IR 05000440/2015007

IR 05000440/2015007; on 07/20/2015 - 08/07/2015; Perry Nuclear Power Plant; Biennial Problem Identification and Resolution (Pi&R) Inspection

ML15264B078
Person / Time
Site:	Perry
Issue date:	09/21/2015
From:	Michael Kunowski NRC/RGN-III/DRP/B5
To:	Harkness E FirstEnergy Nuclear Operating Co
References
IR 2015007
Download: ML15264B078 (31)

Text

UNITED STATES mber 21, 2015

SUBJECT:

PERRY NUCLEAR POWER PLANT - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000440/2015007

Dear Mr. Harkness:

On August 7, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Problem Identification and Resolution biennial inspection at your Perry Nuclear Power Plant and discussed the results of this inspection with Mr. Terry Brown and other members of your staff.

The inspection team documented the results of this inspection in the enclosed inspection report.

Based on the inspection sample, we determined that your staffs implementation of the Corrective Action Program supported nuclear safety. In reviewing your program, the team assessed how well your staff identified problems at a low threshold, your staffs implementation of the stations process for prioritizing and evaluating these problems, and the effectiveness of corrective actions taken by the station to resolve these problems. In each of these areas, the team determined that your staffs performance was adequate to support nuclear safety.

The team also evaluated other processes your staff used to identify issues for resolution.

These included your use of audits and self-assessments to identify latent problems and your incorporation of lessons learned from industry operating experience into station programs, processes, and procedures. The team determined that your stations performance in each of these areas supported nuclear safety.

Finally, the team determined that your stations management maintains a safety-conscious work environment adequate to support nuclear safety. Based on the teams observations, your employees were aware of and willing to raise concerns related to nuclear safety through at least one of the several means available.

The NRC inspectors documented one NRC-identified finding of very low safety significance (Green) in this report. The finding involved a violation of NRC requirements. The NRC is treating the violation as a Non-Cited Violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy. If you contest the violation or the significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the 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 Perry Nuclear Power Plant.

If you disagree with the 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 Perry Nuclear Power Plant.

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/

Michael A. Kunowski, Chief Branch 5 Division of Reactor Projects Docket No. 50-440 License No. NPF-58

Enclosure:

Inspection Report 05000440/2015007 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-440 License No: NPF-58 Report No: 05000440/2015007 Licensee: FirstEnergy Nuclear Operating Company (FENOC)

Facility: Perry Nuclear Power Plant Location: Perry, OH Dates: July 20 through August 07, 2015 Inspectors: J. Jandovitz, Project Engineer, Team Lead M. Marshfield, Senior Resident Inspector J. Bozga, Reactor Inspector A. Shaikh, Reactor Engineer J. Seymour, Resident Inspector (acting)

K. Grant-Leanna, Reactor Engineer (observer)

Approved by: M. Kunowski, Chief Branch 5 Division of Reactor Projects Enclosure

SUMMARY

Inspection Report (IR) 05000440/2015007; 07/20/2015 - 08/07/2015; Perry Nuclear Power

Plant; Biennial Problem Identification and Resolution (PI&R) Inspection.

This inspection was performed by three regional-based inspectors and the Perry Nuclear Power Plant resident inspectors. One (Green) finding was identified by the inspectors, with an associated Non-Cited Violation (NCV) 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, dated February 2014.

Problem Identification and Resolution Based on the samples selected for review, the team concluded that implementation of the Corrective Action Program (CAP) at the Perry Nuclear Power Plant was effective. The licensee had a low stated threshold for identifying problems and entering them in the CAP. Items entered into the CAP were generally screened and prioritized in a timely manner using established criteria. With a few exceptions documented by the team, issues in the CAP were evaluated and corrective actions were generally implemented in a timely manner, commensurate with their safety significance. The team noted that the licensee reviewed operating experience (OE) for applicability to station activities. The audits and assessments reviewed were thorough and effective in identifying site performance deficiencies, programmatic concerns, and improvement opportunities. Based on interviews conducted during the inspection, licensee staff expressed freedom to raise nuclear safety concerns and to enter nuclear safety concerns into the CAP. The 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.

The team identified one Green finding with an associated violation during the inspection.

The finding involved an incorrect evaluation conducted on a degraded condition on a Control Rod Drive Mechanism flange. The inspection team identified examples that were consistent with licensees observations that certain elements of the CAP, at times, displayed a lack of rigor in evaluation and corrective actions.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Barrier Integrity

Green.

The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to adequately evaluate a non-conforming safety-related component prior to return to service. Specifically, during the inspectors identified that the licensee had misapplied a generic vendor evaluation on June 18, 2013, to evaluate the surface damage on control rod drive (CRD) 30-15 and, therefore, failed to adequately evaluate the Use As-Is disposition on the damage to the flange surface of CRD 30-15 prior to returning it to service. The licensee entered the finding into its corrective action program as Condition Report (CR) 2015-10109. As part of the licensees immediate correction actions, the licensee performed a prompt operability determination of CRD 30-15 flange which adequately documented the basis for acceptance of Use As-Is for the flange.

The inspectors determined that the failure to perform an adequate evaluation for the return to service of the damaged CRD 30-15 was a performance deficiency (PD). The PD was determined to be more than minor, and thus a finding, because it is associated with the Barrier Integrity Cornerstone attribute of reactor coolant system equipment and barrier performance and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding could not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, nor could the finding have likely affected other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. Therefore, the finding was determined to be of very low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, for the licensees failure to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to recognize the inherent risks and latent issues associated with the application of a generic vendor evaluation to the evaluation of the damaged CRD 30-15 flange surface

[H.12]. (Section 4OA2.1b.(2).2)

REPORT DETAILS

4OA2 Problem Identification and Resolution

This inspection constituted one biennial sample of Problem Identification and Resolution as defined in Inspection Procedure (IP) 71152, Problem Identification and Resolution.

Documents reviewed are listed in the Attachment to this report.

.1 Corrective Action Program Effectiveness

a. Inspection Scope

Problem Identification - The inspectors reviewed the licensees Corrective Action Program (CAP) implementing procedures and attended meetings to assess the implementation of the CAP by licensee staff. The inspectors reviewed risk and safety significant issues in the licensees CAP since the last U.S. Nuclear Regulation Commission (NRC) Problem Identification and Resolution (PI&R) team inspection in November 2013. The selection of issues ensured an adequate review of issues across all NRC cornerstones. Included in this scope were condition reports generated by the corporate office. The inspectors also interviewed licensee staff about their use of the CAP. Specifically, the team reviewed a sample of open Condition Reports (CRs); CRs initiated for inadequate or ineffective corrective actions; and plant trips, unplanned power reductions, licensee event reports, and limiting conditions for operations (LCOs). Also, reviewed were audits and assessments of the CAP and the self-assessment performed in preparation for this inspection.

Problem Prioritization and Evaluation - The inspectors reviewed open and closed condition. The inspectors also reviewed a selection of work orders (WOs),self-assessment results, audits, and performance indicator reports. The inspectors also reviewed a selection of completed investigations from the licensees various investigation methods, including root cause evaluations (RCEs), full apparent cause evaluations (ACEs), limited apparent cause evaluations (LACEs), and common cause analyses (CCAs).

Corrective Actions - The inspectors planned and completed corrective actions.

The inspectors also reviewed a selection effectiveness reviews, and self-assessment results, audits, performance indicator reports. The team also reviewed work orders, including rework issues and repeat failures.

General - The inspectors selected the measuring and test equipment (M&TE) process to review in detail because this process has had numerous condition reports and corrective actions in recent years. The intent of the review was to determine whether the licensee staff were properly implementing, monitoring, and evaluating it through effective implementation of station monitoring programs, such as the CAP, audits, and assessments.

During the reviews, the inspectors determined whether the licensees actions were in compliance with the licensees CAP and Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, requirements. Specifically, the inspectors determined whether licensee personnel were identifying plant issues at the proper threshold, entering the plant issues into the stations CAP in a timely manner, and assigning the appropriate prioritization for resolution of the issues. The inspectors also determined whether the licensee staff assigned the appropriate investigational method to ensure the correct determination of root, apparent, and contributing causes. The inspectors also evaluated the timeliness and effectiveness of corrective actions for selected issue reports, completed investigations, and NRC previously identified findings that included principally non-cited violations. The inspectors performed walkdowns, as needed, to verify the resolution of issues.

b. Assessment

(1) Effectiveness of Problem Identification Based on the information reviewed, including initiation rates of CRs and information from interviews, the inspectors determined that the licensee has a low threshold for initiating CRs, and from the CRs reviewed, the threshold was appropriate. The inspectors did not identify any safety significant item that was not entered into the CAP. The inspectors assessed the effectiveness of problem identification as adequate to support nuclear safety.

Observations Review of the CAP performance indicators showed that the number of CRs generated has remained relatively constant over the last five years, except for 2013, ranging from 5741 to 6841. In 2013, 7477 CRs were written and the increase is explained by a refueling outage and the completion of the NRC IP 95002 inspection.

Licensee assessments of the CAP, including fleet and oversight assessments, have been self-critical and rated the CAP as effective. A number of corrective actions were developed to provide continuous improvement of CAP implementation. Interviews with personnel from across most of the site organizations supported the view that the CAP was widely used at a low threshold and was thought to have improved over the last year.

Previous NRC PI&R inspections found that issues were identified in work order Work-In-Progress (WIP) logs and not in the CAP. Improvement was noted in the 2013 PI&R and was no issues were identified in this inspection. The licensee did identify one instance where a condition adverse to quality was entered in the warehouse system and not the CAP. CR 2014-12120 was generated to address this issue and it provided the team additional confidence the licensee has instituted changes to ensure the CAP is used for all conditions adverse to quality.

Findings No findings were identified.

(2) Prioritization and Evaluation of Issues The inspectors concluded that the licensees overall performance in the prioritization and evaluation of issues was generally appropriate. In particular, the inspectors observed that while the majority of issues identified were at a low level of significance, those issues and issues of more significance were assigned a review and action level appropriate for the identified condition evaluation and in accordance with governing procedures. Issues were being appropriately screened by the originating departments, the Management Review Board, and Operations shift management for items potentially impacting equipment operability. Evaluations in apparent cause and root cause reports reviewed by the inspectors were generally adequate. The weaknesses identified and discussed below do not reflect a trend or common organizational trait compared to the number of samples completed. However, all the observations, including the finding, revealed a lack of technical rigor and review during the evaluation process.

Observations During review of the licensees implementation of the operability determination process, the inspectors reviewed prompt operability determination, CR 2015-02361, detailing a licensee-identified safety-related Emergency Service Water (ESW) pipe support 1P45-H0354, which was supported from a nonconforming embedment plate. The inspectors identified the prompt operability determination had a weakness in that the operability of the embedment studs was performed; however, the operability evaluation of the embedment plate was not performed. The licensee had restored conformance of the embedment plate. The licensee performed an immediate operability determination when identified by the inspectors that one did not exist.

Findings

.1 (Opened) Unresolved Item 05000440/2015007-01: ASME Code Pressure Tests of

Reactor Vessel Flange Seal Leak-Off Line Were Not Performed

Introduction:

The inspectors identified an unresolved item (URI) concerning the licensees failure to properly classify the reactor vessel flange seal leak-off line and perform the pressure tests in accordance with the applicable editions of the American Society of Mechanical Engineers (ASME) Code,Section XI.

Description:

While reviewing the licensees evaluation of generic communication Information Notice (IN) 2014-02, Failure To Properly Pressure Test Reactor Vessel Flange Leak-Off Lines, the inspectors identified that the licensee had determined that the pressure testing requirements described in IN 2014-02, did not apply to the reactor vessel seal leak-off line because this line was not part of the reactor coolant pressure boundary and, therefore, the licensee continued to exclude this line from ASME pressure testing requirements for Code Class 2 components.

IN 2014-02 states, in part, that 10 CFR 50.55a(g)(4) requires that ASME Code Class 1, 2, and 3 components (including supports) meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in Section XI of the ASME Code, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The ASME Code,Section XI, Tables IWB-2500-1, IWC-2500-1, and IWD-2500-1 require that a system leakage test be conducted in accordance with IWB-5220, IWC-5220, and IWD-5220 for Class 1, 2, and 3 pressure retaining components, respectively. Subarticle IWA-5200 of Section XI of the ASME Code specifies the system test requirement for pressure-retaining components and states that the system leakage tests shall be conducted at the pressure and temperature for Class 1, Class 2, and Class 3 components specified in IWB-5000, IWC-5000, and IWD-5000, respectively.

The inspectors determined that the licensee had never performed the ASME Code required system leakage test on the reactor vessel flange seal leak-off line. The licensee documented the inspectors concern in CR 2015-10559. The licensee is developing additional information to support its current position. This issue is a URI pending inspector review of the licensees additional information (URI 05000440/2015007-01, ASME Code Pressure Tests of Reactor Vessel Flange Seal Leak-Off Line Were Not Performed).

.2 Failure to Adequately Evaluate Damaged CRD Flange

Introduction:

A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to adequately evaluate a non-conforming safety-related component prior to return to service. Specifically, during the inspectors identified that the licensee had misapplied a generic vendor evaluation on June 18, 2013, to evaluate the surface damage on CRD 30-15 and therefore, failed to adequately evaluate the Use As-Is disposition on the damage to the flange surface of control rod drive (CRD) 30-15 prior to returning the CRD to service.

Description:

During review of CR 2013-09353, Damage Found on Flange of CRD 30-15, the inspectors identified that the licensee had dispositioned the damage on the flange surface for Use As-Is. The damage on CRD 30-15 flange surface was identified in CR 2013-09353 as surface marks due to a metal wire instrument tag that had been lodged in between the upper flange and lower flange surfaces. The CRD 30-15 flange was a reactor coolant pressure boundary component and classified as an ASME Code Class 1 component. ASME Code Section XI, IWB 3142.4 allowed the licensee to accept the damaged/flawed Code Class 1 component for continued service provided the licensee performed an analytical evaluation of the damage/flaws wherein the evaluation method and acceptance criteria shall be specified by the licensee. In CR 2013-09353, the licensee performed an analytical evaluation for acceptance (Use As-Is) of the damaged CRD for continued service. The basis for the Use As-Is disposition by the licensee in CR 2013-09353 was a reference to a document by Reedy Associates titled, Specification for Evaluation and Acceptance of Local Area of Materials, Parts and Components that Are Less Than the Specified Thickness. This document provided a basis for acceptance of ASME Code Section III designed and fabricated components that exhibit surface thinning such that the localized stress at areas of material thinning do not exceed 110 percent of the design specified stress. The document further stated that the material areas exhibiting localized thinning such as those identified on the CRD flange surface could accept a reduction in thickness of up to 10 percent the design specified thickness. However, this provision of an acceptable reduction in thickness of 10 percent was only to be applied in conditions where there was a direct linear inverse relationship between the membrane stress on the component and thickness of the component. Specifically, the basis of the Reedy Associates document assumed that a 10 percent decrease in thickness should correspond to a 10 percent increase in membrane stress such that the total stress at the area of thinning would not exceed 110 percent of the specified design stress.

The inspectors identified that the licensee had failed to establish the stress profile of the CRD flange surface vs. flange thickness in its analytical evaluation in CR 2013-09353.

Therefore, the inspectors determined that the licensee failed to establish the prerequisite inverse linear relationship between the flange surface membrane stress and thickness required for use of the Reedy Associates document provision of a 10 percent allowable reduction in material thickness.

The licensee documented the inspectors concern in CR 2015-10109. As part of the licensees immediate correction actions, the licensee performed a prompt operability determination of CRD 30-15 flange which adequately documented the basis for acceptance of Use As-Is for the flange. The inspectors reviewed the prompt operability determination and did not identify additional concerns. The licensee also intended to perform an extent of condition review to identify other examples of misapplication of the Reedy Associates evaluation for acceptability of components for continued service and provide training to its engineering and operations staff on the use of the Reedy Associates evaluation and understanding the evaluations basis.

Analysis:

The inspectors determined that the failure to perform an adequate evaluation for the return to service of the damaged CRD 30-15 was a performance deficiency (PD).

The PD was determined to be more than minor, and thus a finding, because it is associated with the Barrier Integrity Cornerstone attribute of reactor coolant system equipment and barrier performance and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of Findings, Table 3 - SDP Appendix Router, dated June 19, 2012. The inspectors answered No to the question in Section A of Table 3 and therefore the finding was evaluated using the SDP in accordance with IMC 0609, The Significance Determination Process (SDP) for At-Power Operations, Appendix A, dated June 19, 2012. The inspectors determined that this finding could not result in exceeding the reactor coolant system leak rate for a small loss-of coolant accident, nor could the finding have likely affected other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. Therefore, the finding was determined to be of very low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, for the licensees failure to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee staff failed to recognize the inherent risks and latent issues associated with the use of a generic vendor evaluation for the evaluation of the damaged CRD 30-15 flange surface and therefore, failed to recognize potential misapplications of the vendor evaluation

[H.12].

Enforcement:

Title 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, on June 18, 2013, the licensees design control measures failed to verify the adequacy of the design of the damaged CRD 30-15 flange before returning the CRD for continued service. Specifically, the licensee failed to perform stress calculations vs. material thickness to determine suitability of its evaluation methodology and subsequent basis for the acceptance of the damaged CRD 30-15 flange.

The licensee subsequently took immediate corrective actions which included a prompt operability determination of the damaged CRD flange surface. Because this violation was of very low safety significance (Green) and it was entered into the licensees CAP (as CR 2015-10109), it is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000440/2015007-02, Failure to Adequately Evaluate Damaged CRD Flange).

(3) Corrective Actions In general, the inspectors concluded that the corrective actions were generally appropriate for the identified issues. For selected NRC documented violations, corrective actions were determined to be effective and timely.

During the review of the previous five years of the licensees efforts to address issues with the M&TE program implementation, the inspectors found that a 2013 Fleet Oversight Audit identified that issues persist in the areas of usage tracking, issue/calibration delinquencies, and lost M&TE. The licensee completed an effectiveness review in 2014 and determined that the M&TE program had progressed from a marginally effective rating to effective rating, the gaps identified by prior assessments and internal auditing were closed, and the program strength had returned and exceeded prior performance. The basis for this conclusion was tied to the addition of a sign-off step to all work orders that required the worker to document that the usage of the M&TE be entered into the database. However, the inspectors found that the M&TE coordinator continued to identify delinquent users, overdue M&TE, and overdue out of calibration reports. These data were captured in the Perry site performance indicators (PIs). In 2015, there were approximately ten condition reports that resulted the site identifying the potential trend in the misuse of the M&TE. While no regulatory concerns were found, the inspectors concluded previous corrective actions and effectiveness reviews were weak. The licensee continues to monitor this process in the CAP and through internal PIs.

Observations The inspectors identified the following weaknesses in corrective actions reviewed; however, no common aspects or organization trends were noted.

CR 2013-07797, documented a degraded Reactor Pressure Vessel level indication.

Action was initiated to rework the system to its as-designed condition and the CR was closed to Notification 600835747 for troubleshooting and repair. When troubleshooting could not identify a failed component, the spiking originally observed was attributed to process noise. The post-maintenance test was satisfactory; however, the vessel level indication continued to spike. No additional action were performed and the action was closed without further addressing the issue.

CR 2014-18664, was initiated for a change to an Emergency Action Level based on an NRC Finding. It was discovered during management review that corrective action 001 was not performed. Originally, it was scheduled to be completed by the fourth quarter of 2014. A new corrective action was assigned with a new due date of June 29, 2017.

The inspectors questioned this new due date based on the basis for the NRC Finding.

CA 2014-16331-001 was initiated to evaluate historical surveillance data and develop a site position on preconditioning. However, it does not address the agreement between the NRC and the Nuclear Energy Institute on preconditioning.

Findings No findings were identified.

.2 Use of Operating Experience

a. Inspection Scope

The inspectors reviewed the licensees implementation of the facilitys Operating Experience (OE) program. Specifically, the inspectors reviewed implementing OE program procedures, attended CAP meetings to observe the use of OE information, reviewed completed evaluations of OE issues and events. The intent of the review was to:

(1) determine whether the licensee was effectively integrating OE experience into the performance of daily activities,

(2) determine whether evaluations of issues were appropriate and conducted by qualified individuals,

(3) determine whether the licensees program was sufficient to prevent future occurrences of previous industry events, and

(4) determine 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 OE experience, were identified and implemented in an effective and timely manner.

b. Assessment Overall, the inspectors determined that the licensee was adequately evaluating industry OE for relevance to the facility. The licensee had entered all applicable items in the CAP in accordance with the licensees procedures. Both internal and external OE were being incorporated into lessons-learned for training and pre-job briefs. The inspectors concluded that the licensee was evaluating industry OE when performing root cause and apparent cause evaluations.

Observations The inspectors reviewed the licensees OE response to a 10 CFR Part 21 Staad.Pro Error Notice issued in May 2015. The licensee had performed an operating experience review; however, the licensee had not documented it in a condition report until questioned by the inspectors. The condition report documented that the computer software error existed in a calculation of record for several Emergency Closed Cooling System pipe supports. The licensees Procedure No. NOP-OP-1009, Operability Determinations and Functionality Assessments, Revision 5, identified a nonconforming condition as operating experience review which identifies a design inadequacy. The inspectors identified a weakness in the licensees operating experience review in that an operability determination should have been performed once the nonconforming condition (computer software error) was identified. The licensee performed an immediate operability determination.

The URI discussed in section 4OA2.1b.(2), Failure to Perform Required Examinations of Reactor Vessel Flange Seal Leak-Off Line, was identified during review of operating experience, specifically an NRC-issued IN. This was not considered a weakness in the licensees OE program since the licensee did follow its procedures for review and evaluation of the IN using the CAP. Rather, the evaluation of the issue was questioned and is to be resolved.

Findings No findings were identified.

.3 Self-Assessments and Audits

a. Inspection Scope

The inspectors assessed the licensee staffs ability to identify and enter issues into the CAP, prioritize and evaluate issues, and implement effective corrective actions. As part of this inspection, the inspectors reviewed audit reports, fleet assessments, site Quality Assurance audits, and departmental self-assessments.

b. Assessment Based on the self-assessments and audits reviewed, 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 audits and self-assessments were completed by personnel knowledgeable in the subject area, and the audits were thorough and critical. In many cases, self-assessments and audits identified issues that were not previously recognized by the licensee. The inspectors observed that CAP items had been initiated for issues identified through audits and self-assessments.

Observations In some of the self-assessments reviewed, the inspectors noted that the assessment was considered complete and thorough with several appropriate corrective actions.

However, the inspectors found a number of corrective actions that were not done, and had minimal justification for not being done. As an example, focused self-assessment FO-SA-2014-0053, Emergent Locked High Radiation Areas, had four corrective actions.

The first corrective action was closed as not needed, the second was closed as no action required and the third closed with no further action required. The fourth action for a site-wide communication was completed. It appeared the value of the self-assessment was minimized by the lack of completed corrective actions.

CR 2015-00906 noted that welding QA records associated with spend fuel storage casks could not be found. No further action was required by the assessment. When questioned by the inspectors where the records were, the licensee determined that the records were with the vendor who performed the work, as specified in the contract.

Several assessments included aspects that found actions not in compliance with procedures. The assessments did not discuss why the procedures were not followed and, therefore, corrective actions to correct the behavior may not be effective.

Findings No findings were identified.

.4 Safety Conscious Work Environment (SCWE)

a. Inspection Scope

The inspectors assessed the licensees SCWE through the review of the licensees employee concerns program (ECP), implementing procedures, discussions with the coordinator of the ECP, interviews with personnel from various departments, reviews of issue reports, and review of safety culture assessments.

The inspectors interviewed approximately 30 people from a cross section of the site departments. The interviews were conducted in the field on an impromptu basis. The inspectors found this method provided answers and feedback that were not coached or prepared. Questions and discussions centered around the Perry SCWE. In addition to assessing individuals willingness to raise nuclear safety issues, the interviews also addressed changes in the CAP and plant environment and management over the past 2 years. Other items discussed included:

• knowledge and understanding of the CAP;

• effectiveness and efficiency of the CAP;

• willingness to use the CAP; and

• knowledge and understanding of ECP.

b. Assessment The inspectors did not identify any issues of concern regarding the licensees SCWE.

Information obtained during the interviews indicated that an environment was established where the licensee personnel felt free to raise nuclear safety issues without fear of retaliation. In fact, personnel interviewed indicated they were encouraged to enter issues of low significance into the CAP and could freely communicate these issues to their supervisors. Licensee personnel were aware of and familiar with the CAP and other processes, including the ECP and the NRCs process, through which concerns could be raised. The inspectors did not observe and were not provided any examples where there was retaliation for the raising of nuclear safety issues. Documents provided to the inspectors regarding surveys and monitoring of the safety culture and SCWE generally supported the conclusions from the interviews.

Observations The inspectors noted that all the interviews provided very consistent answers and opinions, even with the interviews being randomly conducted in the plant on an impromptu basis, as opposed to interviews being arranged beforehand and held in a meeting room in the administration building. Several comments were received that discussed face-to-face discussions with the CAP originator and their CAP coordinators have decreased, not all actions recommended by the initiator were addressed, and frustration that some issues in CRs were trended instead of getting fixed.

Similar to comments received in the 2013 P&IR inspection, individuals again expressed positive comments concerning the visibility and ease of communications with the senior level leadership in the field.

Findings No findings were identified.

4OA6 Management Meeting

.1 Exit Meeting Summary

On August 7, 2015, the inspectors presented the inspection results to Mr. Terry Brown, Performance Improvement Director, and members of the staff.

The licensee acknowledged the issues presented. The inspectors confirmed that proprietary documents were appropriately returned or would be destroyed.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

B. Blair, Manager - Maintenance

K. Brandt, Supervisor, Nuclear Performance Improvement

N. Conicella, Manager - Regulatory Compliance

C. Elliot, Manager, Radiation Protection

J. Ellis, Director - Recovery

D. Hamilton, Director - Site Operations

E. Harkness, Site Vice-President

J. Severino, Sr. Engineering Specialist, Regulatory Compliance

T. Veitch, Manager - Chemistry

L. Zerr, Supervisor - Regulatory Compliance

NRC Personnel

M. Kunowski, Chief, Branch 5, Division of Reactor Projects

LIST OF ITEMS

Opened

05000440/2015007-01 URI ASME Code Pressure Tests of Reactor Vessel Flange Seal Leak-Off Line Were Not Performed (Section 4OA2.1b.(2).1)

05000440/2015007-02 NCV Failure to Adequately Evaluate Damaged CRD Flange (Section 4OA2.1b.(2).2)

Closed

05000440/2015007-02 NCV Failure to Adequately Evaluate Damaged CRD Flange (Section 4OA2.1b.(2).2)

Discussed

None