IR 05000373/2004004

IR 05000373-04-004; 05000374-04-004, on 07/01/2004 - 09/30/2004; LaSalle County Station, Units 1 & 2; Access Control to Radiologically Controlled Areas

ML043010138
Person / Time
Site:	LaSalle
Issue date:	10/26/2004
From:	Burgess B NRC/RGN-III/DRP/RPB2
To:	Crane C Exelon Generation Co
References
IR-04-004
Download: ML043010138 (40)

Text

October 26, 2004

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000373/2004004; 05000374/2004004

Dear Mr. Crane:

On September 30, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your LaSalle County Station, Units 1 and 2. The enclosed report documents the results of this inspection, which were discussed on September 30, 2004, with the Site Vice President, Mr. G. Barnes, 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. Specifically, this inspection focused on reactor safety and radiation protection.

Based on the results of this inspection, there was one finding of very low safety significance identified by our inspectors, and one associated violation of NRC requirements. However, because the violation was non-willful and non-repetitive and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. Additionally, one licensee identified violation of very low safety significance is documented in Section 4OA7 of this report.

If you contest the subject or severity of the Non-Cited Violation in this report, 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, D.C. 20555-0001; with copies 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, D.C. 20555-0001; and the NRC Resident Inspectors Office at the LaSalle County Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Bruce L. Burgess, Chief Branch 2 Division of Reactor Projects Docket Nos. 50-373; 50-374 License Nos. NPF-11; NPF-18

Enclosure:

Inspection Report 05000373/2004004; 05000374/2004004 w/Attachment: Supplemental Information

REGION III==

Docket Nos:

50-373; 50-374 License Nos:

NPF-11; NPF-18 Report No:

05000373/2004004; 05000374/2004004 Licensee:

Exelon Generation Company, LLC Facility:

LaSalle County Station, Units 1 and 2 Location:

2601 N. 21st Road Marseilles, IL 61341 Dates:

July 1 through September 30, 2004 Inspectors:

D. Kimble, Senior Resident Inspector D. Eskins, Resident Inspector M. Miller, Project Engineer, Branch 2 M. Mitchell, Radiation Protection Specialist M. Sheikh, Resident Inspector - Dresden Station H. Walker, Engineering Inspector R. Winter, Engineering Inspector J. Roman, Illinois Dept. of Emergency Management J. Yesinowski, Illinois Dept. of Emergency Management Observers:

M. Franke, Inspector-in-Training Approved by:

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

Enclosure

SUMMARY OF FINDINGS

IR 05000373/2004004, 05000374/2004004; 07/01/2004 - 09/30/2004; LaSalle County Station,

Units 1 & 2; Access Control to Radiologically Significant Areas.

This report covers a 3-month period of baseline resident inspection and an announced baseline inspection in radiation protection. The inspection was conducted by resident inspectors and Region III inspectors. One Green finding and one associated Non-Cited Violation were identified. 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-Revealing Findings

Cornerstone: Occupational Radiation Safety

Green.

The inspectors identified a Green finding and associated Non-Cited Violation (NCV) when they observed operations personnel entering a posted neutron dose area without proper neutron monitoring, contrary to the licensee's Technical Specifications. This finding was considered NRC-identified as radiation protection personnel were unaware of this issue until questions by inspectors indicated a lack of proper neutron dose control for both this event and similar past occurrences.

The cause of the error was a failure of communication between the operations and radiation protection (RP) staff. The finding, under the Occupational Radiation Safety Cornerstone, does not involve the application of traditional enforcement because it did not result in actual safety consequences or potential to impact the NRCs regulatory function, and was not the result of any willful actions. The finding was more than minor as it involves the failure of the licensee to adhere to procedures to monitor and control radiation exposure, a key attribute under the objective of the radiation safety cornerstone to ensure adequate protection of worker health and safety from exposure to radiation. The finding is of very low safety significance because the personnel involved were using electronic dosimeters that alarm to warn workers of higher than expected dose rates or accumulated dose. The issue was a Non-Cited Violation of Technical Specifications 5.4.1(a), which requires written procedures be established, implemented, and maintained in accordance with the requirements of Regulatory Guide 1.33. Section 7.e(7). of Regulatory Guide 1.33 lists the requirement for radiation protection procedures for personnel monitoring.

RP-AA-210, "Dosimetry Issue, Usage, and Control, is the plant procedure governing neutron dose estimation and monitoring.

The licensee conducted a human performance investigation to determine the cause of the event and identified a failure of communication between the RP and operation staffs. The individuals involved were coached, site personnel were informed of the event, and RP staff personnel were provided additional training on the requirements for entering neutron areas. (Section 2OS1)

Licensee-Identified Violations

A violation of very low safety significance that was identified by the licensee has been reviewed by inspectors. Corrective actions planned or taken by the licensee have been entered into the licensees corrective action program. The violation and corrective action tracking number are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 The unit began the inspection period operating at full power. On the following three occasions power was reduced for several hours at the request of the system load dispatcher. In each case, the unit returned to full power operation later that day:

• July 8, 2004: Power was reduced to approximately 85 percent.

• July 18, 2004: Power was reduced to approximately 95 percent.

• August 30, 2004: Power was reduced to approximately 85 percent.

On September 4, 2004, power was reduced to approximately 48 percent to facilitate a control rod sequence exchange, main steam valve testing, and emergent repairs to a leaking electro-hydraulic control line for the No. 4 main turbine combined-intermediate valve. Full power operation was resumed on September 5, 2004, and the unit remained operating at or near full power for the remainder of the inspection period.

Unit 2 The unit began the inspection period operating at full power. On September 12, 2004, power was reduced to approximately 58 percent to facilitate a control rod sequence exchange, main steam valve testing, and repairs to a leaking valve in the heater drain system. Full power operation was resumed on September 13, 2004, and the unit remained operating at or near full power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R01 Adverse Weather

Review of Site Specific Weather Condition - Tornado Warning/ High Winds

a. Inspection Scope

The inspectors performed a walkdown of the licensees preparations and actions for adverse weather, including conditions that could lead to loss of off-site power and other conditions that could result from high winds or tornado-generated missiles during a tornado warning and high wind condition that occurred on the afternoon of July 13, 2004. The inspectors focused on the licensees procedures and plant specific design features used to mitigate or respond to an approaching tornado and high wind condition.

This review constituted a single inspection sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Semiannual Complete System Alignment Verification

a. Inspection Scope

Due to their high risk significance, the inspectors selected the Unit 1 and Unit 2 emergency diesel generators (EDGs) for complete alignment verifications. The inspectors walked down the EDGs to verify mechanical and electrical equipment lineups, 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.

The inspectors review of EDG alignment constituted a single inspection sample.

b. Findings

No findings of significance were identified.

.2 Quarterly Partial System Alignment Verifications

a. Inspection Scope

The inspectors performed a partial alignment verification of the following equipment trains:

• 1A and 1B EDGs during a Division 1 work week

• Unit 1 high pressure core spray (HPCS) during reactor core isolation cooling (RCIC) system work Operability and proper equipment lineup were verified by physical walk downs of the selected equipment. These systems/trains were selected based upon risk significance, plant configuration, system work or testing, or inoperable or degraded conditions on redundant equipment. The inspectors verified the position of critical redundant components and looked for any discrepancies between the existing equipment lineups and the required lineups.

These partial equipment alignment verifications constituted two inspection samples.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

Quarterly Fire Protection Zone Inspections

a. Inspection Scope

The inspectors walked down the following risk significant areas looking for any fire protection issues:

• Fire zone 4E2; Unit 2 auxiliary equipment room - elevation 731'0"

• Fire zone 4E3; Unit 1 Division 2 essential switchgear room - elevation 731'0"

• Fire zone 4E4; Unit 2 Division 2 essential switchgear room - elevation 731'0"

• Fire zone 4F1; Unit 1 Division 1 essential switchgear room - elevation 710'6"

• Fire zone 4F2; Unit 2 Division 1 essential switchgear room - elevation 710'6"

• Fire zone 5D1; Unit 1 HPCS switchgear zone - elevation 687'0"

• Fire zone 5D1; Unit 2 HPCS switchgear zone - elevation 687'0"

• Fire zone 7C5; Division 2 Residual Heat Removal (RHR) service water pump room - elevation 674'0"

• Fire zone 7C6; Division 1 RHR service water pump room - elevation 674'0" The inspectors selected areas containing systems, structures, or components that the licensee identified as important to reactor safety. The inspectors reviewed the control of transient combustibles and ignition sources, fire detection equipment, manual suppression capabilities, passive suppression capabilities, automatic suppression capabilities, barriers to fire propagation, and any compensatory measures the licensee had enacted due to degraded fire protection features.

These reviews constituted nine inspection samples.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

The inspectors observed a training crew during an evaluated simulator scenario and reviewed licensed operator performance in mitigating the consequences of events. The scenario included several failed instruments as distractions, a failure in the offgas system that forced a power reduction, and a reactor scram complicated by a stuck open safety-relief valve. The evaluated drill scenario also resulted in the declaration of an Alert emergency plan classification. Areas observed by the inspectors included: clarity and formality of communications, timeliness of actions, prioritization of activities, procedural adequacy and implementation, control board manipulations, managerial oversight, emergency plan execution, and group dynamics.

The observation of this simulator scenario constituted a single inspection sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the licensee's handling of performance issues and the associated implementation of the Maintenance Rule (10 CFR 50.65) to evaluate maintenance effectiveness for the selected systems. The following systems were selected based on being designated as risk significant under the Maintenance Rule, being in the increased monitoring (Maintenance Rule category a(1)) group, or due to an issue or problem that potentially impacted system work practices, reliability, or common cause failures:

• Unit 2 control rod position indication system (RPIS)

• Unit 1 control rod position indication system

• Unit 1 reactor core isolation cooling (RCIC) system F028 check valve The inspectors review included verification of the licensee's categorization of specific issues. These involved evaluation of the performance criteria, appropriate work practices, identification of common cause errors, extent of condition, and trending of key parameters. Additionally, the inspectors reviewed the licensee's implementation of the Maintenance Rule requirements, including a review of scoping, goal-setting, performance monitoring, short-term and long-term corrective actions, functional failure determinations associated with the condition reports reviewed, and current equipment performance status.

These quarterly Maintenance Rule effectiveness reviews constituted three inspection samples.

b. Findings

No findings of significance were identified. One unresolved item (URI) was identified.

On August 2, 2004, the licensee identified that previous local leak rate tests (LLRTs)performed on RCIC system containment isolation check valves 1(2)E51-F028 to satisfy the requirements of 10 CFR 50, Appendix J, were invalid. During a routine review, the licensee had determined that a head correction calculation used to establish test pressure was in error, and that the test pressure that both the Unit 1 and Unit 2 F028 valves had been subjected to during their last refueling outage testing was outside the specified value and nonconservative. The licensee immediately declared the 1(2)E51-F028 check valves inoperable, and shut and deenergized the companion 1(2)E51-F069 motor-operated containment isolation valves for that containment penetration on each unit to comply with Technical Specifications.

On September 9-10, 2004, the licensee performed LLRTs on both units RCIC F028 check valves in series attempts to restore the valves to an operable status. The Unit 2 E51-F028 check valve was tested satisfactorily and the Unit 2 RCIC system was returned to a normal alignment with the 2E51-F069 motor-operated valve (MOV)energized and open. The Unit 1 E51-F028 check valve was determined during its LLRT to be stuck open, and disassembly and repair were required. On September 17, 2004, the licensee disassembled the 1E51-F028 containment isolation check valve. Licensee maintenance personnel, as well as inspectors observing the work activity, noted a significant amount of corrosion product buildup inside the valve, to the extent that maintenance technicians had to twist the valves plug repeatedly in order to facilitate its removal. The corrosion product buildup was cleaned up in accordance with licensee maintenance procedures for check valve overhauls, and the valves plug and operating spring were replaced. Later that day, the licensee preformed a LLRT on the 1E51-F028 check valve and obtained satisfactory results. The 1E51-F028 check valve was declared operable, and the Unit 1 RCIC system was subsequently returned to a normal alignment with the 1E51-F069 MOV energized and open.

Inspectors following up on the 1E51-F028 LLRT failure have identified numerous similar failures dating back to at least 1991. In each case, it appears that the licensee performed fairly similar corrective actions to obtain satisfactory LLRT results and return the valve to service. The 2E51-F028 check valve, although identical in design, installed in a similar configuration on Unit 2, and subjected to the same kind of service environment, does not appear to have the same problematic performance history.

Inspectors are presently examining the 1E51-F028 performance history against the requirements of the Maintenance Rule (10 CFR 50.65), and against the licensees quality assurance program corrective action requirements (10 CFR 50, Appendix B, Criterion XVI). The issue is considered unresolved pending the inspectors receipt and review of the licensees equipment apparent cause evaluation (EACE), which is in progress. The licensee has entered this issue into their corrective action program as CR 253839. (URI 05000373/2004004-01)

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

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

• Unit 1 A and B reactor building closed cooling water pump (RBCCW) repairs

• Unit 2 RPIS troubleshooting and repairs

• Unit 1 A service water pump maintenance and repairs

• Unit 1 core standby cooling system (CSCS) service water tunnel robotic and diver inspections

• Unit 1 main generator output breaker 9-10 troubleshooting

• Unit 1 main generator fault (GIX) relay problems

• Unit 1 RPIS troubleshooting and repairs

• 2B EDG unexpected trip during unloading following surveillance testing The inspectors also reviewed the licensee's evaluation of plant risk, risk management, scheduling, and configuration control for these activities in coordination with other scheduled risk significant work. The inspectors verified that the licensee's control of activities considered assessment of baseline and cumulative risk, management of plant configuration, control of maintenance, and external impacts on risk. In-plant activities were reviewed to ensure that the risk assessment of maintenance or emergent work was complete and adequate, and that the assessment included an evaluation of external factors. Additionally, the inspectors verified that the licensee entered the appropriate risk category for the evolutions.

These reviews constituted eight inspection samples.

b. Findings

No findings of significance were identified.

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

a. Inspection Scope

The inspectors monitored the execution of the following deep power reductions and control rod sequence exchanges during the inspection period:

• Unit 1 load drop and control rod sequence exchange, September 4-5, 2004

• Unit 2 load drop and control rod sequence exchange, September 12, 2004 The inspectors reviewed operator and reactor engineering performance during periods of power maneuvering, and verified that personnel actions were in accordance with approved plant procedures. Additionally, the inspectors reviewed the changes to the stations on-line risk profile that resulted from the events.

The inspectors review of these evolutions constituted two inspection samples.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the technical adequacy of the following operability evaluations to determine the impact on Technical Specifications, the significance of the evaluations, and to ensure that adequate justifications were documented:

• EDG ventilation system configuration during damper temperature controller replacement

• CSCS pump room maximum temperatures and ventilation system operation

• Divisional wire separation between redundant trains of stack wide-range gas monitoring instrumentation

• RCIC operation without the barometric condenser vacuum pump

• Unit 1 RPIS degraded condition with intermittent fault

• Lisega snubbers nonconforming condition

• Unit 1 primary containment chiller system dual loop operation

• Auxiliary power transformer 236Y degraded cooling fans Operability evaluations were selected based upon the relationship of the safety-related system, structure, or component to risk.

These reviews constituted eight inspection samples.

b. Findings

No findings of significance were identified.

1R16 Operator Workarounds

a. Inspection Scope

The inspectors reviewed the following operator workarounds to determine each workarounds potential to increase the possibility of an initiating event, affect multiple mitigating systems, or impact the operators ability to respond to accidents or transients:

• Unit 1 and Unit 2 RCIC F028 check valve inoperability

• Unit 1 rod worth minimizer and RPIS system trips These reviews constituted two inspection samples.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

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

• Unit 1 & Unit 2 RCIC system LLRT for check valve F028

• 1A RBCCW pump testing following planned maintenance

• Unit 2 RPIS probe data processing card testing following replacement

• Unit 1 RPIS multiplexer card testing following replacement of 10 multiplexer cards

• Unit 1 Division 1 emergency core cooling system (ECCS) water leg keep-fill pump testing following emergent repairs

• 2B EDG testing following replacement of a failed fuse and diode in the generator control circuitry The inspectors verified by witnessing the test or reviewing the test data that post-maintenance testing activities were adequate for the above maintenance or repair activities. The inspectors reviews included, but were not limited to, integration of testing activities, applicability of acceptance criteria, test equipment calibration and control, procedural use and compliance, control of temporary modifications or jumpers required for test performance, documentation of test data, Technical Specification applicability, system restoration, and evaluation of test data. Also, the inspectors verified that maintenance and post-maintenance testing activities adequately ensured that the equipment met the licensing basis, Technical Specifications, and Updated Final Safety Analysis Report (UFSAR) design requirements.

These post-maintenance testing reviews constituted six inspection samples.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

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

• 2A emergency diesel generator monthly surveillance run

• Unit 2 Division 2 residual heat removal service water (RHRSW) balance test

• Unit 1 breaker trip test for 1VQ051, N2 makeup upstream isolation valve

• 1A service water pump surveillance test

• 2B diesel generator room CO2 system functional test

• Unit 1 and Unit 2 reactor core isolation cooling F028 local leak rate test

• A auxiliary electric equipment room ventilation (VE) system purge mode testing The inspectors observed the performance of surveillance testing activities, including reviews for preconditioning, integration of testing activities, applicability of acceptance criteria, test equipment calibration and control, procedural use, control of temporary modifications or jumpers required for test performance, documentation of test data, Technical Specification applicability, impact of testing relative to performance indicator reporting, and evaluation of test data.

These surveillance test reviews constituted seven inspection samples.

b. Findings

No findings of significance were identified. One unresolved item (URI) was identified.

On September 14, 2004, the licensee was conducting a surveillance test of the A train of the auxiliary electric equipment room (AEER) ventilation (VE) system utilizing the purge mode of operation per an approved testing procedure. In this mode, 100 percent of the air being supplied to the AEER is outside air. Approximately 10 minutes after the start of the test, the A VE compressor tripped. With the air being supplied to the AEER no longer being cooled and dehumidified, moisture began to condense on the coolest equipment panels in the AEER, and multiple control room annunciator alarms began to actuate.

Approximately 18 minutes after the trip of the A VE compressor, the licensee completed shifting to the B train of VE. The spurious control room alarms subsided as temperature and humidity conditions in the AEER were restored to normal. Three control room annunciator alarms (Unit 1 drywell equipment drain sump trouble; Unit 2 post-loss of coolant accident (LOCA); and the Unit 2 power supply alarm for Panel 2P05J in the control room) remained actuated and would not reset. The licensee performed appropriate compensatory measures, as required by procedure, until these were repaired the following day via replacement of the applicable alarm cards.

A prompt investigation by the licensees engineering and operations staff revealed that the A VE compressor had tripped on high discharge pressure due to being overloaded.

Further, the licensees investigation uncovered the fact that the VE system was never designed to operate in purge mode and maintain AEER environmental conditions under the ambient air temperature and humidity conditions in place at the time of the test. In short, the licensees approved test procedure for the surveillance attempted to operate the A VE system and maintain AEER environmental conditions in a configuration that was outside of the design capabilities of the VE compressor.

The inspectors are presently examining this event against the licensees quality assurance program procedure quality requirements (10 CFR 50, Appendix B, Criterion V), as well as other regulatory requirements. The issue is considered unresolved pending the inspectors receipt and review of the licensees equipment apparent cause evaluation (EACE), which is in progress. The licensee has entered this issue into their corrective action program as CRs 253769 and 252847.

(URI 05000373/2004004-02; 05000374/2004004-02)

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the following temporary modifications:

• Disable emergency light (CKT 242D) Unit 2 drywell elevation 807' at 230 degree azimuth (EC 341156/TCCP 341156)

• Setpoint change - seismic monitor trigger level (TCCP 347872)

The inspectors reviewed the safety screening, design documents, UFSAR, and applicable Technical Specifications to determine that the temporary modification was consistent with modification documents, drawings, and procedures. The inspectors also reviewed the post-installation test results to confirm that tests were satisfactory and that the actual impact of the temporary modification on the permanent system and interfacing systems were adequately verified.

These reviews constituted two inspection samples.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The resident inspectors reviewed an emergency preparedness (EP) training drill involving an operations crew in the plant simulator and an EP response team in the sites technical support center (TSC). The inspectors evaluated drill conduct and the adequacy of the licensees critique of the EP teams performance to identify weaknesses and deficiencies. The selected drill scenario provided 6 opportunities for input to the EP Drill/Exercise NRC Performance Indicator. The inspectors observed the classification of events, notifications to off-site agencies, protective action recommendation development, and drill critique. Observations were compared to the licensees observations and corrective action program entries. The inspectors verified that there were no discrepancies between observed performance and performance indicator reported statistics. The scenario observed resulted in emergency declarations involving an alert classification, as well as an upgrade to a general emergency classification.

This drill observation constituted a single inspection sample.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Control to Radiologically Significant Areas (71121.01) Radiation Worker Performance

a. Inspection Scope

During job performance observations, the inspectors evaluated radiation worker performance with respect to stated radiation protection work requirements and evaluated whether workers were aware of the significant radiological conditions in their workplace, the RWP controls and limits in place, and that their performance had accounted for the level of radiological hazards present.

This review represented one inspection sample.

b. Findings

Introduction The inspectors identified a Green finding and associated Non-Cited Violation (NCV)when they observed operations personnel entering a posted neutron dose area without proper neutron monitoring, contrary to the licensee's Technical Specifications. This finding was considered NRC-identified as radiation protection personnel were unaware of this issue until questions by inspectors indicated a lack of proper neutron dose control for both this event and similar past occurrences.

Description On June 1, 2004, inspectors observed LOS-RH-Q1, a routine quarterly residual heat removal (RHR) system run. A portion of the procedure requires venting the 2A RHR low pressure coolant injection (LPCI) line, which is normally accessed via a raised platform within a neutron radiation area. Inspectors observed an operator climb the platform and enter a posted neutron radiation area to perform this venting. However, because inspectors could not recall a mention of this platform or a neutron radiation area in the pre-job radiation protection (RP) brief, they questioned the operators as to the requirements for entry into a neutron radiation area.

Inspectors returned to the RP desk to follow up on this issue and when questioned as to the requirements for entry into a neutron radiation area, RP technicians individually provided different guidance. In some cases they were unable to reference RP-AA-210, Dosimetry Issue, Usage, and Control, the plant procedure governing neutron dose estimation and monitoring. On June 2, 2004, inspectors requested the neutron dose cards for operators performing LOS-RH-Q1, however the RP staff was unable to locate them. It was later determined by the licensee that, contrary to RP-AA-210, dose cards were not filled out for this activity and, furthermore, had not been filled out in the past 6 months for this and similar recurring surveillance activities in neutron radiation areas.

The licensee calculated the total neutron dose for an individual entry to the area, based on time and dose rate, as 0.2 millirem for each entry.

The licensee's initial human performance investigation determined a failure of communication between the operations and RP staff was the cause of these errors.

Specifically, both the operations staff and RP technicians were confused about the requirements for entry into neutron radiation area, failed to adequately communicate during RP briefs about areas in which work would be performed, and on multiple occasions did not monitor neutron dose as required. The operations staff involved were coached on the need to be specific when communicating with RP, site personnel were notified of this event through a station communication, RP training was conducted for the requirements for access to a neutron radiation areas, and a pre-job brief was prepared for use in all further work involving neutron radiation areas.

Analysis The performance deficiency associated with this event was failure to properly monitor neutron dose, as required per licensee procedure RP-AA-210, Dosimetry Issue, Usage, and Control. The inspectors determined that performing activities, not in accordance with the procedure, warranted a significance evaluation in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening. The finding, which is under the Occupational Radiation Safety Cornerstone, does not involve the application of traditional enforcement because it did not result in actual safety consequences or have potential to impact the NRC's regulatory function, and was not the result of any willful actions. The finding was more than minor as it involves the Occupational Radiation Safety Cornerstone attributes of Program and Process, specifically, procedures and exposure/contamination control and monitoring.

Additionally, it affects the Cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation.

The inspectors determined that the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. For the Phase 1 screening, the inspectors answered no to all the questions under the Occupational Radiation Safety column. The finding does not relate to ALARA or work planning, was not an overexposure, there was not a substantial potential for excessive exposure, and it did not affect the licensee's ability to assess dose. Therefore, the finding is considered to be of very low safety significance, or Green. Because the inspectors assessed the main cause for the finding to involve the cross-cutting aspect of human performance, the finding is also discussed in Section 4OA4, Cross-Cutting Aspects of Findings, in this report.

Enforcement Technical Specifications 5.4.1.(a) requires that written procedures shall be established, implemented, and maintained covering specific procedures recommended in Regulatory Guide 1.33. Section 7.e

(7) of Regulatory Guide 1.33 lists the requirement for radiation protection procedures for personnel monitoring. RP-AA-210, Dosimetry Issue, Usage, and Control, is the plant procedure governing neutron dose estimation and monitoring.

Section 4.7, Neutron Dose Estimation and Monitoring, requires neutron dose estimate cards be completed prior to allowing work in areas of known neutron dose.

Contrary to the above, on June 1, 2004, a member of the operations staff entered a neutron radiation area without proper neutron monitoring or neutron dose estimate controls.

Because work conducted in normally accessible neutron radiation areas usually involve relatively low dose rates and short periods of time, and long term neutron doses are accounted for by worker personal dosimetry, the event is of very low safety significance and the finding is within the licensee's response band, or Green. The licensee had entered the issue into their corrective action system as Condition Report 225735. The associated violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000373/2004004-03; 05000374/2004004-03)2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)

.1 Inspection Planning

a. Inspection Scope

The inspectors reviewed the plant UFSAR to identify applicable radiation monitors associated with transient high and very high radiation areas including those used in remote emergency assessment. The inspectors identified the types of portable radiation detection instrumentation used for job coverage of high radiation area work, other temporary area radiation monitors currently used in the plant, continuous air monitors associated with jobs with the potential for workers to receive 50 mrem CEDE, whole body counters, and the types of radiation detection instruments utilized for personnel release from the radiologically controlled area.

This review represented one inspection sample.

The inspectors verified calibration, operability, and alarm setpoint (if applicable) of the following four instruments:

• Eberline PM-7;

• Canberra Fastscan Whole Body Counter;

• Small Articles Monitor (SAM); and

• RSO-50E.

These reviews represented one inspection sample.

The inspectors determined what actions were taken when, during calibration or source checks, an instrument was found significantly out of calibration (>50 percent),determined possible consequences of instrument use since last successful calibration or source check, and determined if the out of calibration result was entered into the corrective action program. There were no instances where the instrument was found significantly out of calibration. The inspectors also considered the licensees 10 CFR 61 source term reviews to determine if the calibration sources used are representative of the plant source term.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

.2 Problem Identification and Resolution

a. Inspection Scope

The inspectors reviewed the licensees self-assessments, audits, Licensee Event Reports, and Special Reports that involved personnel contamination monitor alarms due to personnel internal exposures to verify that identified problems were entered into the corrective action program for resolution. All event reports involving internal exposures

>50 mrem CEDE were reviewed to determine if the affected personnel were properly monitored utilizing calibrated equipment and if the data was analyzed and internal exposures properly assessed in accordance with licensee procedures.

This review represented one inspection sample.

The inspectors reviewed corrective action program reports related to exposure significant radiological incidents that involved radiation monitoring instrument deficiencies since the last inspection in this area. Staff members were interviewed and corrective action documents were reviewed to verify that follow-up activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk based on the following:

• Initial problem identification, characterization, and tracking;

• Disposition of operability/reportability issues;

• Evaluation of safety significance/risk and priority for resolution;

• Identification of repetitive problems;

• Identification of contributing causes;

• Identification and implementation of effective corrective actions;

• Resolution of NCVs tracked in the corrective action system; and

• Implementation/consideration of risk significant operational experience feedback.

This review represented one inspection sample.

The inspectors determined if the licensees self-assessment activities were identifying and addressing repetitive deficiencies or significant individual deficiencies in problem identification and resolution.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Radiation Protection Technician Instrument Use

a. Inspection Scope

The inspectors verified the calibration expiration and source response check currency on radiation detection instruments staged for use and observed radiation protection technicians for appropriate instrument selection and self-verification of instrument operability prior to use.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

.4 Self-Contained Breathing Apparatus (SCBA) Maintenance and User Training

a. Inspection Scope

The inspectors reviewed the status and surveillance records of SCBAs staged and ready for use in the plant and inspected the licensees capability for refilling and transporting SCBA air bottles to and from the control room and operations support center during emergency conditions. The inspectors determined if control room operators and other emergency response and radiation protection personnel were trained and qualified in the use of SCBAs (including personal bottle change-out). The inspectors verified that three individuals on each control room shift crew, and three individuals from each designated department were currently assigned emergency duties.

This review represented one inspection sample.

The inspectors reviewed the qualification documentation for at least 50 percent of the onsite personnel designated to perform maintenance on the vendor-designated vital components, and the vital component maintenance records over the past 5 years for three SCBA units currently designated as ready for service. The inspectors also ensured that the required, periodic air cylinder hydrostatic testing was documented and up to date, and that the DOT required retest air cylinder markings were in place for these three units. The inspectors reviewed the onsite maintenance procedures governing vital component work, including those for the low-pressure alarm and pressure-demand air regulator, and licensee procedures and the SCBA manufacturers recommended practices to determine if there were inconsistencies between them.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

Cornerstones: Mitigating Systems and Barrier Integrity

.1 Reactor Safety Strategic Area

a. Inspection Scope

The inspectors sampled the licensees submittals for the performance indicators (PIs)listed below. The inspectors reviewed Licensee Event Reports (LERs), licensee data reported to the NRC, plant logs, condition reports and NRC inspection reports to verify the following performance indicators from the 3rd Quarter of 2003 to the 2nd Quarter of 2004:

• Safety System Unavailability - Emergency AC Power; Unit 1 and Unit 2

• Safety System Unavailability - High Pressure Injection System; Unit 1 and Unit 2

• Safety System Unavailability - Heat Removal System; Unit 1 and Unit 2

• Safety System Unavailability - Residual Heat Removal System; Unit 1 and Unit 2

• Reactor Coolant System (RCS) Identified Leak Rate; Unit 1 and Unit 2 The inspectors verified that the licensee accurately reported performance as defined by the applicable revision of Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline.

These reviews constituted ten inspection samples

b. Findings

No findings of significance were identified.

.2 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the data submitted by the licensee for the 2nd Quarter 2004 performance indicators for any obvious inconsistencies prior to its public release in accordance with IMC 0608, Performance Indicator Program.

b. Findings

No findings of significance were identified.

The inspectors noted that the licensees submittal for the RCS leakage PI had not been accurately reported on multiple occasions. The source data for this PI consists of the sum of identified and unidentified leakage. Unidentified leakage is recorded per Technical Specifications as the highest of two methods, the preferred and alternate. For greater than 2 years, only the alternate method had been used to compute the value for the RCS leakage PI. Because on many occasions the preferred method of calculating unidentified leakage was higher than the alternate method, data reported to the NRC was erroneously low. The licensee determined, during a review of data submitted from September 2002 to July 2004, that RCS leakage PI data was in error for Unit 1 for 6 months and for Unit 2 for 11 months. However, because these corrected values do not cause the licensees PI to cross a color threshold and the licensee has entered these errors into their corrective action program (CR 244247), this is considered a minor issue. The licensee plans to submit corrected PI values and to revise their process for collecting RCS leakage PI data.

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the routine inspections documented above, the inspectors verified that the licensee entered the problems identified during the inspection into their corrective action program. Additionally, the inspectors verified that the licensee was identifying issues at an appropriate threshold and entering them in the corrective action program, and verified that problems included in the licensee's corrective action program were properly addressed for resolution. Attributes reviewed included: complete and accurate identification of the problem; that timeliness was commensurate with the safety significance; that evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program (CAP) Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program (CAP). This review was accomplished through inspection of the stations daily condition report packages.

b. Findings

No findings of significance were identified.

.3 Selected Issue Follow-Up Inspection:

Licensee Corrective Actions for Recent NRC Performance Indicator (PI) Data Submission Errors Introduction Over the past several months, there have been several instances in which the licensee has submitted inaccurate PI data to the NRC. The inspectors selected these errors and their corrective actions as an annual sample to review the licensees problem identification and resolution for this issue:

• August 12, 2003. Inspectors identified inaccuracies in reported data for the 2nd Quarter of 2003 associated with the 0 and 1A emergency diesel generators (EDGs). Specifically, a safety system functional failure was not reported and unavailability time for both EDGs was under reported. (CR 171232)

• October 15, 2003. Inspectors identified that the unavailability time for the 2B EDG was under reported during the 3rd Quarter of 2002. An extent-of-condition review by the licensee uncovered an inaccuracy in the reported reactor coolant system (RCS) activity PI for the 2nd Quarter of 2003.

(CR 181122)

• January 23, 2004. During preparations for a biennial NRC graded exercise, the licensee determined that some prior emergency preparedness drill and exercise participation PI data submitted to the NRC was in error. (CR 197159)

• March 22, 2004. The inspectors identified that the licensees data gathering practices excluded RCS dose equivalent iodine (DEI) samples taken during non-power transients. Subsequently, the RCS activity PI data reported to the NRC for August of 2003 was found to be in error due to a sample that was not reported. (CR 210248)

• August 12, 2004. The inspectors identified that licensee practices for recording RCS leakage rate data excluded the most conservative calculational values for total leakage. Over a 2 year period, the Technical Specification value was not consistently used for PI data submissions to the NRC, as required by NEI 99-02, Revision 2. (CR 244247)

Although in each of the above cases, the errors noted were not significant enough to have caused a PI color threshold to have changed, each case represents a potential for such a condition to have occurred. As discussed in the NRCs Enforcement Policy, NUREG-1600, PI data submission errors that result in a color threshold change are considered to be of more than minor significance.

The review of this issue by the inspectors constituted a single inspection sample.

a.

Effectiveness of Problem Identification

(1) Inspection Scope The inspectors reviewed the condition reports (CRs) and follow-up actions for the above issues to verify that the licensees identification of the problems were complete, accurate, and timely, and that the consideration of extent-of-condition review, generic implications, common cause, and previous occurrences were adequate.

(2) Issues The inspectors reviewed CR 171232, Inaccurate Unavailability Data in April 2003 NEI Monthly PI; CR 181122, NRC Performance Indication Data Entry Error; CR 197159, EP PI Data Submittal Error Found During FASA; CR 210248, DEI Sample Result Not Used in NRC PI for Unit 2 Aug 03; and CR 244247, Inconsistencies in Data Used to Determine NRC RCS Leakage PI; and associated documents. Up until the final PI issue identified by the inspectors and documented in CR 244247, the licensee had not identified that a repeat or similar condition was occurring. The scope of the licensees corrective actions were narrowly focused, and did not take into account the broader possibility that a more generic problem with PI data assembly and verification was potentially occurring.

b.

Prioritization and Evaluation of Issues

(1) Inspection Scope The inspectors reviewed the PI data submission CRs noted above to assess the licensees evaluation and disposition of performance issues, and application of risk insights for prioritization of issues.

(2) Issues In independently reviewing the PI condition reports for common cause issues, previous occurrences, and similar conditions, the inspectors concluded that the licensees assessments in this area were somewhat narrowly focused. Interviews with licensee personnel by the inspectors indicated a common attitude with respect to the gathering of PI data was one in which the function was considered routine. Contributing to this was a weakness in supervisory oversight to certain parts of the PI data gathering process.

For example, in the issues involving CRs 210248 and 244247, the licensees PI data steward was solely responsible for gathering the appropriate data and entering that data into the licensees database system. Supervisory oversight to ensure independent validation of the source data for these PIs was not effective, and as a result, errors were introduced into the PI reporting process.

By characterizing these PI data errors as unrelated events, the licensees corrective action program assessments missed an opportunity to broadly address several programmatic deficiencies.

c.

Effectiveness of Corrective Actions

(1) Inspection Scope As discussed above, the inspectors noted that the corrective actions taken for the earlier PI errors were narrowly focused, and tended to treat the individual issues as isolated events. As a result, in this section of the inspection the inspectors focused on the effectiveness of corrective actions from the most recent issue, identified by the inspectors in August of 2004.

The inspectors reviewed the related condition reports to determine if the corrective action program addressed generic implications. Additionally, the inspectors verified that corrective actions were appropriately focused to correct the problem.

(2) Issues Subsequent to the initiation of CR 244247, licensee management ordered a comprehensive review of procedures and responsibilities for developing and verifying PIs by all PI data stewards. Further, the licensee has also scheduled a review of the PI data gathering, verification, and submission process as part of an upcoming focused area self assessment (FASA). The inspectors reviewed these actions, both accomplished and planned, and determined that they appeared to be adequate to correct the deficiencies noted and prevent recurrence.

4OA4 Cross-Cutting Aspects of Findings

Cornerstone: Occupational Radiation Safety

Human Performance One of the findings described elsewhere in this report had human performance deficiencies as its major causal element.

• A finding described in Section 2OS1 involved the failure of operations and radiation protection personnel to follow established plant procedures and radiological practices with respect to entry into neutron dose areas. A plant operator performing a routine RHR system surveillance accessed an elevated platform posted as a neutron dose area without first having the neutron dose estimate for the entry calculated and recorded by radiation protection personnel, as was required by plant procedure. Poor communications between operations and radiation protection personnel was attributed to be the primary cause for the issue.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to the Site Vice President, Mr. G. Barnes, and other members of licensee management on September 30, 2004.

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

An interim exit meeting was conducted for:

• An occupational radiation safety radiological access control and radiation monitoring instrumentation and protective equipment inspection with the Plant Manager, Ms. S. Landahl, on August 18, 2004.

4OA7 Licensee-Identified Violation

Cornerstone: Initiating Events

The following violation of very low significance was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.

• On June 10, 2004, an hourly fire watch patrol of the Unit 2 cable spreading room was in effect to compensate for a disabled fire protection deluge system. A review of the logs for this fire watch patrol by the licensee revealed that, on one occasion, a period of 77 minutes elapsed between patrols. This elapsed period between patrols exceeded the allowances in procedure OP-MW-201-007, Fire Protection System Impairment Control, and constituted a violation of Technical Specification 5.4.1(c), which requires that written procedures be established, implemented, and maintained for the licensees fire protection program. A review of the issue by the inspectors determined that it was of very low safety significance because no actual fire watch patrols were missed, and the gap between patrols at issue was of a very short period of time. The licensee had entered this issue into their corrective action program as CR 232835.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

G. Barnes, Site Vice President

S. Landahl, Plant Manager

T. Connor, Maintenance Director

L. Coyle, Operations Director

D. Czufin, Site Engineering Director

A. Ferko, Nuclear Oversight Manager

F. Gogliotti, System Engineering Manager

B. Kapellas, Radiation Protection Manager

J. Lindsey, Operations Training Manager

J. Rappeport, Nuclear Oversight

W. Riffer, Emergency Planning Manager

T. Simpkin, Acting Regulatory Assurance Manager

C. Wilson, Station Security Manager

Nuclear Regulatory Commission

B. Burgess, Chief, Region III, Reactor Projects Branch 2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000373/2004004-01 URI Unit 1 RCIC F028 Containment Isolation Check Valve Maintenance History Indicates Repeated Failures Since 1991 (Section 1R12)

05000373/2004004-02;

05000374/2004004-02 URI A Train VE Compressor Tripped During Purge Mode Surveillance Testing Causing High Humidity Condition and Multiple Control Room Annunciator Alarms (Section 1R22)

05000373/2004004-03;

05000374/2004004-03 NCV Entry into a Neutron Radiation Area by Operations Personnel without Procedurally Required Neutron Radiation Dose Estimates (Section 2OS1)

Closed

05000373/2004004-03;

05000374/2004004-03 NCV Entry into a Neutron Radiation Area by Operations Personnel without Procedurally Required Neutron Radiation Dose Estimates (Section 2OS1)

Discussed

None.