IR 05000247/2006006

IR 05000247-06-006; 09/18/2006 - 10/06/2006; Indian Point Nuclear Generating Unit 2; Problem Identification and Resolution

ML063560335
Person / Time
Site:	Indian Point
Issue date:	12/21/2006
From:	David Lew Division Reactor Projects I
To:	Dacimo F Entergy Nuclear Operations
References
EA-06-311, FOIA/PA-2007-0080, FOIA/PA-2016-0148 IR-06-006
Download: ML063560335 (55)

Text

December 21, 2006

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNITS 2 AND 3 PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT NOS.

05000247/2006006 AND 05000286/2006006 AND NRC REQUEST FOR RESPONSE

Dear Mr. Dacimo:

On October 6, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed concurrent biennial problem identification and resolution team inspections at the Indian Point Nuclear Generating Units 2 and 3. The enclosed inspection reports document the inspection observations and findings which were discussed with Entergy management during an exit meeting onsite on December 5, 2006, and during a teleconference meeting on December 14, 2006.

The inspections were examinations of activities conducted under your licenses as they relate to the identification and resolution of problems, and compliance with the Commission's rules and regulations and the conditions of your licenses. Within these areas, the inspections involved examination of selected procedures and representative records, observations of activities, and interviews with personnel.

The inspection teams concluded that Entergys implementation of the corrective action program at the Indian Point Nuclear Generating Units 2 and 3 was consistent across both units and generally effective. The teams determined that Entergy staff had a low threshold for identifying problems, and issues were prioritized and evaluated commensurate with their safety significance. Corrective actions were typically implemented in a timely manner and addressed the identified causes of the problems. Lessons learned from industry operating experience were reviewed and applied when appropriate, and audits and assessments were critical with appropriate actions taken to address identified issues in most cases.

During the inspection, the team conducted interviews and reviewed specific concerns to understand the application and effectiveness of the corrective action program in support of ensuring an environment where employees feel free to raise concerns. In the context of the safety conscious work environment onsite, the inspection teams observed that most workers

2 indicated that they would raise issues that they recognized as nuclear safety issues. The NRC has become aware of incidents through insights gained during these inspections and from the allegation program where workers perceived that individuals were treated negatively by management for raising issues. As a result of these incidents, some workers expressed reluctance to raise issues under certain circumstances. While most workers made a distinction between nuclear safety issues and other concerns, the teams noted that some of the illustrative examples provided by plant workers could have nuclear safety implications. However, the teams did not identify any more than minor issues, which had not been raised.

Additionally, in June 2006, the NRC referred concerns to you for your information and action involving an alleged potential chilling effect in the Maintenance department. This referral specifically referenced issues identified in a teamwork assessment of the Instrumentation and Controls department, conducted in 2005, as well as the preliminary results of an independent safety culture assessment sponsored by Entergy in early 2006. Our followup during these inspections found that you had deferred action on the referred concerns pending evaluation of the Entergy contracted safety culture assessment, and as a result, you had not taken substantive action at the time of the inspection. Therefore, the team was unable to review or evaluate your actions to address the potential adverse impact on the safety conscious work environment within the Maintenance department.

We recognize that the information that we received and developed during the inspections regarding the willingness of workers to raise issues is generally consistent with the results of the independent safety culture assessment conducted for Entergy at Indian Point in 2006. We understand you have taken actions to improve the general plant culture at Indian Point and there are ongoing initiatives at the site dealing with expectations for workforce performance.

However, we are concerned that at the conclusion of the inspection you had not fully evaluated the results of the Entergy contracted safety culture assessment to understand the causes of the negative responses and declining trends related to the safety conscious work environment onsite. As a result, the NRC requests that Entergy provide its plan for evaluating the potential chilling effect onsite and its plan of action for addressing this matter to the NRC. Based on our discussions of December 14, 2006, it is our understanding that Entergy agrees to provide this information within 30 days of the date of this letter. Following receipt and review of the Entergy response, we will determine if a meeting is needed to discuss your approach, schedule, and NRC oversight. This issue will be included as an input to our assessment of plant performance as described in Inspection Manual Chapter 0305, "Operating Reactor Assessment Program."

In addition to the above observations, there were three Green findings identified by the inspectors during these inspections: two findings at Unit 2 and one finding common to both units. Two of the findings were determined to be violations of NRC requirements. However, because of their very low safety significance and because they were entered into your corrective action program, the NRC is treating these findings as Non-Cited Violations (NCVs), in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you deny any of these NCVs, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the U.S. Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC, 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001; and the NRC Resident Inspector at the Indian Point facility.

3 In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and your response 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/NRC/ADAMS/index.html (the Public Electronic Reading Room).

Because your response will be made available electronically for public inspection in the NRC Public Document Room or from the PARS component of ADAMS, to the extent possible it should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your withholding claim (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

If you have any questions regarding these matters, please contact Eugene Cobey at (610) 337-5171.

Sincerely,

/RA/

David C. Lew, Director Division of Reactor Projects Region I Docket Nos.

50-247 and 50-286 License Nos. DPR-26 and DPR-64

Enclosures:

Inspection Report Nos. 05000247/2006006 and 05000286/2006006 w/Attachments: Supplemental Information

REGION I==

Docket No:

50-247 License No:

DPR-26 Report No:

05000247/2006006 Licensee:

Entergy Nuclear Northeast (Entergy)

Facility:

Indian Point Nuclear Generating Unit 2 Location:

295 Broadway, Suite 3 Buchanan, NY 10511-0308 Dates:

September 18 through October 6, 2006 Team Leader:

T. Walker, Senior Project Engineer, Division of Reactor Projects (DRP)

Inspectors:

M. Cox, Senior Resident Inspector, DRP S. McCarver, Project Engineer, DRP J. Benjamin, Resident Inspector, DRP C. Long, Project Engineer, DRP Observer:

S. Smith, Reactor Engineer, DRP Approved by:

Eugene W. Cobey, Chief Projects Branch 2 Division of Reactor Projects

Enclosure ii

SUMMARY OF FINDINGS

IR 05000247/2006-006; 09/18/2006 - 10/06/2006; Indian Point Nuclear Generating Unit 2;

Problem Identification and Resolution.

This team inspection was performed by three regional inspectors and two resident inspectors.

Three findings of very low safety significance (Green) were identified, two of which were also non-cited violations (NCVs). 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.

Identification and Resolution of Problems The inspectors concluded that the implementation of the corrective action program at Indian Point Unit 2 was generally effective. The inspectors noted that Entergy staff had a low threshold for identifying problems and entering them in the corrective action program. The inspectors also noted that once entered into the system, items were screened, prioritized, and evaluated commensurate with their significance using established criteria. The inspectors determined that corrective actions addressed the identified causes and were typically implemented in a timely manner. In addition, the team noted that Entergy was generally effective in reviewing and applying lessons learned from industry operating experience. The inspectors found that audits and assessments were critical and, in most cases, appropriate actions were taken to address identified issues. However, the inspectors also found that the results of an independent safety culture assessment were not entered into the corrective action program for timely evaluation and appropriate action.

The inspectors found that most workers indicated that they would raise issues that they recognized as nuclear safety issues. However, the inspectors also found that a number of workers interviewed indicated that they were aware of individuals they perceived as having been treated negatively by management for raising issues; most of these workers were in the Instrumentation and Controls (I&C) department. Some workers expressed reluctance to raise issues under certain circumstances due to a number of reasons, including fear of disciplinary action and concerns with the efficacy of the corrective action program. While most workers made a distinction between nuclear safety issues and other concerns, the inspectors noted that some of the illustrative examples provided by plant workers could have nuclear safety implications. However, the inspectors did not identify any more than minor issues, which had not been raised.

There were two Green NCVs and one Green finding identified by the inspectors during this inspection. One of the NCVs was associated with a failure to identify a condition adverse to quality associated with the auxiliary feedwater (AFW) system. The second NCV was associated with a failure to fully evaluate leakage into a steam generator. The finding was associated with the failure to enter adverse conditions into the corrective action program for evaluation and appropriate action.

iii a.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B,

Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to quality associated with improper internal clearances on BFD-68, an auxiliary feedwater check valve, in the corrective action program. Specifically, upon inspection in September 2006, the gasket between the valve's body to bonnet seal was found over-crushed causing the gasket to partially unwind, potentially impacting valve operation. Gasket damage was noted in work orders during internal valve inspections of BFD-68 performed in 1997 and 2002; however, the deficiencies were not identified in the corrective action program. Consequently, the problem was not evaluated and corrected prior to reassembly of the valve. Entergy entered this issue into the corrective action program, evaluated the condition, and conducted repairs to the valve to ensure the proper gasket crush was obtained.

The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events. (Section 4OA2a(3)(a))

Green.

A self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI,

Corrective Action, was identified, in that, Entergy failed to adequately evaluate leakage into the 22 steam generator. During the Indian Point Unit 2 reactor trip on August 23, 2006, main feedwater low flow bypass valve FCV-427L leaked excessively and resulted in an uncontrolled rise in 22 steam generator level; operator response to isolate feedwater to the steam generator in accordance with emergency operating procedures; and automatic actuation of the feedwater isolation system. The excessive leakage condition into the 22 steam generator was identified on April 4, 2006, prior to Indian Point Unit 2 refueling outage 2R17, but was not fully evaluated or corrected prior to the reactor trip on August 23, 2006. This issue was entered into the corrective action program, and FCV-427L was repaired and retested satisfactorily.

The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of the finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not iv result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.

The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolutions addressed the causes and extent of condition of the problem. (Section 4OA2a(3)(b))

Cornerstone: Not Applicable

Green.

The NRC inspectors identified a finding when Entergy failed to initiate condition reports in accordance with EN-LI-102, Corrective Action Process, for the adverse conditions identified in the 2006 Safety Culture Assessment. Consequently, the adverse conditions were not evaluated and appropriate corrective actions were not identified in a timely manner. The contractor who performed the independent safety culture assessment presented the site specific results to Entergy management in June 2006.

The negative responses and declining trends identified in the assessment constituted adverse conditions that should have been entered into the corrective action program. At the time of the inspection, Entergy had not initiated condition reports for the assessment results. Consequently, the results had not been fully evaluated to understand the causes and identify appropriate actions to address the identified issues. Additionally, organizations identified by the contractor as needing management attention had not developed departmental action plans at the time of the inspection. Entergy entered this issue into the corrective action program and initiated a learning organization condition report to track development and implementation of action plans to address the assessment results.

The inspectors determined that the finding was more than minor because if left uncorrected it would become a more significant safety concern. Without appropriate action, the weaknesses in the safety culture onsite would continue, increasing the potential that safety issues would not receive the attention warranted by their significance. The finding was not suitable for SDP evaluation, but has been reviewed by NRC management and has been determined to be a finding of very low safety significance. The finding was not greater than very low safety significance because the inspectors did not identify any issues that were not raised which had an actual impact on plant safety or were of more than minor safety significance.

The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not identify issues with the potential to impact nuclear safety in the corrective action process for evaluation and resolution in a timely manner. (Section 4OA2c(3))

b.

Licensee-Identified Violations

None.

REPORT DETAILS

4OA2 Problem Identification and Resolution (PI&R) (Biennial - IP 71152B)

a.

Assessment of the Corrective Action Program

(1) Inspection Scope The inspection team reviewed the procedures describing the Entergy corrective action program (CAP). Indian Point Unit 2 identified problems for evaluation and resolution by initiating condition reports (CRs) in the Paperless Condition Reporting System (PCRS).

The team evaluated the methods for assigning and tracking issues to assure that issues were screened for operability and reportability, prioritized for evaluation and resolution in a timely manner commensurate with their safety significance, and tracked to identify adverse trends and repetitive issues. In addition, the team interviewed plant staff and management to determine their understanding of and involvement with the corrective action program. The condition reports and other documents reviewed, as well as key personnel contacted, are listed in the Attachment to this report.

The team reviewed condition reports selected across the seven cornerstones of safety in the NRCs Reactor Oversight Program (ROP) to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. The team selected items from the maintenance, operations, engineering, emergency preparedness, physical security, radiation protection, and oversight programs to ensure that Entergy was appropriately addressing problems identified in each functional area. The team selected a risk-informed sample of condition reports that had been issued since the last NRC problem identification and resolution inspection, which was conducted in June 2005.

The team selected items from other processes at Indian Point to verify that they were appropriately considered for entry into the corrective action program. Specifically, the team reviewed a sample of engineering requests (ERs), operability determinations, maintenance work orders (WOs), engineering system health reports, and completed surveillance tests. The team also reviewed completed work packages to determine if issues identified during the performance of preventive maintenance were entered into the corrective action program. In addition, the team attended operations shift turnover meetings and accompanied auxiliary operators during rounds in the plant.

The team considered risk insights from both the NRCs and Entergy's risk assessments for Indian Point Unit 2 to focus the sample selection and plant tours on risk-significant components. The team determined that the systems with the highest risk significance were 480V AC, 125V DC, component cooling water, service water, reactor protection, emergency core cooling system recirculation, safety injection accumulators, and auxiliary feedwater (AFW). Inspector samples focused on these systems, but were not limited to them. The review was expanded to five years for evaluation of selected check valves in the auxiliary feedwater, safety injection and residual heat removal systems.

The inspection team reviewed condition reports to assess whether Entergy adequately evaluated and prioritized identified problems. The issues reviewed encompassed the full range of evaluations, including root cause analyses, apparent cause evaluations, and common cause analyses. The review included the appropriateness of the assigned significance, the scope and depth of the causal analysis, and the timeliness of resolution. The team observed meetings of the Condition Review Group (CRG), in which Entergy personnel reviewed new condition reports for prioritization and assignment, and the Corrective Action Review Board (CARB) where Entergy personnel evaluated root cause evaluations, as well as selected apparent cause evaluations and corrective action assignments. The team also reviewed equipment operability determinations, reportability assessments, and extent-of-condition reviews for selected problems.

The team reviewed the corrective actions associated with selected condition reports to determine whether the actions addressed the identified causes of the problems. The team reviewed condition reports for repetitive problems to determine whether previous corrective actions were effective. The inspectors also reviewed Entergy's timeliness in implementing corrective actions and their effectiveness in precluding recurrence for significant conditions adverse to quality. The team also reviewed condition reports associated with selected NCVs and findings to determine whether Entergy properly evaluated and resolved the issues.

(2) Assessment Identification of Issues One Green NCV was identified in the area of identification of issues for failure to identify improper internal valve clearances on an auxiliary feedwater check valve in the corrective action program for evaluation and resolution.

In general, the team considered the identification of problems at Indian Point to be appropriate. The computer-based condition reporting process, PCRS, facilitates the initiation, tracking and trending of condition reports. Approximately 6,500 condition reports were written each year. There was a low threshold for the identification of issues and, in most cases, problems identified during plant activities were entered into PCRS when appropriate. However, the team found that problems identified in 1997 and 2002 during internal inspections of an auxiliary feedwater check valve were not entered into the corrective action program for evaluation and resolution. As a result, upon inspection in September 2006, the gasket between the valve's body to bonnet seal was found over-crushed causing the gasket to partially unwind, potentially impacting valve operation.

This finding is discussed in detail in Section 4OA2a(3)(a).

Prioritization and Evaluation of Issues One Green NCV was identified in the area of prioritization and evaluation of issues for an inadequate evaluation of leakage into the 22 steam generator.

The team determined that, in general, Entergy appropriately prioritized and evaluated issues commensurate with the safety significance of the issue. Condition reports were screened for operability and reportability, categorized by significance (A through D), and assigned to a department for evaluation and resolution. The Condition Review Group appropriately considered human performance issues, radiological safety concerns, repetitiveness, and adverse trends in their reviews. There were no operability or reportability determinations with which the team disagreed. However, the team did identify a condition report that was improperly categorized which led to insufficient evaluation of the issue. Specifically, the inspectors identified that a condition report which documented a concern regarding security guard readiness was categorized as a

'D' track and trend CR and closed without evaluation. Following discussions with the inspectors, Entergy wrote a new condition report to evaluate and address the issue.

Although the failure to evaluate the condition when it was first raised in 2003 does not comply with NRC requirements, the inspectors determined that due to the nature of the issue it constitutes a violation of minor significance that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

The inspectors found that causal analyses were thorough and appropriately considered extent of condition, generic issues, and previous occurrences. The Corrective Action Review Board reviews were detailed and ensured that corrective actions addressed the identified causes. For significant conditions adverse to quality, corrective actions were identified to prevent recurrence. However, in one case, the inspectors found that Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolution addressed the causes and extent of condition of the problem which adversely impacted the operators ability to control steam generator water level following a reactor trip on August 23, 2006. This finding is discussed in detail in Section 4OA2a(3)(b).

Entergy reviews condition reports site-wide and at the department level to identify adverse conditions occurring at an unacceptable rate or changes in the frequency or severity of events or precursors. The team determined that the monthly reviews and quarterly trend reports provided an effective method for identifying adverse or emerging trends so that actions could be taken in a timely manner to address the issues.

However, the team identified that some departments did not include 'D' condition reports in the trending process. Although the 'D' condition reports were included in the site-wide reviews performed by the Corrective Action and Assessment (CA&A) department, and system engineers tracked all CRs for their assigned systems, adverse or emerging trends within a department could have been missed without trending the 'D' condition reports. Failure to track 'D' condition reports does not comply with Entergy procedure EN-LI-121, "Entergy Trending Process," but the inspectors did not identify any adverse or emerging trends that were not identified. Therefore, the inspectors determined that the issue constitutes a violation of minor significance that is not subject to enforcement action in accordance with the NRC's Enforcement Policy.

Although the evaluation of issues and determination of required corrective actions was generally good, the team identified examples of condition descriptions, dispositions (causal evaluations), and corrective action responses that did not provide clear and complete documentation of the issues and actions taken. This issue had also been identified by Entergy and they were taking corrective actions to improve the stand alone quality of CRs. In the identified cases, the inspectors were able to gather additional information to support the CR packages or the licensee had self-identified the specific issues in the CAP for resolution.

Effectiveness of Corrective Actions No findings of significance were identified in the area of effectiveness of corrective actions.

The team concluded that identified corrective actions were generally appropriate to resolve identified issues, and were typically completed in a timely manner. The inspectors also noted a decreasing trend in the number of items in the backlog of actions to be completed by engineering. However, the inspectors identified a few instances of incomplete or inadequate corrective actions. For example, on October 5, 2006, NRC inspectors noted that the positioner feedback arm for Indian Point Unit 2 high pressure steam dump valve PCV-1121 was not attached to the valve. Industry operating experience information from 1993 and 1997 identified the need to incorporate the verification of tightness of valve positioner feedback arms in preventive maintenance programs due to several incidents caused by feedback arms falling off. The Indian Point Unit 2 condition report response to this operating experience indicated that planned maintenance was performed to verify tightness and that mechanisms were "double nutted" to ensure tightness. NRC inspectors identified that planned maintenance was not accomplished to verify tightness of positioner feedback arms and that many of the positioner arms on similar valves were not double nutted. A condition report was written to address the incomplete corrective actions in response to this operating experience information. Because the failure of PCV-1121 would not be risk significant, this issue was determined to be of minor significance.

In the second quarter 2006, the NRC identified several procedure adequacy findings as documented in IR 50-247/2006-003. In partial response to these findings, Entergy issued CR IP2-2006-3930 to evaluate the concern and determine appropriate corrective actions. Entergy efforts to fully scope, prioritize, establish a timeline and take actions to improve the bulk of operations procedures do not appear to be timely. At the time of this inspection, Entergy's plan for standardizing operations procedures between the two units had not been finalized. The completion dates for standardization of the plant operating procedures appeared to be reasonable; however, the scope and timeline for the bulk of operations procedures (system operating procedures (SOPs), alarm response procedures (ARPs), etc.) were not yet defined and preliminarily were targeted to be completed in calendar year 2010 if the upgrades to SOPs, abnormal operating procedures, and ARPs were performed in series.

The NRC continued to identify procedure adequacy issues such as incorrect acceptance criteria in the surveillance procedure for the 22 auxiliary boiler feed pump overspeed trip test. Specifically, following failure of the auxiliary boiler feed pump during an overspeed trip test conducted in 2002, the licensee conducted an operability evaluation and determined that the pump was operable and the surveillance acceptance criteria should be changed. The pump failed the test again in 2006, because timely corrective action was not taken to revise the surveillance criteria. This issue was of minor significance because it did not impact the ability of the auxiliary boiler feed pump to perform its safety function.

The team also identified instances of improper closure of corrective actions to other processes. Condition reports IP2-2006-0876, IP2-2006-1134, and IP2-2006-2120 were written to document radiation monitor calibration procedure deficiencies. Corrective actions for these deficiencies were improperly characterized as enhancements and were closed out to the procedure feedback process. Closure to this process is not allowed per EN-LI-102, "Corrective Action Program," unless the change is an enhancement.

Some of the changes to the procedures were required to complete the calibrations; therefore, the team did not consider the changes to be enhancements. When the procedures were performed, the procedure deficiencies were handled in accordance with site administrative procedures; therefore, this issue constitutes a violation of minor significance that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. Entergy initiated a condition report to address this issue.

(3) Findings

(a) Failure to Identify a Degraded Condition of an Auxiliary Feed Water Check Valve in the Corrective Action Program

Introduction:

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to quality associated with improper internal clearances on BFD-68, an auxiliary feedwater check valve, in the corrective action program.

Description:

On September 1, 2006, Entergy conducted a quarterly test of the 22 auxiliary boiler feed pump (ABFP). During the test, operators observed that an ultrasonic flow meter indicated no cooling water flow to the pump bearing. The cooling water flow indication ramped to a normal reading of approximately 32 gallons per minute after lift check valve BFD-68 was mechanically agitated by the operator. Following the test, Entergy conducted an internal inspection of the check valve and found that the spiral-wound gasket for the valves body to bonnet seal had been over-crushed. This resulted in the gasket becoming partially unwound with a portion of the gasket material inside the valve cage area, potentially impacting valve operation. It was identified that the clearance between the valve body and bonnet sealing area was not sufficient to allow proper gasket crush. Entergy corrected the condition by machining the valve bonnet to ensure the clearance was appropriate to allow proper gasket crush.

The inspectors observed the field work and reviewed the apparent cause analysis conducted by Entergy. This evaluation determined the gasket material which intruded into the valve cage would not likely have prevented valve operation based on the internal clearances between the valve piston and cage. In addition, no marks were identified on the internal components which would be indicative of valve misalignment. Based on this, Entergy concluded that the most likely cause of the no flow indication was an intermittent failure of the ultrasonic flow meter.

The inspectors reviewed the work history associated with BFD-68 and noted that gasket damage was identified during internal valve inspections performed in 1997 and 2002. In addition, measurements were taken on the valve body and bonnet during the work in 1997 which indicated the internal clearances were not acceptable. Notes were placed in the work order packages identifying the gasket damage; however, these deficiencies were not entered into the licensees corrective action program. Consequently, the condition was not evaluated and corrected prior to September 2006.

Analysis:

The inspectors determined that Entergys failure to identify this degraded condition and place it in their corrective action program was a performance deficiency, in that, the improper clearance was a condition adverse to quality that had the potential to impact operation of a safety-related component. It was reasonable that Entergy should have identified this deficiency in the corrective action program since the degraded condition was found during work on the valve and noted in the associated work packages. Traditional enforcement did not apply since there were no actual safety consequences or potential for impacting the NRCs regulatory function, and the finding was not the result of any willful violation of NRC requirements or Entergys procedures.

The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 22 ABFP required approximately three hours of unplanned unavailability time to conduct repairs to ensure the correct gasket crush when the valve was reassembled. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.

Enforcement:

10 CFR 50 Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment and nonconformances are promptly identified and corrected. Contrary to this, in 1997 and 2002, Entergy failed to identify the improper internal clearances on valve BFD-68 in their corrective action program. Consequently, the condition was not evaluated and corrected prior to reassembly of the valve following maintenance in 1997 and 2002. Entergy subsequently entered this issue into the CAP (CR-IP2-2006-05241), evaluated the condition, and conducted repairs to the valve to ensure the proper gasket crush was obtained. Because this issue was of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000247/2006006-02, Failure to Identify a Degraded Condition of an Auxiliary Feed Water Check Valve in the Corrective Action Program.

(b) Inadequate Evaluation of Leaking 22 Steam Generator Low Flow Bypass Valve FCV-427L

Introduction:

A Green, self-revealing, non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified when Entergy failed to adequately evaluate leakage into the 22 steam generator. The potential that main feedwater low flow bypass valve FCV-427L was leaking was identified on April 4, 2006, prior to the Indian Point Unit 2 refueling outage, but was not fully evaluated or corrected prior to a reactor trip on August 23, 2006. During the reactor trip on August 23, 2006, FCV-427L leaked excessively and resulted in actuation of the feedwater isolation system on high water level in the 22 steam generator.

Description:

On April 4, 2006, Entergy identified that steam generator level traces during several reactor trips dating back to November 2004 showed level increasing at a rate much faster in 22 steam generator than the other steam generators. Because the auxiliary feedwater flow and generator steaming rates were identical, it was concluded the main feedwater regulating valve (FCV-427) and/or low flow bypass valve (FCV-427L)were leaking at greater than their design leakage rates. Entergy decided to address potential leakage across FCV-427 during the refueling outage in April and May 2006, since the valve was already planned for overhaul at that time, and to further evaluate FCV-427L with diagnostic testing in the normal work schedule with the plant on-line following the refueling outage.

During shutdown on April 19, 2006, for the refueling outage, a large level perturbation was again noted on the 22 steam generator. Inspection of valve FCV-427 during the overhaul identified galling, and a blue check revealed that the valve was not seating properly. Leakage past FCV-427L was not evaluated further at this time and diagnostic testing of the valve was planned for November 2006, in conjunction with planned maintenance.

On August 23, 2006, the Indian Point Unit 2 reactor was manually tripped during a plant transient. Approximately ten minutes after the reactor trip, 22 steam generator level was noted to be rising rapidly. Operators evaluated the condition, isolated the main and bypass feedwater valves due to apparent leak-by, and entered Technical Specification 3.7.3 Conditions A.1 and B.1, for FCV-427 and FCV-427L being inoperable. Due to the rapid increase in steam generator level, a high-high steam generator level signal was received for the 22 steam generator, resulting in an automatic main feedwater isolation.

An 8-Hour non-emergency event notification report was made per 10 CFR 50.72(b)(3)(iv)(A).

FCV-427 and FCV-427L were inspected during the forced outage and it was identified that the leakage was from the low flow bypass valve (FCV-427L). The valve stem stroke was adjusted for FCV-427L and the valve was checked for seat leakage satisfactorily during the subsequent startup.

Analysis:

The inspectors determined that Entergys failure to fully evaluate and correct the excessive leakage into the 22 steam generator was a performance deficiency, in that, the leakage past the main feedwater low flow bypass valve FCV-427L was a condition adverse to quality that impacted the ability of the operators to maintain steam generator water level. It was reasonable that Entergy should have fully evaluated the source of the leakage during the refueling outage in April and May 2006 prior to the plant trip in August 2006. Traditional enforcement did not apply since there were no actual safety consequences or potential for impacting the NRCs regulatory function, and the finding was not the result of any willful violation of NRC requirements or Entergys procedures.

The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The leaking low flow bypass valve forced operators to evaluate and respond to the rapidly rising 22 steam generator level by taking actions to isolate feed streams to the steam generator during a reactor trip response, and culminated in an automatic actuation of the feedwater isolation system. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.

The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolution addressed the causes and extent of condition of the problem.

Enforcement:

10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, Entergy did not fully evaluate and correct the cause of excessive leakage into the 22 steam generator in a timely manner which complicated the response to a plant transient. The valve stem stroke was adjusted for FCV-427L and the valve was checked for seat leakage satisfactorily during the subsequent startup. Because this issue was of very low safety significance (Green) and was entered into the licensees corrective action program (CR-IP2-2006-05082), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000247/2006006-03, Inadequate Evaluation of Leaking 22 Steam Generator Low Flow Bypass Valve FCV-427L.

b.

Assessment of the Use of Operating Experience

(1) Inspection Scope The team selected a sample of operating experience issues to confirm that Entergy had evaluated the operating experience information for applicability to Indian Point Unit 2 and had taken appropriate actions, when warranted. Operating experience (OE) documents were reviewed to ensure that underlying problems associated with the issues were appropriately considered for resolution via the corrective action process. A list of the specific documents reviewed is included in the Attachment to this report.

(2) Assessment No findings of significance were identified in the area of operating experience.

The inspectors found that operating experience information was appropriately considered for applicability, and corrective and preventive actions were taken as needed. Site OE coordinators screened issues from various sources for applicability to Indian Point Unit 2 and wrote CRs for additional reviews and corrective actions as necessary. Operating experience information has been integrated into routine activities, such as pre-job briefs, procedures, and training material. The inspectors noted several positive examples in which plant personnel considered operating experience information in addition to material provided by the Operating Experience Program. In a few cases, the inspectors found that OE-related CRs had been closed without a closure review by the site OE coordinator.

c.

Assessment of Self-Assessments and Audits

(1) Inspection Scope The team reviewed a sample of CA&A audits, including the most recent audit of the corrective action program, CAP trend reports, Quality Assurance (QA) audits, departmental self-assessments, and assessments conducted by independent organizations. A specific list of documents reviewed is included in the attachment to this report. These reviews were performed to determine if problems identified through these assessments were entered into the CAP, when appropriate, and whether corrective actions were initiated to address identified deficiencies. The effectiveness of the audits and assessments was evaluated by comparing audit and assessment results against self-revealing and NRC-identified findings and observations made during the inspection.

The team also reviewed the 2006 Nuclear Safety Culture Assessment, dated March 2006. This was a fleet-wide assessment, conducted by an independent contractor in early 2006. The inspectors reviewed the assessment report and discussed actions taken and planned with managers and staff. The inspectors also reviewed corporate assessments and a departmental teamwork assessment that evaluated similar areas in order to determine if appropriate action had been taken to address identified issues.

(2) Assessment One Green finding was identified for failure to enter adverse conditions, which were identified in an independent safety culture assessment, into the corrective action program for evaluation and appropriate action.

The team observed that, overall, audits and assessments were critical and, in most cases, appropriate actions were taken to address identified issues. The inspectors noted that thorough follow-up reviews were conducted by CA&A, the Self Assessment Review Board (SARB) and corporate offices. In a few cases, the inspectors found that appropriate corrective actions were not taken for issues identified during assessments.

For example, an area for improvement (AFI) identified during a self-assessment of lubrication programs was not captured in a CR for evaluation and tracking. Condition reports for two other AFIs from the lubrication program assessment were closed without taking the indicated action. In another case, a corrective action for a radiation protection QA audit finding related to survey results from personnel contamination events was closed without addressing the issue. In these cases, the inspectors determined that the failure to complete these corrective actions were of minor significance due to the nature of the issues. The inspectors also determined that the results of the 2006 Nuclear Safety Culture Assessment were not entered into the corrective action program; and as a result, timely action was not taken to evaluate the results and identify appropriate corrective actions. This finding is discussed in detail in Section 4OA2c(3).

(3) Findings

Introduction:

A Green finding was identified by the NRC inspectors for failure to initiate condition reports in accordance with EN-LI-102, Corrective Action Process, for adverse conditions identified by the 2006 Nuclear Safety Culture Assessment.

Consequently, the adverse conditions were not evaluated and appropriate corrective actions were not identified in a timely manner.

Description:

An independent contractor provided the preliminary results of the 2006 Nuclear Safety Culture Assessment to Entergy in April 2006 and presented the site-specific results to Entergy management in June. In the assessment report, the contractor made recommendations to address the negative responses and declining trends, some of which included management attention to reinforce safety conscious work environment expectations and behaviors site-wide and in a number of specific organizations at Indian Point. The contractor also recommended actions to assure the corrective action program was effective, and immediate action for some organizations based on comparison of the Indian Point results with industry-wide results.

The General Manager of Plant Operations held a meeting with site managers in July to discuss the results for Indian Point and actions needed. Managers were directed to discuss the results of the assessment with their staffs, but the results of the assessment were not entered into the CAP and no specific followup actions were assigned at that time.

At the time of the inspection, an action plan was being developed at the corporate level which would be customized by each site based on the site specific results and identified areas for improvement. The draft action plan for the site indicated that an assessment of safety culture performance indicators from across various disciplines should be done at Indian Point to understand the causes for the identified issues so that effective corrective actions could be taken. Entergy management planned to perform another independent assessment in early 2007 to complete this action. The inspectors considered this assessment to be untimely, in that, it would not provide insight into the causes of the issues identified by the 2006 assessment since the data would be collected more than a year after the original assessment was performed.

The inspectors found that specific actions to address the organizations identified by the contractor as needing management attention were not initiated until mid-September.

The draft corporate action plan also indicated that departmental action plans would be developed for these organizations based on review of the department-specific safety culture assessment results in conjunction with an assessment of the department's safety culture performance indicators. At the time of the inspection, most of these organizations had only recently reviewed the department-specific safety culture assessment results and were in the early stages of developing department action plans.

Entergy Procedure EN-LI-102, Corrective Action Program, requires the initiation of condition reports for adverse conditions, which are defined as conditions that detract from safe nuclear plant operation or that could credibly impact nuclear safety. The inspectors concluded that the negative responses and declining trends identified by the safety culture assessment could impact nuclear safety because the assessment results were precursors and indicators of a possible reluctance to raise safety issues by site employees, particularly in certain organizations. Therefore, these results should have been considered adverse conditions that warranted initiation of a condition report.

Failure to enter the assessment results into the corrective action program resulted in a delay in evaluating the results to understand the causes and identify appropriate corrective or mitigative actions.

Analysis:

The inspectors determined that Entergy's failure to enter the adverse conditions identified during the 2006 Nuclear Safety Culture Assessment into the corrective action program and evaluate the results and identify appropriate corrective actions in a timely manner was a performance deficiency that was reasonably within Entergy's ability to foresee and correct. Traditional enforcement did not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Entergy's procedures.

The inspectors determined that this finding was more than minor because if left uncorrected it would become a more significant safety concern. Without appropriate action, the weaknesses in the safety culture onsite would continue, increasing the potential that safety issues would not receive the attention warranted by their significance. The finding was not suitable for SDP evaluation, but has been reviewed by NRC management and has been determined to be a finding of very low safety significance (Green). The finding was not greater than very low safety significance because the inspectors did not identify any issues that were not raised which had an actual impact on plant safety or were of more than minor safety significance.

The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not identify adverse conditions with the potential to impact nuclear safety in the corrective action process for evaluation and resolution in a timely manner.

Enforcement:

No violation of NRC regulatory requirements was identified. Although Entergy did not initiate condition reports for the adverse conditions identified by the safety culture survey, application of EN-LI-102 for these conditions does not fall under NRC regulatory requirements. After identification by the team, Entergy entered this issue into the CAP (CR IP2-2006-06105) and initiated a Learning Organization (LO)condition report to track development and implementation of site and department action plans to address the assessment results. Because this finding does not involve a violation of regulatory requirements and has very low safety significance, it is identified as FIN 05000247/2006006-01, Failure to Enter Safety Culture Assessment Results into Corrective Action Program.

d.

Assessment of Safety Conscious Work Environment

(1) Inspection Scope During interviews and discussions with station personnel, the team assessed the safety conscious work environment (SCWE) at Indian Point. Specifically, the inspectors assessed whether workers were willing to enter issues into the corrective action program or raise safety concerns to their management and/or the NRC. The inspectors conducted individual interviews and held discussions with staff and supervisors regarding use of the corrective action program, work processes, and other problem identification and resolution activities. The team reviewed the Indian Point Employee Concerns Program (ECP) to assess whether employees were willing to use the program as an alternate path for raising concerns. The team also reviewed a sample of the ECP files to ensure that issues were appropriately addressed.

(2) Assessment No findings of significance were identified.

The team found that most workers indicated that they would raise issues that they recognized as nuclear safety issues. However, the inspectors also found that a number of workers interviewed indicated that they were aware of individuals they perceived as having been treated negatively by management for raising issues; most of these workers were in the Instrumentation and Controls (I&C) department. Some workers expressed reluctance to raise issues under certain circumstances due to a number of reasons, including fear of disciplinary action and concerns with the efficacy of the corrective action program. While most workers made a distinction between nuclear safety issues and other concerns, the inspectors noted that some of the illustrative examples provided by plant workers could have nuclear safety implications (i.e., procedure quality and staff qualification issues). In one case, a worker indicated that he/she would not raise issues under any circumstances. In another case, a worker indicated that he/she had not raised a specific nuclear safety issue. The inspectors determined that although the issue was a nuclear safety issue, it did not have an actual impact on safe plant operation in this particular instance due to the specific circumstances surrounding the issue.

The team determined that the reluctance to raise issues expressed by the I&C staff was the result of several factors, primarily the fear of disciplinary action compounded by unclear expectations and standards, and to some extent a lack of confidence in the corrective action program. The majority of the I&C staff interviewed described instances which they perceived to be either a negative reaction from management or employee discipline for raising issues. The inspectors observed that expectations for writing CRs were not clearly understood within the I&C department which may have contributed to the perception that individuals were disciplined for raising issues. The inspectors also found that expectations and standards in other areas, such as qualification and procedure requirements, were also unclear and contributed to the negative views expressed by some of the individuals. A number of interviewees also believed that issues that did not directly impact plant operations, such as personnel or industrial safety issues, would not be resolved or corrected by the corrective action program.

The team also determined that negative perceptions similar to those in the I&C department existed in other site organizations. For example, within the Operations department there was some apprehension about the perceived increase in disciplinary actions within the department. Additionally, a number of individuals did not have confidence that the corrective action program would resolve issues of lesser significance, particularly repeat issues. Based on a limited review, the team found similar issues, but to a lesser extent, in other departments. Consequently, the team was concerned that the lack of confidence in the corrective action program and the apprehension about disciplinary action could challenge the free flow of information and result in reluctance to raise issues in other departments.

Entergy has self-identified areas for improvement intended to enhance employee confidence in the corrective action program and the ECP, and has taken actions to address negative employee perceptions. However, the team determined that these efforts have not been fully effective in establishing employee confidence in these programs. For example, the Corrective Action and Assessment department has taken actions to improve the quality of feedback to employees, but the inspectors found several examples of corrective action responses that did not provide appropriate documentation of how the issue was resolved. As described above, the team found that many employees still have the perception that lower level issues will not be resolved by the corrective action program. In addition, during interviews, very few workers identified the ECP as an alternate path for raising issues, and most of those that referenced the ECP did not view the program as a viable path for raising issues primarily due to concerns about confidentiality of the program.

4OA6 Meetings, including Exit

On December 5, 2006, the team presented the inspection results to Mr. F. Dacimo and and other Entergy personnel, who acknowledged the findings. The inspection findings and observations were also discussed with Entergy management during a teleconference on December 14, 2006. The inspectors confirmed that proprietary information reviewed during the inspection would be handled in accordance with 10 CFR 2.390, "Public Inspections, Exemptions, Request for Withholding."

ATTACHMENT: Supplemental Information In addition to the documentation that the inspectors reviewed (listed in the attachment),copies of information requests given to the licensee are in ADAMS, under accession number ML063490222.

ATTACHMENT -

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

V. Andreozzi, Electrical Systems Manager

J. Balla, Employee Concerns Program Manager

R. Buckley, Corrective Actions Self Assessment Coordinator

R. Burroni, Assistant Operations Manager - Operations Support

V. Cambigianis, Mechanical Design Manager

S. Carpenter, Maintenance Department Corrective Actions Coordinator

J. Comiotes, Director of Nuclear Safety Assurance

J. Conforti, Maintenance Procedure Coordinator

F. Dacimo, Site Vice President

A. Deland, QA Self Assessment & Corrective Actions Coordinator

J. Donnelly, Director of Maintenance

R. Hansler, Reactor Engineering Manager

M. Hornyak, Project Manager, Operations Support ENN, Operating Experience Department

L. Kelly, Planning, Scheduling & Outage Corrective Actions Coordinator

D. Loope, Radiation Protection Manager

S. Meighan, Radiation Protection CA&A Supervisor

E. O'Donnell, Manager - Unit 2 Operations

D. Parker, Maintenance Superintendent

J. Perotta, Quality Assurance Manager

B. Ray, Assistant Superintendent - I&C

P. Rubin, General Manager Plant Operations

A. Small, Manager - Planning, Scheduling and Outage

B. Taggert, Employee Concerns Program Coordinator

M. Tumicki, CA Corrective Actions Coordinator

S. Verrocki, Systems Manager

T. Vitale, Operations Manager

R. Walpole, Corrective Actions Manager

Contractor Personnel

H. Levin, Synergy Consulting Services Corporation

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000247/2006006-01 FIN Failure to Enter Safety Culture Assessment Results into Corrective Action Program (Section 4OA2c(3))

05000247/2006006-02 NCV Failure to Identify a Degraded Condition of an Auxiliary Feed Water Check Valve in the Corrective Action Program (Section 4OA2.a(3)(a))

05000247/2006006-03 NCV Inadequate Evaluation of Leaking 22 Steam Generator Low Flow Bypass Valve FCV-427L (Section 4OA2.a(3)(b))