IR 05000456/2005010
| ML060380242 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 02/06/2006 |
| From: | Richard Skokowski NRC/RGN-III/DRP/RPB3 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| FOIA/PA-2006-0115, FOIA/PA-2010-0209 IR-05-010 | |
| Download: ML060380242 (37) | |
Text
ary 6, 2006
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000456/2005010; 05000457/2005010
Dear Mr. Crane:
On December 31, 2005, the U. S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Braidwood Station, Units 1 and 2. The enclosed report documents the inspection findings which were discussed on January 9, 2006, with Mr. G. Boerschig 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.
There was one finding of very low safety significance identified in this report. This issue, which was reviewed under the NRC traditional enforcement process, was determined to be a Severity Level IV violation of NRC requirements. Because this violation was a Severity Level IV violation and it was entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation 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, D.C. 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, D.C. 20555-0001; and the Resident Inspector Office at the Braidwood facility. In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Richard A. Skokowski, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77
Enclosure:
Inspection Report 05000456/2005010; 05000457/2005010 w/Attachment: Supplemental Information
REGION III==
Docket Nos: 50-456; 50-457 License Nos: NPF-72; NPF-77 Report No: 05000456/2005010; 05000457/2005010 Licensee: Exelon Generation Company, LLC Facility: Braidwood Station, Units 1 and 2 Location: Braceville, IL Dates: October 1 through December 31, 2005 Inspectors: N. Shah, Senior Resident Inspector G. Roach, Resident Inspector E. Bonano, Health Physicist M. Holmberg, Senior Engineering Inspector J. House, Senior Radiation Specialist R. Jickling, Emergency Preparedness Analyst R. Ng, Resident Inspector S. Orth, Health Physics Team Lead B. Palagi, Senior Operations Engineer T. Ploski, Senior Emergency Preparedness Inspector W. Slawinski, Senior Radiation Specialist W. Snell, Senior Health Physicist M. Wilk, Reactor Engineer J. Roman, Illinois Emergency Management Agency Approved by: R. Skokowski, Chief Branch 3 Division of Reactor Projects Enclosure
SUMMARY OF FINDINGS
IR 05000456/2005010, 05000457/2005010; 10/01/2005 - 12/31/2005; Braidwood Station,
Units 1 & 2; Emergency Action Level and Emergency Plan Changes.
This report covers a 3-month period of baseline resident inspection and announced baseline inspections on emergency preparedness. The inspection was conducted by the resident inspectors, regional emergency preparedness inspectors, regional health physicists, and regional engineering inspectors. One finding associated with a Severity Level IV Non-Cited Violation was 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 NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A. Inspector-Identified and Self-Revealed Findings
Cornerstone: Emergency Preparedness
- Severity Level IV. The inspectors identified that the licensee had changed its standard emergency action level (EAL) scheme by revising one EALs criteria for an Unusual Event declaration that addressed an unplanned radiological release in excess of effluent radiation monitor readings unless the release could be determined to be below Offsite Dose Calculation Manual limits within 15 minutes for releases that could not be terminated in 60 minutes or less. The inspectors determined that this EAL change decreased the effectiveness of the emergency plan, and that the licensee did not obtain prior NRC approval for this change, contrary to the requirements of 10 CFR 50.54(q).
The licensee is evaluating the options to correct the EAL.
This finding was more than minor because extending the time period required for the appropriate emergency classification of a radiological release could adversely affect the performance of both onsite and offsite emergency actions. Because the issue affected the NRCs ability to perform its regulatory function, it was evaluated with the traditional enforcement process as specified in Section IV.A.3 of the Enforcement Policy.
According to Supplement VIII of the Enforcement Policy, this finding was determined to be a Severity Level IV because it involved a failure to meet a requirement not directly related to assessment and notification. Further, this problem was isolated to one EAL and was not indicative of a functional problem with the EAL scheme. Additionally, because the violation was a Severity Level IV and the licensee entered this issue into its corrective action program this finding is being treated as a Severity Level IV Non-Cited Violation of 10 CFR 50.54(q). (Section 1EP4)
Licensee-Identified Violations
No findings of significance were identified.
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near full power throughout the inspection period.
Unit 2 operated at or near full power throughout the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
a. Inspection Scope
The inspectors performed a walkdown of selected plant areas to review the licensees preparations for cold weather conditions. The areas selected were those considered at risk based on a review of the Updated Final Safety Analysis Report (UFSAR), Technical Specifications (TS) and other design basis documents. The specific areas observed were:
- Units 1 and 2 rod control spot cooler systems;
- Units 1 and 2 condensate and refueling water storage tanks (RWST) heating systems; and
- turbine building ventilation louvers.
The inspectors also reviewed Issue Reports (IRs) generated since January 1, 2004, for the station heating system and for the diesel generator, turbine building, auxiliary building and miscellaneous ventilation systems. Specifically, the inspectors noted whether there were any adverse trends for components associated with these systems that potentially impacted the licensees cold weather preparations.
The inspectors verified that minor issues identified during these inspections were entered into the licensees corrective action program. Documents reviewed as part of this inspection are listed in the Attachment. This review constituted one sample of this inspection requirement.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
Partial Walkdowns
a. Inspection Scope
The inspectors performed partial walkdowns of the accessible portions of risk-significant system trains during periods when the train was of increased importance due to redundant trains or other equipment being unavailable. The inspectors utilized the valve and electric breaker checklists listed to determine whether the components were properly positioned and that support systems were aligned as needed. The inspectors also examined the material condition of the components and observed operating parameters of equipment to determine whether there were any obvious deficiencies.
The inspectors reviewed IRs associated with the train to determine whether those documents identified issues affecting train function. The inspectors used the information in the appropriate sections of the TS and the UFSAR to determine the functional requirements of the system. The inspectors also reviewed the licensees identification of and the controls over the redundant risk-related equipment required to remain in service. Documents reviewed during this inspection are listed in the
.
The inspectors completed three samples of this requirement by walkdowns of the following trains:
- 2A containment spray pump;
- 2B emergency diesel generator (EDG); and
- 1B residual heat removal system.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
Quarterly Inspection
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of fire fighting equipment, the control of transient combustibles and ignition sources, and on the condition and operating status of installed fire barriers. The inspectors selected fire areas for inspection based on their overall contribution to internal fire risk, as documented in the Individual Plant Examination of External Events with later additional insights or their potential to impact equipment which could initiate a plant transient or be required for safe shutdown. The inspectors used the Fire Protection Report, Revision 21, to determine: whether fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and that fire doors, dampers, and penetration seals appeared to be in satisfactory condition.
The inspectors completed nine samples of this inspection requirement during the following walkdowns:
- Unit 1 and 2 remote shutdown panel spaces;
- Unit 1 auxiliary electrical equipment space;
- Unit 2 auxiliary electrical equipment space;
- auxiliary building 364' elevation general area;
- Units 1 and 2 turbine and motor driven feedwater pump areas;
- Units 1 and 2 main control room;
- Units 1 and 2 condensate and condensate booster pump areas;
- Units 1 and 2 turbine building 401' elevation general area; and
- auxiliary building radwaste spaces on 401' and 426' elevations.
The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program. Documents reviewed during this inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program
.1 Annual Operating Test Results
a. Inspection Scope
The inspector reviewed the overall pass/fail results of the annual operating examination which consisted of Job Performance Measure and simulator operating tests (required to be given per 10 CFR 55.59(a)(2)) administered by the licensee. The operating testing was conducted in August, September, and October 2005. In addition, the inspectors reviewed the overall pass/fail results for the biennial written examination (also required to be given per 10 CFR 55.59(a)(2)) administered by the licensee. The written tests were administered in June, and July 2005. The overall results were compared with the significance determination process in accordance with NRC Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process. This represented one sample
b. Findings
No findings of significance were identified.
.2 Quarterly Review of Testing/Training Activity
a. Inspection Scope
The inspectors observed operating crew performance during evaluated simulator out-of-the-box scenario, Braidwood Station Licensed Operator Requalification Simulator Scenario Number 0561, Design Basis SGTR / Faulted Steam Generator, Revision 0.
The inspectors evaluated crew performance in the following areas:
- clarity and formality of communications;
- ability to take timely actions in the safe direction;
- prioritization, interpretation, and verification of alarms;
- procedure use;
- control board manipulations;
- oversight and direction from supervisors; and
- group dynamics.
Crew performance in these areas was compared to licensee management expectations and guidelines.
The inspectors verified that the crew completed the critical tasks listed in the simulator guide. The inspectors also compared simulator configurations with actual control board configurations. For any weaknesses identified, the inspectors observed the licensee evaluators to determine whether they also noted the issues and discussed them in the critique at the end of the session. Those documents reviewed during this inspection are listed in the Attachment. This review constituted one sample of this inspection requirement.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
Routine Inspection
a. Inspection Scope
The inspectors reviewed the licensees overall maintenance effectiveness for selected plant systems. This evaluation consisted of the following specific activities:
- observing the conduct of planned and emergent maintenance activities where possible;
- reviewing selected IRs, open Work Orders (WOs), and control room log entries in order to identify system deficiencies;
- reviewing licensee system monitoring and trend reports;
- attending various meetings throughout the inspection period where the status of maintenance rule activities was discussed;
- a partial walkdown of the selected system; and
- interviews with the appropriate system engineer.
The inspectors also reviewed whether the licensee properly implemented Maintenance Rule, 10 CFR 50.65, for the chosen systems. Specifically, the inspectors determined whether:
- the system was scoped in accordance with 10 CFR 50.65;
- performance problems constituted maintenance rule functional failures;
- the system had been assigned the proper safety significance classification;
- the system was properly classified as (a)(1) or (a)(2); and
- the goals and corrective actions for the system were appropriate.
The above aspects were evaluated using the maintenance rule program and other documents listed in the Attachment. The inspectors also verified that the licensee was appropriately tracking reliability and/or unavailability for the systems.
The inspectors completed two samples in this inspection requirement by reviewing the following systems:
- Unit 1 and Unit 2 reactor coolant systems subsequent to 1A reactor coolant pump excessive seal leakage outage; and
- instrumentation and control systems subsequent to repeated feedwater flow control system transients.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed the licensees management of plant risk during emergent maintenance activities or during activities where more than one significant system or train was unavailable. The activities were chosen based on their potential impact on increasing the probability of an initiating event or impacting the operation of safety-significant equipment. The inspections were conducted to determine whether evaluation, planning, control, and performance of the work were done in a manner to reduce the risk and minimize the duration where practical, and that contingency plans were in place where appropriate.
The licensees daily configuration risk assessment records, observations of operator turnover and plan-of-the-day meetings, and observations of work in progress, were used by the inspectors to verify that the equipment configurations were properly listed, that protected equipment were identified and were being controlled where appropriate, that work was being conducted properly, and that significant aspects of plant risk were being communicated to the necessary personnel.
In addition, the inspectors reviewed Braidwood Station Operator Annual Aggregate Review, dated November 2005, to determine whether problems were being entered into the corrective action program with the appropriate characterization and significance.
The inspectors completed four samples by reviewing the following activities:
- 1A EDG engine trip circuit check valve leak-by resulting in engine trip during cooldown mode of shutdown cycle;
- 2A EDG start-up to rated frequency and voltage in excess of 10 seconds and subsequent common cause review;
- direct current bus 111 intermittent grounds; and
- online replacement of Unit 2 digital electrical hydraulic control differential pressure switch 2PDS-TO091.
Those documents reviewed during this inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R14 Operator Performance During Non-Routine Evolutions and Events
a. Inspection Scope
The inspectors completed two samples by observing the following events:
- Unit 2 turbine digital electrical hydraulic control system power supply failure; and
- grounded circuit cards in the Unit 2 2PM05J main control room reactor plant status panel annunciator system.
The inspectors observed the control room response, interviewed plant operators and reviewed plant records including control room logs, operator turnovers, and IRs. The inspectors verified that the control room response was consistent with station procedures and that identified discrepancies were captured in the corrective action program. Corrective action documents reviewed as part of this inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors evaluated plant conditions and selected IRs for risk-significant components and systems in which operability issues were questioned. These conditions were evaluated to determine whether the operability of components was justified. The inspectors compared the operability and design criteria in the appropriate section of the UFSAR to the licensees evaluations presented in the IRs and documents listed in the Attachment to verify that the components or systems were operable. The inspectors also conducted interviews with the appropriate licensee system engineers and conducted plant walkdowns, as necessary, to obtain further information regarding operability questions.
The inspectors completed four samples by reviewing the following operability evaluations and conditions:
- review for Braidwood applicability of Palo Verde Nuclear Generating Station Operating Experience (OPEX) concerning potential air binding of emergency core cooling system (ECCS) injection pumps during transition from refueling water storage tank to ECCS sump recirculation;
- impact of Lake Screen House forebay silting and Bryozoa formation on circulating water (CW), essential service water, and fire protection systems;
- sensing line leak on Unit 1 high pressure turbine first stage pressure transducer 1PT-505; and
- Unit 1A steam generator power operated relief valve seat leakage.
b. Findings
No findings of significance were identified.
1R16 Operator Workarounds
Semi-annual Review of Cumulative Effect of Operator Workarounds
a. Inspection Scope
The inspectors completed a semi-annual review of the cumulative effect of operator workarounds. This inspection consisted of observing plant operators performing routine rounds during plant walkdowns and attending a Plant Operations Review Committee meeting on November 30, 2005. This review constituted one sample of this inspection requirement.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed post-maintenance testing activities associated with important mitigating systems, barrier integrity, and support systems to ensure that the testing adequately demonstrated system operability and functional capability. The inspectors used the appropriate sections of the TS and UFSAR, as well as the WOs for the work performed, to evaluate the scope of the maintenance and to determine whether the post-maintenance testing was performed adequately, demonstrated that the maintenance was successful, and that operability was restored. The inspectors determined whether the testing met the frequency requirements; that the tests were conducted in accordance with the procedures, including establishing the proper plant conditions and prerequisites; that the test acceptance criteria were met; and that the results of the tests were properly reviewed and recorded. The activities were selected based on their importance in demonstrating mitigating systems capability and barrier integrity. The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program by reviewing the documents in the
.
Six samples were completed by observing post-maintenance testing of the following components:
- 1A EDG test run following local control panel 1PL07J annunciator maintenance and engine pneumatic trip system check valve replacement;
- 2A EDG test run following repair of starting air system solenoid valves;
- 2B containment spray pump operability testing following preventative maintenance outage;
- 1PT-505 pressure switch calibration testing following sensing line repair;
- 1B EDG test run following repair of trip circuit check valves; and
- 1B containment spray K644B slave relay test post maintenance outage.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors reviewed surveillance testing activities associated with important mitigating systems, barrier integrity, and support systems to ensure that the testing adequately demonstrated system operability and functional capability. The inspectors used the appropriate sections of the TS and UFSAR, as well as the WOs for the work performed, to evaluate the scope of the maintenance and to determine whether the surveillance testing was performed adequately, demonstrated that the maintenance was successful, and that operability was restored. The inspectors determined whether the testing met the frequency requirements; that the tests were conducted in accordance with the procedures, including establishing the proper plant conditions and prerequisites; that the test acceptance criteria were met; and that the results of the tests were properly reviewed and recorded. The activities were selected based on their importance in demonstrating mitigating systems capability, barrier integrity and the initiating events cornerstone. The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program by reviewing the documents in the Attachment.
Six samples were completed by observing and evaluating the following surveillance tests:
- 2A EDG bypass of auto engine/generator trips in emergency mode;
- 2A auxiliary feedwater pump suction pressure switch calibration verification;
- calibration of the 1RF010 unit 1 reactor cavity drain leak detection sump;
- in-service test (American Society of Mechanical Engineers test) of the 2A safety injection pump; and
- Unit 2 containment emergency hatch local leak rate test and administrative review of selected Unit 2 post outage local leak rate tests.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications
a. Inspection Scope
The inspectors reviewed the following temporary modification:
- freeze seal of component cooling water system piping in support of 0CC9432 valve repair.
For the above modification, the inspectors reviewed the associated design change paperwork, attended applicable prejob briefings and observed installation and/or removal. The inspectors also reviewed contingency plans, as applicable, for modifications supporting continued component operability or reliability. The inspectors also verified that minor issues identified during the inspection were entered into the licensees corrective action program. Those documents reviewed during this inspection are listed in the Attachment. This review constituted one sample of this inspection requirement.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level (EAL) and Emergency Plan Changes
a. Inspection Scope
The inspectors performed a screening review of Revisions 14 and 15 of the Braidwood Station Emergency Plan Annex to determine whether the changes made in Revisions 14 and 15 decreased the effectiveness of the licensees emergency planning. This screening review of Revisions 14 and 15 was not documented in a Safety Evaluation Report and does not constitute an approval of the changes. Therefore, the changes are subject to future NRC inspection to ensure that the emergency plan continues to meet NRC regulations.
These activities completed one inspection sample.
b. Findings
Introduction:
The licensee changed one Emergency Action Level (EAL) that addressed events related to unplanned radiological releases. This change was determined to decrease the effectiveness of the licensees emergency plan, however, the licensee did not submit this change to NRC for prior approval. This is a violation of 10 CFR 50.54(q)and, because it impacted the regulatory process, traditional enforcement was applied.
Since this issue was entered into the licensees corrective action program and because this item involved a failure to meet a regulatory requirement not directly related to assessment or notification, this issue was determined to be a Severity Level IV Non-Cited Violation (NCV).
Description:
The licensees site-specific EALs were based on the guidance in NUMARC/NRSP-007. In 1995, the licensee upgraded the RU2 EAL threshold value to include criteria for confirming the validity of the effluent radiation monitor release indications within 15 minutes by comparison with greater than or equal to two times the Offsite Dose Calculation Manual limit. An Unusual Event would not be declared if the comparison did not support the effluent monitors indication of a release. Revision 13 to the Braidwood Station Emergency Plan Annex reflected this 15-minute criteria and appeared as follows:
Revision 13 RU2 EAL Threshold Value in Part:
Unplanned Radiological release in excess of Table R1 Unusual Event value unless releases can be determined to be below available Table R2 Unusual Event thresholds within 15 minutes.
Revisions 14 and 15 RU2 Threshold Value in Part:
Unplanned radiological release in excess of Table R1 Unusual Event threshold for >60 minutes UNLESS release can be determined to be below available Table R2 Unusual Event thresholds within this period.
Discussions with the licensee emergency preparedness staff and inspection of the 10 CFR 50.54(q) review records indicated this change was made to rearrange the EAL with the more accurate indicators first and due to control room crews interpretation that they had 75 minutes to declare an Unusual Event in this EAL. Also, the licensees 10 CFR 50.54(q) review indicated that the change did not decrease the effectiveness of the emergency plan.
In contrast, the inspectors determined that the change to this indicator represented a decrease in effectiveness of the emergency plan because the re-worded EAL threshold removed the NRCs 1995 approved 15-minute requirement and replaced it with a 60-minute requirement for determining whether releases were below specified effluent monitor thresholds.
The requirements of 10 CFR 50.54(q) allows the licensee to make changes to the emergency plan without Commission approval as long as the change does not decrease the effectiveness of the emergency plan. The inspectors noted that this change could potentially delay the declaration of an Unusual Event by as much as 45 minutes.
However, since the licensee had concluded in its 10 CFR 50.54(q) review that the change to this EAL threshold did not decrease the effectiveness of the emergency plan, this change was not submitted to the NRC for review prior to implementation of the revised EAL threshold.
Analysis:
The inspectors determined that the failure to request NRC approval of the EAL change was a performance deficiency. Furthermore the failure to request NRC approval of the EAL change potentially impeded the NRCs regulatory process and was therefore, in accordance with Section 2.2.e of Appendix B to NRC Manual Chapter 0609, evaluated using the guidance in Section IV of NUREG-1600, General Statement of Policy and Procedure for NRC Enforcement Actions (Enforcement Policy),rather than the NRC Significance Determination Process (SDP). This finding was more than minor because extending the time period required for the appropriate emergency classification of a radiological release could adversely affect the performance of both onsite and offsite emergency actions. The finding was not suitable for SDP evaluation, but have been reviewed by NRC management. The finding was therefore dispositioned as a Severity Level IV violation according to Supplement VIII (Emergency Preparedness) of the Enforcement Policy because it involved licensee failure to meet an emergency planning requirement (namely, 10 CFR 50.54(q)) not directly related to assessment of and notification.
Enforcement:
10 CFR 50.54(q) states, in part, that the licensee may make changes to these plans without Commission approval only if the changes do not decrease the effectiveness of the plans. Proposed changes that decrease the effectiveness of the approved emergency plans may not be implemented without application to and approval by the Commission. Contrary to this, in Revision 14 of the Braidwood Station Emergency Plan Annex, the licensee made a change to its standard EAL scheme that reduced the effectiveness of the emergency plan. This change was not submitted to the NRC for approval prior to implementation. The licensee entered this issue into their corrective action program as Condition Report (CR) 00437193.
Changing an emergency plan commitment without prior NRC approval impacts the NRCs ability to perform its regulatory function and is therefore processed through traditional enforcement, as specified in Section IV.A.3 of the Enforcement Policy, issued May 1, 2000, (65 FR 25388). According to Supplement VIII of the Enforcement Policy, this finding was determined to be a Severity Level IV because it involved a failure to meet a requirement not directly related to assessment and notification. Further, this problem was isolated to one EAL and was not indicative of a functional problem with the licensees EAL scheme. Additionally, because this was a Severity Level IV violation and the licensee entered this issue into its corrective action program, this finding is being treated as Non-Cited Violation (Severity Level IV) consistent with Section VI.A.1 of the Enforcement Policy. (NCV 05000456/2005010-01; 05000457/2005010-01)
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors observed licensee performance during an unannounced, evaluated emergency response drill. Observations included manning of the Technical Support Center, turnover of command and control to the Technical Support Center, event classification and notification, and development of protective action recommendations.
The inspectors also observed Operations Support Center activities. The inspectors checked to see that deficiencies noted during the drill, by either the inspectors or licensee evaluators, were entered into the licensees corrective action program. The inspectors also attended portions of the post drill critique for the Technical Support Center and Operations Support Center crews. Documents reviewed as part of this inspection are listed in the Attachment. This activity constituted one inspection sample.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to determine whether they were being entered into the licensees corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Minor issues entered into the licensees corrective action program as a result of the inspectors observations are generally denoted in the Attachment. These activities were part of normal inspection activities and were not considered separate samples.
b. Findings
No findings of significance were identified.
Introduction The inspectors reviewed the adequacy of the licensees corrective actions regarding several apparent, recurring issues associated with the containment spray (CS) and CW systems, occurring since November 2004. The inspectors reviewed the cause of multiple events to determine if the licensees corrective actions were effective at preventing repeat failures. Those documents reviewed during this inspection are listed in the Attachment. This activity completed one sample.
a.
Prioritization and Evaluation of Issues
- (1) Inspection Scope The inspectors assessed the licensees evaluation and disposition of performance issues, and application of risk insights regarding the tripping of the 1B CW pump on January 28 and the 2A CW pump on February 17, 2005. The inspectors also reviewed the licensees maintenance rule a(1) determination for the CW system dated May 27, 2005, and a July 2005 common cause analysis addressing several pump and motor issues associated with various systems.
- (2) Observations The inspectors noted that each CW pump trip event was captured in an IR and was assigned a priority consistent with its significance. The inspectors also noted that similar CW pump trips had occurred in 1998 and in October 2004. The licensee had identified several possible event initiators for these trips, but had not determined which ones were most probable. The licensee was planning to perform a root cause evaluation to determine the most likely cause of the CW pump trips. This root cause was part of the licensees a(1) evaluation and was being tracked by IRs 295475 and 298584.
The inspectors noted that the July 2005 common cause evaluation had appropriately reviewed the pump and motor issues, including whether similar issues had occurred in the industry. Other than the CW pump tripping issues discussed above, there were no common cause issues related to the other pumps or motors. Additionally, the corrective actions taken to address each of the other pump and motor problems were considered adequate. The results of this evaluation were presented to the station Plant Health Committee for review. Overall, the inspectors determined that the licensee was adequately addressing this issue.
b.
Effectiveness of Corrective Actions
- (1) Inspection Scope The inspectors reviewed the licensees actions to address recurring issues with CS pump limit switch failures and a potential common cause failure mechanism involving the CW pump motor ammeters.
- (2) Observations On March 28, 2005, during post-maintenance testing of the 1A CS pump, a failure of a limit switch prevented the testing from being completed, resulting in additional, unplanned unavailability. Although the failed limit switch prevented manual start of the pump during testing, it did not prevent the pump from performing its safety-function (i.e.,
auto-start) during an actual event. The licensee was aware of previous, similar problems with CS limit switches, occurring in 1996 and 2004, that had also resulted in unplanned unavailability. The inspectors noted that for these earlier occurrences, the licensee had not performed a cause analysis and had therefore missed an opportunity to prevent recurrence. Following the March 2005 event, the licensee performed an apparent cause evaluation and had addressed the extent of condition for all containment spray valve limit switches.
On November 14, 2004, the 2A CW pump tripped due to a failure of the motor ammeter.
The cause of the failure was overheating of the ammeter lead which caused it to become disconnected. However, in investigating the issue, the licensee identified that the ammeters currently installed in the CW pumps were significantly less reliable than those originally installed. In 2001, the licensee replaced the originally installed ammeters with those made by another manufacturer. This replacement was necessary due to the unavailability of the original Electric ammeters. Since it was considered a like-for-like replacement, no parts evaluation was performed for this change. In investigating the above failure, the licensee identified that the new ammeters had a service life of about 3 years, compared to 10 years for the original ammeters. In particular, the licensee discovered that the new ammeters had an industry history of overheating, resulting in the failure of soldered connections. It was possible that this information would have been identified earlier, had a parts evaluation been performed prior to replacement. The licensee was evaluating the adequacy of the preventative maintenance program for these ammeters and whether more reliable ammeters were available in the industry. These actions were being tracked under IR 273286. The inspectors concluded that the licensee was reasonably addressing this issue.
.3 Semiannual Review for Trends
a. Inspection Scope
The inspectors reviewed all IRs generated during the time period between June 1 and November 30, 2005, in an attempt to identify potential trends involving adverse human or equipment performance. This inspection was part of the requirements of Inspection Procedure 71152 for monitoring plant status but was not considered an inspection procedure sample. Documents reviewed which indicated previously unrecognized trends are listed in the Attachment. The inspectors verified that minor issues identified during this inspection were entered into the licensees corrective action program.
The screening was accomplished by grouping IRs into broad categories during daily screening. These groups included, but were not limited to, items involving the same issue, same equipment/components, or the same program. For the period of review, the inspectors also obtained lists of all completed or ongoing licensee common cause investigations, all IRs where the title indicated a trend or potential trend, all systems currently in the maintenance rule (a)(1) status, and the licensees most recent System Health Indicator Program (SHIP) report. These documents were considered licensee-identified trends. The following items were eliminated from the scope of this inspection:
- IRs dealing with company policies, administrative issues, and other minor issues;
- IRs associated with established licensee trending programs/processes, such as the rework program, that were previously reviewed during the semi-annual trend evaluation discussed in Inspection Report 2004-04;
- single IRs with no repeat occurrences or common issues;
- IRs that discussed NRC-identified trends from previous inspection activities;
- IRs that discussed strictly programmatic problems, as the inspection specifically focused on human and equipment performance issues;
- IRs involving Security, Radiation Protection, ISI and Emergency Preparedness issues, that were reviewed by regional specialists during ongoing inspection activities;
- IRs that were duplicates of other IRs involving the same event or failure; and
- IRs generated as a result of a special licensee initiative to specifically look for issues in a certain area.
The daily review of equipment issues included all safety related systems and those systems and components identified as risk significant by the licensees probabilistic risk assessment model. In addition a focused review of the safety injection, CW, and rod control systems was performed by the inspectors. Potential trends identified by the inspectors were provided to the licensee for discussion and additional followup.
b. Finding and Observations The inspectors determined that licensee employees were writing IRs with a low threshold that employees at all levels of the organization were writing IRs, and that IRs were written for all issues of significance. The inspectors noted a large number of IRs for employee identified equipment issues. Collectively, this provided one indication of an effective safety conscious work environment.
The majority of the trends were identified by the licensee. Each trend was documented in an IR and evaluated to determine if a common cause evaluation was necessary. The licensee-identified trends were identified by a combination of the work groups involved with the issues, department or station corrective action program coordinators, department managers, and the nuclear oversight group, indicating that multiple groups were looking for trends.
The following potential trends were identified wholly or in part by NRC inspection:
- poor control of temporary power cords (IR 352191);
- several instances of inadequate engineering technical rigor (IR 390585);
- several instances of station heaters requiring repair (IR 426859).
All NRC identified trends have been entered into the licensees corrective action program. No violations of NRC requirements were identified.
4OA3 Event Followup
The inspectors completed two inspection samples in this area.
.1 Offsite Tritium Contamination From Prior Circulating Water Vacuum Breaker Leaks
a. Inspection Scope
On November 30, 2005, the inspectors were notified that the licensee had measured tritium levels as high as 58,000 picoCuries per Liter (pCi/L), in shallow, monitoring wells located at the northern edge of the owner controlled area. The inspectors evaluated the extent and possible cause of these unexpected sample results. Additionally, the NRC performed an independent analysis of split samples taken from some of the licensees monitoring wells and collected independent samples from some residents nearest to the site boundary.
b. Findings
Introduction:
The inspectors noted an unresolved item regarding whether the licensee had fully characterized the extent of the tritium contamination, whether the source of the contamination was properly identified, whether the licensee had correctly evaluated the integrity of the blowdown line, whether corrective actions were appropriately developed to prevent future releases from the blowdown line, and whether the licensee adequately evaluated potential mitigative actions for the tritium already released. Because this inspection effort was contingent on the licensees planned actions as stated above, this issue is being tracked as an Unresolved Item.
Description:
On November 30, 2005, the inspectors were notified that the licensee had measured tritium levels as high as 58,000 pCi/L in shallow, monitoring wells located at the northern edge of the owner controlled area. Three homes and a larger parcel of vacant land with undeveloped lots were located near the area where the tritium was identified. Additional home sites were also located to the North and Northeast. On December 12, 2005, the licensee held a public meeting to discuss the tritium contamination with those residents living in or near the potentially affected areas. This meeting was attended by NRC staff from the resident and regional offices.
The licensee contacted several of the homeowners with drinking water wells and drilled monitoring wells to determine the extent of the tritium contamination. The NRC performed an independent analysis of split samples taken from some of the licensees monitoring wells and collected independent samples from some residents nearest to the site boundary. The NRC initial sample results were consistent with the licensees results. As of December 30, 2005, the licensee had identified levels of tritium between 1400 - 1600 pCi/L in one residential drinking water well. The tritium levels found in the residential drinking water well were below the Environmental Protection Agency (EPA)drinking water standard of 20,000 pCI/L which equates to an annual dose of 4 millirem and is below NRC dose limits. The other residential well samples had no measurable tritium above normal background. In addition, tritium levels as high as 225,000 -
250,000 pCi/L were measured in non-residential, deep wells (about 25 feet) both onsite and offsite.
The inspectors discussed the potential origin of the tritium contamination with licensee staff. The tritium likely originated from past leakage of the vacuum breakers on the CW blowdown line. This line normally carried non-radioactive CW discharge back to the Kankakee River, but also served as a dilution pathway for planned liquid radioactive releases. The line was about five miles long and had eleven vacuum breakers. These breakers compensated for potential voiding from liquid surges. In November 1998 and December 2000 respectively, significant leakage from two of these breakers flooded a portion of the plant site with several million gallons of water in each occurrence. Each leak occurred over a period of several days coincident with ongoing, liquid radioactive releases through the blowdown line. The licensee sampled, collected, and returned the water from the 2000 event back into the blowdown line, but at the time of the inspections was unable to determine if any action was taken for the 1998 event. A small leak of a few gallons also occurred from another vacuum breaker in May 2005, but the licensee recovered most of this leakage before it could escape to the environment.
The inspectors reviewed the licensees root cause report for the 2000 vacuum breaker release event. This event was caused by damage to the float assembly of the vacuum breaker valve due a lack of surge protection and an inadequate preventative maintenance program. Specifically, the valves were missing an internal surge check valve, which protected the float assembly from damage due to periodic liquid surges inside the pipe. These check valves had never been installed and subsequent preventative maintenance activities had not identified this condition. Although the vacuum breakers were inspected on an annual basis, there was little to no guidance regarding the inspection requirements or documentation of the inspection results. The licensee subsequently replaced all the vacuum breaker valves (including installing the internal surge check valves) and developed a more formal program for inspection of the vacuum breaker valves. The inspection frequency was also changed from annually to semi-annually.
Since the groundwater contamination was identified on November 30, 2005, the licensee has suspended liquid radioactive discharges. During that time, radioactive liquids were being stored onsite in temporary storage tanks. The radioactive discharges will not resume until the blowdown line integrity is verified. On December 21, 2005, the inspectors observed the transfer of radioactive liquid into the temporary tanks.
The licensee planned to perform monitoring of the blowdown line integrity using a vendor supplied and operated acoustical leak detection system. On December 22, 2005, a regional engineering specialist reviewed the licensees testing procedure and observed the installation of test equipment to perform the acoustic leak detection for the portion of the blowdown line between vacuum breakers 0CW-060 and 0CW-138. The licensee expected to complete the testing of the complete blowdown line by the end of February 2006.
The licensee continued to perform groundwater monitoring to properly characterize the extent of the offsite tritium contamination. In addition to the monitoring wells already installed, the licensee planned to install additional wells to evaluate the vertical characterization of the aquifer. Monitoring wells were also planned along the length of the blowdown line towards the river. Concurrent with these actions, the licensee was also performing a root cause evaluation to confirm that the source of the tritium was the vacuum breaker leaks and to evaluate the adequacy of corrective actions for prior leaks, and was developing plans for potential mitigation of the tritium. The actions were being tracked by licensee IR 428868, dated November 30, 2005.
The inspectors planned further reviews to determine whether the licensee had fully characterized the extent of the tritium contamination, whether the source of the contamination was properly identified, whether the licensee had correctly evaluated the integrity of the blowdown line, whether corrective actions were appropriately developed to prevent future releases from the blowdown line, and whether the licensee adequately evaluated potential mitigative actions for the tritium already released. Because this inspection effort was contingent on the licensees planned actions as stated above, this issue is being tracked as an Unresolved Item (URI 05000456/2005010-02; 05000457/2005010-02).
.2 (Closed) Licensee Event Report (LER) 05000456/2005002-00: Feedwater Isolation
Valve 1FW039A Fails to Stroke In the Required Time Due to Failure of Valve Air Regulator to Maintain Set Pressure.
On August 3, 2005, the licensee identified that feedwater isolation valves 1FW039A-D had not been stroke time tested in accordance with the In-Service Testing Program. In 1998, the testing frequency of these valves was changed from every cold shutdown to quarterly (i.e., every 92 days). However, the station predefined surveillance program was not changed to reflect this, resulting in these valves only being tested every cold shutdown. Subsequently the predefined surveillance program was revised to reflect the new testing frequency. Surveillance Requirement 3.0.3 was also entered, which required that the valves be stroke timed within 92 days from the date of discovery of the missed surveillance.
On September 6, 2005, the 1FW039A failed a stroke time test, with a measured stroke time of 6.58 seconds vs. the TS limit of 6.0 seconds. The other feedwater valves met their stroke time requirements. In accordance with TS 3.6.3, Condition C, the 1FW039A valve was closed and declared inoperable, the upstream isolation valve (1FW041A) was closed, and administrative actions were taken to ensure that both valves remained closed. The cause of the failed stroke time was foreign material (i.e., small piece of wire and hard plastic) found inside the regulator main seat area. This material was believed to have been introduced during prior maintenance on the valve in October 2001. The material caused the regulator supply pressure to exceed its setpoint, resulting in additional time for the actuator and solenoid valves to exhaust the supplied air allowing the valve to close. The regulator was subsequently replaced and on September 10, 2005, the 1FW039A passed its time test.
These valves were required to isolate upon receipt of a phase A containment isolation signal. These valves also automatically closed when a safety injection signal was used to initiate feedwater isolation to mitigate the effects of high energy line breaks both inside and outside containment. The valves also fail closed on loss of electrical control power or air. These valves were normally maintained closed, as they were not required to be opened for any required functions. Since these valves were normally closed, the failure to perform the required surveillance testing on the 1FW039A-D and the subsequent stroke time failure of 1FW039A, did not have an affect on either the containment or feedwater isolation functions.
The failure to perform required surveillance testing of the 1FW039A-D valves and the subsequent failure of the 1FW039A valve to meet its required stroke time were considered violations of minor significance that are not subject to enforcement action in accordance with Section IV of the NRCs Enforcement Policy. The licensee documented this problem in IRs 359689 and 370649. This LER is closed.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. G. Boerschig and other members of licensee management at the conclusion of the inspection on January 9, 2006. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
.2 Interim Exit Meetings
Interim exit meetings were conducted for:
- Operator Requalification Program Examination Result Review via telephone conversation with Mr. C. Dunn on October 27, 2005.
- Emergency Preparedness inspection with Mr. Scott McCain and Ms. Kim Aleshire by telephone call on December 28, 2005.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- K. Polson, Site Vice President
G. Boerschig Plant Manager
- K. Aleshire, Emergency Preparedness Manager
- D. Ambler, Regulatory Assurance Manager
- M. Andrews, Chemistry Supervisor
- J. Bauer, Licensing Manager
- S. Butler, Licensing Engineer
- M. Cichon, Regulatory Assurance
- S. Clark, Maintenance Planning Manager
- G. Dudek, Operations Director
- C. Dunn, Training Manager
- C. Gayheart, Operations Training Director
- G. Heisterman, Mechanical Maintenance Manager
- J. Kuczynski, Chemistry Manager
- R. Leasure, Radiation Protection Manager
- F. Lentine, Design Engineering Manager
- S. McCain, Corporate Emergency Preparedness Manager
- J. Moser, Radiation Protection Manager
- M. Olson, Simulator Coordinator
- A. Ronstadt, Site Maintenance Rule Coordinator
- J. Ruth, Examination Developer
- M. Sears, Steam Generator Program Manager
- M. Smith, Site Engineering Director
- P. Summers, Nuclear Oversight Manager
- M. Trusheim, Work Control Manager
Nuclear Regulatory Commission
- R. Skokowski, Chief, Reactor Projects Branch 3
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened
- 05000456/2005010-01 NCV 10 CFR 50.54(q) Violation for Decreasing the Effectiveness of the Emergency Plan by Changing EAL RU2 Threshold That Address Radiological Effluents Without Prior NRC Approval or Adequate CFR 50.54(q) Review
- 05000456/457/2005010-02 URI Tritium Contamination from Past Vacuum Breaker Leaks on Circulating Water Blowdown Line Attachment
Closed
- 05000456/2005010-01 NCV 10 CFR 50.54(q) Violation for Decreasing the Effectiveness of the Emergency Plan by Changing EAL RU2 Threshold That Address Radiological Effluents Without Prior NRC Approval or Adequate CFR 50.54(q) Review
- 05000456/2005002-00 LER Feedwater Isolation Valve 1FW039A Fails to Stroke In the Required Time Due to Failure of Valve Air Regulator to Maintain Set Pressure
Discussed
None.
Attachment