IR 05000315/2008007
| ML082810023 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/06/2008 |
| From: | Ross Telson Reactor Projects Region 3 Branch 4 |
| To: | Rencheck M Indiana Michigan Power Co |
| References | |
| FOIA/PA-2010-0209 IR-08-007 | |
| Download: ML082810023 (20) | |
Text
ber 6, 2008
SUBJECT:
D. C. COOK NUCLEAR PLANT, UNITS 1 AND 2, PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION (05000315/2008-007; 05000316/2008-007)
Dear Mr. Rencheck:
On August 29, 2008, the U. S. Nuclear Regulatory Commission (NRC) completed a routine biennial Problem Identification & Resolution (PI&R) inspection at your D.C. Cook Nuclear Power Plant. The enclosed report documents the inspection results, which were discussed on August 29 with 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.
On the basis of the sample selected for review, there were no findings of significance identified during this inspection. The team concluded that problems were properly identified, evaluated, and resolved within the Corrective Action Program (CAP). However, during the inspection, several examples of minor problems were identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Ross Telson, Chief Projects Branch 4 Division of Reactor Projects Docket Nos. 50-315; 50-316 License Nos. DPR-58; DPR-74 Enclosure: Inspection Report 05000315/2008-007; 05000316/2008-007 w/Attachment: Supplemental Information cc w/encl: L. Weber, Site Vice President J. Gebbie, Plant Manager G. White, Michigan Public Service Commission Michigan Department of Environmental Quality Planning Manager, Emergency Management and Homeland Security Division, Michigan State Police Department T. Strong, State Liaison Officer
SUMMARY OF FINDINGS
IR 05000315/2008-007; 05000316/2008-007; 08/11/2008 - 08/29/2008; D. C. Cook Nuclear
Power Plant, Routine Biennial Problem Identification and Resolution Inspection.
This inspection was conducted by four regional inspectors. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,
Reactor Oversight Process, Revision 4, dated December 2006.
Identification and Resolution of Problems The inspection team concluded that, based on the samples reviewed, the corrective action (CA)program was capable of effectively identifying, evaluating, and resolving issues. The licensee staffs actions were in compliance with the facilitys CAP and 10 CFR Part 50, Appendix B requirements. Specifically, the inspectors concluded that licensee personnel were identifying plant issues at a low threshold, entered the plant issues into the stations CA program in a timely manner, performed an adequate evaluation of the issue and implemented corrective actions in an effective manner. Minor examples of inadequate implementation of the processes were observed and the inspection record indicated that several issues were self-revealed or identified by external organizations. Licensee performance with operating experience, self assessments, audits and maintaining a safety conscious work environment was effective.
REPORT DETAILS
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
Completion of sections a. through d. constitutes 1 biennial sample of problem identification and resolution as defined by Inspection Procedure 71152.
a.
Assessment of the CAP Effectiveness
- (1) Inspection Scope The inspectors reviewed the licensees CA program implementing procedures and attended CA program meetings to assess the implementation of the CA program by site personnel. The inspectors reviewed risk significant issues in the licensees CA program that overlap the last NRC PI&R inspection completed in August 2006. The selection of issues ensured an adequate review of issues across NRC cornerstones.
The inspectors reviewed condition reports (CRs) generated as a result of daily plant activities. In addition, the inspectors reviewed CRs that included a selection of completed investigations from the licensees various investigation methods, which included root cause, in depth apparent cause, apparent cause, and trending performance investigations. The inspectors evaluated the timeliness and effectiveness of corrective actions for selected condition reports, completed investigations, and NRC findings, including non-cited violations (NCVs). The inspectors also reviewed issues identified through NRC generic communications, department self assessments, licensee audits, operating experience reports, and NRC documented findings.
The inspectors performed a five year review of diesel generator issues documented in corrective action documents to assess the licensee staffs efforts in monitoring for any system degradation due to aging. The inspectors performed partial system walkdowns of the diesels.
Inspectors reviewed the classification of condition reports for resolution to determine if the licensee appropriately determined the investigation methods. Inspectors also attended Initial Screening Committee and Management Screening Committee meetings to observe the review of CRs for classification.
- (2) Assessment - Effectiveness of Problem Identification Based on the information reviewed, the inspectors concluded that the threshold for initiating condition reports was low and well below the plant procedural requirements.
Nevertheless, licensee expectations for CR initiation were not being consistently met as evidenced by CRs prompted by NRC inspectors. In addition, due to errors in coding of the condition reports, the inspectors could not determine how many condition reports were the result of event-driven issues.
The inspectors concluded that the program was effective at identifying issues. However, a review of the NRC findings identified in the last two years which were self revealing or identified by external organizations illustrated that improvement could be made.
For example, the residents identified that the licensee failed to perform adequate maintenance on the turbine building sump overflow check valve.
OBSERVATIONS During the inspection, the team requested data on the number of condition reports that were coded as event-driven, external driven and self identified. The Cook process describes event-driven as those that are self evident and require no active or deliberate observation by the licensee. Self-identified conditions are those that are discovered through licensee programs or those that are found through deliberate observation.
External-driven conditions are those found by an outside agency, e.g. the NRC. The inspectors noted that the number of event-driven condition reports was abnormally low.
For example, in the first quarter of 2007, only 11 CRs out of 1553 were event-driven.
The inspectors reviewed CRs in the review process and identified multiple examples where a CR was incorrectly classified as self identified. For example, a CR where a worker had snapped a bolt head had been classified as self identified.
No findings of significance were identified.
- (3) Assessment - Effectiveness of Prioritization and Evaluation of Issues During Initial Screening Committee and Management Screening Committee meetings observed by the inspectors, all CRs were assigned appropriate prioritization and evaluation levels.
The inspectors determined that the evaluations in root cause reports and apparent cause reports that were reviewed were adequate. The corrective actions addressed the identified problems and the timeliness of corrective actions was appropriate to the safety significance. However, the inspectors noted some minor weaknesses in evaluations and identification of corrective actions.
OBSERVATIONS During review of the licensees condition reports related to NCVs 05000315/2006-004-01 and 02, the inspectors noted that the licensee had not fully considered the impact of a seiche on equipment in the service water screen house. The NCVs addressed, in part, protection of safety-related equipment from the effects of a seiche up to an elevation of 595 feet above sea level. Although the licensee relocated vulnerable equipment to above the 595 level, the licensee did not consider the potential for spray or sloshing to affect equipment. In addition, the off-normal procedure in use by the facility did not accurately describe the means of plant notification if conditions were favorable for a seiche. The licensee documented these issues in CRs 00836575 and 00836451, respectively.
The inspectors reviewed CR 06090039 that addressed a significant condition adverse to quality involving the drop of a bulkhead section in containment. The corrective actions in the evaluation were too narrowly defined. As a corrective action, the licensee revised procedures to require use of a load cell in instance where a load could bind. The inspectors concluded that this corrective action was too narrowly focused because they did not address other conditions which could overload rigging. However, subsequent to the root cause, the licensee revised the governing procedures to use a load cell for all lifts in excess of four tons.
The inspectors reviewed Action Request (AR) 00804579 which evaluated a condition prohibited by Technical Specifications (TS). During performance of diesel generator load testing, commonly known as Loss of Offsite Power/Loss of Coolant Accident testing, the licensee rendered containment isolation inoperable with containment purge in operation and fuel moves in progress. The violation occurred when the licensee took the solid state protection system switch to the test position. In the in-depth apparent cause evaluation, the licensee recognized several failed barriers that resulted in the technical specification violation. However, the licensee did not recognize weaknesses in the infrequently performed test and evolution briefs. Specifically, the brief did not address the impact of the testing on Technical Specifications. In addition, the operating experience included in the brief was not adequate in that it did not discuss an earlier instance when power was removed from the solid state protection system switch.
During review of AR 00806546, which addressed the blow out of packing on an instrument isolation valve when mechanics attempted to adjust the packing, the inspectors noted that the licensee completed the root cause prior to evaluation of the material condition of the valve. Since the valve could not be fully evaluated for several months, the inspectors concluded that completion of the non-material portion of the root cause prior to completion of valve diagnostics was reasonable. However, once the licensee completed the material evaluation, the information was not re-evaluated for additional insights regarding the failure of the valve. When the inspectors discussed the condition with Boric Acid Control program owner, the program owner stated that when similar valves had indications of a packing leak, the valves would be scheduled for replacement during the next outage. In addition, operations had informally agreed to prohibit packing adjustments on similar valves. The inspectors concluded that the informal practices lacked sufficient rigor to ensure future implementation of these actions. The licensee initiated AR 00837146 to modify plant procedures to document these practices.
- (4) Assessment - Effectiveness of Corrective Actions The inspectors concluded that over the two year period encompassed by the inspection, the licensee implemented effective corrective actions. The inspectors identified no significant examples were problems recurred.
No findings of significance were identified.
b.
Assessment of the Use of Operating Experience
- (1) Inspection Scope The inspectors reviewed a sample of operating experience (OE) issues for applicability to D.C. Cook and the associated actions American Electric Power implemented to address the potential issues. The inspectors selected the samples from NRC Generic Communications, industry OE sources, reports made pursuant to 10 CFR 21, and noteworthy issues from other reactor sites. The inspectors reviewed the method in which OE was communicated throughout the station, and where appropriate, verified that applicable issues were entered into the CA program. In the review of associated evaluations, the inspectors evaluated whether the problems associated with each issue were appropriately considered for resolution in accordance with the CAP process.
- (2) Assessment The inspectors observed that the licensee transmitted industry OE daily to each department OE Coordinator. Regulatory Affairs staff reviewed NRC Information Notices, Generic NRC correspondence, and Westinghouse advisories and bulletins for applicability. The inspectors identified no examples where the licensee failed to take appropriate action based on the OE. The inspectors observed a daily screening meeting and concluded that the licensee adequately screened OE.
No findings of significance were identified.
c.
Assessment of Self-Assessments and Audits
- (1) Inspection Scope Inspectors reviewed samples of the governing procedures, schedules, plans, reports, and resulting CRs for licensee self assessments and quality assurance audits. A sample of corrective actions generated for issues was also reviewed.
- (2) Assessment The inspectors reviewed the self-assessment of the CAP performed by the Performance Assurance organization. The inspectors determined that the assessment results were similar to the observations of the inspectors. The inspectors also reviewed other self-assessments as well as Excellence Plan Health Indices listed in the references. The inspectors identified no areas of concern.
No findings of significance were identified.
d.
Assessment of Safety Conscious Work Environment
- (1) Inspection Scope The inspectors interviewed approximately 25 plant workers, across all major work groups and all levels of responsibility to assess the safety-conscious work environment at the station. The inspectors conducted the interviews using the guidance provided in Appendix 1 of NRC Inspection Procedure 71152, Suggested Questions for Use in Discussions with Licensee Individuals Concerning PI&R Issues. The licensee recently promulgated a new policy that would discipline employees that were injured as a result of at risk behavior. Therefore, the inspectors specifically asked questions about the employees willingness to raise issues associated with industrial safety. Employees remained willing to report safety issues.
In addition to the interviews, the inspectors reviewed documents and observed activities to determine if plant employees might be reluctant to raise safety concerns. As part of the inspection, the inspectors reviewed the latest safety culture self-assessment. This report had been previously reviewed in IR 05000315/2008502, 05000316/2008502. The results of the self-assessment were consistent with the inspectors observations. The inspectors also reviewed the station procedures related to the Employee Concerns Program, and discussed implementation of the program with the program manager.
- (2) Assessment The inspectors concluded that employees report issues at a low level. Many employees expressed concern that at-risk behavior had not been sufficiently defined. Workers indicated that they felt comfortable identifying issues and discussing concerns with supervision without fear of reprisal.
The inspectors observed that personnel interviewed were aware of the different avenues through which they could express concerns including the CAP, informing their supervision, contacting the Employee Concerns Program coordinator, or contacting the NRC. The majority of workers interviewed preferred utilizing the CAP as their first avenue to raise both nuclear and industrial safety concerns.
No findings of significance were identified.
e.
Other Activities (Closed) Licensee Event Report (LER) 05000315/2008-003-00, Failure to Comply with Technical Specification Limiting Condition for Operation 3.0.6 On March 5, 2008, while in the process of preparing to remove the Unit 1 East Essential Service Water (ESW) Pump from service for scheduled maintenance, the Control Room Senior Reactor Operators questioned the need for a Safety Function Determination and documentation as required by procedures and Technical Specifications (TS). This action delayed the pump maintenance based on the potential consequences that would have existed as a result of having the Unit 2 East Motor Driven Auxiliary Feedwater (AFW) Pump Room Cooler out of service in conjunction with the Unit 1 East ESW Pump.
More specifically, removal of U1 East ESW Pump from service would have rendered redundant trains of AFW inoperable since the Unit 1 ESW East Train is cross tied with the Unit 2 ESW West train. With the system configured as it was, with cross ties open between units, the unit 2 west train would have been required to be declared inoperable.
Upon the suspension of U1 maintenance activities, prior to removing U1 East ESW pump from service, the licensee identified that, as noted in the licensees evaluation, a safety function determination is required any time an ESW system is taken out of service, unless all TS Conditions and Required Actions for the systems supported by the ESW train are entered. Contrary to procedural requirements, which require formal documentation, undocumented reviews were performed in this case and in approximately 115 other cases, to verify that a loss of safety function would not occur as a result of planned maintenance. In all cases, the licensee was able to demonstrate that the safety function was retained.
The Licensee entered this issue into their CAP, and as a result has implemented 6 corrective actions to restore their facility within compliance when a safety function determination becomes necessary. These actions include: implementing training to address knowledge and proficiency issues related to this event; incorporating of the Safety Function Determination Program into Operations Head Procedure; briefing by Shift Management to all operations crews on the rules of usage for TS LCO 3.0.6 and the requirements of the safety function determination procedure; clarifying PMP-7030-SFD-001 to reinforce its use a guide for evaluating upcoming maintenance activities for potential loss of safety function; benchmarking of safety function determination Procedure by Operations to ensure administrative burden is minimized in the control room for pre-planned maintenance activities; and ensuring that power log standard entries are changed to meet the requirements of PMP-7030-SFD-001 and TS LCO 3.0.6 when applicable.
This failure to comply with TS LCO 3.0.6 constitutes a violation of minor significance and is not subject to enforcement action in accordance with Section IV of the NRCs enforcement policy because the issue was administrative in nature and had no associated loss in safety function.
This LER is closed.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- R. Crane, Learning Organization Manager
- J. Gebbie, Plant Manager
- J. Hammond, Corrective Action Specialist
- C. Hutchinson, Emergency Preparedness manager
- C. Lane, Engineering programs Manager
- R. Niedzielski, Regulatory Affairs Specialist
- P. Schoepf, Regulatory Compliance Supervisor
- L. Weber, Site Vice President
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Closed
- 05000215/2008-003-00 LER Failure to Comply with Technical Specification Limiting Condition for Operation 3.0.6 (4OA2)
Attachment