discharge of contaminated water from the RHR system onto the reactor building floor. The finding was more than minor for the following reasons: 1) high contamination levels in the reactor building resulted from the spill, 2) the potential loss of RHR cooling water from the system, and 3) the potential challenge to electrical equipment wetted from the spill. The event was of very low safety significance because neither personnel contamination nor personnel overexposure occurred, electrical equipment was not damaged, and the RHR system was not required at the time of the event.
Technical Specification 5.4.1.a requires written procedures be established, implemented, and maintained covering the activities specified in Regulatory Guide 1.33, Appendix A. Regulatory Guide 1.33, Appendix A, Item 4h requires procedures for
emergency core cooling water systems. On November 24, 2001, operations personnel failed to fully complete the draining and venting steps listed in Section 7.6 of procedure 23.205. Consequently, Division 2 RHR supply to thermal recombiner water spray cooler T4804001B vent valve E1100F255 was left open while filling the Division 2 RHR system, an emergency core cooling water system. Failure to fully implement the procedure is an apparent violation. However, because of the very low safety significance and because the issue is in the licensees corrective action program, it is being treated as an example of a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-341/01-17-02). This Non-Cited Violation is addressed in CARD 01-19330.
.2 Restoration of 5 North Feedwater Heater a. Inspection Scope On November 28, 2001, operators found that main turbine extraction steam feedwater heater No. 5N drain valve N3016F355 had been mispositioned opened during plant startup. The inspectors reviewed work packages, safety tagging records and Level 2 CARD 01-19702 and interviewed operators who were involved in the event.
b. Findings No findings of significance were identified.
.3 Restoration of North Reactor Feedwater Pump (N2102D010)
a. Inspection Scope On November 28, 2001, water was unexpectedly sent to the reactor when operators restored the north reactor feedwater pump (NRFP) to service after repairing a leaking suction strainer. The inspectors reviewed logs and CARDs, and interviewed operators in response to a feedwater transient event that occurred while restoring the NRFP following the repair of a leaking suction strainer.
b. Findings The inspectors identified one Green finding involving a second example of a Non-Cited Violation of Technical Specification 5.4.1.a for implementing an inadequate procedure while placing the NRFP in service. Operators used the incorrect procedure for conducting this activity and opened the NRFP discharge valve. The event caused water level to rise to the Level 8 trip setpoint, causing isolation signals to be sent to the high pressure coolant injection (HPCI) system, reactor core isolation cooling (RCIC) system, and south reactor feedwater pump turbines. The south reactor feedwater pump (SRFP)
tripped and a half scram was generated from intermediate range monitor E. The event occurred as follows:
The licensee had developed a safety tagging record to isolate the NRFP to work on the associated strainer. The record listed the following positions for the valves:
+ NRFP Suction Valve N2000F634 - closed
+ NRFP Discharge Line Isolation Valve N2100F607 - closed
+ NRFP Discharge Hydraulic Stop Valve N2100F045A - was not listed closed or open but remained open
+ NRFP to Reactor Pressure Vessel Startup Level Control Isolation Valve N2100F611 - closed Operators performed this lineup and mechanics tightened nuts on a strainer to stop the leak. At the same time, the SRFP was in operation, providing flow to the reactor through the startup level control valve and associated piping. Reactor power was about four percent, reactor pressure was approximately 650 psig and reactor water level was about 197 inches. Four of the intermediate range monitors (IRMs) were on Range 9 (scale 0-40) and three were on Range 10 (scale 0-125). The trip setpoint for the 0-40 percent range IRMs is 38 percent and the trip setpoint for the 0-125 percent range IRMs is 120 percent.
Operators selected procedure 23.107, Reactor Feed Pump Operation, Section 7.2, Restoring an Isolated Reactor Feed Pump Turbine with Condenser Vacuum Established, to restore the NRFP. However, prerequisites for this procedure were reactor pressure at 940 psig and both the north and south reactor feed pump turbines started. The operators failed to recognize that the plant was not in this condition while attempting to restore the system. The plant was at 650 psig with only the SRFP running and on startup level control. No procedure existed to restore the NRFP under the existing plant condition.
The operators decided to begin placing the NRFP in service and performed the valve lineup in procedure 23.107, Section 7.2. Operators opened valve N2100F611, which diverted some flow from the line containing the startup level control valve N2100F403 to the discharge line of the NRFP. When the operators continued performing the lineup in procedure 23.107, they failed to recognize the impact of opening NRFP discharge line isolation valve N2100F607. Once they opened the valve, operators saw that feed flow had increased and that the reactor water level had increased and closed the valve immediately. About 1.8 million pounds mass per hour of cold water went into the reactor.
Because the cold water (moderator) entered the reactor, reactor power increased from 4 to 11 percent. Consequently, because only IRM E exceeded the trip setpoint, a half scram occurred. Reactor water level exceeded the Level 8 trip setpoint at 215 inches and sent isolation signals to steam isolation valves for the RCIC and HPCI systems, the main turbine and the reactor feedwater pumps. The main turbine was not on line at the time. The SRFP tripped as designed. Also, isolation valves for the RCIC and HPCI turbines were already closed. Maximum water level reached about 225 inches, which was well below the main steam line nozzles to the main turbine. Level remained at 225 inches for greater than 5 minutes. Reactor power dropped quickly to the initial level of four percent. Operators subsequently restarted the SRFP to maintain water flow to
the reactor. The plant remained in a steady state condition following the event.
Operators initiated CARD 01-22208.
The performance deficiency associated with this event is the use of an inadequate procedure for the existing plant conditions while restoring the NRFP after repairs, which led to the unexpected transient. This finding was greater than minor because it had an actual impact of an unexpected small power excursion and an unexpected level increase that isolated safety equipment and stopped water flow to the vessel during startup. The event was of very low safety significance because the transient occurred during reactor startup at low power level and the power level excursion was not significant.
Technical Specification 5.4.1.a requires that written procedures be established, implemented and maintained covering the activities specified in Regulatory Guide 1.33, Appendix A. Appendix A of Regulatory Guide 1.33, Item 4o requires procedures for operating the reactor feedwater system. Contrary to Technical Specification 5.4.1.a and Regulatory Guide 1.33, Procedure 23.107, Reactor Feed Pump Operation, Section 7.2, Restoring an Isolated Reactor Feed Pump Turbine with Condenser Vacuum Established, provided inadequate instructions for placing the NRFP in service with the plant at 650 psig and four percent power and only the SRFP in service. The procedure required that the plant be at 950 psig with both north and south feedwater pump turbines operating. This is an apparent violation. However, because of the very low safety significance and because the issue is in the licensees corrective action program, it is being treated as another example of Non-Cited Violation NCV 50-341/01-17-02, consistent with Section VI.A.1 of the NRC enforcement policy.
1R14 Nonroutine Plant Evolutions (71111.14)
1. Level 8 Trip During Startup a. Inspection Scope On November 28, 2001, the inspectors observed how control room personnel responded to an unexpected increase in reactor power (4 to 11 percent), an unexpected increase in level to the Level 8 trip setpoint, and a half scram initiated by an IRM during reactor startup. The inspectors interviewed operators involved in the event, reviewed abnormal operating procedures, standard operating procedures, drawings, plant parameter strip chart recorder traces, and General Electric Transient Analysis Report System data.
b. Findings No findings of significance were identified. However, the specifics of the emergent work issues that caused the event were discussed in Section 1R13 of this report.
2. Loss of Stator Cooling Water Pressure Causes Manual Reactor Scram a. Inspection Scope On December 6, 2001, the inspectors observed how control room personnel responded to an unexpected decrease in stator cooling water pressure that resulted in a manual scram of the reactor from 100 percent power. The inspectors interviewed operators involved in the event, reviewed abnormal operating procedures, standard operating procedures, drawings, plant parameter strip chart recorder traces, and General Electric Transient Analysis Report System data.
b. Findings No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope On November 28, 2001, while performing surveillance tests of the HPCI system at 165 psig reactor steam pressure, a fluid transient occurred. The resulting flow peak of 1546 gallons per minute was less than the magnitude of previous non-damaging transients. A walkdown of the HPCI system was performed by the licensee. No damage to the piping or pipe supports was noted. Based on these observations and the bounded analysis there was no operability concern. A future modification to add a keep-fill system will prevent this condition from occurring.
b. Findings No findings of significance were identified.
1R16 Operator Work-Arounds (71111.16)
a. Inspection Scope The inspectors reviewed the aggregate assessment of operator work-arounds for the third quarter of 2001. The inspectors reviewed the workaround impacts on reliability, availability, and potential for misoperation of any systems listed in the aggregate assessment of operator work-arounds. The review included the cumulative effects of operator work-arounds that could increase an initiating event frequency or that could affect multiple mitigating systems and the ability of operators to respond in a correct and timely manner to plant transients and accidents.
b. Findings No findings of significance were identified.
1R17 Permanent Plant Modifications (71111.17)
a. Inspection Scope Engineering design package 29068A, for replacing the emergency diesel generator 12 exciter, was reviewed and selected aspects were discussed with engineering personnel.
This document was reviewed for adequacy of the safety evaluation and consideration of design parameters. The modifications were for equipment upgrades of existing equipment.
b. Findings No findings of significance were identified.
1R19 Post Maintenance Testing (71111.19)
a. Inspection Scope The inspectors reviewed the test data for the components for piping supports dedicated as American Society of Mechanical Engineering snubbers to ensure compliance with the code and Technical Specifications. The inspectors verified that the testing demonstrated that the snubbers were capable of performing their intended function.
b. Findings No findings of significance were identified.
1R20 Refueling and Outage (71111.20)
a. Inspection Scope The inspectors directly observed and verified whether operators appropriately followed standard operating procedures, implemented Technical Specifications, and conducted briefings correctly during the following activities:
+ Reactor Criticality during Reactor Startup after Refueling Outage 8
+ Main Turbine Generator Synchronization after Refueling Outage 8
+ Reactor Criticality Point of Adding Heat during Reactor Startup after Forced Outage 01-01
+ Main Turbine Generator Synchronization after Forced Outage 01-01
+ Infrequently Performed Test or Evolution for Startup after Refueling Outage 8
+ Drywell Inspection after Refueling Outage 8 b. Findings No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope The inspectors witnessed and reviewed test data for the emergency diesel generator 14 loss of power, loss of coolant accident surveillance test. The inspectors reviewed the Technical Specifications to confirm that the surveillance activities had verified the equipment would perform its intended functions. The inspectors observed staffing levels of the control room and relay room, and in the field.
b. Findings No findings of significance were identified.
Cornerstone: Emergency Preparedness 1EP6 Drill Evaluation (71114.06)
a. Inspection Scope The inspectors reviewed the following drill critiques to evaluate the adequacy of the licensees critique of performance in identifying weaknesses and deficiencies. The inspectors verified that the weaknesses were placed in the corrective action system and that all corrective actions for identified weaknesses were resolved for closed CARDs:
+ Scenario 30.2, Radiological Emergency Response Preparedness Team Blue, Control Room Shift 5, May 1, 2001
+ Scenario 31, Radiological Emergency Response Preparedness Team Red, Control Room Shift 1, July 17, 2001 b. Findings There were no findings of significance identified.
2. RADIATION SAFETY Cornerstone: Occupational Radiation Safety (OS)
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 High Risk Significant, High Dose Rate Locked High Radiation Areas and Very High Radiation Areas a. Inspection Scope The inspectors reviewed the stations implementation of physical and administrative controls over access to High Radiation Areas, High Dose Rate Locked High Radiation Areas, and Very High Radiation Areas, including a discussion of these controls with the
Radiation Protection (RP) Manager and first line RP supervisors, to verify that revisions to procedures implementing these controls did not reduce the effectiveness and level of worker protection. Additionally, the inspectors selectively walked down the boundaries of Locked High Radiation Areas and Very High Radiation Areas reestablished since the completion of the stations recent refueling outage to verify adequate controls were in place.
b. Findings No findings of significance were identified.
.2 Identification and Resolution of Problems a. Inspection Scope The inspectors reviewed several CARDs completed during the final month of the stations recent refueling outage related to radiation worker performance and RP technician proficiency. The inspectors reviewed these documents to assess the licensees ability to identify repetitive problems, contributing causes, the extent of conditions, and corrective actions intended to achieve lasting results. Additionally, though the inspectors reviewed the licensees High Radiation Area controls as discussed in Section 2OS1.1, the licensee did not identify any additional High Radiation Area access control issues during the inspection cycle.
b. Findings No findings of significance were identified.
2OS2 As-Low-As-Reasonably-Achievable (ALARA) Planning and Controls (71121.02)
.1 Post-Outage ALARA Reviews a. Inspection Scope Due to the close proximity in time between the completion of the licensees refueling outage and the inspection, the inspectors were only able to review two audits and self-assessments that focused on overall ALARA performance during the outage rather than assessing individual job ALARA performance. However, the inspectors reviewed the audit and self-assessment to assess the licensees ability to identify repetitive problems, contributing causes, the extent of conditions, and corrective actions intended to achieve lasting results.
b. Findings No findings of significance were identified.
Cornerstone: Public Radiation Safety (PS)
2PS2 Radioactive Material Processing and Transportation (71122.02)
.1 Review and Walkdowns of Radioactive Waste Systems a. Inspection Scope The inspectors reviewed the liquid and solid radioactive waste system description in the Updated Final Safety Analysis Report and the most recent radiological effluent release report (for calendar year 2000) for information on the types and amounts of radioactive waste (radwaste) generated for disposal.
The inspectors performed walkdowns of the liquid and solid radwaste processing systems located in the Radwaste and Onsite Storage Facilities to verify that the systems were as described in the Updated Final Safety Analysis Report and the Process Control Program, and to assess the material condition and operability of the systems. The inspectors also discussed the current operation of the system with members of the radioactive waste operations crew. In the case of abandoned radwaste equipment (i.e., asphalt extruder solidification system), the inspectors reviewed the licensees administrative and physical controls implemented to isolate these systems to verify the equipment would not contribute to an unmonitored radioactive material release path and would not inadvertently affect operating systems.
b. Findings No findings of significance were identified.
.2 Waste Characterization and Classification a. Inspection Scope The inspectors reviewed the licensees method and procedures for determining the classification of radioactive waste shipments, including the licensees use of scaling factors to quantify difficult-to-measure radionuclides (e.g., pure alpha or beta emitting radionuclides). Specifically, the inspectors reviewed the licensees spring 2001 radio-chemical analysis results for the condensate resin, bead resin/charcoal, dry active waste, fuel pool cooling cleanup, and reactor water cleanup waste streams. The inspectors reviewed the report to verify that the licensees scaling factors were accurately determined such that waste shipments were classified in accordance with the requirements contained in 10 CFR Part 61 and the licensees Process Control Program.
The inspectors also reviewed the procedure for transferring waste materials into shipping containers to determine if appropriate waste stream mixing and/or sampling procedures were utilized for the purposes of waste classification per 10 CFR 61.55.
The inspectors additionally reviewed the licensees processes employed to ensure that changes in operating parameters, which may result in changes to the waste stream composition, are identified between the annual or biennial scaling factor updates.
b. Findings No findings of significance were identified.
.3 Shipment Preparation a. Inspection Scope The inspectors were unable to directly observe shipments of radioactive material as the licensee was not conducting any radioactive material shipments during the inspection.
Therefore, to ensure that the shipping activities were performed in accordance with the requirements of 49 CFR Parts 172 and 173, the inspectors examined the shipping packages described in Section 2PS2.4. For these shipments, the inspectors reviewed the final radiological surveys, labeling, placarding, vehicle inspections, and instructions to the driver. Additionally, the inspectors examined the training program provided to personnel responsible for the conduct of radioactive waste processing and radioactive material shipment preparation activities to assess the licensees compliance with 49 CFR Part 172, Subpart H requirements. Specifically, the inspectors reviewed the lesson plans, student handouts, and course completion documentation for licensee and vendor-provided courses to ensure that personnel (shippers, RP technicians, and fuel handlers) had adequately completed both the awareness/safety training and function specific training applicable for their individual job functions.
b. Findings No findings of significance were identified.
.4 Shipping Records a. Inspection Scope The inspectors reviewed a selection of non-excepted package shipments completed during calendar years 1999 - 2001 to verify compliance with NRC and Department of Transportation requirements (i.e., 10 CFR Parts 20 and 71; 49 CFR Parts 172 and 173).
Specifically, the inspectors reviewed the following radioactive materials/waste shipment records:
+ 99-001 Irradiated Reactor Hardware Liner (Type B, January 11, 1999)
+ 99-041 Cs-137 Calibration Source (Type A, August 19, 1999)
+ 00-013 Irradiated Hardware Liner (Type B, May 10, 2000)
+ 00-040 High Pressure Turbine Rotor (Surface Contaminated Object II, April 18, 2000)
+ 00-089 Dewatered Powdered, Charcoal, and Bead Resin (Low Specific Activity
[LSA] II, September 27, 2000)
+ 01-022 Powdered and Bead Resin (Unprocessed) Liner (LSA-II, May 8, 2001)
+ 01-030 13 High Rad Drums (Compacted Dry Active Waste) (LSA-II, June 12, 2001)
+ 01-077 Laundry (LSA-II, November 21, 2001)
b. Findings No findings of significance were identified.
.5 Identification and Resolution of Problems a. Inspection Scope The inspectors reviewed self-assessments and CARDs completed during the previous 18 months which concerned the areas of radioactive waste processing and radioactive waste/material shipping. The inspectors reviewed these documents to assess the licensees ability to identify repetitive problems, contributing causes, the extent of conditions, and corrective actions intended to achieve lasting results.
b. Findings No findings of significance were identified.
4. OTHER ACTIVITIES (OA)
4OA1 Performance Indicator Verification (71151)
Mitigating System and Initiating Events Performance Indicator Verification a. Inspection Scope The inspectors reviewed licensee event reports, licensee memoranda, unit logs, and NRC inspection reports to verify the following performance indicators for second quarter 2001 through third quarter of 2001.
+ Unplanned Scrams per 7000 Critical Hours
+ Reactor System Activity
+ Scrams with Loss of Normal Heat Removal
+ Unplanned Power Changes per 7000 Critical Hours
+ Safety System Unavailability, High Pressure Injection System
+ Safety System Unavailability, RCIC
+ Safety System Unavailability, RHR System
+ Safety System Functional Failures
+ Safety System Unavailability, Emergency AC Power b. Findings There were no findings of significance identified.
4OA5 Other The inspectors reviewed the interim report for the May 2001 Plant Evaluation performed by an inspection team from the World Association of Nuclear Operators. No further inspection was deemed necessary by NRC inspectors, and no assessment was made of the results of the inspection.
.4OA6 Management Meetings
.1 Exit Meeting Summary The inspectors presented the inspection results to M and other members of licensee management at the conclusion of the inspection on December 21, 2001. The licensee acknowledged the findings presented. No proprietary information was identified.
Specific Area Exits Radiation Protection Senior Official at Exit: D. Cobb, Plant Manager Date: December 7, 2001 Proprietary (explain yes): No Subject: Occupational Radiation Safety (Access Control and ALARA); Public Radiation Safety (Radwaste and Transportation)
Change to Inspection Findings: No
KEY POINTS OF CONTACT Licensee H. Arora, Nuclear Licensing M. Brown, Engineer, Nuclear Licensing J. Carter, Supervisor, Radwaste D. Cobb, Plant Manager D. Craine, Supervisor, Radiological Engineering J. Davis, Manager, Outage T. Dong, Manager, In-Service Inspection Q. Duong, Manager, Plant Support Engineering S. Hassoun, Principle Engineer, Nuclear Licensing R. Johnson, Supervisor, Nuclear Licensing E. Kokosky, Manager, Radiation Protection M. Kramer, Shift Manager, Operations A. Mann, Manager, Operations J. Moyers, Manager, Nuclear Assessment D. Noetzel, Manager, System Engineering W. OConnor, Vice President, Nuclear Generation N. Peterson, Manager, Nuclear Licensing M. Philippon, Shift Technical Advisor, Operations J. Priest, Nuclear Quality Assurance S. Stasek, Director, Nuclear Assessment J. Tibai, Manager, Maintenance Rule B. Weber, Supervisor, Radwaste J. Werner, Manager, Training D. Williams, Assistant Manager, Radiation Protection
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened URI 50-341/01017-01 Fire Detection Equipment May not be Tested per Operability Test Procedures NCV 50-341/01017-02 Inadequate Use of Procedures During System Restoration Closed NCV 50-341/01017-02 Inadequate Use of Procedures During System Restoration Discussed None LIST OF ACRONYMS USED AC Alternating Current ALARA As-Low-As-Reasonably-Achievable CARD Condition Assessment Resolution Document CFR Code of Federal Regulations HPCI High Pressure Coolant Injection IRM Intermediate Range Monitor LSA Low Specific Activity NRFP North Reactor Feedwater Pump SRFP South Reactor Feedwater Pump RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RP Radiation Protection
LIST OF DOCUMENTS REVIEWED The following documents were selected and reviewed by the inspectors to accomplish the objectives and scope of the inspection and to support any findings.
1R01 Adverse Weather Procedure Freeze protection lineup verification Revision 21 27.000.04 Procedure Shiftly, Daily and Weekly required surveillances Revision 100 24.000.02 1R04 Equipment Alignment Dwg. Condensate System Functional Operating Revision AB 6M721-5714-1 Sketch UFSAR Condensate and Feedwater System Revision 8 Section 10.4.7 Procedure 23.107 Reactor Feedwater and Condensate Systems Revision 92 Procedure 23.104 Condensate Storage and Transfer System Revision 60 Procedure 23.107 Condensate Filter Demin system Revision 57 Alarm Response Condensate System Low Flow Revision 7 Procedure 5D108 Alarm Response North Hotwell Level Hi/Low Revision 9 Procedure 5D130 Procedure HPCI System Offline Auto Initiation Time Revision 38 24.202.04 Response Test Procedure Standby Feedwater System Revision 29 23.107.01 Emergency Reactor Pressure Vessel Control, Sheet 1 Revision 9 Operating Procedure Dwg. 6M721-5707 Core Spray Functional Operating Sketch Revision Z UFSAR Section Core Spray System Revision 9 6.2.3.3.2 Procedure Division 2 CSS Pump and Valve Operability and Revision 40 24.203.03 Automatic Actuation
Procedure Core Spray Pump and Valve Operability and Revision 27 24.203.04 Position Verification Test Procedure 24.203 Core Spray System Revision 33 Procedure Div 2 CSS Leakage Monitoring Test Revision 25 43.203.005 Procedure ECCS - Core Spray System Division 2 Logic Revision 34 44.030.002 Functional Test Procedure Plant Startup Master Checklist Revision 48 22.000.01 Annunciator Div 2 CSS Actuated Revision 6 Response Procedure 2D3 Annunciator Div I/II Fill Line Pressure Low Revision 8 Response Procedure 2D90 1R05 Fire Protection UFSAR Section Fire hazard Analysis: Reactor Building, Third Revision 10 9A.4.1.8.1 Floor, Zone 7, El. 641 Ft 6 In.
Drwg 6A721-2400 Fire Protection Evaluation Plan Plot Revision M Drwg 6A721-2407 Fire Protection Evaluation Reactor and Auxiliary Revision Q Buildings Third Floor Plan El-641"-6" and 643"-6" UFSAR Section Fire Hazard Analysis: Reactor Building, Fifth Revision 8 9A.4.1.10 Floor, Zone 9, El. 684 Ft 6 In Drwg 6A721-2409 Fire Protection Evaluation Reactor and Auxiliary Revision R Buildings Fifth Floor Plan El-677"-6" and 684"-6" Drwg 6A721N- Fire Protection Evaluation Residual Heat Revision C 2042 Removal Complex Upper Floor Plan El-617'-0" Drwg 6A721N- Fire Protection Evaluation Residual Heat Revision E 2041 Removal Grade Floor Plan El-590'-0" Drwg 6A721-2401 Fire Protection Evaluation Reactor Building Revision K Subbasement Plan El-540'-0" UFSAR Section Fire Hazard Analysis: Control Room, Zone 9, El. Revision 10 9A.4.2.10 643 Ft 6 In, 655 Ft 6 In and 677 Ft 6 In Procedure Plant Fires Revision 31 20.000.22
UFSAR Section Radwaste Building general description Revision 11 9A.4.4.1 Fire Brigade Drill 1st Floor RAD Waste Chemical Lab storage Scenario No. 6 Area - El. 583'6" 1R07 Heat Sink Performance 47.205.02 Residual Heat Removal Division 1 (South) Heat Revision 6 exchanger Performance Test Job TG25010930 Perform 47.205.002, RHR Division 2 HX November 11, 2001 Performance Test UFSAR Section Residual Heat Removal Revision 5 5.5.7 UFSAR Section Recirculation Line Break Long Term Response Revision 6 6.2.1.3.3 CARD 01-13239 Log Mean Temperature Differential (LMTD) July 16, 2001 Correction Factor Used in RHR Heat Exchanger Test Analysis CARD 01-13240 RHR Heat Exchanger Test Acceptance Criteria August 2, 2001 CARD 01-13241 RHR Heat Exchanger Design Fouling Less than August 2, 2001 Allowed in Heat Exchanger Performance Test CARD 01-14727 NRC Concern - RHR Heat exchanger Monitoring May 4, 2001 1R12 Maintenance Rule Implementation NUMARC 93-01 Nuclear Energy Institute Industry Guideline for Revision 2 Monitoring Effectiveness at Nuclear Power Plants April 1996 Maintenance Rule Desk Top Reference July 2, 2001 PRA Ranking Probabilistic Importance Measure Table 4.1 Log 98-002 Maintenance Rule position Paper: Bases Revision O, Summary for Maintenance Rule Performance October 2, 1998 Criteria, Table 1 Log 96-01 Maintenance Rule Position Paper: Development Revision 1, of Conditional Probability for SSCs Modeled in October 2, 1998 the Fermi 2 PSA
Log 96-002 Maintenance Rule Position Paper: Development Revision 1, of Train and Divisional Level Conditional October 2, 1998 Probability, Allowed Number of Failures and Out-of-Service Hours, and Redundancy Factor MR06, Section Establishing Performance Criteria Revision 6 5.2.1 MR06, Appendix Performance Criteria Summary Revision 8 H
Control Room Logs for Condensate Storage December 31, 1998 Tank (P1100), Mechanical Draft Cooling Towers - December 5, 2001 (E1156), and the Turbine Building Closed Cooling Water System (P4300)
Condition Assessment Resolution Documents for December 31, 1998 the Condensate Storage System (P1100), - December 5, 2001 Mechanical Draft Cooling Towers (E1156), and the Turbine Building Closed Cooling Water System (P4300)
Work Requests and Preventive Maintenance December 31, 1998 Task for the Condensate Storage System - December 5, 2001 (P1100), Mechanical Draft Cooling Towers (E1156), and the Turbine Building Closed Cooling Water System (P4300)
Control Room Logs for the Safety Relief Valves December 31, 1998 (B2104), Residual Heat Removal Service Water - December 19, (E1151), and Condensate System (N2000) 2001.
Condition Assessment Resolution Documents for December 31, 1998 the Safety Relief Valves (B2104), Residual Heat - December 19, Removal Service Water (E1151), and 2001.
Condensate System (N2000)
Work Requests and Preventive Maintenance December 31, 1998 Task for the Safety Relief Valves (B2104), - December 19, Residual Heat Removal Service Water (E1151), 2001.
and Condensate System (N2000)
Critical Performance Evaluation Data for December 31, 1998 Maintenance Rule Functional Failures - December 5, 2001 000Z004113 Minor Maintenance Form: Shaft Seal Leaking November 28, 2000 for Emergency Hotwell Pump STR 00-4163 Safety Tagging Record to Replace Shaft Seal for November 28, 2000 the Emergency Hotwell Pump
6M721-5721-1 Condensate Storage and Transfer System Revision U Operating Sketch 6M721-2006 Condensate Storage and Transfer System Revision AZ Diagram CARD 99-11515 CST Level Indication Lost Due to Freezing January 5, 1999 000Z2990526 Change Oil in MDCT Fan C Gear Reducer March 31, 1999 STR 99-0296 Safety Tagging Record to Change Oil in MDCT March 31, 1999 Fan C Gear Reducer CARD 00-17280 North TBCCW Pump Failed to Start May 17, 2000 WR V293960311 Refurbish 480 Volt Breaker 72M-2D, Test March 3, 2001 Relays, Power Shield and Ammeter STR 01-0219 Safety Tagging Record to Refurbish 480 Volt March 3, 2001 Breaker 72M-2 D, Test Relays, Power Shield and Ammeter 1R13 Maintenance Risk Assessment and Emergent Work WR 000Z991909 Perform ASME As-Found & As-Left Relief November 16, 2001 Valve Testing Per 43.000.002 System Operating Residual Heat Removal System, Attachment 1B, Revision 73 Procedure 23.205 Div 2 RHR Initial Valve Lineup STR 2001-006932 Safety Tagging Record for E1100 Division 2 November 24, 2001 RHR System Outage CARD 01-19330 Nuclear Operator Finds and Stops Leak From November 24, 2001 Division 2 RHR (Mispositioned Valve)
WR 000Z013640 5N Heater Tube Leak Identified During FW November 6, 2001 Heater Integrity Check During S/D System Operating Extraction Steam and Heater Drains Revision 55 Procedure 23.108 STR 2001-006851 Safety Tagging Record for N2003B003 November 20, 2001 CARD 01-19702 Mispositioned Valve N3016F355 Found Open November 28, 2001 During Investigation of High Offgas In Flow Drawing 6M721- Main Turbine Extraction Steam System Revision R 5717-2, Functional Operating Sketch System Operating Reactor Feedwater and Condensate Systems Revision 91 Procedure 23.107
STR 2001-007059 Safety Tagging Record for N. Reactor November 28, 2001 Feedwater Pump CARD 01-22208 Level 8 Trip While Unisolating NRFP November 28, 2001 1R14 Nonroutine Plant Evolutions MLS 11 Licensing/Safety Engineering Conduct Manual, Revision 10 Chapter 11 - Post Event investigations CARD 01-22208 Level 8 Trip While Unisolating the NRFP November 28, 2001 ST-OP-315-0046- Figure 2: Feedwater System 001 Procedure 23.107 System Operating Procedure, Reactor Revision 65 Feedwater System GETARS General Electric Transient Analysis Recording November 28, 2001 System Data: Wide Range Reactor Pressure, Narrow Range Reactor Level, Wide Range Level DCS Digital Control System Data: Feedwater Level November 28, 2001 Control Summary DCS Digital Control System Data: Feedwater Control November 28, 2001 System Flow Summary DCS Digital Control System Data: North Reactor November 28, 2001 Feedwater Pump Flows and Pressures Scram 01-01 Post Scram Data Evaluation December 6, 2001 Scram 01-01 Sequence of Events Recorder Data December 6, 2001 Scram 01-01 Average Power Range Monitor Traces December 6, 2001 Scram 01-01 Traces for Reactor Vessel Level December 6, 2001 Scram 01-01 Traces B21-R623A, Reactor Vessel December 6, 2001 Level/Pressure Scram 01-01 Traces B21-R623B, Reactor Vessel December 6, 2001 Level/Pressure Scram 01-01 Traces B21-R613, Core Flow December 6, 2001 Scram 01-01 Traces C32-R609, Vessel Pressure December 6, 2001 Scram 01-01 Traces C32-R607, Main Steam Flow December 6, 2001 Scram 01-01 Traces N30-R824, Condenser Vacuum December 6, 2001
Scram 01-01 Traces B21-R007, Vessel Metal Temperature December 6, 2001 Scram 01-01 General Electric Transient Analysis System Data December 6, 2001 CARD 01-22371 Manual Scram due to loss of Stator Water December 6, 2001 Cooling System 1R15 Operability Evaluations CARD 01-20890 HPCI fluid transient during performance of November 28, 2001 24.202.02 1R16 Operator Work-Arounds NPOP-01-0199 Aggregate Assessment of Operator Work September 17, 2001 Arounds TMIS-01-0155 Risk Assessment of Revised Operator Work September 17, 2001 Arounds - September 2001 ODE-006 Operator Work Arounds (ODE-006) October 2001 1R17 Permanent Plant Modifications EDP 29068 Exciter-Regulator Replacement for EDG 11, 12 Revision A and 14 ECRs 29068-1 Changes for Packages to The Exciter-Regulator Revisions A, B, C through 9 Replacement for EDG 11, 12 and 14 and O MES 19 Preparation and Control of Engineering Design Revision 13 Packages 1R19 Post Maintenance Testing Log No.01-048 ISI/NDE-IST Program Evaluation Sheet: November 8, 2001 Functionality of Snubber E11-3158-G30 Log No.98-008 Pacific Scientific Snubber Visual Examination for September 8, 1998 Snubber B21-4093-G13 CARD 01-21931 SST Job Performance Records Do Not Reflect December 5, 2001 Actual Completed Performance CARD 01-21930 Missing Documentation for Testing December 5, 2001 Snubber 810064
Form NIS-2 ASME Section XI Owners Report for Repairs or November 10, 2001 Replacements of EESW Pipe Support P45-3353-G14 DER 91-0010 Deviation Event Report: Generic Letter 90-09, December 29, 1990 Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions TSR-3155 Technical Service Request: EESW Column July 13, 2001 Separation Mitigation RID 70971 Replacement Installation Document: Restore November 8, 2001 Function of Strut P45-3353-G14 Log 01-030 Pacific Scientific Snubber Visual Examination for November 2, 2001 Snubber B21-2593-G13 Procedure Snubber Functional Test Revision 37 43.000.011 ISI-NDE Program Part C: Inservice Inspection-Nondestructive Revision 2 Examination (ISI-NDE) Program (Plan) for Snubbers Log No.98-008 Pacific Scientific Snubber Visual Examination for September 7, 1998 Snubber B21-2593-G13 WR 000Z932245 Test Reactor Core Isolation Cooling (RCIC) E51- June 16, 1993 3174-G33 Snubber WR 000Z930020 Remove Snubbers, Install Struts and Modify January 11, 1993 Insulation on Drywell Piping WR 000Z973173 Refurbish Snubbers as Required During RF-06 December 22, 1997 Job ID Rebuild Hydraulic Snubbers During Refuel to May 11, 1994 A334930628 Satisfy EQ, TS and Perform Preventive Maintenance WR 000Z951759 Snubber E11-3152-G21 Has Cracked Tack Weld February 17, 1995 but is Operable WR 000Z968255 Rebuild Hydraulic Snubbers During Refuel 06 to December 6, 1996 Satisfy EQ, TS and Perform Preventive Maintenance WR 000Z946846 Snubber E11-3164-G26 is Leaking Oil At September 26, 1994 Noticeable Rate (Puddle on Floor)
WR 000Z013579 Pipe Clamp Yoke is Bent and Clamp is Loose, November 1, 2001 and Snubber Bushing is Wedged
WR 000Z953605 Replace the Other 3 Clamp Studs and Nuts For May 26, 1995 N30-3529-G36 WR 000Z945303 Investigate and Repair Snubber N3059G084 June 11, 1994 Job ID Embedded Pipe plate for Snubber P1166G009 is November 11, 1989 009C891031 Damaged Generic Letter Alternative Requirements for Snubber Visual December 11, 1990 90-09 inspection Intervals and Corrective Actions 1R20 Refueling and Outages Operations Reactivity Management Revision 0 Conduct Manual MOP19 Procedure 23.623 Reactor Manual Control System Revision 45 Procedure Plant Startup to 25% Power Revision 53 22.000.02 Infrequently Cycle 9 Startup Test Program Revision 0 Performed Test or Evolution IPTE 01-01 1R22 Surveillance Testing Procedure EDG 14 Loss of Offsite Power and ECCS Start Revision 32 24.307.04 with Loss of Offsite Power Test.
Technical 3.8.1 AC Sources - Operating Specifications 3.8.2 AC Sources - Shutdown 1EP6 Drill Evaluation Scenario 30.2 Drill/Exercise Critique Summary, RERP Blue May 31, 2001 Team, Shift 5, May 1, 2001 CARD 01-10171 EOF Related Followup Actions As a Result of March 19, 2001 the March 7, 2001 RERP Drill CARD 01-10190 Problems with Medical Response When the May 18, 2001 Plant Nurse or First Responder is not Available Scenario 30.2 Drill/Exercise Critique Summary, RERP Red August 9, 2001 Team, Shift 1, July 17, 2001
CARD 01-10195 RERP: Evaluate RP Concerns for Security January 11, 2001 Personnel Response During Emergencies CARD 01-16624 RERP Telephone System July 31, 2001 2OS1 Access Control to Radiologically Significant Areas EF2 Radiation Protection Organization November 12, Self-Assessment of Plant Conditions and 2001 Personnel Performance during RF08 from Tours Conducted by RPO Personnel CARD 01-17799 RRA Access Denial November 9, 2001 CARD 01-21706 Worker Leaves Areas After Alarming PCM November 16, 2001 CARD 01-21708 Accessing LHRA Gates December 2, 2001 MRP06 Accessing and Control of High Radiation, Locked Revision 4 High Radiation, and Very High Radiation Areas 2OS2 As-Low-As-Is-Reasonably-Achievable (ALARA) Planning and Controls EF2 Radiation Protection Organization November 12, 2001 Self-Assessment of Plant Conditions and Personnel Performance During RF08 from Tours Conducted by RPO Personnel Audit Report Nuclear Quality Assurance Audit Report 01-0115 October 22 -
01-0115 - Radiation Protection Program November 26, 2001 2PS2 Radioactive Material Processing and Transportation Fermi 2 UFSAR Sections 11.2 and 11.5 Revision 7 and Revision 8 Radioactive Material Shipment Logs 1999 - 2001 CARD 00-12105 HAZMAT Training Requirements January 3, 2000 CARD 01-12016 Liner LH-01-001 has 4 Defective Dewatering February 27, 2001 Elements CARD 01-17909 RWCU Demin A Would Not Go Into Service November 25, 2001 CARD 01-19082 Lockout Occurred Unexpectedly November 23, 2001 Fermi 2 Technical Fermi 2 Process Control Program Manual Revision 19 Manual LP-GN-528-0003 Hazardous Material (HAZMAT) Orientation, Revision 0 Function Specific Training - Level 1 MRP24 Fermi 2 10CFR61 Compliance Manual Revision 1
NRC-01-0031 Annual Radioactive Effluent Release and May 1, 2001 Radiological Environmental Operating Reports NRPC-01-0166 Scaling Factors Report Dated April 25, 2001 May 29, 2001 NRPC-01-0168 Validation of Stainless Steel Laundry Container May 30, 2001 Shipment Using DAW Scaling Factors, Sample Reference Date - January 12, 2001 Plant Technical Transportation Accidents Involving Radioactive Revision 7 Procedure Material from Fermi 2 20.000.27 Plant Technical Shipping Low Specific Activity (LSA) Radioactive Revision 16 Procedure Material 65.000.506 Plant Technical Shipping Less Than or Equal to A1, A2 Quantities Revision 11 Procedure of Radioactive Material 65.000.508 Plant Technical Shipping Greater Than A1, A2 Quantities of Revision 14 Procedure Radioactive Materials 65.000.509 Plant Technical Receipt, Storage, Inventory, Inspection and Revision 9 Procedure Packing of Radioactive Material Shipping 65.000.515 Packages Plant Technical Shipping Surface Contaminated Object Revision 4 Procedure Radioactive Material 65.000.522 Plant Technical Radwaste Shipments Revision 4 Procedure 65.000.523 Radioactive Irradiated Reactor Hardware Liner #95455-6-3/4 January 11, 1999 Material Shipment 99-001 Radioactive Cs-137 Calibration Source (L-96-0027) August 19, 1999 Material Shipment 99-041 Radioactive Irradiated Hardware Liner L91-001 May 10, 2000 Material Shipment 00-013 Radioactive HP Turbine Rotor April 18, 2000 Material Shipment 00-040 Radioactive Dewatered Powdered, Charcoal, & Bead Resin September 27, Material Shipment LH-94-009 2000 00-089 Radioactive Powdered and Bead Resin (Unprocessed) Liner May 8, 2001 Material Shipment LH-00-005 01-022
Radioactive 13 High Rad Drums (Compacted DAW) June 12, 2001 Material Shipment 01-030 Radioactive Laundry November 21, Material Shipment 2001 01-077 RWP 01-1006 Survey, Segregate, and Compact Dry Active Revision 1 Waste. Perform Maintenance, Handling, Preparation, and Shipping of Radioactive Material RWP 01-1014 Transfer and Process Water, Oil, Filter Media, Revision 2 and Filters. Hook-up, Tear Down, and Repair Equipment Associated with Dewatering and Solidifying Liners Vendor Procedure Set Up and Operating Procedure for the Revision 20 FO-OP-032-483 RDS-1000 Unit at Detroit Edison Fermi-2 4OA1 Performance Indicator Verification Second and Third Quarter Performance Indicators for HPCI, RCIC, RHR, and Emergency AC Power Safety System Unavailability Second and Third Quarter Performance Indicators for Safety System Functional Failures TMTE-01-0125 NRC Performance Indicators for HPCI, RCIC, July 17, 2001 RHR, and Emergency AC Power Systems Second Quarter 2001 Safety System Unavailability TMTE-01-0186 NRC Performance Indicators for HPCI, RCIC, October 10, 2001 RHR, and Emergency AC Power Systems Third Quarter 2001 Safety System Unavailability Dwg Residual Heat Removal (RHR) Division II Revision X 6M721-5706-1 Functional Operating Sketch Dwg Residual Heat Removal (RHR) Division I Revision V 6M721-5706-2 Functional Operating Sketch Dwg RHR Service Water Make Up Decant and Revision U 6M721-5706-2 Overflow Systems Functional Operating Sketch Dwg High Pressure Coolant (HPCI) Injection System Revision AC 6M721-5708-1 Functional Operating Sketch Dwg HPCI Lube Oil/Control Oil System Functional Revision H 6M721-5708-2 Operating Sketch
Dwg Reactor Core Isolation Cooling (RCIC) System Revision AC 6M721-5709-1 Functional Operating Sketch Dwg RCIC Lube Oil/Control Oil System Functional Revision E 6M721-5709-2 Operating Sketch Procedure ECCS - HPCI Torus Level Functional Test Revision 34 44.030.155 Procedure ECCS - HPCI/RCIC Condensate Storage Tank Revision 21 44.030.400 Level Loop, E41-N061B Calibration/Functional Procedure NSSSS -HPCI Exhaust Diaphragm Pressure, Revision 27 44.020.219 Division I Functional Test Procedure NSSSS - HPCI and RCIC Room Area Revision 29 44.020.227 Temperature Channel A Functional Test Unit Logs for HPCI (E41), RCIC (E51) and RHR April 1, 2001 -
(E11) September 30, 2001 CARDs for HPCI (E41), RCIC (E51) and RHR April 1, 2001 -
(E11) September 30, 2001 STR 2001-005498 Safety Tagging Record: Repack HPCI Cooling April 23, 2001 Water to Lube Oil Pressure Control Valve STR 2001-005472 Safety Tagging Record: Adjust Torque Switch April 20, 2001 for HPCI turbine Exhaust Stop Check Valve STR 2001-005758 Safety Tagging Record: Clean Orifices D008 April 16, 2001 and 009 on HPCI Barometric Condenser STR 2001-006431 Safety Tagging Record: Test/ inspect MOV August 7, 2001 MCC E5150F001 STR 2001-006644 Safety Tagging Record: Implement EDP 30202 September 24, 2001 to Replace GEMAC Flow Control Station STR 2001-007122 Safety Tagging Record: Troubleshoot Cause for December 17, 2001 Valve E5150F044 not Opening STR 2001-005659 Safety Tagging Record: Test Thermal April 10, 2001 Overloads for Valve E1150F611A STR 2001-005935 Safety Tagging Record: Repack Valve April 27, 2001 E1100F086 STR 2001-005964 Safety Tagging Record: Repack Pump May 7, 2001 E1156C003 STR 2001-006041 Safety Tagging Record: Electrical Maintenance May 18, 2001 on Pump B
STR 2002-006041 Safety Tagging Record: Electrical Maintenance May 18, 2001 on Pump D STR 2001-006022 Safety Tagging Record: Test Thermal May 15, 2001 Overloads for Valves E1150F024B and F027B STR 2001-06122 Safety Tagging Record: Check Torque on Blade June 4, 2001 Clamping Hardware Bolting, Clean Blades, lubricate motor STR 2001-006423 Safety Tagging Record: Test Thermal July 31, 2001 Overloads for E1150F007A and F027A and Test MCC Positions and Valves E1150 F004A, F004C and F016A. Lubricate E1150F034A and C 34