IR 05000263/2004002

IR 05000263-04-002, on 01/01/2004 - 03/31/2004; Monticello Nuclear Generating Plant; Fire Protection and Operability Evaluations

ML041180332
Person / Time
Site:	Monticello
Issue date:	03/31/2004
From:	Burgess B NRC/RGN-III/DRP/RPB2
To:	Thomas J. Palmisano Nuclear Management Co
References
IR-04-002
Download: ML041180332 (41)

Text

April 23, 2004

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT NRC INTEGRATED INSPECTION REPORT 05000263/2004002

Dear Mr. Palmisano:

On March 31, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Monticello Nuclear Generating Plant. The enclosed integrated inspection report documents the inspection findings which were discussed on April 2, 2004, with Mr. Jack Purkis and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, there were four NRC-identified findings of very low safety significance, of which three involved a violation of NRC requirements. However, because these violations were of very low safety significance and because the issues were entered into the licensees corrective action program, the NRC is treating these violations as Non-Cited Violations in accordance with Section VI.A.1 of the NRCs Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -

Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Monticello Nuclear Generating Station. In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA by Geoffrey Wright Acting for/

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

Enclosure:

Inspection Report 05000263/2004002 w/Attachment: Supplemental Information

REGION III==

Docket No:

50-263 License No:

DPR-22 Report No:

05000263/2004002 Licensee:

Nuclear Management Company, LLC Facility:

Monticello Nuclear Generating Plant Location:

2807 West Highway 75 Monticello, MN 55362 Dates:

January 1 through March 31, 2004 Inspectors:

S. Burton, Senior Resident Inspector R. Orlikowski, Resident Inspector D. McNeil, Reactor Engineer J. Bond, Regional Inspector D. Chyu, Regional Inspector M. Parker, Regional Inspector Observers:

None Approved by:

B. L. Burgess, Chief Branch 2 Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000263/2004002; 01/01/2004 - 03/31/2004; Monticello Nuclear Generating Plant; Fire

Protection and Operability Evaluations.

This report covers a 3-month period of baseline resident inspection. The inspections were conducted by Region III reactor inspectors and the resident inspectors. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.

The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A.

Inspector-Identified and Self-Revealed Findings Cornerstones: Initiating Events and Mitigating Systems

Green.

A finding of very low safety significance was identified by the inspectors for a violation of Technical Specification for failing to follow Fire Protection Program procedures which required that changes made to the Fire Protection Program be evaluated for impacts to safe-shutdown capabilities. The Engineering Department failed to evaluate the replacement of two dry chemical fire extinguishers with two pressurized water extinguishers in the intake structure area. The licensee has instituted corrective actions including a formal root cause evaluation to assess this issue.

This issue was more than minor because an unsuppressed electrical or oil fire could affect both trains of emergency service water. The issue was of very low safety significance because the 20-foot separation between two trains did not contain any combustibles and because the automatic fire suppression system was not affected by the finding. The issue was a Non-Cited Violation of Technical Specification 6.5.A, which requires written procedures covering the Fire Protection Program. (Section 1R05(1))

Green.

Three (3) examples of a finding of very low safety significance were identified by the inspectors for a violation of 10 CFR 50, Appendix B, Corrective Action requirements for failing to take prompt and adequate corrective actions to correct pre-fire strategies.

The licensee has instituted corrective actions including a formal root cause evaluation to assess this issue.

This issue was more than minor because pre-fire strategies are used by the fire brigade to identify additional equipment needed and to determine the fire hazards in the fire zones. Failure to have updated and accurate pre-fire strategies could impair the fire brigades ability to promptly and properly respond in the event of a fire. The issue was determined to be of very low safety significance as a result of an SDP evaluation which provided credit for the robustness of the fire protection methodology and the automatic fire suppression system for the fire zone. A Non-Cited Violation of 10 CFR 50,

Appendix B, Criterion XVI, Corrective Action was identified for failure of the licensee to take prompt actions to correct conditions adverse to quality. (Section 1R05(2))

Cornerstone: Mitigating Systems

Green.

A finding of very low safety significance with no associated violation was identified by the NRC inspectors associated with the non-safeguards 13 diesel generator (DG) output breaker. The finding was associated with the failure of the Electrical Maintenance Department to identify and correct a damaged output breaker, resulting in increased plant risk. During a monthly surveillance test in January 2004 the 13 DG output breaker failed to shut. An investigation was performed and no apparent cause of the breakers failure to shut was identified prior to returning the 13 DG to service. During the February surveillance test, the 13 DG output breaker again failed to shut for monthly testing. Further investigation identified a bent linkage in the breaker, which was the cause of the breakers failure to shut. The Electrical Maintenance Department repaired the bent linkage and returned the 13 DG to service.

Since the 13 DG has a cumulative impact over time on the plants safety due to its contribution to core damage frequency (CDF), the inspectors concluded that the finding was more than minor because this finding would become a more significant safety concern if left uncorrected. This finding was of very low safety significance because there was no design deficiency, no actual loss of safety function, no single train loss of safety function for greater than the Technical Specification allowed outage time, and no risk due to external events. (Section 1R15(1))

Cornerstones: Mitigating System and Barrier Integrity

Green.

A finding of very low safety significance was identified by the Engineering Department, but because the finding required a Phase 2 significance determination, the finding was treated as an NRC-identified finding. The finding was associated with the failure to maintain the qualification of switchgear when non-safety related alarm modules were installed on the Division I and Division II 250 VDC buses without an appropriate interface. The alarm re-flash units were installed without safety-related fuses as the interface between the safety and non-safety components. The licensee instituted corrective actions to install an appropriate interface and review certain past modifications for similarities.

The issue was more than minor because it directly impacted the design control attributes for both the Mitigating Systems and Barrier Integrity objectives. The results of the SDP process found the issue to be Green after consideration of the robust design of the modification and because the fuses had in the past blown to protect the source and adequately isolated the non-safety equipment from the bus. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, Design Control was issued for failure to maintain the safety qualification of safety-related switchgear. (Section 1R15(2))

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

Monticello operated at full power for the entire assessment period except for brief down-power maneuvers to accomplish rod pattern adjustments and to conduct planned surveillance testing activities.

REACTOR SAFETY

Cornerstone: Initiating Events, Mitigating Systems, Barrier Integrity, and

Emergency Preparedness

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

The inspectors performed partial walkdowns of accessible portions of trains of risk-significant mitigating systems equipment. The inspectors reviewed equipment alignment to identify any discrepancies that could impact the function of the system and potentially increase risk. Identified equipment alignment problems were verified by the inspectors to be properly resolved. The inspectors selected redundant or backup systems for inspection during times when equipment was of increased importance due to unavailability of the redundant train or other related equipment. Inspection activities included, but were not limited to, a review of the licensees procedures, verification of equipment alignment, and an observation of material condition, including operating parameters of equipment in-service. As part of this inspection, the documents in 1 were utilized to evaluate the potential for an inspection finding.

The inspectors selected the following equipment trains to verify operability and proper equipment line-up for a total of two samples:

• hard pipe vent system with Division II residual heat removal (RHR) out-of-service for maintenance, during the week ending January 7, 2004; and

• Division I residual heat removal service water (RHRSW) system with Division II RHR out-of-service for maintenance, during the week ending January 7, 2004.

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown

The inspectors performed a complete walkdown of equipment for one risk significant mitigating system. The inspectors walked down the system to verify mechanical and electrical equipment line-ups, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of past and outstanding work orders (WO) was performed to verify that any deficiencies did not significantly affect the system function. In addition, the inspectors reviewed the condition report (CR) database to verify that any system equipment alignment problems were being identified and appropriately resolved. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors selected the following system to verify operability and proper equipment line-up for a total of one sample:

• reactor core isolation cooling (RCIC), for the weeks ending March 6, 2004, and March 13, 2004.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

The inspectors walked down risk significant fire areas to assess fire protection requirements. The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and had implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events (IPEEE), the potential to impact equipment which could initiate or mitigate a plant transient, or the impact on the plants ability to respond to a security event. The inspection activities included, but were not limited to, the control of transient combustibles and ignition sources, fire detection equipment, manual suppression capabilities, passive suppression capabilities, automatic suppression capabilities, compensatory measures, and barriers to fire propagation. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors selected the following areas for review for a total of six samples:

• Fire Zone 7-A, 125V Division I battery room, during the week ending January 17, 2004;

• Fire Zone 10, administration building, during the weeks ending January 17, 2004 and January 24, 2004;

• Fire Zone 12-A, lower 4 kv bus area (11, 13 and 15), during the week ending January 24, 2004;

• Fire Zone 12-B, hydrogen seal area, during the week ending January 24, 2004;

• Fire Zone 23-A, intake structure pump room, during the week ending March 20, 2004; and

• Fire Zone 13-B, RX feedwater pump and lube oil reservoir area, during the week ending March 20, 2004.

b. Findings

(1) Failure to Properly Evaluate Fire Protection Strategy and Program Changes Introduction The inspectors identified a Non-Cited Violation (NCV) of Technical Specifications (TS)having very low safety significance (Green) for failing to follow Fire Protection Program procedures, which require that changes made to the Fire Protection Program be evaluated for impact on safe-shutdown abilities.

Description While performing a fire protection inspection of the intake structure area (Fire Zone 23-A), the inspectors noted that the area contained two pressurized water extinguishers intended to extinguish small Class A fires. The licensees pre-fire strategies identified the combustible loads in Fire Zone 23-A as lubricating oil, cable insulation, and the contents of a storage locker for flammables in the area. The combustible loads were not Class A fire hazards.

The National Fire Protection Association (NFPA) Code No. 10, Standards for Portable Extinguishers, which identifies the proper selection of extinguishers by the class of hazards, does not identify pressurized water extinguishers for protection from Class B hazards (oil and flammable liquids). The NFPA Fire Protection Handbook states that the extinguishers in any one area should correspond to the hazards of that area. The handbook also states that if non-foam water base extinguishers are used on Class B fires the fire may flare up, spread, or injure the operator. The inspectors determined that the pressurized water extinguishers placed in the intake structure area were not best suited for controlling the fires associated with the fire hazards in the area.

On December 19, 2003, the licensee issued CR 03011892 which assessed an external operating experience (OE) document, titled, ABC Dry Chemical Fire Extinguishers Incompatible with Chlorine-Based Oxidizers. The OE document advised against the use of dry chemical and Halon fire extinguishers in certain areas, warning that ammonium based compounds typically found in multipurpose (ABC) dry chemical fire extinguishers can react violently, igniting or exploding, on contact with strong oxidizers such as the chlorine or bromine based chemicals used in circulating water treatment systems. The corrective measure outlined in the OE document consisted of staging water-filled extinguishers in these areas to supplement the existing dry chemical extinguishers. In response to the CR, the licensee removed two dry chemical fire extinguishers from the intake structure area because sodium hypochlorite interfaced with circulating water through polyvinyl chloride (PVC) piping. The licensee replaced the two dry chemical extinguishers with two pressurized water extinguishers. The licensee generated CR 04003245 to acknowledge that a thorough evaluation had not been completed at the time the extinguishers were replaced.

Analysis The inspectors determined that a performance deficiency existed because the Engineering Department failed to follow Fire Protection Program procedures which required that changes made to the Fire Protection Program be evaluated for impacts to safe-shutdown capabilities. The inspectors concluded that the finding was greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of protection against external factors (fire) and could have effected the mitigating systems objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences, because an unsuppressed electrical or oil fire could affect both trains of emergency service water.

The inspectors completed a significance determination of this issue using IMC 0609, Significance Determination Process (SDP), dated April 30, 2002, Appendix F, Determine Potential Risk Significance of Fire Protection and Post-Fire Safe Shutdown Inspection Finding, dated February 2, 2001, Scheme 3. As part of the Phase 2 evaluation the inspectors considered the potential impact on equipment located in the affected fire zone. The inspectors determined that there were electrical cabinets which could ignite the intervening cable trays in the overhead and propagate fire to both trains of emergency service water (ESW) system. The inspectors used the electrical fire as the most limiting scenario, with ignition frequencies of 2.4E-3 per reactor year for all of the electrical cabinets in the intake structure as referenced in the licensees IPEEE (log 10(IF)= -2.62). The 20-foot separation between the redundant trains was not degraded (FB= -2). The automatic fire suppression capability was assumed to be in a normal operating state because no finding was identified within this capability (AS= -1.25). This finding affected the manual effectiveness and was conservatively considered highly degraded (MS=-0.25). Since the exposure time for the degraded condition existed for more than 30 days, the estimated likelihood rating for the postulated fire event was determined to be less than 1E-6 occurrences per reactor year.

A fire in the intervening cable trays could cause direct damage to the cabling for ESW pumps A and B. These pumps are required to support the operation of the emergency diesel generators (EDG). However, in this case, the EDGs were not needed because a fire in the intake structure would not cause a loss of offsite power. Therefore, two SDP worksheets, Transients and Transients without Power Conversion System, were used to evaluate the finding. Other redundant safe shutdown equipment would remain available to mitigate the consequences of a fire in that area. Based upon the inspectors evaluation of the Fire Protection SDP using these inputs, the finding screened as a finding of very low safety significance (Green).

Enforcement Technical Specification 6.5.A requires written procedures be established, implemented and maintained. Subsection A.1 requires procedures recommended in Regulatory Guide 1.33, Revision 2, February 1978, and Subsection A.2 requires procedures for the Fire Protection Program Implementation. Appendix A of Regulatory Guide 1.33 requires written procedures for the Plant Fire Protection Program. Administrative Work Instruction 4AWI-08.01.00, Fire Protection Program Plan, Section 4.11.2 requires that changes be evaluated to meet the conditions of the license which states, in part, that changes shall be evaluated against the ability to achieve and maintain safe shutdown in the event of a fire and that the change will not alter specific features of the NRC approved program. Contrary to the above, the Engineering Department failed to follow the Fire Protection Program procedures when they changed the class of extinguishers in a safe-shutdown area. Specifically, the Engineering Department failed to properly evaluate the change for adverse effects on the ability to achieve and maintain safe shutdown in the event of a fire. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, as noted below, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000263/2004002-01). The licensee has entered this into their corrective action program as CR 04002930 and CR 04003245. The licensee also initiated CR 04003007, which required a formal root cause investigation for the potential programmatic breakdown emerging within the Fire Protection Program and/or 10 CFR 50, Appendix R areas.

(2) Inadequate Corrective Action Impair Fire Brigade Response Capabilities Introduction The inspectors identified a Non-Cited Violation (NCV) having very low safety significance (Green) for three

(3) examples where inadequate corrective actions resulted in incorrect pre-fire strategies which could impair the fire brigades ability to respond to a fire.

Description In April 2002, CR 02003229 was generated to address an adverse trend in discrepancies of fire extinguishers, alarm bells, emergency lights, and phones in the pre-fire strategies. The CR identified the cause of the discrepancies as a lack of attention to detail when fire strategy maps were revised. The associated corrective action generated was to revise the fire drill procedure to include a step which would direct the fire brigade members to verify the accuracy of the fire strategy maps in the area drilled. Because the Operations Department has reviewed only 8 of 87 fire zones for accuracy during post-drill activities in the past 2 years and because they do not drill in all of the 87 fire zones, the inspectors concluded that the corrective actions would not have been timely or comprehensive. The inspectors concluded that the change made to the fire drill procedure was not an adequate action to verify and correct discrepancies in the pre-fire strategies because the limited number of drills performed and limited areas of the plant in which drills were performed would not be timely and failed to assess many areas. The licensee documented the inadequate corrective action in CR 04003239.

In August 2003, CR 03008727 was generated to capture inspector observations which identified that the pre-fire strategy for the reactor feed pump (RFP) area only provided direction to isolate a combustible hydrogen gas source at an isolation valve located inside the RFP area fire zone. The NFPA Fire Protection Handbook advises that all combustible gas sources should be isolated prior to entering a fire zone to combat a fire, further stating that no attempt to extinguish pressurized fuel fires should be made unless the source of fuel can be promptly shut off, otherwise the fuel may explode. The proposed corrective action was to add a statement to the pre-fire strategy regarding the isolation of the hydrogen gas from outside the fire zone. The allowed completion time to correct the strategy was 300 days. The inspectors noted that the Engineering Department had not started corrective actions after approximately 220 days and the Engineering Department indicated that the plans were to start fire map evaluations in June, less than 60 days prior to the completion due date. Because the time to complete the action was delayed and assumed to take less than 60 days the inspectors determined that waiting over 220 days to complete the action did not constitute prompt action to correct this condition adverse to quality. In an attachment to CR 04003007, the licensee acknowledged that the action to revise the RFP area pre-fire strategy should have been more timely. Corrective actions to be taken were under evaluation by the licensee.

In its response to an external operating experience documented in CR 03011892, which discussed the toxic gas, ignition, and explosion hazards associated with the use of Halon and dry chemical fire extinguishers near circulating water treatment systems containing chlorine-based chemicals, the licensee took actions to remove dry chemical extinguishers from the intake structure area and created a training document which cautioned the fire brigade members about using dry chemical extinguishers in the intake structure area. The Engineering Department failed to take prompt and adequate actions to correct conditions adverse to quality when they did not properly update the fire strategies and maps to include the toxic gas, ignition, and explosion hazards in the intake structure area when they were discovered. The corrective actions did not include a step to revise the pre-fire strategy maps to reflect the potential toxic hazard and the fire extinguishants best suited for the intake structure area as specified in 10 CFR 50, Appendix R.III.K.

The fire brigade uses pre-fire strategies to identify additional equipment needed, to identify adjacent resources available, and to determine the hazards in the fire zones during a fire. Failure to have updated and accurate pre-fire strategies could impair the fire brigades ability to promptly and properly respond to a fire. Therefore for each of the above examples, the actions taken to correct the pre-fire strategies were not considered prompt or adequate.

Analysis The inspectors determined that a performance deficiency existed because higher priority should have been given to implementing prompt and adequate changes to the fire strategies. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002, because the fire brigade utilizes the pre-fire strategies to assess hazards, locate alternate equipment, and prepare to combat fires. The lack of adequate strategies impacts the brigades timeliness and mitigating capabilities.

The inspectors completed a significance determination of this issue using IMC 0609, Significance Determination Process (SDP), dated April 30, 2002, Appendix F, Determine Potential Risk Significance of Fire Protection and Post-Fire Safe Shutdown Inspection Finding, dated February 2, 2001, Scheme 1 (RFP Area) and Scheme 3 (intake structure area). Because the protective scheme for the RFP area utilizes a 3-hour fire barrier separation that provides wall-to-wall and floor-to-floor separation and the barrier was not affected by the degradation of the fire brigade effectiveness, the finding screened out as Green. The protective scheme for the fire zone utilized more than 20 feet of combustible-free horizontal separation between the redundant safe-shutdown trains and an automatic fire suppression system as part of its fire protection methodology. However, since the finding could have impacted the fire brigades effectiveness, the finding screened out of Phase 1 and a Phase 2 evaluation of IMC 0609, Appendix F, Scheme 3 was needed. A Phase 2 analysis was performed.

The result of the Phase 2 analysis is documented in the analysis section of 1R05(1).

Enforcement The licensees Fire Protection Program, including the Fire Protection Plan and pre-fire strategies, is committed to 10 CFR 50 Appendix B, Criterion XVI, Corrective Action which requires that measures be established to promptly identify and correct deficiencies and other conditions adverse to quality. Contrary to the above, the licensee failed to promptly correct NRC-identified deficiencies within its pre-fire strategies as evidenced by the following examples: inadequate corrective actions to address incorrect pre-fire strategies, untimely actions to update the pre-fire strategies to include a combustible gas isolation statement, and inadequate corrective actions to identify additional hazards in the Area fire zone. This finding is a violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, as noted below, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000263/200402-02). The licensee entered this issue into their corrective action program as CR 04003007, which required a formal root cause investigation of the potential programmatic breakdown emerging within the Fire Protection Program and/or 10 CFR 50, Appendix R areas.

1R11 Licensed Operator Requalification Program

.1 Licensed Operator Simulator Exercise

a. Inspection Scope

The inspectors performed a quarterly review of licensed operator requalification training.

The inspection assessed the licensees effectiveness in evaluating the requalification program, ensuring that licensed individuals operate the facility safely and within the conditions of their license, and evaluated licensed operator mastery of high-risk operator actions. The inspection activities included, but were not limited to, a review of high risk activities, emergency plan performance, incorporation of lessons learned, clarity and formality of communications, task prioritization, timeliness of actions, alarm response actions, control board operations, procedural adequacy and implementation, supervisory oversight, group dynamics, interpretations of TSs, simulator fidelity, and licensee critique of performance. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors observed the following requalification activity for a total of one sample:

• a training crew during an evaluated simulator scenario that included a loss of Bus 15, a loss of circulating water and bypass valves coupled with a failure to scram, which resulted in the operators entering applicable abnormal response procedures including emergency operating procedures and the emergency plan, during the week ending March 13, 2004.

b. Findings

No findings of significance were identified.

.2 Written Examination and Operating Test Results

a. Inspection Scope

The inspectors reviewed the pass/fail results of individual written tests, operating tests, and simulator operating tests (required to be given per 10 CFR 55.59(a)(2))

administered by the licensee during calender year 2004. This represents one sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed systems to assess maintenance effectiveness, including maintenance rule activities, work practices, and common cause issues. Inspection activities included, but were not limited to, the licensee's categorization of specific issues including evaluation of performance criteria, appropriate work practices, identification of common cause errors, extent of condition, and trending of key parameters. Additionally, the inspectors reviewed implementation of the Maintenance Rule (10 CFR 50.65)requirements, including a review of scoping, goal-setting, performance monitoring, short-term and long-term corrective actions, functional failure determinations associated with reviewed condition reports, and current equipment performance status. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors performed the following maintenance effectiveness reviews for a total of two samples:

a function-oriented review of the 11 and 12 EDG system because it was designated as risk significant under the Maintenance Rule, during the weeks ending January 31 through February 14, 2004; and

an issue-oriented review of the RCIC system because it was designated as risk significant under the Maintenance Rule, during the weeks ending February 21, 2004, and February 28, 2004.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed maintenance activities to review risk assessments (RA) and emergent work control. The inspectors verified the performance and adequacy of RAs, management of resultant risk, entry into the appropriate licensee-established risk bands, and the effective planning and control of emergent work activities. The inspection activities included, but were not limited to, a verification that licensee RA procedures were followed and performed appropriately for routine and emergent maintenance, that the RAs for the scope of work performed were accurate and complete, that necessary actions were taken to minimize the probability of initiating events, and that activities to ensure that the functionality of mitigating systems and barriers were performed.

Reviews also assessed the licensee's evaluation of plant risk, risk management, scheduling, configuration control, and coordination with other scheduled risk significant work for these activities. Additionally, the assessment included an evaluation of external factors, the licensee's control of work activities, and appropriate consideration of baseline and cumulative risk. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors observed maintenance or planning for the following activities or risk significant systems undergoing scheduled or emergent maintenance for a total of three samples:

• investigate and repair 13 diesel generator (DG) lockout, during the weeks ending January 24, 2004, and January 31, 2004;

• failure of control room emergency ventilation system compressor seal, during the week ending February 28, 2004; and

• service water piping corrosion, during the weeks ending February 28, 2004, and March 6, 2004.

b. Findings

No findings of significance were identified.

1R14 Personnel Performance During Non-Routine Plant Evolutions and Events

a. Inspection Scope

The inspectors reviewed personnel performance to planned evolutions to review operator performance and the potential for operator contribution to the evolution. The inspectors observed or reviewed records of operator performance during the evolution.

Reviews included, but were not limited to, operator logs, pre-job briefings, instrument recorder data, and procedures. As part of this inspection, the documents in 1 were utilized to evaluate the potential for an inspection finding.

The inspectors observed the following evolution for a total of one sample:

• planned back-seating of the high pressure coolant injection (HPCI) inboard steam isolation valve MO-2034 to reduce drywell leakage, during the week ending February 7, 2003.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors performed operability evaluations of degraded or non-conforming systems that potentially impacted mitigating systems or barrier integrity. The inspectors reviewed operability evaluations affecting mitigating systems or barrier integrity to ensure that operability was properly justified and that the component or system remained available. The inspection activities included, but were not limited to, a review of the technical adequacy of the operability evaluations to determine the impact on TS, the significance of the evaluations to ensure that adequate justifications were documented, and that risk was appropriately assessed. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors reviewed the following operability evaluations for a total of four samples:

• operability of HPCI with MO-2063 in the closed position, during the week ending January 21, 2004;

• operability of 250 VDC buses with non-safety fuses, during the weeks of February 14, 2004, through March 30, 2004;

• relay device lockout on the non-safeguards 13 diesel generator output breaker, during the week ending February 28, 2004; and

• drywell containment air monitor alarm and loss of flow, during the weeks ending February 14, 2004, and February 28, 2004.

b. Findings

(1) Introduction The inspectors identified a finding of very low safety significance (Green) with no associated violation for failure of the Electrical Maintenance Department to identify and correct a damaged output breaker for the non-safeguards 13 DG. Since the 13 DG has a cumulative impact over time on the plants safety due to its contribution to core damage frequency (CDF), the inspectors concluded that the finding was more than minor because this finding would become a more significant safety concern if left uncorrected.

Description The 13 DG is a non-TS and non-safety related system. The 13 DG does provide a function to backfeed equipment, including 125 VDC and 250 VDC battery chargers, during a station blackout (SBO) event. Because of this function, the 13 DG was determined to be risk-significant in the Monticello Probabilistic Risk Assessment (PRA)model and, therefore, affects the CDF of the Monticello plant. The 13 DG is also included under the Maintenance Rule of 10 CFR 50.65 based on its impact on risk for a SBO event.

On January 22, 2004, while attempting to synchronize the 13 DG to load center LC-107 during a monthly test, the operator received a relay device lockout on the output breaker and the output breaker failed to shut. The 13 DG was shut down and the Operations Department wrote CR 04000750 to document the issue. The Electrical Maintenance Department performed an investigation which included a walk-down of the 13 DG, interviews with operations personnel, a review of 13 DG related technical manuals and drawings, and discussions with Cummins/Ziegler diesel generator vendor representatives. A work order was written to remove the output breaker from service for inspection and maintenance at a later time. On January 23, 2004, the 13 DG was declared functional with the cause not being identified and the work order to inspect the output breaker still outstanding.

On February 18, 2004, the 13 DG output breaker again failed to shut during synchronization with load center LC-107 during a monthly test. The failure of the output breaker to shut was caused by a relay device lockout of the breaker. The Operations Department wrote CR 04001791 to document the issue. On February 20, 2004, an investigation of the output breaker revealed a damaged interlock bracket that was determined to be the cause of the relay device lockout events on both January 22, and February 18, 2004. The Electrical Maintenance Department repaired the bent interlock bracket and returned the 13 DG to service on February 20, 2004.

Analysis The inspectors determined that the failure to identify and correct the damaged 13 DG output breaker after the first failure was a performance deficiency warranting further evaluation. The inspectors reviewed this finding using the guidance contained in Appendix B, Issue Disposition Screening, of Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports. As a result, the inspectors compared this performance deficiency to the minor questions contained in Section 3, Minor Questions, to Appendix B of IMC 0612. Since the 13 DG has a cumulative impact over time on the plants safety due to its contribution to CDF, the team concluded that the finding was more than minor because this finding would become a more significant safety concern if left uncorrected.

The inspectors reviewed this finding in accordance with IMC 0609, Significance Determination Process (SDP)," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding affected the mitigation systems cornerstone; however, the finding was not a design or qualification deficiency, did not represent an actual loss of safety function of a system or the loss of safety function of single train TS equipment for greater than the allowed outage time, or the loss of safety function of non-TS equipment, nor was there risk due to external events. Therefore, the finding was considered to be of very low safety significance (Green).

Enforcement The 13 DG is a non-safeguards, non-TS system and is not required to cope with a station blackout; therefore, no violations of regulatory requirements occurred. This issue was considered to be a finding of very low safety significance (FIN 05000263/2004002-03). The licensee entered the issue into its corrective action program as CR 040001791, Received 86 lockout on 52-710 [DG is output breaker]

while trying to synchronize 13 DG to LC-107 during monthly operability test of 13 DG, on February 19, 2004, and repaired the bent interlock bracket on the 13 DG output breaker.

(2) Introduction The inspectors identified a Non-Cited Violation (NCV) having very low safety significance (Green) of 10 CFR 50, Appendix B, Criterion III, Design Control for failure to maintain the qualification of switchgear when non-safety related under-voltage re-flash alarm modules were installed on the Division I and Division II 250 VDC buses without an appropriate interface. This finding impacted both the barrier integrity and mitigating systems cornerstones.

Description On February 18, 2004, while performing a review of the under-voltage alarm units for 250 VDC motor control centers (MCC) D311, D312, and D313, the Engineering Department discovered that alarm modules were isolated from the safety-related bus with non-safety related fuses. The inspectors reviewed the equipment supported by the affected MCCs and found that HPCI, RCIC, and containment isolation valves in both divisions, all 10 CFR 50 Appendix B components, were powered from the buses.

Analysis Because both the mitigating systems cornerstone and the barrier integrity cornerstone were affected, the inspectors recognized that a Phase 2 significance determination would be required and the inspectors could not readily ascertain the significance of the finding. The inspectors consulted Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, which indicated that a licensee-identified finding appearing to have more than a very low safety significance should be treated as an NRC-identified finding, analyzed using the SDP found in IMC 0609, and dispositioned in accordance with the Enforcement Policy.

The inspectors reviewed the finding and determined that a performance deficiency existed because the installed alarm units failed to provide an appropriate safety-related interface between the safety and non-safety systems. The inspectors determined that the issue was more than minor because it directly impacted the design control attributes for both the mitigating systems and barrier integrity objectives. Because both the mitigating systems and barrier integrity cornerstones were affected, the SDP Phase 1 worksheet required a Phase 2 analysis.

The initial Phase 2 risk assessment characterized this finding as potentially risk significant, using the benchmarked site specific Risk-Informed Inspection Notebook.

However, a Phase 3 analysis, performed by a senior reactor analyst, determined the issue was a Green finding, after providing additional consideration for robust design and installation of the modification, and because the fuses had in the past blown to protect the source and adequately isolated the non-safety equipment from the bus. Therefore, after assessing the licensees operability evaluation, the senior reactor analyst confirmed the licensees conclusion that the qualification deficiency did not result in a loss of function.

The Engineering Department determined that the design of the system was robust and preserved the operability of the equipment for the following reasons: the wiring for the non-safety components was equivalent to safety grade; the wiring was installed in conduit that would maintain the environmental qualifications of the buses; the alarm modules, construction, and installations were of a similar design to the MCC, which preserved the environmental and seismic qualifications; and the non-safety related fuses were Underwriters Laboratories qualified, providing reasonable expectation for the fuses to appropriately isolate faults on the equipment. Additionally, the fuses had in the past blown to protect the source and adequately isolated the non-safety equipment from the bus.

Enforcement Criterion III, Design Control, of 10 CFR 50, Appendix B, states, in part, that design changes shall be subject to design control measures commensurate with those applied to the original design. Contrary to this requirement, the licensee failed to maintain the qualification of safety-related switchgear when they installed non-safety related alarm modules on the Division I and Division II 250 VDC buses without an appropriate interface. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000263/2004002-04). The licensee has entered this into their corrective action program as CR 04001787. Completed corrective actions included establishing a WO to install safety-related fusing with acceptable interrupt ratings. Additionally, the licensee initiated actions to review similar past design changes associated with this condition.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors verified that the post-maintenance test procedures and activities were adequate to ensure system operability and functional capability. Activities were selected based upon the structure, system, or components ability to impact risk. The inspection activities included, but were not limited to, witnessing or reviewing the 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, 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 USAR design requirements. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors selected the following post-maintenance activities for review for a total of four samples:

• post-maintenance test for replacement of failed average power range monitor (APRM) card, during the week ending January 17, 2004;

• post-maintenance testing of No. 12 EDG engine driven fuel oil pump following realignment of the pump coupling, during the week ending January 31, 2004;

• post-maintenance testing of two control rods after speed adjustments to compensate for potentially stuck open check valve, during the weeks ending February 14, 2004, and February 28, 2004; and

• post-maintenance test for replacement of anticipated transient without SCRAM (ATWS) electrical relay, during the week ending February 28, 2004.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed surveillance testing activities to assess operational readiness and to ensure that risk-significant structures, systems, and components were capable of performing their intended safety function. 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. The inspection activities included, but were not limited to, a review 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. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors selected the following surveillance testing activities for review for a total of six samples:

• emergency core cooling system (ECCS) high drywell pressure sensor test, during the week ending January 17, 2004;

• technical support center-emergency ventilation system (TSC-EVS) quarterly operability test, during the weeks ending January 17, 2004, through January 31, 2004;

• reactor core isolation cooling (RCIC) quarterly pump and valve tests, during the week ending February 14, 2003;

• anticipated transient without SCRAM (ATWS) - recirculation trip for reactor pressure and level trip unit test and calibration, during the week ending February 28, 2004;

• reactor high pressure scram functional test and instrument calibration, during the week ending March 13, 2004; and

• local power range monitor (LPRM) calibration, during the week ending March 6, 2004.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The inspectors selected emergency preparedness exercises that the licensee had scheduled as providing input to the Drill/Exercise Performance Indicator. The inspection activities included, but were not limited to, the classification of events, notifications to off-site agencies, protective action recommendation development, and drill critiques.

Observations were compared with the licensees observations and corrective action program entries. The inspectors verified that there were no discrepancies between observed performance and performance indicator reported statistics. As part of this inspection, the documents in Attachment 1 were utilized to evaluate the potential for an inspection finding.

The inspectors selected the following emergency preparedness activity for review for a total of one sample:

• the inspectors observed a licensed operator weekly examination scenario that was performed on March 8, 2004, in conjunction with licensed operator requalification training. Drill notifications were made with state, county, and local agencies for an alert classification.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Reactor Safety Strategic Area

a. Inspection Scope

The inspectors review of performance indicators (PI) used PI guidance and definitions contained in Nuclear Energy Institute (NEI) Document 99-02, Revision 2, Regulatory Assessment Performance Indicator Guideline, to verify the accuracy of the PI data.

The inspectors review included, but was not limited to, conditions and data from logs, licensee event reports, condition reports, and calculations for each PI specified. As part of the inspection, the documents listed in Appendix 1 were utilized to evaluate the accuracy of PI data.

The following PIs were reviewed for a total of three samples:

• unplanned scrams per 7000 critical hours, for the period of January 2003 through December 2003;

• unplanned scrams with loss of normal heat removal, for the period of January 2003 through December 2003; and

• unplanned power changes per 7000 critical hours, for the period of January 2003 through December 2003.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

Cornerstone: Initiating Events, Mitigating Systems, Barrier Integrity, and

Emergency Preparedness

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

For inspections performed and documented in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees corrective action system at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Minor issues entered into the licensees corrective action system as a result of inspectors observations are included in the list of documents reviewed attached to this report.

b. Findings

A Green finding of low safety significance and associated violation for inadequate corrective actions were identified when the licensee failed to promptly and adequately correct pre-fire strategies. (Section 1R05)

.2 Review of Open Work Orders with an Age Greater than 30 Days

Introduction Administrative Procedure 4AWI-10.01.01, Corrective Action Program, states that the work control process, including work orders, is a corrective action process. Additionally, 4AWI-04.05.05, WO Closeout and Disposition, states that the work order preparer shall review completed work orders to determine if conditions adverse to quality exist.

The inspectors noted that the open work order list contained many work orders where the work had been completed yet the final review had not been done in excess of 700 days. This condition raised the concern that issues documented in the work process may exist which contain conditions adverse to quality, yet the condition had not been entered into the corrective action program or corrected.

a. Inspection Scope

The inspectors selected approximately 30 work orders from the list of open work orders related to risk significant systems for review. From their review the inspectors identified and followed-up on six work orders that had technician comments which potentially impacted quality.

b.

Issues The inspectors identified that the technicians had written condition reports for identified problems and annotated such on the work orders. Further, the licensee had developed a procedure change to perform the post-work review when the work was completed.

The proposed procedure change appeared to be an effective means to ensure issues were properly evaluated and if appropriate entered into the corrective action program.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. Jack Purkis and other members of licensee management on April 2, 2004. 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.

.2 Interim Exit Meetings

Interim exits were conducted for:

• Licensed Operator Requalification Testing for Calendar Year 2004 and Applicability of NRC Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process (SDP),"

with Mr. G. Lashinski on March 10, 2004;

4OA7 Licensee-Identified Violations

None.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Palmisano, Site Vice President

J. Purkis, Plant Manager

R. Baumer, Licensing

G. Bregg, Manager, Quality Services

K. Jepsen, Radiation Protection Manager

D. Neve, Regulatory Affairs Manager

E. Sopkin, Director of Engineering

Nuclear Regulatory Commission

B. Burgess, Chief, Reactor Projects Branch 2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000263/2004002-01 NCV

Failure to Follow Fire Protection Program Procedures Which Require that Changes Made to the Fire Protection Program be Evaluated for Impacts to Safe-Shutdown Capabilities (Section 1R05(1))

05000263/2004002-02 NCV Failure to take Prompt and Adequate Corrective Actions to Correct Pre-Fire Strategies (Section 1R05(2))

05000263/2004002-03 FIN Failure to Identify and Correct a Damaged 13 DG Output Breaker Results in Increased Plant Risk (Section 1R15(1))

05000263/2004002-04 NCV Failure to Maintain the Qualification of Safety-Related Switchgear when Non-Safety Related Alarm Modules were Installed on the Division I and Division II 250 VDC Buses Without an Appropriate Interface (Section 1R05(2))

Closed

05000263/2004002-01 NCV Failure to Follow Fire Protection Program Procedures Which Require that Changes Made to the Fire Protection Program be Evaluated for Impacts to Safe-Shutdown Capabilities (Section 1R05(1))

05000263/2004002-02 NCV Failure to take Prompt and Adequate Corrective Actions to Correct Pre-Fire Strategies (Section 1R05(2))

05000263/2004002-03 FIN Failure to Identify and Correct a Damaged 13 DG Output Breaker Results in Increased Plant Risk (Section 1R15(1))

05000263/2004002-04 NCV Failure to Maintain the Qualification of Safety-Related Switchgear when Non-Safety Related Alarm Modules were Installed on the Division I and Division II 250 VDC Buses Without an Appropriate Interface (Section 1R05(2))

Discussed

None.