IR 05000263/2004007

IR 05000263-04-007 (Drs), on 10/18/2004 - 11/05/2004; Monticello Nuclear Generating Plant; Safety System Design and Performance Capability

ML043380314
Person / Time
Site:	Monticello
Issue date:	12/03/2004
From:	Julio Lara NRC/RGN-III/DRS/EEB
To:	Thomas J. Palmisano Nuclear Management Co
References
IR-04-007
Download: ML043380314 (47)

Text

ber 3, 2004

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION 05000263/2004007(DRS)

Dear Mr. Palmisano:

On November 5, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed a baseline inspection at your Monticello Nuclear Generating Plant. The enclosed report documents the inspection findings which were discussed on November 5, 2004, with you and on November 22, 2004, with Mr. N. Haskell, and other members of your staff.

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

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Specifically, this inspection focused on the design and performance capability of the 250 Vdc and high pressure coolant injection systems.

Based on the results of this inspection, four NRC-identified findings of very low safety significance were identified, which involved violations of NRC requirements. However, because these violations were of very low safety significance and because they were entered into your corrective action program, the NRC is treating the issues 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 Plant. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's 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/

Julio Lara, Chief Electrical Engineering Branch Division of Reactor Safety Docket No. 50-263 License No. DPR-22 Enclosure: Inspection Report 05000263/2004007(DRS)

cc w/encl: J. Cowan, Executive Vice President and Chief Nuclear Officer Manager, Regulatory Affairs J. Rogoff, Vice President, Counsel, and Secretary Nuclear Asset Manager, Xcel Energy, Inc.

Commissioner, Minnesota Department of Health R. Nelson, President Minnesota Environmental Control Citizens Association (MECCA)

Commissioner, Minnesota Pollution Control Agency D. Gruber, Auditor/Treasurer, Wright County Government Center Commissioner, Minnesota Department of Commerce Manager - Environmental Protection Division Minnesota Attorney Generals Office

SUMMARY OF FINDINGS

IR 05000263/2004007(DRS); 10/18/2004 - 11/05/2004; Monticello Nuclear Generating Plant;

Safety System Design and Performance Capability.

The inspection was a three-week baseline inspection of the design and performance capability of the high pressure coolant injection and 250 Vdc systems. The inspection was conducted by regional engineering inspectors and a mechanical consultant. Four issues of very low safety significance were identified. The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green, or be assigned a severity level after NRC management review. The NRC's 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

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, in that, the design requirement to ensure the high pressure coolant injection (HPCI) pump discharge piping was kept full to maintain system operability was not adequately translated into procedures. Specifically, the effect of a known void in the HPCI discharge piping was not evaluated for its impact with the HPCI pump aligned with suction from the torus in the standby mode. As such, adequate acceptance criteria was not provided to ensure the operability of the HPCI system during this mode of operation. The licensees corrective actions included, as an interim action, placing a Temporary Information Tag on the control room switch for the HPCI suction valve from the condensate storage tank that states if HPCI suction is swapped to the torus, to evaluate HPCI for operability.

This finding was more than minor because it was associated with the attributes of configuration control and procedural quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the HPCI system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance based on the results of the SDP Phase 1 screening worksheet.

(Section 1R21.2.b.1)

Green.

The inspectors identified a Non-Cited Violation of Technical Specification 6.5.A.2,

Procedures, associated with an inadequate procedure to return the suction of the HPCI pump from the torus to the condensate storage tank during an anticipated transient without scram (ATWS) condition to ensure the self-cooled HPCI pump lube oil and control oil temperatures would remain within limits to prevent pump damage and ensure continued operation. The licensees corrective actions included a procedural change to allow continued operation of the HPCI system during an ATWS event.

This finding was more than minor because it was associated with the attribute of procedure quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the HPCI system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance based on the results of the SDP Phase 1 screening worksheet. (Section 1R21.2.b.2)

Green.

The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Action, associated with not promptly identifying and evaluating a condition adverse to quality. Specifically, the licensee did not replace aging electrolytic capacitors in the six Division I and Division II, 250 Vdc battery chargers, in a timely manner, allowing them to go beyond the service life specified by the vendor and the plants preventative maintenance (PM) program. In addition, routine PM activities for all six 250 Vdc battery chargers have not been performed since February 2000. The licensees corrective actions included: performing an operability evaluation; placing a purchase order for the capacitors; and initiating plans to replace the capacitors on an accelerated schedule.

The finding was more than minor because it was associated with the attribute of equipment performance, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the 250 Vdc system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance based on the results of the SDP Phase 1 screening worksheet.

(Section 1R21.2.b.3)

Green.

The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, regarding the emergency diesel generators ability to operate following a design basis tornado as portions of the exhaust and intake air piping located on the emergency diesel generator building roof were not adequately supported to withstand tornado wind forces. As part of the licensees corrective actions, the diesel exhaust piping was modified so that the piping design basis was met.

This finding was more than minor because it was associated with the attribute of design control, which affected the mitigating systems cornerstone objective of ensuring the capability of the emergency diesel generators to respond to natural phenomena to prevent undesirable consequences. The finding was of very low safety significance based on the results of an SDP Phase 3 analysis. (Section 4OA5.1)

Licensee-Identified Violations

None.

REPORT DETAILS

1R21 Safety System Design and Performance Capability

Introduction:

Inspection of safety system design and performance verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected systems to perform design bases functions. As plants age, the design bases may be lost and important design features may be altered or disabled. The plant risk assessment model is based on the capability of the as-built safety system to perform the intended safety functions successfully. This inspectable area verifies aspects of the mitigating systems cornerstone for which there are no indicators to measure performance.

The objective of the safety system design and performance capability inspection is to assess the adequacy of calculations, analyses, other engineering documents, and operational and testing practices that were used to support the performance of the selected systems during normal, abnormal, and accident conditions.

The systems and components selected were the high pressure coolant injection (HPCI)and 250 Vdc systems (two samples). These systems were selected for review based upon:

• having high probabilistic risk analysis rankings;

• considered high safety significant maintenance rule systems;

• not having received recent NRC review; and

• being complementary systems.

The criteria used to determine the acceptability of the systems performance was found in documents such as:

• licensee technical specifications;

• applicable updated safety analysis report (USAR) sections; and

• the systems' design documents.

The following system and component attributes were reviewed in detail:

System Requirements Process Medium - water; Energy Source - electrical power, steam, air; Control Systems - initiation, control, and shutdown actions; Operator Actions - initiation, monitoring, control, and shutdown; and Heat Removal - ventilation.

System Condition and Capability Installed Configuration - elevation and flow path operation; Operation - system alignments and operator actions; Design - calculations and procedures; and Testing - flow rate, pressure, temperature, voltage, and levels.

Component Level Equipment Qualification - temperature and radiation; and Equipment Protection - seismic and electrical.

.1 System Requirements

a. Inspection Scope

The inspectors reviewed the USAR, technical specifications, system design basis documents, lesson plans, drawings, and other available design basis information, as listed in the attached List of Documents, to determine the performance requirements of HPCI and 250 Vdc systems, and their associated support systems. The reviewed system attributes included process medium, energy sources, control systems, operator actions, and heat removal. The rationale for reviewing each of the attributes was:

Process Medium: This attribute required review to ensure that the HPCI system would supply the required amount of water to the reactor following normal transients and design basis events.

Energy Sources: This attribute needed to be reviewed to ensure that the HPCI system would start when called upon, and that appropriate valves would have sufficient power to change state when so required. This attribute also needed to be reviewed to ensure that the 250 Vdc system would provide sufficient power to the components it supplied.

Controls: This attribute required review to ensure that the automatic controls for the HPCI and 250 Vdc systems were properly established. Additionally, review of alarms and indicators was necessary to ensure that operator actions would be accomplished in accordance with the design.

Operations: This attribute was reviewed because the emergency operating procedures permitted the operators to manually control HPCI operation to maintain desired reactor water level. Therefore, operator actions played an important role in the ability of the HPCI system to achieve its functions.

Heat Removal: This attribute required review to ensure that the heat generated while the HPCI system was running can be effectively removed.

b. Findings

No findings of significance were identified.

.2 System Condition and Capability

a. Inspection Scope

The inspectors reviewed design basis documents and plant drawings, abnormal and emergency operating procedures, requirements, and commitments identified in the USAR and technical specifications. The inspectors compared the information in these documents to applicable electrical, instrumentation and control, and mechanical calculations, setpoint changes, and plant modifications. The inspectors also reviewed operational procedures to verify that instructions to operators were consistent with design assumptions.

The inspectors reviewed information to verify that the actual system condition and tested capability were consistent with the identified design bases. Specifically, the inspectors reviewed the installed configuration, the system operation, the detailed design, and the system testing, as described below.

Installed Configuration: The inspectors confirmed that the installed configuration of the HPCI and 250 Vdc systems met the design basis by performing detailed system walkdowns. The walkdowns focused on the installation and configuration of piping, components, and instruments; the placement of protective barriers and systems; the susceptibility to flooding, fire, or other environmental concerns; physical separation; provisions for seismic and other pressure transient concerns; and the conformance of the currently installed configuration of the systems with the design and licensing bases.

Operation: The inspectors performed a procedure walk-through of selected manual operator actions to confirm that the operators had the knowledge and tools necessary to accomplish actions credited in the design basis.

Design: The inspectors reviewed the mechanical, electrical, and instrumentation design of the HPCI and 250 Vdc systems to verify that the systems and subsystems would function as required under design conditions. This included a review of the design basis, design changes, design assumptions, calculations, boundary conditions, and models as well as a review of selected modification packages. Instrumentation was reviewed to verify appropriateness of applications and setpoints based on the required equipment function. Additionally, the inspectors performed limited analyses in several areas to verify the appropriateness of the design values.

Testing: The inspectors reviewed records of selected periodic testing and calibration procedures and results to verify that the design requirements of calculations, drawings, and procedures were incorporated in the system and were adequately demonstrated by test results. Test results were also reviewed to ensure automatic initiations occurred within required times and that testing was consistent with design basis information.

.1 HPCI System Void In Piping Not Analyzed When Initially Aligned with Suction from the

Torus

Introduction:

The inspectors identified a finding involving a Non-Cited Violation (NCV)of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) for failure to ensure the operability of the HPCI system when aligned with suction from the torus in the standby mode.

Description:

Technical Specification (TS) Bases 3.5/4.5 stated that the HPCI pump discharge piping was to be maintained full to prevent water hammer damage and to provide cooling at the earliest moment. The licensee previously identified that a steam void of approximately one cubic foot has existed since 1998 in the HPCI discharge piping just upstream of HPCI injection isolation valve MO-2068. This issue was discussed in various licensee corrective action program documents (CAP016153 and CAP016628 and their associated apparent cause evaluations and corrective action descriptions). The licensee evaluated the steam void and found, in part, the following:

(1) the void was caused by the fact that MO-2068 was at nearly feedwater temperature due to feedwater recirculation in the dead-end HPCI injection line to the feedwater system;

(2) water hammers have occurred as a result of this void, but no adverse effects have been detected on piping, valves, or supports;

(3) water hammer displacements were measured in mid-1998, and the displacements were used to perform piping analysis of the effect of the water hammer (the piping analysis results met American Society of Mechanical Engineers Code allowables);

(4) void size was self-limited to its current size by heat transfer conditions and HPCI discharge static pressure; and

(5) void size could only get larger if leakage were to occur past MO-2068 (the temperature of the piping on the HPCI side of MO-2068 was monitored for leakage on a weekly basis).

On October 21, 2004, the inspectors noted that the licensees evaluation of the void addressed the normal system alignment when the HPCI pump suction was aligned from the condensate storage tank (CST). The inspectors questioned whether the size/effect of the steam void in the HPCI pump discharge would be impacted with the HPCI pump suction aligned in standby condition from the torus, since this alignment would result in a lower static pressure in the HPCI discharge line. The licensee stated that the size/effect of the steam void had not considered with HPCI aligned with suction from the torus.

The inspectors also questioned whether the acceptance criterion contained in Procedure 1047-03, Operations Reactor Side Checklist Weekly Procedure, for the temperature used to monitor for the formation of steam voids in the HPCI discharge piping was low enough to alert operators of a potential change in void size when HPCI was aligned to the torus. The acceptance criterion used in Procedure 1047-03 was a maximum of 200 degrees Fahrenheit (oF). However, with HPCI aligned to the torus, the pressure at the location of the HPCI discharge temperature measurement could be as low as 5.6 pounds per square inch absolute (psia). The saturation temperature of a steam void at 5.6 psia would be approximately 165 oF. Consequently, the void size could grow and yet not be detected by the current method with HPCI aligned to the torus.

The licensee reviewed electronic operator logs that have been in effect since January 1, 2003, to determine the dates and duration of HPCI alignment to the torus. HPCI was aligned with suction from the torus (for CST level switch surveillance testing) on February 10, 2003, for a duration of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 5 minutes, on February 12, 2003, for a duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 30 minutes, and on February 13, 2004, for a duration of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 50 minutes with the HPCI system considered operable during those time periods.

There was also a HPCI alignment with suction from the torus (for maintenance associated with the 12 CST) between August 13, 2001, and August 17, 2001, for about 97 hours0.00112 days <br />0.0269 hours <br />1.603836e-4 weeks <br />3.69085e-5 months <br /> with the HPCI system considered operable.

The licensee entered this issue into the corrective action program as CAP035380.

As an interim action, the licensee issued a Temporary Information Tag on the control room switch for the HPCI suction valve from the CST, that stated if HPCI suction is swapped to the torus to evaluate HPCI for operability and to see CAP035380. The licensee was evaluating a recommendation from CAP035380 to issue a temporary procedure change to Ops Man B.03.02-05, Section G.3, Manual Switchover of HPCI Suction from Condensate Storage Tanks to the Torus, to require entry into the limiting condition for operation for an inoperable HPCI system when this procedure section was entered.

Analysis:

The inspectors determined that the failure to ensure the operability of the HPCI system with a known void when aligned in standby mode with suction from the torus was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, because it was associated with the attributes of configuration control and procedure quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the HPCI system to respond to initiating events to prevent undesirable consequences. Inadequate acceptance criteria to ensure the operability of the HPCI system when initially aligned with suction from the torus could potentially render the HPCI system incapable of performing its required safety function.

The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding screened as Green because it was not a design issue resulting in loss of function per Generic Letter (GL) 91-18, did not represent an actual loss of a systems safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, procedures, and instructions.

Contrary to this requirement, on October 22, 2004, inspectors identified that the design requirement to ensure the HPCI pump discharge piping was maintained full was not correctly translated into specifications, procedures, and instructions. Specifically, the effect of a known void in the HPCI discharge piping was not evaluated for its impact on the HPCI system with the HPCI pump aligned with suction from the torus in the standby mode. As a result, adequate acceptance criteria to ensure the operability of the HPCI system when aligned with suction from the torus in standby was not provided. In addition, at various times between August 13, 2001, through August 17, 2001, on February 10, 2003, on February 12, 2004, and on February 13, 2004, the HPCI system was aligned with suction from the torus with the HPCI system considered operable.

However, because this violation was of very low safety significance and because the issue was entered into the licensees corrective action program (CAP035380), this violation is being treated as a NCV, consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000263/2004007-01). As part of its corrective actions, the licensee issued a Temporary Information Tag to evaluate HPCI for operability prior to operating the HPCI suction valve from the CST.

.2 Inadequate Procedure to Ensure HPCI Function During Anticipated Transient Without

Scram (ATWS) Event

Introduction:

The inspectors identified a finding involving a NCV of TS 6.5.A.2, Procedures, having very low safety significance (Green) for the failure to have an adequate procedure to ensure the continued operation of the HPCI system to maintain reactor pressure vessel (RPV) water level during an ATWS condition in accordance with analysis assumptions.

Description:

On October 21, 2004, the inspectors identified that procedure Ops Man B.03.02-05, Section G.3, Manual Switchover of HPCI Suction from the Torus to the Condensate Storage Tanks, which was used to support actions in ATWS emergency operating procedure (EOP) C.5-2007, Failure to Scram, was inadequate in that the procedure did not ensure the continued operation of the HPCI system to maintain RPV water level during an ATWS condition in accordance with analysis assumptions. Specifically, the procedure would not adequately direct the operators to return the suction of the HPCI pump from the torus to the CST (if the HPCI pump suction had automatically swapped to the torus on high torus level during an ATWS event) to ensure the self-cooled HPCI pump lube oil temperatures would remain within limits to prevent pump damage and ensure continued operation.

During an ATWS event, of the torus water temperature will increase as high reactor pressure will cause the safety relief valves to lift discharging steam into the torus. In the limiting ATWS case, as identified in General Electric letter GLN-99-011, Revised ATWS Evaluation for Monticello Extended Power Uprate Project, the torus water temperature was calculated to be as high as 190oF based on the licensees plant specific re-rate analysis. Monticello has a design feature that automatically swaps the HPCI pump suction from the CSTs to the torus on a high torus level of +2.0 inches. The licensee determined that during the limiting ATWS case, the high torus level transfer of HPCI suction to the torus could occur within a few minutes following event initiation. While C.5-2007 allowed the bypass of the HPCI high torus level suction transfer logic and stated that the CST was the preferred source for the HPCI pump, the licensee noted that due to the short time frame involved (a few minutes), it was likely that a transfer of the suction source for HPCI to the torus would occur during the limiting ATWS case.

Ops Man C.5.1-2007 stated that if the HPCI automatic high torus level transfer occurred, the suction of the HPCI pump should be returned to the CST as soon as practicable. Detail J of C.5-2007 contained a caution that stated, Exceeding 180EF suction temperature may damage system, which was based on the HPCI pump being self-cooled.

The transfer of HPCI suction from the torus back to the CSTs would be accomplished using procedure Ops Man B.03.02-05, Section G.3. However, Prerequisite 1 required torus level to be less than +2 inches. This prerequisite represented a conflict on the use of this procedure during an ATWS event if the automatic transfer of HPCI suction had already occurred on high torus level, since torus level would already be above +2 inches. Thus, the procedure to transfer HPCI suction back to the CSTs was inadequate for this condition. Since the torus water temperature was calculated in the licensees plant specific re-rate analysis to be as high as 190oF, the caution in C.5-2007 would have required the HPCI pump to be shutdown when torus temperature exceeded 180oF.

As a result of this issue, the licensee issued a temporary procedure change (Volume F Memo No. 2174, dated October 21, 2004) to Ops Man B.03.02-05, Section G.3, to change Prerequisite 1 to allow entry into the procedure if the HPCI high torus water level suction transfer was bypassed per the EOPs (even if torus water level is high). The HPCI high torus level suction transfer can be bypassed per procedure Ops Man C.5-3202, Bypass HPCI Signals.

Analysis:

The inspectors determined that the failure to provide an adequate procedure to ensure the continued operation of the HPCI system for maintaining RPV water level during an ATWS condition in accordance with analysis assumptions was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Disposition Screening, because it was associated with the attribute of procedure quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the HPCI system to respond to initiating events to prevent undesirable consequences.

The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding screened as Green because it was not a design issue resulting in loss of function per GL 91-18, did not represent an actual loss of a systems safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation.

Enforcement:

Technical Specification 6.5.A.2, Procedures, required, in part, that written procedures be established, implemented, and maintained covering the EOPs required to implement the requirements of NUREG-0737, Clarification of TMI Action Plan Requirements, and NUREG-0737, Supplement 1. NUREG-0737, Item I.C.1, and NUREG-0737, Supplement 1, Section 7, required, in part, that the EOPs cover multiple failure events including an ATWS event.

Contrary to this requirement, on October 21, 2004, it was discovered that procedure Ops Man B.03.02-05, Section G.3, which was used to support actions in ATWS EOP C.5-2007, was inadequate to return the suction of the HPCI pump from the torus to the CST to ensure the self-cooled HPCI pump lube oil and control oil temperatures would remain within limits to prevent pump damage and ensure continued operation.

However, because this violation was of very low safety significance and because the issue was entered into the licensees corrective action program (CAP035344), this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000263/2004007-02). As part of its corrective actions, the licensee issued a temporary procedure change to resolve this concern.

.3 Missed Preventive Maintenance (PM) Activities for the 250 Vdc Battery Chargers

Introduction:

The inspectors identified a finding involving an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, having very low safety significance (Green) for the failure to identify a condition adverse to quality in that the licensee did not promptly evaluate and implement the vendor recommended and licensee specified periodic replacement of electrolytic filter capacitors and other required periodic PM activities for the six Division I and Division II, 250 Vdc battery chargers.

Description:

The inspectors determined that the Division I (D52, D53, D54) and Division II (D70, D80, D90), 250 Vdc battery chargers routine PM requirements specified in procedures 4525-PM, No. 13 and 14 Battery Charger Preventive Maintenance, and EWI-10.01.01, Electronic Component Aging Management Process Implementation, were not accomplished within the periodicity specified in the procedures. Specifically, Step 4 of the Prerequisites in Procedure 4525-PM stated that the replacement of battery charger capacitors was required on a 7 to 10 year interval.

The inspectors noted that the licensees PM schedule was revised in December 2002 to require replacement of battery charger printed circuit boards and filter capacitors every third cycle. The inspectors determined that the Division I electrolytic capacitors were last replaced in June 1993, and the Division II electrolytic capacitors were last replaced in July 1994, allowing them to go beyond the service life specified by the plants PM program (6 years) and the vendor (7 to 10 years). No documented evaluation or assessment for deferral of this activity was available for review.

In addition, the inspectors determined that the routine PM activities required by Procedure 4525-PM, which included maintenance activities such as verification and adjustments of phase wave form, input and load currents, current limit setpoints, high voltage shutdown setpoint, and float and equalize voltage setpoints were not accomplished within the specified periodicity. The procedure specified that the PM activities be performed every cycle, however, the inspectors determined that the required PM was last performed in February 2000. No documented evaluation or assessment for deferral of this activity was available for review. Also, the vendor informed the licensee that routine checking of the charger ripple levels can be used to determine when the capacitors were reaching end of life.

The inspectors also noted that the C and D Batteries Division vendor manual NX-16647 stated that the capacitors shelf life was limited and normally will not exceed 1 year without being recharged on an annual basis up to five years. The inspectors determined that this was not being accomplished for the spare battery chargers D53 and D80, which might be on standby for an extended period of time.

The inspectors determined that Work Orders 0200581, 0200582, 0200583, 0200585, 0200586, and 0200587 were written in January 2002, to perform the specified routine battery charger PM activity including replacement of selected battery components like the electrolytic capacitors. However, at the time of the inspection, no replacements had occurred and no capacitors have been ordered. The inspectors noted that 4 AWI-05.07.02, Preventive Maintenance Program, stated in section 4.11.3 that all due date deviations and/or deferrals should be documented on Form 3488. Section 4.11.4 of the procedure stated that if a PM task will not be completed by the 25 percent plus grace period, a condition report to document the missed due date should be initiated. None of these requirements were accomplished. Since this issue was not entered into the corrective action program, an evaluation of the capacitor service life was not performed.

In response to questions from the inspectors regarding this issue, the licensee initiated CAP035589 on November 3, 2004, and performed an operability evaluation to assess battery charger operability. The determination concluded that the battery chargers were operable, based on no known failed components or equipment associated with this condition and that the battery condition was being monitored by another surveillance.

The inspectors concluded that the licensees operability evaluation was adequate, but had not been conducted until the inspectors questioned the operability of the battery chargers.

Analysis:

The inspectors determined that the licensee had not followed the replacement frequency of its electrolytic capacitors and had not accomplished the required routine PM activities required for all six Division I and Division II, 250 Vdc battery chargers, which was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Disposition Screening, because it was associated with the attribute of equipment performance, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the 250 Vdc system to respond to initiating events to prevent undesirable consequences.

The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding screened as Green because it was not a design issue resulting in loss of function per GL 91-18, did not represent an actual loss of a systems safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criteria XVI, Corrective Action, required, in part, that measures be established to assure that conditions adverse to quality, such as deficiencies and defective material and equipment, were promptly identified and corrected.

Contrary to the above, on November 3, 2004, the licensee operated the plant with battery chargers electrolytic capacitors that were beyond their failure-based service life, and the specified routine PM activities on the battery chargers passed their due dates without a documented evaluation or assessment for deferral of this activity, a condition adverse to quality. However, because this violation was of very low safety significance and because the issue was entered into the licensees corrective action program (CAP035589), this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000263/2004007-04). As part of its corrective actions, the licensee ordered replacement capacitors and plans to install the capacitors on an accelerated schedule.

.3 Components

a. Inspection Scope

The inspectors examined the HPCI and 250 Vdc systems to ensure that component level attributes were satisfied. Specifically, the following attributes of the HPCI and 250 Vdc systems were reviewed:

Equipment/Environmental Qualification: This attribute verifies that the equipment is qualified to operate under the environment in which it is expected to be subjected to under normal and accident conditions. The inspectors reviewed design information, specifications, and documentation to ensure that the HPCI and 250 Vdc components were qualified to operate within the temperatures and radiation fields specified in the environmental qualification documentation.

Equipment Protection: This attribute verifies that the HPCI and 250 Vdc systems are adequately protected from natural phenomenon and other hazards, such as high energy line breaks, floods or missiles. The inspectors reviewed design information, specifications, and documentation to ensure that the HPCI and 250 Vdc systems were adequately protected from those hazards identified in the USAR which could impact their ability to perform their safety function.

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution

.1 Review of Condition Reports

a. Inspection Scope

The inspectors reviewed a sample of HPCI and 250 Vdc system problems that were identified by the licensee and entered into the corrective action program. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.

b. Findings

Section 1R21.2.b.3 described that vendor and licensee specified periodic replacement of electrolytic filter capacitors and other required periodic PM activities for the 250 Vdc battery chargers that were not performed and the licensee had not entered the issue into the corrective action program. Consequently, the concern was never fully evaluated.

4OA5 Other

The inspectors reviewed items discussed in previous inspection reports to determine if further regulatory action was required to be taken.

.1 (Closed) Unresolved Item 05000263/2003002-10: Inadequate Diesel Generator

Exhaust Piping Protection Against Natural Phenomena (Tornadoes).

a. Inspection Scope

Unresolved Item 05000263/2003002-10 identified that the licensee was unable to provide documentation to confirm that combustion air intake and exhaust piping would not be adversely affected by design basis tornado wind loadings. Based on the absence of design calculations and the incomplete probabilistic risk analysis, the inspectors were unable to evaluate the effect on the emergency diesel generator operation. This item was left unresolved pending licensee preparation of a calculation to ensure the diesel generators could perform their safety function. In followup to the unresolved item, the inspectors reviewed the licensee's operability calculation and performed the risk analysis to evaluate the as-found condition.

b. Findings

Introduction:

The inspectors identified a finding involving a NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance (Green) for failure to adequately design the diesel generator exhaust silencers to withstand the design basis tornado wind loading. As part of resolving Unresolved Item 05000263/2003002-10, the licensee performed analyses and modified the diesel exhaust piping so that the design basis was met.

Description:

During the 2003 safety system design and performance capability inspection, the NRC identified that the diesel generator exhaust piping did not appear to be protected against tornado winds. In addition, the inspectors reviewed a 1992 internal memorandum that noted portions of the exhaust and intake air piping located on the emergency diesel generator building roof were not adequately supported to withstand tornado wind forces. As of the 2003 inspection, the licensee had taken no action to rectify the deficient condition. Following this issue being identified by the NRC, the licensee performed an operability calculation (CA-03-030) and determined that extensive modifications were necessary to bring the exhaust piping back into design conformance.

Although the operability calculation concluded the exhaust piping was always operable, the inspectors identified a number of significant deficiencies within the calculation. Of greatest importance was that the licensee's model would not converge for a specific node. To resolve this problem, the licensee performed a hand calculation. The NRC determined that the model was inadequate and the licensee's hand calculation overly simplistic. The inspectors determined that during a design basis tornado, it was likely that the exhaust piping would bend or crimp, stalling both the diesel generators. As part of the licensees corrective actions, the diesel exhaust piping was modified so that the design basis was met.

Analysis:

Evaluation of this issue concluded that it was a performance deficiency resulting in a finding of very low safety significance (Green). The performance deficiency was that, by the inspectors assessment, the diesel generators would not have been able to perform their safety-related function during a design basis tornado.

The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Disposition Screening, because it was associated with the attribute of design control, which affected the mitigating systems cornerstone objective of ensuring the capability of the diesel generators to respond to natural phenomena to prevent undesirable consequences. No other cornerstones were determined to be degraded as a result of this issue.

The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening and determined that the diesel generator exhaust piping not being designed to withstand a design basis tornado was not a design issue resulting in loss of function per GL 91-18, did not represent an actual loss of a systems safety function, did not result in exceeding a TS allowed outage time, and did screen as potentially risk significant due to a severe weather initiating event.

Because the issue screened as potentially risk significant due to a severe weather initiating event, the inspectors contacted a senior reactor analyst to perform a Phase 3 analysis. The following information presents the results of that analysis.

Method of

Analysis:

Condition assessment and sensitivity analyses using the Idaho National Engineering and Environmental Laboratory (INEEL) Graphical Evaluation Module (GEM) software and the Standardized Plant Analysis Risk (SPAR) Revision 3i model for Monticello plant.

Assumptions:

(a) Duration time = 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of direct tornado impact;

(b) Probability truncation = 1E-15; and

(c) Diesel generators 11 and 12 become unavailable and no recovery of the diesel generators is assumed during the tornado impact. Additional assumptions regarding number of tornadoes and affected square mileage were obtained from the licensee.

Model Changes:

(a) Initiating event frequency estimates for Loss of Offsite Power scenario were modified for the three cases of tornado event frequency estimates; and

(b) All other initiating events and their probabilities were set to "FALSE" logic, and zero probabilities.

Results: The conditional core damage probability (CCDP) and conditional core damage frequency (CDF) estimates, assuming the plant is operating 85 percent of the time during the year, in the three case studies are summarized as follows:

A.

33 events of F4 winds for an impact area of 0.25 sq. mi., 54 years of data, 79,610 square miles; F4 initiating frequency = (33 x 0.25)/(54 x 79,610) = 1.9E-6 CCDP = 3.8E-11 per hour; Conditional CDF = (3.8E-11x 0.85 x 8760) = 2.8E-7 B.

6 events of F5 winds for an impact area of 0.5 square miles, 54 years of data, 79,610 square miles; F5 initiating frequency = (6 x 0.5)/(54 x 79,610) = 7.0E-7 CCDP = 1.4E-11 per hour; Conditional CDF = (1.4E-11x 0.85 x 8760) = 1.0E-7 C.

33 events of F4 winds for an impact area of 0.25 square miles and 6 events of F5 winds for an impact area of 0.5 square miles, 54 years of data, 79,610 square miles; Combined F4 and F5 initiating frequency = (33 x 0.25)/(54 x 79,610) + (6 x 0.5)/

(54 x 79,610) = 2.6E-6 CCDP = 5.2E-11 per hour; Conditional CDF = (5.2E-11x 0.85 x 8760) = 3.9E-7 The above results show GREEN, or very low, risk significance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"

required, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of March 7, 2003, the licensee failed to assure that the design basis for the emergency diesel generators was correctly translated into specifications, drawings, procedures, and instructions. Specifically, the diesel generator exhaust piping was not designed to withstand the 300 mile per hour wind loadings of a design basis tornado. However, because this violation was of very low safety significance and because the issue was entered into the licensees corrective action program (condition report 03001909), the issue is being treated as a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000263/2004007-04). As part of its corrective actions, the licensee modified the diesel exhaust piping to meet the design basis requirement.

The unresolved item is closed.

4OA6 Meetings, Including Exits

.1 Exit Meeting

The inspectors presented the inspection results to Mr. T. Palmisano and other members of licensee management at the conclusion of the inspection on November 5, 2004. A follow-up telephone exit was held on November 22, 2004, with Mr. N. Haskell. The inspectors determined that proprietary information was reviewed during the inspection and returned to the licensee at the close of the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Palmisano, Site Vice President

J. Purkis, Plant Manager

R. Baumer, Licensing

S. Brown, Engineering Programs Manager

J. Grubb, Business Support Manager

N. French, Plant Engineering Supervisor

S. Hammer, Principal Engineer, Operations

N. Haskell, Design Engineering Manager

B. MacKissock, Operations Manager

R. Neulk, System Engineer

D. Neve, Regulatory Affairs Manager

R. Olsen, General Supervisor Electrical and I&C Maintenance

D. Pennington, HPCI System Engineer

S. Porter, Engineering Supervisor

D. Seestrom, 250 Vdc System Engineer

S. Sharp, Director of Engineering

A. Stover, Nuclear Oversight Manager

A. Williams, Projects Manager

D. Zercher, Design Engineer

Nuclear Regulatory Commission

J. Lara, Chief, Electrical Engineering Branch, Division of Reactor Safety

S. Burton, Senior Resident Inspector

Attachment

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000263/2004007-01 NCV Failure to Provide Adequate Guidance to Ensure the Operability of the HPCI System When Aligned with Suction from the Torus (Section 1R21.2.b.1)

05000263/2004007-02 NCV Failure to Provide Adequate Procedural Guidance to Ensure the Continued Operation of the HPCI System During an ATWS (Section 1R21.2.b.2)

05000263/2004007-03 NCV Failure to Evaluate and Implement the Replacement of Electrolytic Capacitors (Section 1R21.2.b.3)

05000263/2004007-04 NCV Failure to Design Emergency Diesel Generator Exhaust Silencers for Tornado Wind Loading (Section 4OA5.1)

Closed

05000263/2003002-10 URI Effect of Tornado Wind Loading on Emergency Diesel Generator Exhaust Silencers (Section 4OA5.1)

05000263/2004007-01 NCV Failure to Provide Adequate Guidance to Ensure the Operability of the HPCI System When Aligned with Suction from the Torus (Section 1R21.2.b.1)

05000263/2004007-02 NCV Failure to Provide Adequate Procedural Guidance to Ensure the Continued Operation of the HPCI System During an ATWS (Section 1R21.2.b.2)

05000263/2004007-03 NCV Failure to Evaluate and Implement the Replacement of Electrolytic Capacitors (Section 1R21.2.b.3)

05000263/2004007-04 NCV Failure to Design Emergency Diesel Generator Exhaust Silencers for Tornado Wind Loading (Section 4OA5.1)

Attachment