IR 05000331/2015007

IR 05000331/2015007; on 04/13/2015 - 05/01/2015; Duane Arnold Energy Center (Daec); Identification and Resolution of Problems

ML15156A755
Person / Time
Site:	Duane Arnold
Issue date:	06/05/2015
From:	Mcghee J NRC/RGN-III/DRP/B1
To:	Vehec T NextEra Energy Duane Arnold
References
IR 2015007
Download: ML15156A755 (38)

Text

UNITED STATES une 5, 2015

SUBJECT:

DUANE ARNOLD ENERGY CENTER-NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000331/2015007

Dear Mr. Vehec:

On May 1, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Problem Identification and Resolution (PI&R) inspection at your Duane Arnold Energy Center (DAEC).

The enclosed inspection report documents the inspection results, which were discussed at an exit meeting on May 1, 2015, with Mr. G. Pry and other members of your staff.

The inspectors 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 samples selected for review, the team concluded that the corrective action program (CAP) at DAEC was generally effective in identifying, evaluating and correcting issues, with areas for improvement. The licensee had a low threshold for identifying issues and entering them into the CAP. A risk based approach was used to determine the significance of the issues and priority for issue evaluation and resolution. Corrective actions were generally implemented in a timely manner, commensurate with their safety significance. Operating experience was entered into the CAP when appropriate and evaluated according to procedure.

The use of operating experience was integrated into daily activities and found to be effective in preventing similar issues at the plant. In addition, self-assessments and audits were found to be conducted at appropriate frequencies with sufficient depth for all departments. The assessments reviewed were thorough and effective in identifying site performance deficiencies, programmatic concerns, and improvement opportunities. On the basis of the interviews conducted, the inspectors did not identify any impediment to the establishment of a safety conscious work environment at DAEC. Licensee staff was aware of and generally familiar with the CAP and other station processes, including the employee concerns program, through which concerns could be raised. The team determined that your stations performance in each of these areas supported nuclear safety. Based on the results of this inspection, one NRC identified finding of very low safety significance (Green) was documented in this report. This finding involved a violation of NRC requirements.

The inspectors also documented one Severity Level IV violation under the traditional enforcement process with no associated finding. Additionally, two licensee-identified violations were documented in this report. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or significance of any NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the DAEC. If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III and the NRC resident inspector at the DAEC.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management 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/

James McGhee, Acting Chief Branch 1 Division of Reactor Projects Docket No. 50-331 License No. DPR-49

Enclosure:

Inspection Report No. 05000331/2015007 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 50-331 License Nos: DPR-49 Report No: 05000331/2015007 Licensee: NextEra Energy Duane Arnold, LLC Facility: Duane Arnold Energy Center Location: Palo, IA Dates: April 13, 2015, through May 1, 2015 Team Leader: R. Ng, Project Engineer Inspectors: J. Steffes, Resident Inspector C. Phillips, Project Engineer I. Hafeez, Reactor Inspector A. Schwab, Reactor Engineer Approved by: J. McGhee, Acting Chief Branch 1 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000331/2015007; 04/13/2015-05/01/2015; Duane Arnold Energy Center (DAEC); Identification and Resolution of Problems.

This inspection was performed by four region-based inspectors and the Duane Arnold Resident Inspector. One Green finding and one Severity Level IV violation were identified by the inspectors. These finding and violation were considered non-cited violations (NCVs) of NRC regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310; Aspects Within Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated February 4, 2015. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,

Reactor Oversight Process Revision 5, dated February 2014.

Identification and Resolution of Problems On the basis of the samples selected for review, the team concluded that the CAP at DAEC was generally effective in identifying, evaluating and correcting issues, with areas for improvement.

The licensee had a low threshold for identifying issues and entering them into the CAP. A risk based approach was used to determine the significance of the issues and that informed the determination of priority for issue evaluation and resolution. Corrective actions were generally implemented in a timely manner, commensurate with their safety significance. Operating experience was entered into the CAP when appropriate and evaluated according to procedure.

The use of operating experience was integrated into daily activities and found to be effective in preventing similar issues at the plant. In addition, self-assessments and audits were found to be conducted at appropriate frequencies with sufficient depth for all departments. The assessments reviewed were thorough and effective in identifying site performance deficiencies, programmatic concerns, and improvement opportunities. On the basis of the interviews conducted, the inspectors did not identify any impediment to the establishment of a safety conscious work environment at DAEC. Licensee staff was aware of and generally familiar with the CAP and other station processes, including the employee concerns program, through which concerns could be raised. The team determined that the stations performance in each of these areas supported nuclear safety.

Although implementation of the CAP was determined to be effective overall, the inspectors identified several issues that were of very low safety significance and/or represented potential weakness of the program.

Cornerstones: Mitigating Systems

Green.

The inspectors identified a finding of very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the licensee implemented GENERA-F010-01, 1E053A2 (B2) Flange Inspection, Section W,

Revision 5, Step 5.1.3.3.b as a corrective action to NCV 05000331/2014009-02, in order to ensure proper alignment of the 1E053A2 (B2) flange. The procedure was inappropriate for the circumstances because the instructions, as written, in Step 5.1.3.3.b would not result in meeting the acceptance criteria for flange alignment listed in GENERA-F010-01, 1E053A2 (B2) Flange Inspection, Section W, Revision 5, Attachment 8. The licensee entered this issue into the CAP as condition report (CR) 02041369.

The inspectors determined the licensees failure to provide procedures of a type appropriate to the circumstances to assure that for a significant condition adverse to quality, the cause of the condition was determined and corrective actions were taken to preclude repetition was a performance deficiency warranting further review. The inspectors determined that this finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Misalignment of the flanges could lead to excessive oil leak that rendered the diesel generator inoperable. The inspectors determined the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating system, structure or component and did not result in a loss of operability or functionality. In addition, the finding did not represent a loss of system or function, did not represent an actual loss of function of a least a single train for longer than its technical specification allowed outage time, and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. The inspectors determined this finding had a cross cutting aspect in the area of PI&R, specifically resolution, because licensee personnel failed to take effective corrective actions to ensure that the resolutions address causes and extent of conditions commensurate with their safety significance [P.3].

(Section 4OA2.1.b.3.ii)

• Severity Level IV. The inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e) for failure to assure that the information included in the last update of the updated final safety analysis (UFSAR) report contained the latest information developed. The licensee implemented a change to the UFSAR, in preparation for License Amendment 243 that did not contain the latest information developed. Specifically, Section 5.4.6.1 (page 5.4-30 of Revision 17) was updated with a note that stated the reactor core isolation cooling system was not safety-related. In fact, the reactor core isolation cooling system had always been designated as safety-related. The licensee entered this issue into the CAP as CR 01974995 and prepared an updated final safety analysis report (UFSAR) change that removed the statement that the reactor core isolation cooling system was not safety-related.

The inspectors determined that the update to the UFSAR with incorrect information was a performance deficiency in accordance with IMC 0612, Power Reactor Inspection Reports,

Appendix B, Issue Screening, issued on September 7, 2012. The inspectors concluded that traditional enforcement applied because the failure to correctly update the UFSAR impacted the regulatory process. The Enforcement Policy, dated February 4, 2015, Section 6.1.d.3, gave the example that if, a licensee fails to UFSAR as required by 10 CFR 50.71(e)but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures; then this was a Severity Level IV violation. In this case, the UFSAR was updated incorrectly and did not, result in any unacceptable change to the facility or procedures. The inspectors determined this to be a similar example and therefore was more than minor and a Severity Level IV violation. This violation was not associated with a finding that was evaluated by the significance determination process. Therefore, a cross-cutting aspect was not assigned to this traditional enforcement violation.

(Section 4OA5.1.b)

=

Licensee-Identified Violations===

Violations of very low safety or security significance or Severity Level IV that were identified by the licensee have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees CAP. These violations and CAP tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

4OA2 Problem Identification and Resolution

This inspection constituted one biennial sample of PI&R as defined by Inspection Procedure 71152, Problem Identification and Resolution. Documents reviewed were listed in the Attachment to this report.

Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The inspectors reviewed the procedures and processes that described the CAP at DAEC to ensure, in part, that the requirements of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, were met. The inspectors observed and evaluated the effectiveness of meetings related to the CAP, such as the Management Review Committee meeting and the Corrective Action Review Board meeting. Selected licensee personnel were interviewed to assess their understanding of and their involvement in the CAP.

The inspectors reviewed selected condition reports (CRs) across all seven Reactor Oversight Process cornerstones to determine if problems were being properly identified and entered into the licensees CAP. The majority of the risk-informed samples of CRs reviewed were issued since the last NRC biennial PI&R inspection completed in March of 2013. The inspectors also reviewed selected issues that were more than 5 years old.

The inspectors assessed the licensees characterization and evaluation of the issues and examined the assigned corrective actions. This review encompassed the full range of safety significance and evaluation classes, including root cause evaluations (RCEs),

apparent cause evaluations, common cause evaluations and condition evaluations. The inspectors assessed the scope and depth of the licensees evaluations. For significant conditions adverse to quality, the inspectors evaluated the licensees corrective actions to prevent recurrence and for less significant issues, the inspectors reviewed the corrective actions to determine if they were implemented in a timely manner commensurate with their safety significance.

The inspectors selected the standby gas treatment system/secondary containment system to review in detail based on input from the resident staff. The standby gas treatment system/secondary containment system was a safety-related and risk significant, Maintenance Rule (a)(1) system. The primary purpose of this review was to determine whether the licensee was monitoring and addressing performance issues of this risk significant system. A 5-year review of the reactor protection system/neutron monitoring system and standby liquid control system tank boron temperature controller issues were also performed to assess the licensees efforts in monitoring and correcting system and component level performance issues The inspectors performed walkdowns, as needed, to verify the resolution of issues.

The inspectors examined the results of self-assessments of the CAP completed during the review period. The results of the self-assessments were compared to self-revealed and NRC-identified findings. The inspectors also reviewed the corrective actions associated with previously identified NCVs and findings to determine whether the station properly evaluated and resolved those issues. The inspectors performed walkdowns, as necessary, to verify the resolution of the issues. The inspectors also reviewed the open corrective action items related to the two White findings that were not completed by the end of the associated 95002 supplemental inspection (Inspection Report 05000331/2014009, ADAMS Accession Number ML14241A689).

b. Assessment

(1) Identification of Issues Based on the results of the inspection, the inspectors concluded that, in general, the station was effective in identifying issues at a low threshold and entering them into the CAP. The inspectors determined that normally problems were identified and captured in a complete and accurate manner in the CAP. The station was appropriately screening issues from both NRC and industry operating experience at an appropriate level and entering them into the CAP when applicable to the station. The inspectors also noted that deficiencies were identified by external organizations (including the NRC) that had not been previously identified by licensee personnel. These deficiencies were subsequently entered into the CAP for resolution.

The inspectors determined that the station was generally effective at trending low level issues to prevent larger issues from developing. The licensee also used the CAP to document instances where previous corrective actions were ineffective or were inappropriately closed.

The inspectors performed a 5-year review of the standby liquid control tank boron temperature controller issue. As part of this review, the inspectors interviewed the current system engineer, reviewed a sample of standby liquid control tank boron temperature controller system health reports, CRs, operating experience, test calibration data, life cycle management plan and Maintenance Rule status. The inspectors reviewed licensees CAP and work management system procedures that provided guidance for trending. In addition, the inspectors walked down the standby liquid control tank boron temperature controllers. The inspectors concluded that standby liquid control tank boron temperature controller related concerns were identified and entered into the CAP at a low threshold, and concerns were resolved in a timely manner commensurate with their safety significance.

i) Findings No findings were identified.

(2) Prioritization and Evaluation of Issues Based on the results of the inspection, the inspectors concluded that the station was effective, with areas for improvement, at prioritizing and evaluating issues commensurate with the safety significance of the identified issue, including an appropriate consideration of risk.

The inspectors determined that the Management Review Committee meetings and the Corrective Action Review Board meetings were generally thorough and maintained a high standard for evaluation quality. Members of the Corrective Action Review Board discussed selected issues in sufficient detail and challenged each other regarding their conclusions and recommendations.

The inspectors performed a detailed review of issues related to the reactor protection/neutron monitoring system covering roughly the past 5 years. The two systems entered into the Maintenance Rule (a)(1) category in April 2013 due to exceeding the criteria for scram input failures, primarily due to issues with local power range monitor circuit cards. The Inspectors reviewed action plans approved by the expert panel to replace/refurbish the circuit cards, associated cause evaluations, Maintenance Rule evaluations, and CRs. The inspectors noted that the licensee generally showed no reluctance in placing structure, system, and components into Maintenance Rule (a)(1) status. Appropriate corrective actions to address the maintenance deficiencies were prescribed and completed. A detailed review of the structure, system, and components performance generally occurred before returning such structure, system and components could be returned to (a)(2) status.

The inspectors determined that the licensee usually evaluated equipment functionality requirements adequately after a degraded or non-conforming condition was identified.

In general, appropriate actions were assigned to correct the degraded or non-conforming condition.

However, vulnerabilities and deficiencies were noted in the licensees evaluations of some conditions. These vulnerabilities and deficiencies had the potential to lead to degraded or inoperable conditions not being recognized.

i) Observations Detail in Evaluation The inspectors noted while reviewing RCE 01834595, Secondary Containment Damper Events, that although the root cause identified met all regulatory requirements, more detailed discussion could have been provided about the work planning process if the licensee had taken their root cause a level deeper. The licensee determined that the root cause was, Adjustment to stop rod was made in wrong direction, and took the corrective action to prevent recurrences to revise procedure GMP-MECH-42, HVAC Dampers, to include critical steps for rebuilding and adjusting secondary containment dampers and operators. These critical steps directed the following: Provide guidance for rotation adjustments on each individual damper within the Secondary containment applicable to the procedure. If the licensee had asked why the stop rod adjustments were made in the wrong direction, they could have found a possible programmatic weakness in work order planning and verification. The inspectors considered this a missed opportunity to achieve an overall programmatic improvement.

ii) Findings No findings were identified.

(3) Effectiveness of Corrective Action Based on the results of the inspection, the inspectors concluded that the licensee was generally effective, with areas for improvement, in addressing identified issues and the assigned corrective actions were generally appropriate. The licensee implemented corrective actions in a timely manner, commensurate with their safety significance, including an appropriate consideration of risk. Problems identified using root or apparent cause methodologies were resolved in accordance with the CAP procedural and regulatory requirements. Corrective actions designed to prevent recurrence were generally comprehensive, thorough, and timely. The inspectors sampled corrective action assignments for selected NRC documented violations and determined that actions assigned were generally effective and timely.

The inspectors also performed a 5-year extensive review of the standby gas treatment and secondary containment systems. As part of this review, the inspectors reviewed a sample of standby gas treatment and secondary containment CRs, apparent cause evaluations, and RCEs. The inspectors also walked down several secondary containment dampers, which the licensee has had historical issues with, to make sure these issues were addressed and there were no current visually identifiable problems.

The inspectors concluded that standby gas treatment and secondary containment concerns were identified and entered into the CAP at a low threshold, and concerns were resolved in a timely manner commensurate with their safety significance. An observation related to the details of a RCE was documented in Section 4OA2.1.b.2.i above.

The inspectors also noted vulnerabilities in the thoroughness of corrective actions. The inspectors identified several instances where corrective actions were closed inappropriately and that additional actions were needed to complete the closeout of the corrective actions. The inspectors determined most of these discrepancies were minor compliance issues with the licensees CAP procedure and the licensee had taken actions to address these issues. Two of these instances were determined to be licensee identified violations and were documented in Section

4OA7 of this report.

The inspectors considered these procedural compliances issues noteworthy since the licensee had been working through a human performance substantive cross-cutting issue in applying a consistent process for decision making. These compliance issues represented weaknesses that must be addressed in order to prevent more significant issues from happening. In the case of the radiation survey maps issue documented below in Section 4OA2.1.b.3.i, a more-than-minor violation would have resulted if the circumstances were slightly different.

The inspectors also identified that there were 301 open corrective action items at the time of the inspection. More than 50 of these open corrective action items were greater than 2 years old. The inspectors verified that the sampled CRs were evaluated and actions assigned appropriately. The inspectors reviewed a sample of these corrective action items and determined that most of the remaining actions were minor non-conformances or enhancements and the due dates for the actions had been extended a number of times due to resource limitations or other emergent issues. For those corrective actions that were safety significant, the inspectors verified that the due dates were reasonable and the licensee had appropriate compensatory actions in place.

The inspectors regarded this long term corrective action issue as an improvement opportunity. While the total number of outstanding actions was manageable, they could potentially affect the licensees focus on more important safety issues and complicate resource utilization.

i) Observations Drawing Errors As part of the corrective action for the reactor core isolation cooling White finding (VIO 05000331/2013004-03; RCIC Turbine Overspeed Trip), the licensee had determined that a piping and instrumentation diagram needed to be corrected. The licensee had also determined that due to the similarity of the two systems, the piping and instrumentation diagram for the high pressure coolant injection (HPCI) system also needed to be revised.

Drawing BECH-M124, P&ID Reactor Core Isolation Cooling System (Steam Side),

Revision 59, was in effect at the time of the reactor core isolation cooling immediate operability decision and the licensee concluded an inaccuracy in the drawing significantly contributed to the erroneous decision that the reactor core isolation cooling system remained operable. The senior reactor operator utilized drawing BECH-M124 and observed that sensing element SE-2457 was connected to the shaft of the reactor core isolation cooling turbine and then was shown to be connected to speed indicator SI-2457 but there was no other connections shown. This led the senior reactor operator to erroneously conclude that any component that had failed was related to indication only and was not related to speed control.

Drawing BECH-M124, Revision 60 was issued to show that there was a connection to the reactor core isolation cooling turbine governor through the use of a drawing reference arrow. During the preparation of the engineering change package to add a power indicating meter, licensee personnel determined that Revision 60 failed to show that the reactor core isolation cooling turbine governor was in between sensing element SE-2457 and speed indicator-2457 but in fact still showed that sensing element and the speed indicator were directly connected. Revision 61 of drawing BECH-M124 was issued showing that the speed indicator came off of the reactor core isolation cooling turbine governor and not directly from the sensing element. At this point, the only error remaining on drawing BECH-M124 was a minor error in the direction of the location arrows. The arrows on the drawing that indicated where one part of the drawing leaves the page and where it re-entered either on the same page or a different drawing, were pointed the wrong direction. The signal flow left the sensing element and went to the turbine logic control. The convention on the direction of the arrow should have shown that the signal left the sensing element and entered into the turbine control logic. The arrow on the drawing at the turbine logic control pointed back at the sensing element.

The inspectors had had several discussions with licensee management about this drawing error. This error was minor but still incorrect.

The licensee had included the need to revise HPCI drawing BECH-M122 as part of their corrective action to prevent recurrence (CAPR) for the reactor core isolation cooling White finding. This was due to the great similarity between the HPCI and the reactor core isolation cooling drawings. During their review of the extent of condition and corrective actions, the inspectors observed that while drawing BECH-M122 had been revised similar to Revision 60 of reactor core isolation cooling drawing BECH-M124; it had not been further corrected. The HPCI sensing element SE-2284 was shown directly connected to the speed indicator SI-2284 without the associated HPCI turbine governor in between them. The inspectors determined that the HPCI drawing was inaccurate and issued NCV 05000331/2014009-01, Inadequately Performed Drawing Revision Related to CAPR, as a result of the 95002 supplemental inspection in June 2014.

During this inspection, the inspectors reviewed the corrective actions associated with NCV 05000331/2014009-01. The inspectors identified that while drawing BECH-M122 was updated in Revision 64 to show that the sensing element was connected to the turbine control logic, that drawing change contained the same arrow configuration error that was in BECH-M124, Revision 61. The inspectors also reviewed BECH-M124, Revision 62, and the arrow convention error was still there.

These errors with the arrow conventions were of very minor safety significance and therefore not subject to enforcement action in accordance with the NRC s Enforcement Policy. However, these errors were indicative of a lack of attention to detail in the corrective action process and on behalf of management.

Effectiveness Reviews During the 95002 supplemental inspection performed in June 2014, the inspectors determined that the effectiveness reviews, selected by the licensee to determine if the CAPRs associated with the White Findings, were more focused on whether the root causes had been addressed or whether the White findings had repeated. The licensee failed to address whether the CAPRs had been effective, sustainable and institutionalized. This led to a number of the CAPRs not having a suitable effectiveness review planned. The NRC issued NCV 05000331/2014009-03; Failure to Establish Effectiveness Review Attributes to Assess the Effectiveness of Corrective Actions, for having inappropriate effectiveness reviews.

The inspectors reviewed the new effectiveness review plans associated with NCV 05000331/2014009-03. The inspectors identified that the licensee had performed effectiveness reviews for at least two issues discussed in this report (HPCI drawing changes and standby diesel generator flange alignment) and determined that the corrective actions were effective. The inspectors, however, determined that the corrective actions associated with those issues were incorrect. The HPCI drawing was changed but still had an error and the method described for the standby diesel generator flange alignment could not have met the acceptance criteria established in the procedure. These issues were discussed in details in Section 4OA2.1.b.3.i and Section 4OA2.1.b.3.ii, respectively. The licensees procedure for effectiveness reviews, PI-AA-100-1005, Root Cause Analysis, Revision 12, Attachment 16, Effectiveness Review Detail, Step 5, states that the completed effectiveness review may identify issues with individual corrective actions and still conclude the corrective actions collectively have been effective. If one or more actions were not completed appropriately, but the review determined that collectively, the actions had been effective, the individual issues would be documented on a new CR for resolution. The inspectors did not identify any CRs associated with the issues identified.

The cross-cutting issue, associated with the NCV described in Section 4OA2.1.b.3.ii of this report regarding the standby diesel generator flange alignment, was listed as Evaluation, which captured the failure to adequately perform the effectiveness review.

As described in Section 4OA2.1.b.3.i, the HPCI drawing change issue was minor. For these reasons, another NCV, for failing to perform steps in PI-AA-100-1005 was not issued. However, the inspectors determined that the failure to appropriately perform these effectiveness reviews indicated a weakness in the CAP.

Radiation Surveys Not Performed Per Procedure In 2013, the licensee identified a number of required surveys were not performed per procedure HPP 3103.01, HP Survey Performance and Frequencies. One of the corrective actions at the time was to use a computer system to track completion of these surveys. Shortly after an NRC finding was issued for an unrelated missed calibration of radiation protection instruments in 2013, the missed survey corrective action was closed to the CR for the NRC finding since the corrective action for the finding was also to use the same computer system to track the calibration. However, the corrective action to use the computer system to track survey completion was cancelled in August 2013 with the justification that the surveys were not regulatory requirements. The licensee established additional requirements in procedure HPP 3103.01 that the required surveys would be due before reaching 90 percent of the frequency period. A CR was required to be written identifying the survey that was at-risk of going overdue.

To verify the effectiveness of the licensees corrective action, the inspectors reviewed the surveys completed in 2014. The inspectors identified that the tracking sheet for the biennial surveys was signed off with expired surveys. Four of these surveys were actually performed but not logged in the tracking sheet as required. Two biennial surveys were determined to have exceeded the biennial frequency. No CRs were generated when they reached 90 percent of the due date.

The inspectors determined that this issue was a failure to follow radiation protection procedure HPP 3103.01. The issue was considered a minor procedural violation because there were no personnel entries to these areas using the expired maps.

Therefore, this minor violation was not subject to enforcement action in accordance with the NRC s Enforcement Policy. Although this issue was minor, the licensees justification to cancel the corrective action in August 2013 was weak and relied on radiation protection supervisor to keep track of these surveys. Essentially, the licensee was relying on the same manual tracking mechanism that had failed them previously.

The licensee entered this issue into the CAP as CR 02044506, completed the missed surveys and is evaluating alternative for tracking survey performance.

ii) Findings Emergency Diesel Generator Maintenance Procedure Lacks Appropriate Work Instructions and Acceptance Criteria

Introduction:

The inspectors identified a finding of very low significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances to assure that for a significant condition adverse to quality, the cause of the condition was determined and corrective actions were taken to preclude repetition.

Description:

During the 95002 supplemental inspection performed in June 2014, the inspectors reviewed licensee CRs, procedures, and work documents used following the March 8, 2013, gasket failure of a flange on the A standby diesel generator lube oil cooler that resulted in a White finding. The inspectors reviewed the CAPR recommended per Revision 5 of the RCE 01855032. The inspectors noted that Root Cause 1, CAPR 3, required the licensee to develop a maintenance procedure for the A and B standby diesel generators similar to the model work order for 1E053A2 and 1E053B2, lube oil heat exchangers, to:

• perform visual inspection of flange surfaces;

• perform visual inspection of flange channel head to heat exchanger shell sealing surfaces;

• measure clearances between the shell and tube bundle flanges without the gasket installed;

• measure flatness of shell flanges and re-machine flange surfaces if not within tolerance;

• measure and record vertical and horizontal alignment between the lube oil and jacket water channel heads;

• specify a Garlock multi-swell gasket to be used for the stationary end of the heat exchanger;

• torque to the value identified in updated evaluation, CA 01855032-12; and

• torque the floating end finger-tight plus 1-2 turns.

The inspectors reviewed licensee work packages that incorporated procedure GENERA-F010-01, 1E053A2 (B2) Flange Inspection, Revision 4 and determined the licensee failed to incorporate the CAPR item to Measure and Record vertical and horizontal alignment between the lube oil and jacket water channel heads. The inspectors determined that the Flange Inspection procedure instructed the individual performing the flange inspection to measure horizontal and vertical alignment between 1E053A2 (B2) channel head flange and 1E053A3 (B3) channel head flange visually or using a ruler.

In addition, the licensee determined that they failed to measure and record flange alignment as specified by RCE 01855032 on September 24, 2013, during standby diesel generator lube oil heat exchanger flange inspections. The licensee documented this issue in CR 01955777, dated April 10, 2014, following interviews with site maintenance and engineering staff. The licensee stated that the channel head flange alignment was visually inspected per procedure GENERA-F010-01, and that no visible misalignment issues were noted. The licensee stated the procedure did not require gap and alignment parameters to be documented as requested by RCE 01855032, CAPR-11 or CAPR-15.

The inspectors determined the licensee provided the above assessment and closed this CR and failed to revise procedures to incorporate CAPR-11 and CAPR-15 to measure and record flange alignment in accordance with RCE 01855032. The inspectors concerns were entered into the licensee CAP as CR 01974810, to ensure CAPR items are adequately incorporated into procedures in a measureable, achievable and timely manner in accordance with root cause analysis procedure PI-AA-100-1005, and condition evaluation and corrective action procedure, PI-AA-205.

In addition, the licensee failed to provide qualitative or quantitative acceptance criteria for the assessment of flange alignment issues. The inspectors reviewed the implementing procedure GENERA-F010-01, Revision 0 though Revision 4, for flange inspections and noted the procedure required the flange inspector to:

MEASURE horizontal and vertical alignment between 1E053A2 (B2)channel head flange and 1E053A3 (B3) channel head flange visually or using ruler. Contact Engineering if any excessive misalignment exists.

The inspectors determined that these work instructions failed to adequately implement CAPR-11 and CAPR-15 from RCE 01855032, which required the maintenance personnel to measure and record horizontal and vertical flange alignment.

The inspectors concluded that the procedure allowed maintenance personnel to make an in-field determination of acceptability of alignment, based on judgment, in the absence of qualitative or quantitative acceptance criteria. The inspectors concerns were entered into the Corrective Actions Program as CR 01975553, to ensure that acceptance criteria are developed and included in Flange Inspection procedure, GENERA-F010-01.

The NRC issued NCV 05000331/2014009-02, Incomplete Corrective Actions To Prevent Recurrence, in the 95002 supplemental inspection report 05000331/2014009 to address the issues described above.

During this PI&R inspection on April 15, 2015, the inspectors reviewed the licensees corrective actions to address NCV 05000331/2014009-02. The licensee changed GENERA-F010-01, Section W, Steps 5.1.3.3.b in Revision 5 to require the measurement of alignment by taking a ruler with 1/32 increments and measuring the gap between the outer edges of the two flanges in 4 locations 90-degrees apart from each other. The maximum allowable distance measured (acceptance criteria) was 1/16 at any location. There was a drawing to describe how the measurements were to be taken in GENERA-F010-01, Section W, Appendix 7 and a requirement to record the measurements taken in Attachment 8.

The inspectors also observed the 1EO53A2 flange in the A standby diesel generator room. The bottom half of the flange was not the same size and shape as the top half of the flange. The inspectors observed that the measurements called out in GENERA-F010-01, Section W, Step 5.1.3.3.b could not be performed satisfactorily because of the differences in the two halves of the flange.

Analysis:

The inspectors determined the licensees failure to provide procedures of a type appropriate to the circumstances to assure that for a significant condition adverse to quality, the cause of the condition was determined and corrective actions were taken to preclude repetition was a performance deficiency warranting further review. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Misalignment of the flanges could lead to excessive oil leak that rendered the diesel generator inoperable.

The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions.

The inspectors determined the finding was of very low safety significance (Green)because the finding was not a deficiency affecting the design or qualification of a mitigating system, structure and component and did not result in a loss of operability or functionality. In addition, the finding did not represent a loss of system or function, did not represent an actual loss of function of a least a single train for longer than its TS allowed outage time, and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance.

The inspectors determined this finding had a cross-cutting aspect in the area of PI&R, specifically resolution, because licensee personnel failed to take effective corrective actions to ensure that the resolutions address causes and extent of conditions commensurate with their safety significance [P.3].

Enforcement:

As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Contrary to the above, on July 3, 2014, the licensee implemented GENERA-F010-01, 1E053A2 (B2) Flange Inspection, Section W, Revision 5, Step 5.1.3.3.b as a corrective action to NCV 05000331/2014009-02, in order to ensure proper alignment of the 1E053A2 (B2) flange. The procedure was not appropriate for the circumstances because the instructions, as written, in Step 5.1.3.3.b would not result in meeting the acceptance criteria for flange alignment listed in GENERA-F010-01, 1E053A2 (B2)

Flange Inspection, Section W, Revision 5, Attachment 8. The licensee evaluated the gasket that was in place when the inspection was performed and determined that the installation was acceptable. Inspectors determined that the corrective actions to revise the procedure could reasonably be performed before gasket replacement was required.

This violation was being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very low safety significance (Green) and was entered into licensees CAP as CR 02041369.

(NCV 05000331/2015007-01, Inappropriate Diesel Generator Maintenance Procedure)

.2 Assessment of the Use of Operating Experience

a. Inspection Scope

The inspectors reviewed the licensees implementation of the facilitys Operating Experience program. Specifically, the inspectors reviewed the operating experience program implementing procedures, and completed evaluations of operating experience issues and events. The inspectors also observed meetings and daily activities for the use of operating experience information. The intent was to determine if the licensee was effectively integrating experience into the performance of daily activities, whether evaluations of issues were proper and conducted by qualified personnel, whether the licensees program was sufficient to prevent future occurrences of previous industry events, and whether the licensee effectively used the information in developing departmental assessments and facility audits. The inspectors also assessed if corrective actions, as a result of operating experience experience, were identified and implemented effectively and in a timely manner.

b. Assessment Based on the results of the inspection, the inspectors concluded that in general, operating experience was effectively utilized at the station. The inspectors observed that operating experience was discussed as part of the daily and pre-job briefings. Industry operating experience was effectively disseminated across plant departments and no issues were identified during the inspectors review of selected licensee operating experience evaluations.

c. Findings

No findings were identified.

.3 Assessment of Self-Assessments and Audits

a. Inspection Scope

The inspectors reviewed selected self-assessments, bench markings, and Nuclear Oversight audits, as well as the schedule of past and future assessments. The inspectors evaluated whether these audits and self-assessments were effectively managed, adequately covered the subject areas, and properly captured identified issues in the CAP. In addition, the inspectors interviewed licensee personnel regarding the implementation of the audit and self-assessment programs.

b. Assessment Based on the results of the inspection, the inspectors concluded that self-assessments and audits were typically accurate, thorough, and effective at identifying issues and enhancement opportunities at an appropriate threshold. The inspectors concluded that these audits and self-assessments were completed by personnel knowledgeable in the subject area. In many cases, these self-assessments and audits had identified numerous issues that were not previously recognized by the station. These issues were entered into CRs as required by the CAP procedures.

c. Findings

No findings were identified.

.4 Assessment of Safety Conscious Work Environment

a. Inspection Scope

The inspectors interviewed selected DAEC personnel to determine if there were any indications that licensee personnel were reluctant to raise safety concerns to either their management or the NRC due to fear of retaliation. The inspectors reviewed selected employee concern program (ECP) case files to identify any emergent issues or potential trends. The inspectors also assessed the licensees safety conscious work environment through a review of ECP implementing procedures, discussions with ECP manager, interviews with personnel from various departments, and reviews of CRs. The licensees programs to publicize the CAP and ECP were also reviewed. The inspectors reviewed licensees self-assessments and assessments by external organizations of safety culture to determine if there were any organizational issues or trends that could impact the licensees safety performance.

b. Assessment The inspectors did not identify any issues that suggested conditions were not conducive to the establishment and existence of a safety conscious work environment at DAEC.

Licensee staff members were aware of and generally familiar with the CAP and other station processes, including the ECP, through which concerns could be raised. In addition, a review of the types of issues in the ECP indicated that the licensee staff members were appropriately using the CAP and ECP to identify issues. The licensee staff also indicated that management had been supportive of the CAP by providing time and resources for employee to generate their CRs.

The staff also expressed a willingness to challenge actions or decisions that they believed were unsafe. All employees interviewed noted that any safety issue could be freely communicated to supervision and safety significant issues were being corrected.

Some employees indicated that training and retraining of the CAP process were not consistently performed and they lacked the proficiency to navigate the computerized CAP efficiently. The inspectors considered that an improvement opportunity for the CAP implementation.

Since the beginning of 2013, various safety culture assessments had been performed by contractors, the licensees staff, and a nuclear plant owner/operators organization. The results indicated that there were no impediments to the identification of nuclear safety issues. The inspectors reviewed these surveys and did not identify any adverse trend.

c. Findings

No findings were identified.

4OA5 Other Activities

.1 (Closed) Unresolved Item 05000331/2014009-04; Failure To Correctly Update the

Updated Final Safety Analysis Report

a. Inspection Scope

The issue associated with unresolved item (URI)05000331/2014009-04 was identified during the review of extent of condition as part of the Supplemental Inspection Program Inspection (95002); and was reported in NRC Integrated Inspection Report 05000331/2014-009. Subsequently, the licensee performed a review, in accordance with its CAP and documented the result in CR 01974995. The inspectors reviewed the corrective action documents and interviewed members of plant management.

b. Findings

Introduction:

The inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e) for failure to assure that the information included in the last update of the UFSAR contained the latest information developed.

Description:

During the 95002 supplemental inspection conducted in June of 2014, the inspectors identified a URI regarding the potential failure to update the UFSAR in accordance with 10 CFR 50.71(e). The licensee had provided information to the inspectors, which documented the safety classification of the reactor core isolation cooling system. However, the information provided did not match the safety classification description in the UFSAR.

The inspectors questioned the safety classification of the reactor core isolation cooling system. The licensee provided EC-01.12, Equipment Data Sheet-SQA Level B, which identified the reactor core isolation cooling system as safety-related. This document was last updated in May 2010. The UFSAR, page 5.4-30, Revision 17, had a note at the bottom that stated that reactor core isolation cooling was not safety-related.

The licensee determined that the UFSAR was changed to reflect that the reactor core isolation cooling system was non-safety related in error. The change to the UFSAR was made during an extensive license amendment in 2003 for extended power uprate. Title 10 CFR 50.71(e), states, in part, each person licensed to operate a nuclear power reactor under the provisions of § 50.21 or § 50.22, and each applicant for a combined license under Part 52 of this chapter, shall update periodically, as provided in paragraphs (e)(3) and

(4) of this section, the final safety analysis report (FSAR)originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.

Analysis:

This issue was assessed in accordance with the traditional enforcement path in IMC 0612, Appendix B, Issue Screening, issued on September 7, 2012. The inspectors determined that traditional enforcement did apply because the failure to correctly update the UFSAR impacted the regulatory process. The Enforcement Policy, dated February 4, 2015, Section 6.1.d.3, gave the example that if, a licensee fails to update the FSAR as required by 10 CFR 50.71(e) but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures; then this was a Severity Level IV violation. In this case, the UFSAR was updated incorrectly but did not, result in any unacceptable change to the facility or procedures. The inspectors determined this to be a similar example and therefore more than minor and a Severity Level IV violation. This violation was not associated with a finding that was evaluated by the Significance Determination Process. Therefore, a cross-cutting aspect was not assigned to this traditional enforcement violation.

Enforcement:

As required by 10 CFR 50.71(e) that, Each person licensed to operate a nuclear power reactor under the provisions of § 50.21 or § 50.22, and each applicant for a combined license under Part 52 of this chapter, shall update periodically, as provided in paragraphs (e)(3) and

(4) of this section, the FSAR originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. This submittal shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the applicant or licensee or prepared by the applicant or licensee pursuant to Commission requirement since the submittal of the original FSAR, or as appropriate, the last update to the FSAR under this section. The submittal shall include the effects of all changes made in the facility or procedures as described in the FSAR; all safety analyses and evaluations performed by the applicant or licensee either in support of approved license amendments or in support of conclusions that changes did not require a license amendment in accordance with § 50.59(c)(2) or, in the case of a license that references a certified design, in accordance with § 52.98(c) of this chapter; and all analyses of new safety issues performed by or on behalf of the applicant or licensee at Commission request. The updated information shall be appropriately located within the update to the FSAR.

Contrary to the above, on October 28, 2003, the facility licensee implemented a change to the UFSAR, in preparation for License Amendment 243 that did not contain the latest information developed. Specifically, Section 5.4.6.1 (page 5.4-30 of Revision 17) was updated with a note that stated the reactor core isolation cooling system was not safety-related. In fact, the reactor core isolation cooling system had always been designated as safety-related. The licensee prepared an UFSAR change that removed the statement that the reactor core isolation cooling system was not safety-related.

Because this violation was of very low safety significance, was not repetitive or willful, and was entered into the licensees CAP as CR 01974995, this violation was being treated as a Severity Level IV NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 0500331/2015007-02, Failure to Correctly Update the Updated Final Safety Analysis Report)

URI 05000331/2014009-04 was closed.

4OA6 Management Meetings

Exit Meeting On May 1, 2015, the inspectors presented the inspection result to Mr. G. Pry and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

4OA7 Licensee-Identified Violations

The following violations of very low significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as NCVs.

• As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Procedure PI-AA-205, Condition Evaluation and Corrective Action, Revision 23, Section 4.10, Step 8, requires, in part, that corrective actions shall not be closed to other existing actions unless the description, scope, condition designation, and intent of the action that will remain open is equivalent to that of the action being closed. Contrary to the above, in December 2013, while performing RCE associated with CR 01884408, the licensee inappropriately credited a long term corrective action developed in an apparent cause evaluation associated with CR 01835557 to accomplish the stated CAPR action. Further, procedure PI-AA-104-1000, Corrective Action, Revision 1, Section 4.11, Step 6, requires, in part, that requests for significance level 1 CAPR due date extensions be documented on PI-AA-104-1000-F01, Change Request Form, Revision 1, receive MRC approval for the stated changes, and be attached to the condition report. Contrary to that, on December 5, 2014, the long term corrective action due date was extended 4 months without Management Review Committee approval. The CAPR and long term corrective action were related to the failure of a safety related main steam line temperature instrument.

Because the instrument was maintained in an operable status and did not initiate a transient based upon the deficient condition, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02025444 and 02044053. Immediate corrective actions included adding an assignment to correctly restate that the plant modification to replace the temperature indicators as a CAPR in CR 01884408.

• As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Procedure PI-AA-104-1000, Corrective Action, Revision 3, Section 4.10, Step 1, requires, in part, that closure of corrective actions is not permitted until the corrective actions are completed as prescribed or appropriate justification and approval for intent change or cancellation/nonperformance of the corrective action is documented in the condition report. Contrary to the above, on November 24, 2014, a long term corrective action to replace the northwest corner room cooler coils due to copper tube degradation was closed without replacing the cooler coils or obtaining an approval for cancellation of the correction action.

Because the cooler was maintained in an operable status and did not initiate a transient based upon the deficient condition, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02016918. Immediate corrective actions included reopening the corrective action assignment and obtaining approval for extending the due date.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

G. Pry, Plant General Manager

M. Strope, Assistant Operations Manager

C. Hill, Training Manager

P. Hansen, Performance Improvement Manager

M. Davis, Licensing / Emergency Planning Manager

M. Durbin, Maintenance Director

B. Porter, Radiation Protection Manager

B. Simmons, Nuclear Oversight Manager

J. Schwertfeger, Security Manager

D. Church, Program Engineering Manager

G. Hawkins, ERRT Manager

NRC

J. McGhee, Branch Chief

L. Haeg, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000331/2015007-01 NCV Inappropriate Diesel Generator Maintenance Procedure

(Section 4OA2.1.b.3.ii)

05000331/2015007-02 NCV Failure to Correctly Update the Updated Final Safety

Analysis Report (Section 4OA5.1.b)

Closed

05000331/2015007-01 NCV Inappropriate Diesel Generator Maintenance Procedure

(Section 4OA2.1.b.3.ii)

05000331/2015007-02 NCV Failure to Correctly Update the Updated Final Safety

Analysis Report (Section 4OA5.1.b)

05000331/2014009-04 URI Failure To Correctly Update the Updated Final Safety

Analysis Report (Section 4OA5.1.b)

Discussed

None