IR 05000263/2001010

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IR 05000263/2001-010(DRP), Nuclear Management Company, LLC, Monticello Nuclear Generating Plant, Inspection on 11/15-12/31/2001 Re Post-Maintenance Testing. Two Noncited Violations Noted
ML020160509
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 01/16/2002
From: Burgess B
Division Reactor Projects III
To: Forbes J
Nuclear Management Co
References
IR-01-010
Download: ML020160509 (31)
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Text

ary 16, 2002

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT NRC INSPECTION REPORT 50-263/01-10(DRP)

Dear Mr. Forbes:

On December 29, 2001, the NRC completed an inspection at your Monticello Nuclear Generating Plant. The results of this inspection were discussed on January 3, 2002, with you and 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, radiation safety, inservice inspection, 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 inspectors identified one issue of very low safety significance (Green) which was determined to involve a violation of NRC requirements.

However, because of its very low safety significance and because your staff has entered the issue into your corrective action program, the NRC is treating the issue as a Non-Cited Violation, in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you deny the 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 20555-0001; 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. 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. DPR22

Enclosure:

Inspection Report 50-263/01-10(DRP)

REGION III==

Docket No: 50-263 License No: DPR-22 Report No: 50-263/01-10(DRP)

Licensee: Nuclear Management Company, LLC Facility: Monticello Nuclear Generating Plant Location: 2807 West Highway 75 Monticello, MN 55362 Dates: November 15 through December 29, 2001 Inspectors: S. Burton, Senior Resident Inspector D. Kimble, Resident Inspector M. Mitchell, Regional Health Physics Inspector D. Jones, Regional Engineering Inspector Approved by: Bruce L. Burgess, Chief Branch 2 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000263/01-10(DRP), on 11/15-12/31/2001; Nuclear Management Company, LLC; Monticello Nuclear Generating Plant; Post-Maintenance Testing.

The inspection was conducted by resident and regional inspectors. The report covers a 61/2-week period. The inspection identified one Green finding. The significance of all of the findings are 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 applicable violation. 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 Cornerstones: Barrier Integrity, Mitigating Systems, and Initiating Events

  • Green. The inspectors reviewed the post modification test for the Division II Low Pressure Core Injection 5 Minute Timer Bypass Modification. During the testing evolution, an error associated with a jumper bypass in the test procedure resulted in the loss of shutdown cooling to the reactor vessel. The failure of the licensee to provide an appropriate procedure to test the modification constitutes a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V. The finding was of very low safety significance because of the low decay heat load present in the reactor and the licensee's ability to manually recover shutdown cooling in a short period of time (Section 1R19.2).

B. Licensee Identified Violations None.

Report Details Summary of Plant Status The Unit began the inspection period shutdown for Refuel Outage No. 20. The plant was taken critical on December 13, 2001, with main generator synchronization to the grid occurring on December 15, 2001. Full power was reached on December 18, 2001. A power reduction to approximately 75 percent was performed on December 21, 2001, for rod pattern adjustment, with return to full power operation on December 22, 2001. The Unit remained at or near full power for the remainder of the inspection period.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1R01 Adverse Weather (71111.01)

a. Inspection Scope The inspectors conducted a review of the licensee's preparations for winter conditions to verify that the plant's design features and implementation of procedures were sufficient to protect mitigating systems from the effects of adverse weather. Documentation for selected risk-significant systems was reviewed to ensure that these systems would remain functional when challenged by inclement weather. Cold weather protection, such as heat tracing, was verified to be in operation where applicable.

b. Findings No findings of significance were identified.

1R04 Equipment Alignment (71111.04)

a. Inspection Scope The inspectors performed a partial walkdown of various Division I equipment during the Division II refuel outage work window to verify operability and proper equipment lineup while the counterpart train was disabled due to planned maintenance. These systems were selected due to the increase in core damage frequency, which resulted from rendering other risk significant equipment out-of-service for maintenance. 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 A.3-03-C, Vessel Instrument Rack Area - Elevation 962'
  • Fire Zone A.3-03-E, Contaminated Records Area
  • Fire Zone A.3-04-A, Reactor Building 3rd Floor South
  • Fire Zone A.3-04-B, Reactor Building Closed Cooling Water Heat-Exchanger Area 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.

1R08 Inservice Inspection (ISI) Activities (71111.08)

a. Inspection Scope The inspectors evaluated the implementation of the licensee's ISI program for monitoring degradation of the reactor coolant system boundary and the risk significant piping system boundaries. Specifically, the inspectors verified through observations that in-process ultrasonic and magnetic particle inspections of residual heat removal (RHR)

discharge piping weld ISI-13142-18-B was conducted in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements. The inspectors also reviewed ISI procedures and personnel and equipment certifications.

The inspectors reviewed the NIS-2 forms for Code repairs performed during the outage (Refueling Outage No. 20) and confirmed that ASME Code requirements were met. In addition, the inspectors reviewed reports concerning ISI issues to verify that an appropriate threshold for identifying issues had been established. The inspectors also evaluated the effectiveness of the corrective actions for identified issues.

b. Findings No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11)

a. Inspection Scope The inspectors reviewed the evaluation of simulator annual examination scenarios and reviewed licensed operator performance in mitigating the consequences of events. The scenario included transient and emergency actions and resulted in execution of multiple emergency operating procedures. Areas observed by the inspectors included:

sequence of actions, prioritization of activities, procedural adequacy and implementation, and emergency plan requirements.

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:

  • Alternate Shutdown System
  • Off-Gas Recombiner System
  • Reactor Pressure Relief System 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.

  • Outage Planning and Emergent Work Review 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.

1R16 Operator Workarounds (OWA) (71111.16)

a. Inspection Scope The inspectors reviewed the following operator workarounds. The inspectors assessed each workarounds potential to impact: system function; the operators ability to respond to accident conditions and implement emergency operating procedures; and equipment operability.

  • Operator Workaround 00-073, Loss of CRD [Control Rod Drive] Pump During LOCA [Loss of Coolant Accident] With Fuel Failure Creates a Potential Leakage Pathway
  • Operator Workaround 01-114, Normal Operation of EDG-ESW [Emergency Diesel Generator-Emergency Service Water] Pumps Requires Closure of SW-239-1 and SW-239-2 to Prevent Dead-Heading the Pumps b. Findings No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

.1 Miscellaneous Post-Maintenance Testing Activities 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.

  • Control Rod Drive 10-39 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, Technical Specification 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, Technical Specifications, and Updated Safety Analysis Report (USAR)

design requirements.

b. Findings No findings of significance were identified.

.2 Division II Low Pressure Coolant Injection (LPCI) 5 Minute Timer Bypass Switch a. Inspection Scope The inspectors reviewed the post modification tests associated with the Division II LPCI 5 minute timer bypass switch installation modification. This activity was selected based upon the LPCI system's overall contribution to risk. The inspectors verified by witnessing selected testing activities and by reviewing the test data that the testing activities were adequate for the bypass switch installation. The inspectors' reviews included, but were not limited to, integration of the testing activities, applicability of the acceptance criteria, test equipment calibration and control, procedural use and compliance, control of temporary jumpers required for test performance, documentation of test data, Technical Specification applicability, system restoration, and evaluation of test data. Also, the inspectors verified that the testing activities adequately ensured that the equipment met the licensing basis, Technical Specifications, and USAR design requirements.

b. Findings Inspectors identified one finding of very low safety significance (Green) and an associated NCV.

On November 26, 2001, the licensee was conducting a post modification test for the installed bypass switches for the 5 minute (LPCI) initiation timer on Division II. This modification was an operational enhancement, intended to improve emergency operating procedure performance, and was being conducted during a scheduled Division II equipment work window. At step 8.5.B of the procedure, an inadvertent closure of the Division I LPCI injection valve, MO-2012, occurred and shutdown cooling to the reactor vessel was lost.

The licensees investigation identified that step 8.5.B of the test procedure required installing a jumper in Panel C-33 on Relay 10A-K43B from stud 4 to stud 3. A review of drawing NX-7905-46-9, "Residual Heat Removal (RHR) System Schematic Diagram,"

indicated that the proper studs to jumper were studs 5 and 6. Jumpering studs 4 and 3 on Relay 10A-K43B, as required by the procedure, provided a closure signal to MO-2012. The licensee completed a detailed review of the test procedure, corrected the procedure, and completed the testing satisfactorily.

The inspectors evaluated this finding using NRC Inspection Manual Chapter IMC 0610*,

Appendix B, "Thresholds for Documentation," and determined it to be more than minor in that it had an actual and credible impact on safety. Specifically, the inspectors determined that the loss of RHR Division I injection capability with Division II out-of-service for planned work represented some credible risk to the Unit. Further, the inspectors determined that the finding impacted the mitigating systems cornerstone of nuclear safety in that it affected the operability, availability, reliability, and/or function of a train in a mitigating system, in this case, LPCI. As a result, the inspectors assessed the finding using the SDP for shutdown operations in IMC 0609, Appendix G. During the screening process, the inspectors determined that because the resulting reactor coolant system heatup was small with respect to the available margin to boiling and the licensee had the ability to manually open the Division I injection valve and restore shutdown cooling in a short period of time, the finding was of very low safety significance (Green).

The licensee has entered this issue into their corrective action program as CR 20017538.

Appendix B to 10 CFR 50, Criterion V requires, in part, that activities affecting quality shall be prescribed by documented instructions appropriate to the circumstances.

Contrary to this requirement, the post modification test procedure for the Division II LPCI 5 minute timer bypass switch modification was not appropriate in that it failed to provide the correct terminal studs to jumper. This violation is being treated as a NCV consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-263/01-10-01(DRP)).

1R20 Outage Activities (71111.20)

a. Inspection Scope The inspectors continued evaluation of outage activities for Refueling Outage No. 20, which began on November 3, 2001, and ended on December 13, 2001. 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 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, 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.

  • No. 12 EDG Functional Test From the Alternate Shutdown Panel
  • Emergency Core Cooling System (ECCS) Automatic Initiation Test 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, Technical Specification applicability, impact of testing relative to performance indicator reporting, and evaluation of test data.

Additionally, the inspectors monitored the reactor vessel hydrostatic test and reviewed the results. Particular emphasis was placed upon control rod drive hydraulic (CRDH)

system insert and withdraw lines. Emphasis was placed on the CRDH system due to transgranular stress cracking corrosion that was observed on the CRDH system withdraw lines during the 1998 and 2000 outages. Because multiple condition reports indicated that the contributor was polyvinyl-chloride label tape, which remains installed on piping inside the drywell, the inspectors determined that increased review was warranted. Actions that resulted from inspector observations, including evaluation of the condition, corrective actions, and proposed inspection enhancements for the 2003 refueling outage, were also reviewed.

b. Findings No findings of significance were identified.

2. RADIATION SAFETY Cornerstone: Occupational Radiation Safety 2OS1 Access Control to Radiologically Significant Areas (71121.01)

.1 Plant Walkdowns and Radiation Work Permit Reviews a. Inspection Scope The inspector conducted walkdowns and radiological surveys of radiologically significant areas (radiation and high radiation areas) to verify the adequacy of the licensee's radiological controls (surveys, postings, barricades). Specifically, the inspector walked down radiologically significant areas located in the reactor building, including the drywell, reactor core isolation cooling system (RCIC) room during testing, and turbine building to determine whether prescribed radiation work permits (RWPs), procedure and engineering controls were in place, and whether licensee surveys and postings were complete and accurate. The inspector also reviewed RWPs used to access these areas to verify that work instructions and controls had been adequately specified and that electronic pocket dosimeter set points were in conformity with survey indications.

b. Findings No findings of significance were identified.

.2 Job-In-Progress Reviews a. Inspection Scope The inspector observed the following high exposure or high radiation area work activities performed during the ongoing refueling outage and evaluated the licensee's use of radiological controls:

  • Miscellaneous Valve Repair in Reactor Water Clean-Up Room
  • CV-2790 Valve Work
  • Drywell General Entry
  • Radiation Protection (RP) Coverage in the Drywell
  • Helper/Laborer Entry into the Drywell

The inspector reviewed all radiological job requirements for each activity and observed job performance with respect to those requirements. The inspector reviewed required surveys, including system breach radiation, contamination, and airborne surveys; radiation protection job coverage; and contamination controls to verify that appropriate radiological controls were utilized. The inspector also reviewed surveys and applicable postings and barricades to verify their accuracy. The inspector observed radiation protection technician and worker performance at work sites to determine if the technicians and workers were aware of the significance of the radiological conditions in their workplace, the RWP controls/limits, and that their performance was adequate, given the level of radiological hazards present and the level of their training.

b. Findings No findings of significance were identified.

2OS2 As-Low-As-Is-Reasonably-Achievable (ALARA) Planning and Controls (71121.02)

.1 Exposure Histories a. Inspection Scope The inspector reviewed the station's collective exposure histories for 1998 to the present. The review included collective exposures during the year 2001 forced outages and the 2001 refueling outage. The inspector performed the reviews to evaluate the licensee's ALARA program's strengths and weaknesses. The inspector also reviewed the station's three-year rolling average exposure information and compared it with national boiling water reactor industry data.

b. Findings No findings of significance were identified.

.2 Job Site Inspections and ALARA Control a. Inspection Scope The inspector selected the following high exposure or high radiation area activities performed during the 2001 refueling outage and evaluated the licensee's use of ALARA controls:

  • Miscellaneous Valve Repair in Reactor Water Clean-Up Room
  • Replacement of Flow Elements in RHR Room
  • Drywell General Entry
  • Nozzle ISI and Insulation Work
  • RP Coverage in the Drywell
  • Helper/Laborer Entry into the Drywell

The inspector reviewed ALARA plans for each activity and observed work activities associated with the CV-2790 valve, drywell general entry, RP coverage in the drywell, and helper/laborer entry into the drywell. The inspector evaluated the licensee's use of engineering controls to achieve dose reductions. The inspector also determined if workers were utilizing the low dose waiting areas for each activity and whether the first-line supervisor for each job ensured that the jobs were conducted in a dose efficient manner. The inspector also reviewed individual exposures of selected work groups to determine if any significant exposure variations existed among workers.

b. Findings No findings of significance were identified.

.3 Source Term Reduction and Control a. Inspection Scope The inspector reviewed the status of the licensee's source term reduction program. The inspector did the review to determine what results had been achieved and what effects, if any, those results were having on exposures during the refueling outage.

b. Findings No findings of significance were identified.

.4 Radiological Work Planning a. Inspection Scope The inspector selected the following refueling outage job activities that were expected to exceed five person-rem to assess the adequacy of the radiological controls and work planning:

  • Drywell General Entry
  • ALARA efforts in the Drywell
  • Helper/Laborer Drywell Enter
  • General ISI work in the Drywell For each job activity, the inspector reviewed ALARA evaluations including initial reviews, in-progress reviews, and associated dose mitigation techniques and evaluated the licensee's exposure estimates and performance. The inspector also assessed the integration of ALARA requirements into work packages to evaluate the licensee's communication of radiological work controls.

b. Findings No findings of significance were identified.

.5 Verification of Exposure Goals and Exposure Tracking System a. Inspection Scope The inspector reviewed the methodology and assumptions used for the 2001 refueling outage exposure estimates and exposure goals and compared job dose rate and man-hour estimates for accuracy. The inspector examined job dose history files and dose reductions anticipated through lessons learned to verify that the licensee appropriately forecasted outage doses. The inspector also reviewed the licensee's exposure tracking system to determine if the level of exposure tracking detail, exposure report timeliness, and exposure report distribution was sufficient to support control of collective exposures.

b. Findings No findings of significance were identified.

.6 Identification and Resolution of Problems (71121.01 and 71121.02)

a. Inspection Scope The inspector evaluated the effectiveness of the self-assessment process to identify, characterize, and prioritize problems. The inspector reviewed the 2001 refueling outage related ALARA and access control issues to determine if they were adequately addressed. The inspector also reviewed chemistry and radiation protection effectiveness reports for the year 2001 that, in part, assess the condition reports for adequacy of the licensee's ability to identify problems and make effective corrective actions.

b. Findings No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation (71121.03)

.1 Source Tests and Calibration of Radiological Instrumentation a. Inspection Scope The inspector reviewed the most recent calibration records for radiological instruments associated with transient high and very high radiation areas (area radiation monitors (ARMs)). The inspector also reviewed calibration records for instruments used for providing surveys of high radiation work and/or for air monitoring for jobs with the potential for workers to receive greater than 100 millirem committed effective dose equivalent (CEDE). The inspector reviewed these records to verify that radiological

instrumentation had been calibrated in accordance with procedures and that alarm set-points (if applicable) were properly set. In particular, the inspector reviewed selected ARMs in the spent fuel pool, primary coolant sampling station, off-gas storage building and radioactive waste control room to verify that they had been appropriately calibrated and function and operation tested in calendar year 2001. The inspector reviewed the calibration procedures and calendar year 2001 calibration records to verify that selected portable radiation survey instruments had been properly calibrated consistent with the licensee's procedures. The inspector also reviewed the calibration procedures and calendar year 2001 calibration records for the whole body counter to verify that it had been properly calibrated. The inspector observed the calibration of selected area monitoring instruments to verify that the instruments were calibrated in compliance with the appropriate procedures.

b. Findings No findings of significance were identified.

.2 Self-Contained Breathing Apparatus (SCBA) Program a. Inspection Scope The inspector reviewed R.05.07 "SCBA Inspection and Functional Test", to verify the adequacy of the program to provide SCBA for unknown or emerging conditions. The inspector walked down the available SCBA equipment and filling stations, reviewed the status and surveillance records of SCBA staged for use in the plant, assessed the licensee's capability for refilling and transporting SCBA bottles for use in the control room and support locations in the plant, and reviewed calendar year 2001 training and qualification records of selected individuals to verify compliance with Subpart H of 10 CFR Part 20 and with station procedures.

b. Findings No findings of significance were identified.

.3 Identification and Resolution of Problems a. Inspection Scope The inspector reviewed quarterly 2001 radiation protection department self-assessments of the occupational radiation protection program to evaluate the effectiveness of the self-assessment process to identify, characterize, and prioritize problems and to verify that previous radiological instrumentation and SCBA-related issues were adequately addressed. The inspector also reviewed selected year 2001 condition reports that addressed radiation instrument deficiencies. The review was used to determine if any significant radiological incidents involving radiation instrument deficiencies had occurred during the year 2001. The review was also conducted to verify that the licensee had effectively implemented the corrective action program.

b. Findings No findings of significance were identified.

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

Cornerstone: Occupational Radiation Safety a. Inspection Scope The inspector reviewed the licensee's assessment of its performance indicator (PI) for occupational radiation safety to determine if indicator-related data was adequately assessed and reported. Since no reportable elements were identified by the licensee for the last four quarters, the inspector compared the licensee's data with fourth quarter of 2000 and the first three quarters of 2001 condition reports to verify that there were no occurrences concerning the occupational radiation safety cornerstone.

b. Findings No findings of significance were identified.

4OA6 Meeting Exit Meeting The inspectors presented the inspection results to M and other members of licensee management on January 3, 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.

In addition to the January 3 exit, interim exits were conducted on November 28, November 30, and December 21 with M and Mr. Jepsen to discuss ISI and Radiological Protection areas.

KEY POINTS OF CONTACT Licensee G. Bregg, Manager, Quality Services R. Deopere, Inservice Inspection Supervisor D. Fadel, Director of Engineering J. Forbes, Site Vice-President R. Frederickson, Superintendent Material Inspection and Repair J. Grubb, General Superintendent, Engineering 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, Operations NRC None.

ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-263/01-10-01 NCV Inadequate Test Procedure for LPCI 5 Minute Timer Bypass Switch Modification (Section 1R19.2)

Closed 50-263/01-10-01 NCV Inadequate Test Procedure for LPCI 5 Minute Timer Bypass Switch Modification (Section 1R19.2)

Discussed None.

LIST OF ACRONYMS USED ALARA As-Low-As-Is-Reasonably-Achievable ARM Area Radiation Monitor ASME American Society of Mechanical Engineers AWI Administrative Work Instruction CAM Continuous Air Monitor CEDE Committed Effective Dose Equivalent CFR Code of Federal Requirements CR Condition Report CRD Control Rod Drive CRDH Control Rod Drive Hydraulic DRP Division of Reactor Projects DW Drywell ECCS Emergency Core Cooling System ED Electronic Dosimeter EDG Emergency Diesel Generator ESW Emergency Service Water HPCI High Pressure Core Injection HPGE High Purity Germanium IMC Inspection Manual Chapter ISI Inservice Inspection LER Licensee Event Report LOCA Loss of Coolant Accident LPCI Low Pressure Core Injection MOV Motor-Operated Valve mrem Millirem NCV Non-Cited Violation NIOSH National Institute of Safety & Health NMC Nuclear Management Company NRC Nuclear Regulatory Commission NUMARC Nuclear Management and Resources Council OWA Operator Workaround PI Performance Indicator PMT Post-Maintenance Testing RBCCW Reactor Building Closed Cooling Water RCA Radiologically Controlled Area RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RHRSW Residual Heat Removal Service Water RP Radiation Protection RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup RWP Radiation Work Permit SBLC Standby Liquid Control SBO Station Blackout SCBA Self-Contained Breathing Apparatus SDP Significance Determination Process SRI Safety Review Item

SRV Safety Relief Valve TS Technical Specification USAR Updated Safety Analysis Report WBC Whole Body Count WO Work Order

LIST OF DOCUMENTS REVIEWED 1R01 Adverse Weather USAR: Revision 18 Section 5.3.4 - Reactor Building Heating and Ventilating Systems Section 10.3.2 - Plant Heating, Ventilating, and Air Conditioning Systems M-149 Chilled Water Piping System Drawing Revision N 1151 Winter Checklist Revision 40 1R04 Equipment Alignment Section B.8.1.3 Design Basis Document for RHR Service Water Revision 2 Operations Manual:

Section B.3.1 - Core Spray System Section B.3.4 - Residual Heat Removal System Section B.8.1.3 - RHR Service Water System M-120 [Division 2] Residual Heat Removal System Revision BH M-121 [Division 1] Residual Heat Removal System Revision BK M-112 RHR Service Water and Emergency Service Revision BF Water System M-811 Service Water and Make-up Water Intake Revision C Structure Technical Specifications and Bases:

TS 3/4.5 - Core and Containment Spray/Cooling Systems USAR: Revision 18 Section 6.2.3 - Residual Heat Removal System Section 10.4.2 - Residual Heat Removal Service Water System 4AWI-08.15.01 Risk Management For Outage and On-Line Revision 0 Activities 1R05 Fire Protection NX-16991 Monticello Updated Fire Hazards Analysis Monticello Fire Strategies:

A.3-03-C - Vessel Instrument Rack Area - Elev 962' Revision 4 A.3-03-E - Contaminated Records Area Revision 3*

A.3-04-A - Reactor Building 3rd Floor South Revision 3*

A.3-04-B - RBCCW Hx Area Revision 2*

Procedures and Administrative Work Instructions (AWIs):

4AWI-08.01.01 - Fire Prevention Practices Revision 17 4AWI-08.01.02 - Combustion Source Use Permit Revision 6 QUAD-5-80-009 Quadrex Corporation Report, Specifications for Revision 7 Installation of Electrical and Mechanical Penetration Seals at the Monticello Nuclear Generating Plant 0275-2 Fire Barrier Wall, Damper, and Floor Inspection Revision 16 0275-1 Fire Barrier Penetration Seal Visual Inspection Revision 9 1R08 Inservice Inspection Inservice Inspection (ISI) Examination Summary May 30, 2000 Report-Refueling Outage No. 20 CR 20000209 In-Vessel Inspections Found Indications on Jet Pump Brace CR 20000318 Minor Indications Found on CRD Lines During Eddy Current Examination ISI-MT-1 Dry Powder Magnetic Particle Examination October 18, 2001 ISI-NDE-0 Equipment, Personnel and Material Reporting January 24, 2001 ISI-UT-1A Ultrasonic Examination of Ferritic Piping Welds November 2, 2001 to Appendix VIII ISI-VT-4.0 Visual Examination of Monticello Reactor November 8, 2001 Vessel Interior PDI-ISI-254 Remote Inservice Examination of Reactor October 24, 2001 Vessel Shell Welds 1R11 Licensed Operator Requalification Program RQ-SS-03 Licensed Operator Annual Examination Scenario Revision 20 RQ-SS-15 Licensed Operator Annual Examination Scenario Revision 7 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 Monticello Maintenance Rule Periodic 2nd Quarter - 2001 Assessment Report Operations Manual:

Section B.3.3 - Reactor Pressure Relief System Section B.3.5 - Standby Liquid Control System Section B.7.2.1 - Off Gas Recombiner System Section B.8.1.3 - Residual Heat Removal Service Water Maintenance Rule Program System Basis Document:

Section B.3.3 - Reactor Pressure Relief System Revision 2 Section B.3.5 - Standby Liquid Control System Revision 1 Section B.7.2.1 - Off Gas Recombiner System Revision 3 Section B.8.1.3 - Residual Heat Removal Service Water Revision 1 USAR: Revision 18 Section 6.6 - Standby Liquid Control System Section 6.2 - Residual Heat Removal Technical Specifications and Bases:

Section 3/4.4 - Standby Liquid Control System SBLC System Performance Data Collection November 23, 2001 Worksheet Residual Heat Removal Service Water System December 17, 2001 Performance Data Collection Worksheet Alternate Shutdown System Performance Data December 18, 2001 Collection Worksheet CR 20016731 "B" Off Gas Recombiner Tripped at 1158 on 11/3/01 While Placing the Mechanical Vacuum Pump in Service Per C.3 Shutdown Procedure

CR 20017839 Leakage For "D" SRV Accumulator Was Greater Than Allowed CR 20015126 SBLC TIC-11-48 Switch Failed to Trip Storage Tank Heater CR 20011088 LI-11-66, SBLC Tank Level Indicator Indicates Full Upscale (Control Room), Local Indicator Indicates Normal CR 20014640 Received Second Unexpected Alarm 3-A-17 Auto Blowdown Value Bellows Leaking With White Light on C-03 For "H" SRV CR 20000057 "B" Recombiner Train Tripped When Mechanical Vacuum Pump Was Started During Plant Shutdown CR 20018006 CV-1728 Unable to Be Positioned From the Control Room Due to Interference With the Manual Handwheel CR 20012078 Small Amount of Gel (Emulsified Oil) Was Identified in Number 13 RHRSW Pump Lower Bearing Oil Sight Glass CR 20013383 Failure to Adequately Investigate the Extent of Condition on #13 RHRSW Pump in a Timely Manner CR 20003339 Both Recombiner Trains Will Not Transfer From Warmup to Standby EWI 05.02.01 Monticello Maintenance Rule Program Document Revision 5 M-127 P&ID: Standby Liquid Control System Revision V WO 0000691 Blown Fuses on SBLC System WO 0105529 Replace Relay WO 0106002 Replace RV-11-39B 1R13 Maintenance Risk Assessments and Emergent Work Control Procedures:

4AWI-04.01.01 - General Plant Operating Activities Revision 30 SWI-14.01 - Risk Management of On-line Maintenance Revision 0 Technical Specification and Bases:

TS 3/4.7 - Containment Systems01-100 Jumper Bypass: AO-2-86A & -86B Downstream Piping Plugs 1R16 Operator Workarounds B.8.01.2-05 Operations Manual - Emergency Diesel Revision 11 Generator C.4-B.01.03.A Abnormal Procedure -Loss of CRD pump Flow Revision 6 CR 20013155 Loss of CRD Pump During LOCA With Fuel Failure Creates a Potential Leakage Pathway CR 20014904 Normal Operation of EDG-ESW Pumps Requires Closure of SW-239-1 and SW-239-2 to Prevent Dead-Heading the Pumps OWA 01-114 Operator Workaround - Normal Operation of EDG-ESW Pumps Requires Closure of SW-239-1 and SW-239-2 to Prevent Dead-Heading the Pumps OWA 00-073 Operator Workaround - Loss of CRD Pump During LOCA With Fuel Failure Creates a Potential Leakage Pathway 1R19 Post-Maintenance Testing Operations Manual:

B.08.01.02 - EDG Emergency Service Water System B.08.11 - Diesel Oil System B.09.08 - Emergency Diesel Generators B.03.04 - Residual Heat Removal System Technical Specification and Bases:

Section 3/4.9 - Auxiliary Electrical Systems Administrative Work Instructions:

4AWI-04.05.09 - Foreign Material Exclusion/Cleanliness Control Revision 7 4AWI-05.05.02 - Fuel Integrity Monitoring and Failed Fuel Action Revision 2 Plan 0187-01 No. 11 EDG and No. 11 ESW Pump System Revision 37 Tests 2020 Consumable Items Log Revision 26 3661 Project Request Form: E-Number O1TOZ5, Modify B4300 Control Circuit CA-01-041 Calculation Cover Sheet: B4300 Control Cable Design

CR 20017292 PMT Failure. MOV Failed to Electrically Stroke.

CR 20017522 MO-2012 Intermittent Operation When Handswitch Placed to Open; Valve Remains Closed per C-03 Indicating Lights CR 20001486 In An SBO Event, B4300 Will Attempt To Close With Low Control Power Available, Likely Resulting In A Blown Control Fuse CR 20017361 Diesel Oil Storage Tank Level Not Recorded on 0187-01 Test (11 EDG) as Required by Procedure (TS Required Step)

CR 20017538 MO-2012 (Division I RHR LPCI Injection Valve)

Auto Closed During Pre-op Test of Division II 5 Minute Timer Bypass Modification. Loss of Shutdown Cooling.

CR 20016454 Control Rod 10-39 Withdrawal Motion Stopped Just Past 00. Rod Declared Inoperable.

NRC IN 93-82 Recent Fuel and Core Performance Problems in October 12, 1993 Operating Reactors NX-7905-46-9 RHR System Schematic Diagram Revision N WO 0107354 Modification of B4300 Control Circuit WO 0004075 MO-2012 Diss / Insp Valve (IEIN 89-1) 4900-1, VOTES WO 0109804 MO-2012, Incorporate 01A-017 Alteration WO 0109891 Adjust MO-2012 Anti-rotation Device WO 0109797 MO-2012 Failed to Stroke WO 0105741 Division II LPCI 5 Minute Timer Bypass Installation Per Mod 00Q250 WO 0003810 Pre-op Test of LPCI 5 Minute Timer Bypass Switch WO 0109500 Attempt to Withdraw CRD 10-39 Revision 1 1R20 Outage Activities

Operations Manual:

D.2 - Reactor and Core Components Handling Equipment 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 0074 Control Rod Drive Exercise Revision 29 0118 Reactor Vessel Temperature Monitoring Revision 5 0137 Master Local Leak Rate Test Revision 22 0137-07A Reactor Steam Supply Valves Leak Rate Testing Revision 15 0201 Refueling Interlocks Weekly Test Revision 12 0907 Procedure for Moving Fuel Into, Out of, and Revision 24 Within the Core 1054 Control Rod Drive Normal Drive Timing Test Revision 11 2150 Plant Prestart Checklist Revision 23 2167 Startup Checklist Revision 41 8136-01 Secondary Containment Penetration Work Revision 1 Control Index 8151 Heavy Load Movement Procedure Revision 6 9006 Reactor Well and Dryer-Separator Storage Pool Revision 15 Draining 9007-B Shift Supervisor Refueling Checklist Revision 14*

9010 Refueling Platform Daily Inspection and Auxiliary Revision 13 Bridge Inspection 9026 Refueling Bridge Functional Test Revision 11 WO 0005177 Disassemble, Inspect, Reassemble No. 11 Heat Exchanger 1R22 Surveillance Testing Internal Correspondence, Chemistry to January 26, 2000 Engineering: Red Plastic Label Tape and Tan Foam-Backed Paper 0036-02 ECCS Automatic Initiation Test, Including Loss Revision 21 of Auxiliary Power

0255-20-IIA-1 Reactor Coolant Pressure Boundary Hydrostatic Revision 13 Test 0255-06-IA-4 HPCI-31 Torus Suction Check Valve Operability Revision 3 Test 0255-20-IIC-2 Reactor Coolant Pressure Boundary Leakage Revision 13 Test 0255-20-IIC-2 Reactor Coolant Pressure Boundary Leakage Revision 12 Test 0419-01 Alternate Shutdown System Cycle Functional Revision 5 Test for 12 EDG and EDG Oil Transfer Pump Switches 3186-G-01-03 Quality Control Inspection Record for Revision 5 WO 0107732 4001-11-01 Swing Check Valve Inspection Revision 6 4262 Mechanical Maintenance Pre-job Briefing Revision 5 Checklist for WO 0107732 4AWI-06.07.03 Chemical Compatibility In and On Plant Systems and Components 8041 Stainless Steel Pipe Cleaning and Inspection Revision 0 Procedure 98-003 Licensee Event Report - Transgranular Stress Cracking Corrosion in Control Rod Drive Lines CR 20011423 Inspection Plan for CRD Pipe Cracking per CR 19981023 Did Not Include Inspection of Undervessel Insert / Withdraw Lines CR 20000206 Crack Indication Found On CRD Withdrawal Line Found In Drywell During Eddy Current Examination CR 19981023 Possible Cracked CRD Withdrawal Line in DW Found During RPV Hydro With Attachments CR 19981029 Remove Leaking CRD Withdrawal Line 34-27 and Have a Metallurgical Review Performed to Assist With Cause Eval CR 20000318 Minor Indications Found On CRD Lines During Eddy Current Examination CR 19981023 Possible Cracked CRD Withdrawal Line In DW Found During RPV Hydro

CR 20000206 Crack Indication on CRD Withdrawal Line Found in Drywell During Eddy Current Examination CR 20011432 Inspection Plan for CRD Pipe Cracking Per CR 19981023 Did Not Include Inspection of Undervessel Insert/Withdraw Lines CR 20017718 Further Evaluation of Potential Degradation Mechanism & Locations of CRD Withdrawal Lines Inside Drywell Recommended (including attachments, actions, and references)

LER 50-263/98-03 Transgranular Stress Corrosion Cracking Identified In Control Rod Drive Lines M-111 Reactor Building Cooling Water System Revision AD M-115 Nuclear Boiler Steam Supply System Revision AV M-118 Control Rod Hydraulic System Revision AL M-119 Control Rod Hydraulic System Revision M SIR-99-115 Review of Stainless Steel Pipe Cleaning and Inspection Procedure #8014 (SRI No. 99-07)

SRI 99-007 Safety Review Item: Stainless Steel Pipe Revision 0 Cleaning and Inspection Procedure #8041 WO 0107732 Dis-assemble and Inspect Valve for IST Program 2OS1 Access Control to Radiologically Significant Areas (71121.01)

CR 20016820 Two Individuals Worked in RCA without 11/6/2001 Electronic Dosimetry CR 20016596 Several Items Found Crossing a Contaminated 10/31/2001 Boundary 4 AWI 08.04.03 CR 20016572 Increased Frequency of ED Dose Alarms Since 10/30/2001 October 1, 2001 CR 20017348 Radiation Area Not Posted From All Entry Points 11/20/2001 R.01.01 RWP Preparation and Issuance Revision 33 R.01.03 RWP Revision Revision 9 R.02.01 Dose Rate Surveys Revision 12 R.02.02 Contamination Surveys Revision 16 R.08.06 Contaminated Area Control Revision 5

R.12.02 Radiation Protection Key Control Revision 16 R.13.01 Job Coverage Revision 21 R.13.06 Job Planning Revision 9 RPIP 1621 AM-2 Area Monitor Description, Operation and Revision 9 Calibration 2OS2 As-Low-As-Is-Reasonably-Achievable (ALARA) Planning and Controls (71121.02)

Monticello 2000 Year End ALARA Report 2/1/2001 4 AWI 10.01.03 Condition Report Process Revision 17 4 AWI 08.04.01 Radiation Protection Plan Revision 12 5608-02 WO/Procedure Assignment to Existing Specific Revision 4 RWP CR 20017658 RWPs with Exposure Exceeding Estimate are 11/29/2001 not Reviewed in a Timely Manner CR 20017250 Individual Received Unplanned CEDE Greater 11/16/2001 than 10 mrem CR 20017210 Contamination Levels Inside MO-2398 Higher 11/15/2001 than Expected R.01.06 RWP ALARA Reviews Revision 3 RWP 10527 Helper/Laborer Drywell General Entry Revision 0 RWP 10555 CV-2790 Valve Work Revision 1 RWP 10177 RWCU Miscellaneous Valve Work Revision 0 RWP 10504 Perform RP Surveys and Coverage Revision 1 RWP 10507 ALARA Efforts in the Drywell Revision 0 RWP 10515 Drywell General Entry Revision 1 RWP 10520 Nozzle ISI and Insulation Work Revision 1 2OS3 Radiation Monitoring Instrumentation Calibration of the Canberra Fastscan WBC January 18, 2001 System and the Monticello Nuclear Generation Plant 0068 Spent Fuel Pool and Reactor Building Exhaust Revisions 16 and 17 Plenum Monitor Calibration

1024 Area Radiation Monitor Calibration Revisions 26 and 27 2001-004-05-036 Nuclear Oversight Observation Report December 5, 2001 4093-PM Control Room Air Supply Cylinder Change Out Revision 1 A.2-414 Large Volume Liquid Sample and/or Dissolved Revision 19 Gas Sample Obtained At Post Accident Sampling System CR 20016344 Unexpected Alarm C-252 B-7 Offgas Storage October 21, 2001 Building CAM CR 20018220 Wrong Computer Point Alarm Activated During December 20, 2001 ARM Calibration 1024 on February 22, 2001 CR 20015537 Three Area Radiation Monitors "As Found" September 20, 2001 Readings Out of Tolerance CR 20014766 Neutron Meter (RB2) "As Found" Readings Out August 10, 2001 of Tolerance High CR 20014764 HPGE #3 Incorrectly Returned to Service August 10, 2001 CR 20018188 Control Room Breathing Air Supply System December 19, 2001 Hoses Have No Inventory/Inspection Requirement CR 20017827 Condensate Drain Hose from Offgas Piping December 12, 2001 Routed to Clean Turbine Building Drain CR 20013313 Resolve Comments on Fire Brigade Training June 12, 2001 June 8, 2001 CR 20014492 Fastscan WBC Outside of Acceptance Criteria July 30, 2001 on Quarterly Inter-Lab Comparison for Zn-65 CR 20013971 HPGE #1 Failed Daily Source Check July 8, 2001 CR 20013335 SCBA Fiber Breathing Air Cylinders and Valve June 13, 2001 Assemblies Not NIOSH Approved CR 20013243 HPGE #4 Detector Cs-137 Energy Calibration June 7, 2001 Activities Low M-7704L-007 MSA SCBA Training Revision 0 R.03.01 Instrumentation Requirements Revision 20 R.05.07 SCBA Inspection and Functional Testing Revision 10 R.05.08 Service Air Composition Test Revision 3 R.09/49 NMC Portal Monitor Tests Revision 4

R.09.01 Fastscan Quality Assurance Calibration Check Revision 11 R.09.07 RO-2/RO-2A Tests Revision 12 R.09.10 Johnson Extender Tests Revision 10 R.09.13 NMC Continuous Air Monitors Revision 10 R.09.15 Neutron Instrument Source Check Revision 7 R.09.20 Controlled Area Portal Alarm Functional Test Revision 14 and Posting Verification R.09.37 NMC Friskall Checks Revision 11 4OA1 Performance Indicator Verification CR 20017250 Individual Received Unplanned CEDE Greater November 16, 2001 than 10 mrem 30

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