IR 05000263/2002002

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IR 05000263/2002-002(DRP), on 01/01-02/13/2002; Nuclear Management Company, LLC; Monticello Nuclear Generating Plant; Integrated Inspection Report. No Violations Noted
ML020590139
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 02/27/2002
From: Burgess B
NRC/RGN-III/DRP/RPB2
To: Forbes J
Nuclear Management Co
References
IR-02-002
Download: ML020590139 (22)
v • d • e


Text

ary 27, 2002

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT NRC INSPECTION REPORT 50-263/02-02(DRP)

Dear Mr. Forbes:

On February 13, 2002, the NRC completed an inspection at your Monticello Nuclear Generating Plant. The results of this inspection were discussed on February 12, 2002, with you and other members of your staff. The enclosed report presents the results of that inspection.

The inspection was an examination of activities conducted under your license as they relate to reactor safety, verification of performance indicators, event followup, emergency preparedness, and compliance with the Commission's rules and regulations and with the conditions of your license. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations of activities, and interviews with personnel. Based on the results of this inspection, the NRC did not identify any issues which were categorized as being risk significant.

Immediately following the terrorist attacks on the World Trade Center and the Pentagon, the NRC issued an advisory recommending that nuclear power plant licensees go to the highest level of security, and all promptly did so. With continued uncertainty about the possibility of additional terrorist activities, the Nation's nuclear power plants remain at the highest level of security and the NRC continues to monitor the situation. This advisory was followed by additional advisories and although the specific actions are not releasable to the public, they generally include increased patrols, augmented security forces and capabilities, additional security posts, heightened coordination with law enforcement and military authorities, and more limited access of personnel and vehicles to the sites. The NRC has conducted various audits of your response to these advisories and your ability to respond to terrorist attacks with the capabilities of the current design basis threat (DBT). From these audits, the NRC has concluded that your security program is adequate at this time. In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's 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/

Bruce L. Burgess, Chief Branch 2 Division of Reactor Projects Docket No. 50-263 License No. DPR-22

Enclosure:

Inspection Report 50-263/02-02(DRP)

REGION III==

Docket No: 50-263 License No: DPR-22 Report No: 50-263/02-02(DRP)

Licensee: Nuclear Management Company, LLC Facility: Monticello Nuclear Generating Plant Location: 2807 West Highway 75 Monticello, MN 55362 Dates: January 1 through February 13, 2002 Inspectors: S. Burton, Senior Resident Inspector D. Kimble, Resident Inspector R. Jickling, Regional Emergency Preparedness Inspector Approved by: Bruce L. Burgess, Chief Branch 2 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000263/02-02(DRP), on 01/01-02/13/2002; Nuclear Management Company, LLC; Monticello Nuclear Generating Plant; integrated inspection report.

The inspection was conducted by resident and regional inspectors. The report covers a 61/2 week period of resident inspection. No findings were identified in any cornerstones.

The NRC's 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.

A. Inspector Identified Findings None.

B. Licensee Identified Violations A violation of very low significance which was identified by the licensee has been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. The violation is listed in Section 4OA7 of this report.

Report Details Summary of Plant Status The plant began the inspection period operating at full power. Power was reduced to approximately 63 percent on January 11, 2002, to facilitate post-refuel outage maximum core flow testing and licensee investigation into excessive condenser air in-leakage. The plant returned to full power operation on January 12. On January 21, the plant scrammed from full power on a turbine load rejection signal, which resulted from a failure in the turbine control system (Sections 1R14 and 1R20). The reactor was restarted on January 25, and full power was reached on January 29. On February 8, power was reduced to approximately 75 percent to facilitate routine periodic main steam valve testing. Full power was restored on February 10, 2002 and the plant remained at or near full power for the remainder of the inspection period.

1. REACTOR SAFETY Cornerstones: Initiating events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness 1R04 Equipment Alignment (71111.04)

a. Inspection Scope The inspectors performed a partial walkdown of the following redundant equipment trains or systems to verify operability and proper equipment lineup while a counterpart train or system was disabled due to planned maintenance. These systems or trains were selected due to the increase in core damage frequency caused by rendering another system or train out-of-service for maintenance.

  • RCIC and other ECCS with HPCI out-of-service The inspectors verified the position of critical redundant equipment and looked for any discrepancies between the existing equipment lineup and the required lineup.

b. Findings No findings of significance were identified.

1R05 Fire Protection (71111.05)

a. Inspection Scope The inspectors walked down the following risk significant areas looking for any fire protection issues. The inspectors selected areas containing systems, structures, or components that the licensee identified as important to reactor safety.

  • Fire Zone 7A, Division I 125Vdc Battery Room
  • Fire Zone 7B, Division I 250Vdc Battery Room
  • Fire Zone 7C, Division II 125Vdc Battery Room
  • Fire Zone 24, Diesel Fire Pump Room The inspectors reviewed the control of transient combustibles and ignition sources, fire detection equipment, manual suppression capabilities, passive suppression capabilities, automatic suppression capabilities, and barriers to fire propagation.

b. Findings No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11)

a. Inspection Scope The inspectors observed a training crew during an evaluated requalification exam simulator scenario and reviewed licensed operator performance in mitigating the consequences of events. Areas observed by the inspectors included: clarity and formality of communications, timeliness of actions, prioritization of activities, procedural adequacy and implementation, control board manipulations, managerial oversight, emergency plan execution, and group dynamics.

b. Findings No findings of significance were identified.

1R12 Maintenance Rule Implementation (71111.12)

a. Inspection Scope The inspectors reviewed the licensee's implementation of the Maintenance Rule (10 CFR 50.65) to ensure rule requirements were met for the selected systems.

The following systems were selected based on being designated as risk significant under the Maintenance Rule, or being in the increased monitoring (Maintenance Rule category a(1)) group:

  • Area Radiation Monitoring System
  • Non-Essential Diesel Generator The inspectors verified the licensee's categorization of specific issues, including evaluation of the performance criteria. The inspectors reviewed the licensee's implementation of the maintenance rule requirements, including a review of scoping, goal-setting, and performance monitoring; short-term and long-term corrective actions; functional failure determinations associated with the condition reports reviewed; and current equipment performance status.

b. Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope The inspectors reviewed and observed emergent work, preventive maintenance, or planning for risk significant maintenance activities. The inspectors observed maintenance or planning for the following activities or risk significant systems undergoing scheduled or emergent maintenance.

No. 22-07

  • Diagnoses and Repair of Reactor Pressure Regulation and Turbine Control The inspectors also reviewed the licensee's evaluation of plant risk, risk management, scheduling, and configuration control for these activities in coordination with other scheduled risk significant work. The inspectors verified that the licensee's control of activities considered assessment of baseline and cumulative risk, management of plant configuration, control of maintenance, and external impacts on risk. In-plant activities were reviewed to ensure that the risk assessment of maintenance or emergent work was complete and adequate, and that the assessment included an evaluation of external factors. Additionally, the inspectors verified that the licensee entered the appropriate risk category for the evolutions.

b. Findings No findings of significance were identified.

1R14 Personnel Performance During Nonroutine Plant Evolutions and Events (71111.14)

a. Inspection Scope The inspectors reviewed personnel performance during an unplanned scram on January 21, 2002, that was caused by a turbine control system malfunction and an associated load reject scram. The inspectors independently evaluated the initiating cause of the scram and operator actions in response to the event. This evaluation included a review of operator logs and plant computer data, and personnel interviews.

b. Findings No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope The inspectors reviewed the technical adequacy of the following operability evaluations to determine the impact on Technical Specifications (TS), the significance of the evaluations, and to ensure that adequate justifications were documented.

  • Drywell Floor Drain Sump Pump
  • Auxiliary Transformer 1AR Voltage Below Specification Operability evaluations were selected based upon the relationship of the safety-related system, structure, or component to risk.

b. Findings No findings of significance were identified.

1R16 Operator Workarounds (OWA) (71111.16)

a. Inspection Scope The inspectors reviewed OWA No.01-149, "Repeated High Level Alarms on CRD HCU 22-07 Require Operator Actions." The inspectors reviewed the workaround's potential to impact the operators' ability to adequately assess the CRD accumulator's operational status.

b. Findings No findings of significance were identified.

1R17 Permanent Plant Modifications (71111.17)

a. Inspection Scope

The inspectors reviewed the feed water heater level control modification installed during the December 2001 refueling outage to verify that the design basis, licensing basis, and performance capability of risk significant systems were not degraded by the installation of the modification. The inspectors also verified that the modifications did not place the plant in an unsafe configuration. The inspectors considered the design adequacy of the modification by performing a review, or partial review, of the modifications impact on plant electrical requirements, material requirements and replacement components, response time, control signals, equipment protection, operation, failure modes, and other related process requirements.

b. Findings No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope The inspectors selected the following post-maintenance activities for review. Activities were selected based upon the structure, system, or component's ability to impact risk.

  • Forced Outage MO-2076, RCIC Outboard Steam Isolation Valve, Repair The inspectors verified by witnessing the test or reviewing the test data that post-maintenance testing activities were adequate for the above maintenance activities.

The inspectors reviews included, but were not limited to, integration of testing activities, applicability of acceptance criteria, test equipment calibration and control, procedural use and compliance, control of temporary modifications or jumpers required for test performance, documentation of test data, TS applicability, system restoration, and evaluation of test data. Also, the inspectors verified that maintenance and post-maintenance testing activities adequately ensured that the equipment met the licensing basis, TS, and Updated Safety Analysis Report (USAR) design requirements.

b. Findings No findings of significance were identified.

1R20 Outage Activities (71111.20)

a. Inspection Scope The inspectors evaluated outage activities for an unscheduled outage that began on January 21, 2002, and ended on January 27, 2002. The unplanned outage was the result of a turbine control system malfunction that resulted in a load reject scram. The inspectors reviewed activities to ensure that the licensee considered risk in developing, planning, and implementing the outage schedule. The inspectors observed or reviewed the reactor shutdown and cooldown, outage equipment configuration and risk

management, electrical lineups, selected clearances, control and monitoring of decay heat removal, control of containment activities, startup and heatup activities, corrective actions, and identification and resolution of problems associated with the outage.

b. Findings No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope The inspectors selected the following surveillance test activities for review. Activities were selected based upon risk significance and the potential risk impact from an unidentified deficiency or performance degradation that a system, structure, or component could impose on the unit if the condition were left unresolved.

  • Service Water Pump - Safety Related 480 Vac Circuit Breaker 10-year Maintenance
  • Rod Block Monitor Quarterly Functional Test and Calibration
  • Periodic Reactivity Anomaly Check The inspectors observed the performance of surveillance testing activities, including reviews for preconditioning, integration of testing activities, applicability of acceptance criteria, test equipment calibration and control, procedural use, control of temporary modifications or jumpers required for test performance, documentation of test data, TS applicability, impact of testing relative to performance indicator reporting, and evaluation of test data.

b. Findings No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)

a. Inspection Scope The inspectors reviewed Revisions 19 and 20 of the Monticello Nuclear Generating Plant Emergency Plan to determine whether changes identified in the revisions reduced the effectiveness of the licensees emergency planning, pending onsite inspection of the implementation of these changes.

b. Findings No findings of significance were identified.

1EP6 Drill Evaluation (71114.06)

a. Inspection Scope The resident inspectors reviewed a simulator-based training evolution to evaluate drill conduct and the adequacy of the licensees critique of performance to identify weaknesses and deficiencies. The inspectors selected simulator scenarios that the licensee had scheduled as providing input to the Drill/Exercise Performance Indicator.

The inspectors observed, when applicable, the classification of events, notifications to off-site agencies, protective action recommendation development, and drill critiques.

Observations were compared to the licensees observations and corrective action program entries. The inspectors verified that there were no discrepancies between observed performance and performance indicator reported statistics. The simulator scenario observed resulted in an unusual event and alert classifications.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES 4OA1 Performance Indicator Verification (71151)

Cornerstones: Mitigating Systems Safety System Unavailability a. Inspection Scope The inspectors verified the accuracy and completeness of the Safety System Unavailability - Emergency AC Power and Residual Heat Removal System performance indicator data submitted by the licensee from January 1, 2001, through December 31, 2001. The inspectors reviewed data reported to the NRC since the last verification. The review was accomplished, in part, through evaluation of the TS requirements, plant records, procedural reviews, and reactor coolant sample data.

b. Findings No findings of significance were identified.

4OA3 Event Follow-up (71153)

Cornerstones: Initiating Events and Barrier Integrity

.1 (Closed) Licensee Event Report 50-263/2001-011: "Worker Jarred Sensitive Instrument Rack Causing Scram" a. Inspection Scope

The inspectors evaluated LER 50-263/2001-011, "Worker Jarred Sensitive Instrument Rack Causing Scram."

b. Findings On October 23, 2001, with the plant in coastdown for refueling, a radiation protection specialist in the normal course of his duties inadvertently bumped a sensitive instrument rack in the reactor building. The jarring of the instrument rack subsequently caused a primary containment isolation system (PCIS) Group 1 isolation signal which resulted in a reactor scram signal directly off of main steam isolation valve (MSIV) closure.

The licensee's analysis of the event showed that despite a slight complication caused by the locking of the feedwater regulating valves (FRVs) in the fully open position and the resulting loss of both main feed pumps, the safety significance of the event was low due to the operating crew's ability to successfully complete the reactor scram abnormal operating procedure. The inspectors had previously examined the locking of the FRVs associated with the reactor scram, and findings and enforcement actions are documented in NRC Inspection Report 50-263/01-09, Section 4OA3. Additionally, the inspectors had previously examined human performance issues associated with the post-scram inadvertent partial depressurization of the reactor vessel, and findings and enforcement actions are documented in NRC Inspection Report 50-263/01-16, Section 4OA2.2. The licensee has entered this issue into their corrective action program as CR 20016420.

.2 (Closed) Licensee Event Report 50-263/2001-012: "Refueling Testing Identifies Containment Isolation Valve Leakage Greater than Allowed by the Technical Specifications" a. Inspection Scope The inspectors evaluated LER 50-263/2001-012, "Refueling Testing Identifies Containment Isolation Valve Leakage Greater than Allowed by the Technical Specifications."

b. Findings During the November 2001 refueling outage, local leak rate testing results for containment isolation valves indicated that Technical Specification and 10 CFR 50, Appendix J, limits were exceeded. The licensee identified four containment isolation valves, including two of four outboard MSIVs, with excessive leakage. The licensee's analysis of the condition indicated that even with these valves leaking in excess of established specifications, public health and safety was not adversely impacted because redundant valves in series in each pathway were tested satisfactorily for leakage.

A licensee identified violation associated with this issue is documented in Section 4OA7 of this report. The licensee has entered this issue into their corrective action program under the following CRs: 20017070, 20017090, 20017091, and 20017092.

4OA6 Meeting Exit Meeting The inspectors presented the inspection results to M and other members of licensee management on February 12, 2002. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee Identified Violation The following violation of very low safety significance was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. If you deny this NCV, you should provide a response with the basis of your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the Monticello facility.

NCV Tracking Number Requirement Licensee Failed to Meet NCV 50-263/02-02-01 Technical Specification 3.7.A.2.b.2 requires that the combined maximum flow path leakage rate for all penetrations and valves subject to Type B and C leak rate testing at the calculated peak containment accident pressure of 42 psig, Pa, be less than or equal to 0.6 La, where La is the maximum allowable leakage rate in percent by weight of the containment air volume at Pa. Additionally, Technical Specification 3.7.A.2.b.3 requires that the combined maximum flow path leakage for all MSIVs be less than or equal to 46 scf per hour when tested at 25 psig.

Contrary to these requirements, as-found leak rate testing conducted during the licensee's November 2001 refueling outage revealed valve leakage in excess of the specifications. This violation is being treated as a NCV consistent with Section VI.A of the NRC Enforcement Policy.

KEY POINTS OF CONTACT Licensee G. Bregg, Manager, Quality Services D. Fadel, Director of Engineering J. Forbes, Site Vice-President J. Grubb, General Superintendent, Operations K. Jepson, General Superintendent, Chemistry and Radiation Services B. Linde, Superintendent, Security D. Neve, Acting Licensing Project Manager J. Purkis, Plant Manager B. Sawatzke, General Superintendent, Maintenance C. Schibonski, General Superintendent, Safety Assessment E. Sopkin, General Superintendent, Engineering NRC B. Burgess, Chief, Reactor Projects Branch 2 ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-263/02-02-01 NCV Refueling Testing Identifies Containment Isolation Valve Leakage Greater than Allowed by the Technical Specifications (Section 4OA7)

Closed 50-263/2001-011 LER Worker Jarred Sensitive Instrument Rack Causing Scram (Section 4OA3.1)

50-263/2001-012 LER Refueling Testing Identifies Containment Isolation Valve Leakage Greater than Allowed by the Technical Specifications (Section 4OA3.2)

50-263/02-02-01 NCV Refueling Testing Identifies Containment Isolation Valve Leakage Greater than Allowed by the Technical Specifications (Section 4OA7)

Discussed None.

LIST OF ACRONYMS USED AC Alternating Current AWI Administrative Work Instruction CFR Code of Federal Requirements CR Condition Report CRD Control Rod Drive DBT Design Basis Threat DC Direct Current DG Diesel Generator DRP Division of Reactor Projects EDG Emergency Diesel Generator EPR Electric Pressure Regulator ESW Emergency Service Water FRV Feedwater Regulating Valve FW Feedwater HCU Hydraulic Control Unit HPCI High Pressure Core Injection IR Inspection Report LER Licensee Event Report LLRT Local Leak Rate Testing LPCI Low Pressure Core Injection MPR Mechanical Pressure Regulator MSIV Main Steam Isolation Valve NCV Non-Cited Violation NMC Nuclear Management Company NUMARC Nuclear Management and Resources Council OWA Operator Workaround OWI Operations Work Instruction PCIS Primary Containment Isolation System psig Pounds Per Square Inch Gauge PTL Pull-To-Lock RCIC Reactor Core Isolation Cooling RHRSW Residual Heat Removal Service Water scf Standard Cubic Feet SWI Scheduling Work Instruction TS Technical Specification URI Unresolved Item USAR Updated Safety Analysis Report Vac Volts Alternating Current Vdc Volts Direct Current WO Work Order

LIST OF DOCUMENTS REVIEWED 1R04 Equipment Alignment Operations Manual:

B.09.08 - Emergency Diesel Generators B.03.02 - HPCI System B.08.01.02 - EDG ESW System B.02.03 - RCIC System B.09.06 - 4160 Vac System Drawings:

M-123 - HPCI System (Steam Side) Revision AF M-123-1 - HPCI Hydraulic Control and Lubrication Revision B M-124 - HPCI System (Water Side) Revision Y M-125 - RCIC System (Steam Side) Revision AK M-126 - RCIC System (Water Side) Revision Y M-110-1 - Service Water System Revision BL M-112 - RHR and Emergency Service Water Systems Revision BF M-811-1 - Service Water and Make-Up Intake Structure Revision CD M-133-1 - Diesel Oil System Revision AD NF-36298-1 - Electrical Load Flow Revision M NF-36298-2 - DC Electrical Load Distribution Revision A 1R05 Fire Protection NX-16991 Monticello Updated Fire Hazards Analysis Technical Manual Plant Fire Strategies:

A.3-07A - 125Vdc - Division I Battery Room Revision 3 A.3-07B - 250Vdc - Division I Battery Room Revision 5 A.3-07C - 125Vdc - Division II Battery Room Revision 3 A.3-24 - Diesel Fire Pump Room Revision 4 Procedures and Administrative Work 4AWI-08.01.01 Instructions: Revision 17 4AWI-08.01.02 - Fire Prevention Practices Revision 6 0271 - Combustion Source Use Permit Revision 27

- Fire Hose Station and Yard Hydrant Hose 0274 House Equipment Inspection Revision 19

- Fire Hose Hydrostatic Test Interior Hose 0275-1 Stations Revision 9 0275-2 - Fire Barrier Penetration Seal Visual Inspection Revision 16

- Fire Barrier Wall, Damper, and Floor Inspection QUAD-5-80-009 Specifications for Installation of Electrical and Revision 7 Mechanical Penetration Seals at the Monticello Nuclear Generating Plant by Quadrex Corporation

Operations Manual:

A.3 - Fire Fighting Procedures B.08.05 - Fire Protection 1R11 Licensed Operator Requalification Program RQSS-28 Licensed Operator Annual Examination Scenario Revision 9 1R12 Maintenance Rule Implementation NUMARC [Nuclear Management and Resources Council]:

93-01 - Nuclear Energy Institute Industry Guideline for Revision 2 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants 93-01, Section 11 - Assessment of Risk Resulting from the February 22, 2000 Performance of Maintenance Activities Regulatory Guides:

1.160 - Monitoring the Effectiveness of Maintenance at Revision 2 Nuclear Power Plants 1.182 - Assessing and Managing Risk Before May 2000 Maintenance Activities at Nuclear Power Plants 05.02.01 Engineering Work Instruction, Monticello Revision 5 Maintenance Rule Program Document Maintenance Rule Periodic Assessment Report 2nd Quarter - 2001 Operations Manual:

B.3.1 - Core Spray System B.5.12 - Area Radiation Monitors B.9.15 - Non-Essential Diesel Generator Maintenance Rule Program System Basis Document:

B.3.1 - Core Spray System Revision 1 B.5.12 - Area Radiation Monitors Revision 2 B.9.15 - Non-Essential Diesel Generator Revision 2 Design Basis Document:

B.3.1 - Core Spray System Revision 2 USAR: Revision 18 Section 6.2.2 - Core Spray System Sect. 14.7.2.3.1.3 - Core Spray System Section 7.5.3 - Area Radiation Monitoring System Section 8.4 - Plant Standby Diesel Generator Systems Drawings:

M-122 - Core Spray System Revision AB

CR 20003352 Core Spray Header Break Detection Readings Out of Service with Reactor >212 -F CR 20003767 Core Spray Motor Cooling Flow Less Than As Found Acceptance Criteria in 9/29/00 Performance of 1339 Test CR 20003881 Limiting Stroke Time for MO-1749/1750 Does not Agree with SAFER/GESTR Limit of 20 Seconds to Support Core Spray Initiation CR 20005138 B Core Spray Torus Suction Valve MO-1742 Failed to Open per Key Lock Switch Actuation During Post-Maintenance Testing CR 20017561 Relief Valve RV-1746 Setpoint Found Outside of Section XI Acceptance Band CR 20017260 Relief Valve RV-1745 Setpoint Found Outside of Section XI Acceptance Band CR 20017883 Bearing Flush Tube on B Core Spray Pump Cracked CR 20011801 Area Radiation Monitor RM-F1 Was Found Out-of-Tolerance During Calibration Performed on 2/22/01 CR 20015537 Three Area Radiation Monitors As-Found Out-of-Tolerance During Last Semiannual Calibration CR 20020387 13 DG Fuel Oil Tank Room Heater Does Not Work 1R13 Maintenance Risk Assessments and Emergent Work Control Procedures:

4AWI-04.01.01 - General Plant Operating Activities Revision 28 SWI-14.01 - Risk Management of On-line Maintenance Revision 0 4010PM - HCU Water Accumulator Replacement Revision 9 0074 - Control Rod Drive Exercise Revision 29 2188 - Scram Accumulator Nitrogen Charging Revision 13 0081 - Control Rod Drive Scram Insertion Time Test Revision 35 WO 0110506 Replace HCU 22-07 RWP 146 West CRD Accumulators Revision 2 WO 0200267 Disassemble , Inspect, Replace Components On MPR

WO 0200280 Back Flush the MPR Pressure Sensing Line WO 0200311 Adjust Needle Valve On MPR WO 0200296 Test MPR Dash Pot Bellows Relief Valves WO 0200264 Clean Inlet Screen / Replace EPR Moog valve WO 0200242 Inspect Turbine Front Standard Internal WO 0200277 Reset Primary Valve Limit Stop WO 0200259 Collect MPR Response Data WO 0200291 Install Temporary Test Equipment On MPR WO 0200281 Turbine Pressure Limit Switch Clearance Check CR 20020457 Reactor Scram 113 While At 100% Power CR 20020599 Test 1045 Halted Due to Turbine Control System Oscillations CR 20020573 Undocumented Modifications Made to the Mechanical Pressure Regulator During 1973 Turbine Outage CR 20020516 Calculated As-found Value Outside of Stated Band for MPR During Execution of Test 1045 CR 20020608 Turbine Control System Oscillation During Performance of Test 1045 on 1-25-02 1R14 Personnel Performance During Nonroutine Plant Evolutions and Events Operations Manual:

C.3 - Shutdown Procedure Revision 28 C.4-A - Reactor Scram Revision 19 C.4-B.5.7.A - Loss of Reactor Water Level Control Revision 5 C.4-B.6.5.A - Reactor Feedwater Pump Trip Revision 5 C.1 - Startup Procedure Revision 33 Scram 113 Summary Report 0000-B Operations Daily Log - Part B Revision 82 1R15 Operability Evaluations CR 20020628 Drywell Floor Drain Sump Pump P-24B Started in Standby and Pumped down to 169.5 Gal. Had to Be Manually Shutdown

Operations Manual:

B.7 - Liquid Radwaste Technical Specifications and Bases:

Section 3/4.6 - Primary System Boundary, Coolant Leakage Section 3/4.9 - Auxiliary Electrical Systems CR 20020943 1AR Voltage Indication out of Spec. Declared 1AR Inoperable and Placed 1AR Breaker Feeds to Bus 15 and 16 to PTL USAR: Revision 18 Section 8 - Plant Electrical Systems 1R16 Operator Workarounds Quarterly Operator Work-Around Review and December 17, 2001 Assessment Operations Manual:

B.1.2 - Control Rod Drives B.1.3 - CRD Hydraulic System USAR: Revision 18 Section 3.5.3 - Control Rod Drive System Design Basis Documents:

B.1.2 - Control Rod Drives Revision B B.1.3 - CRD Hydraulic System Revision B CR 19982736 CRD Operability Condition when the Associated Accumulator is Inoperable CR 20020093 NRC Resident Inspector Question on Operability of HCU Accumulators with Frequent High Water Level Alarms CR 20020152 Operability Evaluation Associated with CRD 22-07 Accumulator Alarms Was Limited in Scope, Rigor, and Diligence CR 20018297 Repeated High Level Alarm on CRD HCU 22-07 Require Operator Actions 1R17 Permanent Plant Modifications 00Q265 Feedwater Heater Level Controller Upgrade Revision 1 CR 20020013 FW Htrs E-12A/B Dump Valve LCs Not Fast Enough. Level Went High Offscale & Could Not Tell Margin to Turbine Damage

WO 0108346 Upgrade Level Controllers FW Htrs 12, 13, 14 &

WO 0108395 FW Heater LC Upgrade Pre-Op Test 1R19 Post-Maintenance Testing Drawings and Prints:

M-125 - RCIC (Steam Side) Revision AM NX-8435-150-2 - Turbine Control Diagram Revision G Procedures and Forms:

0137-07A - Reactor Steam Supply Valves Leak Rate Revision 15 1040-01 Testing Revision 42 3069 - Turbine Generator Revision 9 4900-01PM - Post-Maintenance Activities Control Cover Revision 15 EMP 01.01 Sheet Revision 7

- PM For Limitorque Motor Operated Valves

- Motor Operated Valve Testing Using VOTES SCR-02-0055 10 CFR 50.59 Screening for Installing Revision 0 Instrumentation to Monitor MPR Performance Tagout 02-00328 LLRT Isolation for RCIC Valve MO-2076 Version 1 WO 0200296 Test MPR Dash Pot Bellows Relief Valves WO 0200311 Adjust Needle Valve on MPR Control WO 0200325 MO-2076 Repair Valve Actuator WO 0200291 Install Temporary Test Equipment on MPR WO 0200274 Obtain Oil Sample of EPR Oil Reservoir WO 0200316 Set MPR and Associated Computer Alarm Point TRB178 WO 0200328 LLRT Isolation for RCIC Valve MO-2076 WO 0200320 MO-2076 Went to Dual Indication When Stroked Open 1R20 Outage Activities Operations Manual:

C.1 - Startup Procedure Revision 33 C.2 - Power Operation Revision 12 C.3 - Shutdown Procedure Revision 28 C.4-A - Reactor Scram Revision 19 2150 Plant Prestart Checklist Revision 23

2167 Startup Checklist Revision 41 2159 Predicted Critical For Plant Start Up Revision 6 1R22 Surveillance Testing Procedures:

4851-12-PM - ABB K-1600s and K-3000s 450 Volt Breaker Revision 6 Maintenance 0045 - Rod Block Monitor Functional Test and Revision 28 Calibration 0083 - Reactivity Anomaly Check Revision 12 CR 20018369 LCB-024 Breaker Closure Delayed During Start of #11 Service Water Pump WO 0108999 Perform 4851-12-OM on LCB-024 Located in 523-305 WO 0110540 Repair Failure of LCB-024 to Close Promptly Technical Specifications and Bases:

Section 3/4.2 - Protective Instrumentation Section 3/4.3E - Reactivity Anomaly Card 2070971088 Instrument Calibration Worksheet for Recorder NR-7-46C 1EP4 Emergency Action Level and Emergency Plan Changes Monticello Nuclear Generating Plant Emergency Revision 19 Plant Monticello Nuclear Generating Plant Emergency Revision 20 Plan 1EP6 Drill Evaluation RQSS-28 Licensed Operator Annual Examination Scenario Revision 9 4OA1 Performance Indicator Verification Data Collection Sheets for Emergency AC Power 1st to 4th Quarters 2001 Data Collection Sheets for Residual Heat 1st to 4th Quarters Removal 2001 CR 20017628 EDG Was Declared Operable Without Clearing a Conditional Release of Air Line Lubricators

CR 20017688 Min/Max Voltage Slightly Higher than Specified Range in 12 EDG Auxiliary Sys. Test 1052-4 WO 0001534 MO-2014 PM 4900-1 WO 0003809 Pre-op Test of LPCI 5 Min Timer Bypass WO 0105741 Install LPCI Div. 2 5 Min timer Bypass WO 0004079 MO-2029 PM 4900-1 4OA3 Event Follow-up NRC Inspection Reports:

50-263/01-09 - Routine Resident Inspections From October 1st December 4, 2001 through November 14th, 2001 50-263/01-16 - Problem Identification and Resolution December 14, 2001 Inspection Procedures and Forms:

0137 - Master Local Leak Rate Test Revision 22 0137-07A - Reactor Steam Supply Valves Leak Rate Revision 15 Testing CR 20017090 LLRT Failure of MO-2397 CR 20017092 LLRT Failure of CV-2790 CR 20017091 LLRT Failure of MO-2398 CR 20017070 LLRT Failure of MSIVs AO-2-86C and AO-2-86D CR 20016420 Individual Bumped Instrument Rack Resulted in PCIS Group 1 Isolation and Reactor Scram 21

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