IR 05000263/2002002
| ML020590139 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| 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) | |
Text
February 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.
- High pressure coolant injection (HPCI) and other emergency core cooling systems (ECCS) with reactor core isolation cooling (RCIC) out-of-service
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:
Core Spray System
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.
- Replacement of Control Rod Drive (CRD) Hydraulic Control Unit (HCU)
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 Turbine EPR (Electric Pressure Regulator)/MPR (Mechanical Pressure Regulator) Repairs
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
Low Pressure Core Injection
Non-Cited Violation
Nuclear Management Company
Nuclear Management and Resources Council
Operator Workaround
OWI
Operations Work Instruction
Primary Containment Isolation System
psig
Pounds Per Square Inch Gauge
Pull-To-Lock
Reactor Core Isolation Cooling
Residual Heat Removal Service Water
scf
Standard Cubic Feet
SWI
Scheduling Work Instruction
TS
Technical Specification
Unresolved Item
Updated Safety Analysis Report
Vac
Volts Alternating Current
Vdc
Volts Direct Current
Work Order
LIST OF DOCUMENTS REVIEWED
1R04
Equipment Alignment
B.09.08
B.03.02
B.08.01.02
B.02.03
B.09.06
Operations Manual:
- HPCI System
- RCIC System
- 4160 Vac System
M-123
M-123-1
M-124
M-125
M-126
M-110-1
M-112
M-811-1
M-133-1
NF-36298-1
NF-36298-2
Drawings:
- HPCI System (Steam Side)
- HPCI Hydraulic Control and Lubrication
- HPCI System (Water Side)
- RCIC System (Steam Side)
- RCIC System (Water Side)
- Service Water System
- RHR and Emergency Service Water Systems
- Service Water and Make-Up Intake Structure
- Diesel Oil System
- Electrical Load Flow
- DC Electrical Load Distribution
Revision AF
Revision B
Revision Y
Revision AK
Revision Y
Revision BL
Revision BF
Revision CD
Revision AD
Revision M
Revision A
1R05
Fire Protection
NX-16991
Monticello Updated Fire Hazards Analysis
Technical Manual
A.3-07A
A.3-07B
A.3-07C
A.3-24
Plant Fire Strategies:
- 125Vdc - Division I Battery Room
- 250Vdc - Division I Battery Room
- 125Vdc - Division II Battery Room
- Diesel Fire Pump Room
Revision 3
Revision 5
Revision 3
Revision 4
4AWI-08.01.01
4AWI-08.01.02
0271
0274
0275-1
0275-2
Procedures and Administrative Work
Instructions:
- Fire Prevention Practices
- Combustion Source Use Permit
- Fire Hose Station and Yard Hydrant Hose
House Equipment Inspection
- Fire Hose Hydrostatic Test Interior Hose
Stations
- Fire Barrier Penetration Seal Visual Inspection
- Fire Barrier Wall, Damper, and Floor Inspection
Revision 17
Revision 6
Revision 27
Revision 19
Revision 9
Revision 16
QUAD-5-80-009
Specifications for Installation of Electrical and
Mechanical Penetration Seals at the Monticello
Nuclear Generating Plant by Quadrex
Corporation
Revision 7
A.3
B.08.05
Operations Manual:
- Fire Fighting Procedures
- Fire Protection
1R11
Licensed Operator Requalification Program
RQSS-28
Licensed Operator Annual Examination Scenario
Revision 9
1R12
Maintenance Rule Implementation
93-01
93-01, Section 11
NUMARC [Nuclear Management and Resources
Council]:
- Nuclear Energy Institute Industry Guideline for
Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants
- Assessment of Risk Resulting from the
Performance of Maintenance Activities
Revision 2
February 22, 2000
1.160
1.182
Regulatory Guides:
- Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants
- Assessing and Managing Risk Before
Maintenance Activities at Nuclear Power Plants
Revision 2
May 2000
05.02.01
Engineering Work Instruction, Monticello
Maintenance Rule Program Document
Revision 5
Maintenance Rule Periodic Assessment Report
2nd Quarter - 2001
B.3.1
B.5.12
B.9.15
Operations Manual:
- Core Spray System
- Area Radiation Monitors
- Non-Essential Diesel Generator
B.3.1
B.5.12
B.9.15
Maintenance Rule Program System Basis
Document:
- Core Spray System
- Area Radiation Monitors
- Non-Essential Diesel Generator
Revision 1
Revision 2
Revision 2
B.3.1
Design Basis Document:
- Core Spray System
Revision 2
Section 6.2.2
Sect. 14.7.2.3.1.3
Section 7.5.3
Section 8.4
USAR:
- Core Spray System
- Core Spray System
- Area Radiation Monitoring System
- Plant Standby Diesel Generator Systems
Revision 18
M-122
Drawings:
- 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
4AWI-04.01.01
SWI-14.01
4010PM
0074
2188
0081
Procedures:
- General Plant Operating Activities
- Risk Management of On-line Maintenance
- HCU Water Accumulator Replacement
- Control Rod Drive Exercise
- Scram Accumulator Nitrogen Charging
- Control Rod Drive Scram Insertion Time Test
Revision 28
Revision 0
Revision 9
Revision 29
Revision 13
Revision 35
Replace HCU 22-07
RWP 146
West CRD Accumulators
Revision 2
Disassemble, Inspect, Replace Components On
Back Flush the MPR Pressure Sensing Line
Adjust Needle Valve On MPR
Test MPR Dash Pot Bellows Relief Valves
Clean Inlet Screen / Replace EPR Moog valve
Inspect Turbine Front Standard Internal
Reset Primary Valve Limit Stop
Collect MPR Response Data
Install Temporary Test Equipment On MPR
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
C.3
C.4-A
C.4-B.5.7.A
C.4-B.6.5.A
C.1
Operations Manual:
- Shutdown Procedure
- Reactor Scram
- Loss of Reactor Water Level Control
- Reactor Feedwater Pump Trip
- Startup Procedure
Revision 28
Revision 19
Revision 5
Revision 5
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
B.7
Operations Manual:
- Liquid Radwaste
Section 3/4.6
Section 3/4.9
Technical Specifications and Bases:
- Primary System Boundary, Coolant Leakage
- 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
Section 8
USAR:
- Plant Electrical Systems
Revision 18
1R16
Operator Workarounds
Quarterly Operator Work-Around Review and
Assessment
December 17, 2001
B.1.2
B.1.3
Operations Manual:
- Control Rod Drives
- CRD Hydraulic System
Section 3.5.3
USAR:
- Control Rod Drive System
Revision 18
B.1.2
B.1.3
Design Basis Documents:
- Control Rod Drives
- CRD Hydraulic System
Revision B
Revision B
CR 19982736
CRD Operability Condition when the Associated
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
Upgrade Level Controllers FW Htrs 12, 13, 14 &
FW Heater LC Upgrade Pre-Op Test
1R19
Post-Maintenance Testing
M-125
NX-8435-150-2
Drawings and Prints:
- RCIC (Steam Side)
- Turbine Control Diagram
Revision AM
Revision G
0137-07A
1040-01
3069
4900-01PM
EMP 01.01
Procedures and Forms:
- Reactor Steam Supply Valves Leak Rate
Testing
- Turbine Generator
- Post-Maintenance Activities Control Cover
Sheet
- PM For Limitorque Motor Operated Valves
- Motor Operated Valve Testing Using VOTES
Revision 15
Revision 42
Revision 9
Revision 15
Revision 7
SCR-02-0055
10 CFR 50.59 Screening for Installing
Instrumentation to Monitor MPR Performance
Revision 0
Tagout 02-00328
LLRT Isolation for RCIC Valve MO-2076
Version 1
Test MPR Dash Pot Bellows Relief Valves
Adjust Needle Valve on MPR Control
MO-2076 Repair Valve Actuator
Install Temporary Test Equipment on MPR
Obtain Oil Sample of EPR Oil Reservoir
Set MPR and Associated Computer Alarm Point
TRB178
LLRT Isolation for RCIC Valve MO-2076
MO-2076 Went to Dual Indication When Stroked
Open
1R20
Outage Activities
C.1
C.2
C.3
C.4-A
Operations Manual:
- Startup Procedure
- Power Operation
- Shutdown Procedure
- Reactor Scram
Revision 33
Revision 12
Revision 28
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
4851-12-PM
0045
0083
Procedures:
- ABB K-1600s and K-3000s 450 Volt Breaker
Maintenance
- Rod Block Monitor Functional Test and
Calibration
- Reactivity Anomaly Check
Revision 6
Revision 28
Revision 12
CR 20018369
LCB-024 Breaker Closure Delayed During Start
of #11 Service Water Pump
Perform 4851-12-OM on LCB-024 Located in
523-305
Repair Failure of LCB-024 to Close Promptly
Section 3/4.2
Section 3/4.3E
Technical Specifications and Bases:
- Protective Instrumentation
- 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
Plant
Revision 19
Monticello Nuclear Generating Plant Emergency
Plan
Revision 20
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
Removal
1st to 4th Quarters
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
MO-2014 PM 4900-1
Pre-op Test of LPCI 5 Min Timer Bypass
Install LPCI Div. 2 5 Min timer Bypass
MO-2029 PM 4900-1
4OA3 Event Follow-up
50-263/01-09
50-263/01-16
NRC Inspection Reports:
- Routine Resident Inspections From October 1st
through November 14th, 2001
- Problem Identification and Resolution
Inspection
December 4, 2001
December 14, 2001
0137
0137-07A
Procedures and Forms:
- Master Local Leak Rate Test
- Reactor Steam Supply Valves Leak Rate
Testing
Revision 22
Revision 15
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