IR 05000266/2015007

IR 05000266/2015007; 05000301/2015007; 06/22/2015 - 07/10/2015; Point Beach Nuclear Plant, Units 1 and 2; NRC Problem Identification and Resolution Inspection Report

ML15231A778
Person / Time
Site:	Point Beach
Issue date:	08/19/2015
From:	Jamnes Cameron Reactor Projects Region 3 Branch 4
To:	Mccartney E Point Beach
References
IR 2015007
Download: ML15231A778 (31)

Text

UNITED STATES ust 19, 2015

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000266/2015007; 05000301/2015007

Dear Mr. McCartney:

On July 10, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Problem Identification and Resolution (PI&R) biennial inspection at your Point Beach Nuclear Plant.

The enclosed report documents the results of this inspection, which were discussed on July 10, 2015, with Mr. DeBoer and other members of your staff.

Based on the inspection sample, the inspection team determined that your staffs implementation of the corrective action program and associated processes generally supported nuclear safety. In reviewing your programs, 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. Discussed in the enclosed report is an issue identified by plant personnel and the NRC associated with the technical rigor in some of your cause analyses and assessments.

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 are 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. This finding involved a violation of NRC requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy. If you contest the violation or 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 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 Point Beach Nuclear Plant.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 Point Beach Nuclear Plant.

In accordance with Title 10 of the Code of Federal Regulations (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 Bruce L. Bartalett for/

Jamnes L. Cameron, Chief Branch 4 Division of Reactor Projects Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27

Enclosure:

IR 05000266/2015007; 05000301/2015007 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 05000266; 05000301 License Nos: DPR-24; DPR-27 Report No: 05000266/2015007; 05000301/2015007 Licensee: NextEra Energy Point Beach, LLC Facility: Point Beach Nuclear Plant, Units 1 and 2 Location: Two Rivers, WI Dates: June 22, 2015 through July 10, 2015 Inspectors: J. Rutkowski, Project Engineer, Team Lead J. Neurauter, Senior Reactor Inspector, DRS K. Barclay, Resident Inspector Approved by: J. Cameron, Branch Chief Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report (IR) 05000266/2015007; 05000301/2015007; 06/22/2015 - 07/10/2015;

Point Beach Nuclear Plant, Units 1 and 2; Effectiveness of Prioritization and Evaluation of Issues This inspection was performed by two NRC regional inspectors and the Point Beach Nuclear Plant (PBNP) resident inspector. One Green finding was identified by the inspectors. The finding was considered a 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 the Cross-Cutting Areas effective date December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated February 4, 2015. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

Problem Identification and Resolution On the basis of the sample selected for review, the team determined that implementation of the corrective action (CA) program and associated processes at the PBNP generally support nuclear safety. The licensee demonstrated a low threshold for identifying problems and entering them in the CA program. Items entered into the CA program were screened and prioritized in a timely manner using established criteria; were generally evaluated commensurate with their safety significance; and corrective actions were generally implemented in a timely manner, commensurate with the safety significance. The team noted that the licensee reviewed operating experience for applicability to station activities. Audits and self-assessments were performed at a level sufficient to identify most deficiencies. On the basis of interviews conducted during the inspection, workers at the site expressed freedom to enter safety concerns directly into the CA program or through their supervisors. The team identified one Green finding during the inspection. The finding involved the licensees failure to take timely and appropriate corrective action to ensure that a valve could be opened as specified by procedure for mitigating the consequences of a fire. The inspection team identified examples that were consistent with licensees observations that certain elements of the CA program at times displayed a lack of rigor in evaluation and corrective actions.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green: The inspectors identified a finding of very low safety significance and an associated NCV of license condition 4.F for the licensees failure to demonstrate the capabilities of systems needed to perform a design function for Appendix R cold shutdown. Specifically, none of the licensees tests, inspections, or maintenance activities demonstrated that CC-722A, the component cooling water pump suction cross-tie valve, was capable of being opened as required in AOP-10B, Safe to Cold Shutdown in Local Control. The licensee corrective actions included entering the issue into their CA program, declaring CC-722A non-functional, and commencing four-hour fire rounds.

The inspectors determined the finding to be more than minor because the failure to demonstrate the capabilities of systems needed to perform a design function for Appendix R safe shutdown was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process,

Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding affected the ability to reach and maintain safe shutdown, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1-A, because the inspectors determined that the finding would not prevent the reactor from reaching and maintaining hot shutdown. This finding has a cross-cutting aspect of Resolution (P.3), in the area of problem identification and resolution, because the licensee did not take effective corrective actions to address the issue in a timely manner. Specifically, in 2007, the licensee identified that they had not been testing the valve as specified in their Fire Protection Evaluation Report and as of July 2015 had still not corrected it. (Section 4OA2b.(2))

Licensee-Identified Violations

None

REPORT DETAILS

4OA2 Problem Identification and Resolution

The activities documented in Sections

.1 through .4 constituted one biennial sample of

problem identification and resolution as defined in IP 71152.

.1 Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The inspectors reviewed the licensees corrective action (CA) program implementing procedures and attended CA program meetings to assess the implementation of the CA program by site personnel.

The inspectors reviewed risk and safety significant issues in the licensees CA program since the last NRC Problem Identification and Resolution (PI&R) inspection in August 2013. The selection of issues ensured an adequate review of issues across NRC cornerstones. The inspectors used issues identified through NRC generic communications, department self-assessments, licensee audits, operating experience reports, and NRC documented findings as sources to select issues. Additionally, the inspectors reviewed action requests/condition reports (CRs) generated as a result of facility personnels performance in daily plant activities. The inspectors also reviewed a selection of work orders (WOs), self-assessment results, audits, performance indicator reports, system health reports, and completed investigations from the licensees various investigation methods, which included root cause evaluation (RCE) and apparent cause evaluation (ACE).

The inspectors selected the direct current (DC) to 120 volt alternating current (AC)inverters and associated components to review in detail. The inspectors review was to determine whether the licensee staff were properly monitoring and evaluating the performance of these and associated components through effective implementation of station monitoring programs. A five year review of the inverters was also undertaken to assess the licensee staffs efforts in monitoring for system degradation due to aging aspects. The inspectors also performed a partial system walkdown of the component cooling water system to determine whether identifiable issues with the system were adequately described in the CA program and system health documents.

During the reviews, the inspectors determined whether the licensee staffs actions were in compliance with the facilitys CA program and 10 CFR Part 50, Appendix B requirements. Specifically, the inspectors determined whether licensee personnel were identifying station issues at the proper threshold, entering the station issues into the stations CA program 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 investigation method to ensure the proper 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 eight NRC previously identified findings that included principally non-cited violations.

The inspectors also reviewed corrective actions from licensees RCEs 01883633, White Finding-Flooding, and 01896156, Common Cause Degraded Cornerstone-Mitigating Systems, Two White Findings, which were not completed by the licensee as of February 2015.

Documents reviewed are listed in the Attachment to this report.

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 CA program. The inspectors assessed the effectiveness of problem identification as adequate to support nuclear safety.

Observations The inspectors found that issues were being identified and captured in the licensees CA program. The licensee documented that in calendar year 2014 approximately 12,210 CRs were initiated. During the assessment period of approximately 17 months, approximately 6,382 CRs were closed to maintenance WOs.

In the previous biennial problem identification and resolution inspection conducted in 2013, the inspectors documented that the inspectors noted instances where conditions were not captured historically in the CA program. The licensee, in response to that observation, adopted an expectation of when in doubt, fill it out. During this inspection the inspectors did not identify problems with initial documentation of identified issues.

However, the licensee in their focused self-assessment to review readiness for the NRC inspection (PBSA-PI-15-01), found instances where CRs should have been written per site expectations and were not written.

The inspectors reviewed open corrective WOs, open corrective action items, system health reports for the last five years, and walked the DC to 120 volt AC inverters with the systems engineer. The inspectors also walked portions of the component cooling water system with that systems engineer and reviewed the current status of WOs and CA program documents. The inspectors found the systems to be in overall good health with a reasonable number of open corrective WOs and corrective action items even though the inverters had several past issues that resulted in inverter failures. The inspectors did not identify any major conflicts between actual system conditions and the condition of the systems as represented in WOs, system health reports, and CA program documents. The inspectors did identify one issue with the component water cooling system which is addressed in Section 4OA2.2.b(2) below.

Findings No findings were identified.

(2) Effectiveness of 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 with some weaknesses including lack of rigor. The inspectors identified one weak RCE (1911809: Entry into an Unusual Event Due to a Service Water Leak from Zurn Strainer SW-2911-BS) where the cause listed in the evaluation was not the cause of the identified condition.

The inspectors identified no items in the backlogs of the CA program or maintenance WO system that were risk significant, either individually or collectively, although the inspectors and the licensee identified a large backlog of CA program activities in the Engineering Department. The inspectors reviewed the licensees WO backlog and associated performance metric data and concluded that equipment issues were generally being addressed appropriately.

Observations Licensee data indicated that in May 2015 the CA program had approximately 815 open conditions reports and approximately 490 CA program open actions with an additional 107 items classified as long term corrective actions. That data indicated that there were seven open CRs that were initiated prior to 2006 and an additional 44 that were initiated prior to 2010. For CA program open items, the average age was 114 days with 150 open items greater than 120 days old. The average age of long term corrective actions was shown as 1346 days. The Engineering Department was responsible for the majority of the backlogged items. The licensee had previously identified the backlog issue in the Engineering Department.

Licensee data indicated that, as of May 2015, the station had approximately 3,500 open WOs. The station uses the industry classification scheme in AP-928, Work Management Process Description, for grouping WOs. In May 2015, the station had approximately 25 WOs classified as corrective maintenance and 1,250 classified as deficient maintenance. The inspectors review concluded that the numbers were consistent with industry averages and from the documents reviewed did not identify any current significant corrective maintenance issues.

The inspectors reviewed several available RCEs and ACEs. ACEs reviewed varied from detailed and intrusive to quick reviews that used the Why Staircase, as the predominant and sometimes the only analysis tool. The inspectors questioned that practice and the resultant conclusions for ACE 1903531, Workers Loosened Bolts on Wrong Pump, and ACE 1946969, NCV-Inadequate Effectiveness Reviews. The licensee stated that for less significant ACEs there was no specific written guidance for the detail expected; the licensees present practice is for the Management Review Committee to inform the evaluating organization on the amount of detail expected.

The licensee stated that there is procedural guidance for the expected detail for RCEs and more significant ACEs.

Root Cause Evaluation Deficiencies During the review of RCE 1911809, Entry into an Unusual Event Due to a Service Water Leak from Zurn Strainer SW-2911-BS, the inspectors identified several deficiencies with the evaluation and determined that the licensees conclusion was incorrect. Specifically, the licensee found that the packing bolts were too short and concluded that the cause of the packing gland bolts becoming disengaged was an inadequate design of the packing gland bolting. The inspectors review found no regulatory concerns with the design and based on the information presented in the RCE concluded that the design, if maintained properly and correct replacement parts were used, was adequate. When the inspectors questioned a statement in the RCE about an original strainer Bill of Materials specifying seven-eighth inch cap screws for the packing gland, the licensee was unable to locate a document with those specific details.

Additionally, during the licensees review of the issue to answer the inspectors questions, an engineer identified that the bolts recovered after the packing gland leak were not the type specified by the vendor and were likely replaced sometime in the past.

The licensee entered the issue into their CA program as CR 02056763, SW Zurn Strainer RCE1911809 Inconsistency, and planned to revise the RCE.

G-03/G-04 Emergency Diesel Generator Operability The licensee identified that wind could adversely affect the ability of radiators to remove heat for emergency diesel generators G-03 and G-04. The effect of wind had not been previously evaluated in the design of the radiators. As part of CR 01980588, Basis for Operability of G-03/G-04 Questioned, the licensee utilized computer software to evaluate the effect of wind on the diesel generator radiators to conclude the diesel generators were operable. However, the licensee documented that to benchmark the software as required for use in a safety-related design basis calculation would be expensive. Therefore, the licensee decided to install a physical barrier so the effect of wind on the radiators could be neglected consistent with the original design.

The inspectors reviewed the licensees operability determination associated with CR 01980588 that concluded the radiators were operable considering the effect of wind.

The inspectors questioned why that the radiators and the diesel generators, while classified as operable, were not classified non-conforming to the existing design basis until the wind barrier was installed. Specifically, to be fully operable in the interim, did the licensee need a design basis evaluation that considered the effect of the newly identified wind load? The inspectors verified that a physical wind barrier was installed.

Therefore, the effect of wind on the G-03/G-04 and its radiators does not need to be considered consistent with the original design basis.

Component Service Life Grace Period The inspectors identified that PBNP utilized a 25 percent grace period with respect to vendor recommended component replacement frequency. In discussion with the licensee, the grace period is utilized at PBNP to provide component replacement flexibility, not to exceed the vendor recommended replacement frequency. For example, a component with a vendor recommended 10-year replacement frequency had a PBNP target replacement frequency of 7.5 years; component replacement less than the vendor recommended10 years utilizing the 25 percent grace period. However, the inspectors found that NextEra Energy fleet procedure MA-AA-204-1000, Preventive Maintenance and Surveillance Procedure, allowed the licensee to extend component service life, the vendor recommended replacement frequency, by up to 25 percent without an engineering evaluation. The licensee initiated condition report CR 02059320, PM Process Requires Improved Guidance, to address this concern.

The inspectors informed the licensee that the NRC has a documented regulatory position on design life of safety-related structures, systems, and components (SSCs) in Task Interface Agreement (TIA) 2014-01, Final Task Interface Agreement - Regulatory Position of Design Life of Safety-Related Structures, Systems, and Components Related to Unresolved Items at Donald C. Cook Nuclear Power Plant, Monticello Nuclear Generating Plant, and Palisades Nuclear Plant (TIA 2014-01), dated May 7, 2015 (ADAMS Accession No. ML15127A569). Specifically, the NRC staffs position expressed in this TIA is, in-part, if a safety-related SSC exceeds its specified service life or the licensee has information that challenges the presumption that a safety-related SSC can perform its specified function(s), the licensee must promptly address the nonconforming condition in accordance with their operability/functionality and CA programs. The inspectors noted that the NRC had previously issued non-cited violations at PBNP for equipment that failed after its stated design life and had not received an appropriate engineering evaluation; the inspectors did not identify any new current issues with equipment operating beyond its design life.

Findings Failure to Demonstrate the Functionality of a Credited Safe Shutdown Component

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated non-cited violation (NCV) of license condition 4.F for the licensees failure to demonstrate the capabilities of systems needed to perform a design function for 10 CFR Part 50, Appendix R safe shutdown. Specifically, the licensees tests, inspections, or maintenance activities did not demonstrate that CC-722A, the component cooling water pump suction cross-tie valve, was capable of being opened as required in AOP-10B, Safe to Cold Shutdown in Local Control.

Description:

During the assessment of the component cooling water system, the inspectors identified a condition report related to the licensee extending a preventative maintenance task for cycling CC-722A from five years out to nine years. The inspectors performed a detailed review of the issue and found that the licensee had identified, in 2004, that they should be cycling manual valves credited for Appendix R safe shutdown, and had identified as early as 2005, that CC-722A needed to be cycled. The licensee created a work request in 2007, which resulted in a work order to perform the necessary testing. In May 2010, the licensee canceled the work order after they determined that opening CC-722A and cross-connecting both units component cooling water (CC)systems would render both units CC systems inoperable. The inspectors review of Plant Health Committee meeting minutes from September 21, 2010, found that the licensee acknowledged that they lacked the ability to test a safe shutdown valve that was likely in a degraded condition from over-torqueing it shut to stop intersystem leakage. The meeting minutes also acknowledged that the valve had not been tested since 1997. The Plant Health Committee approved a modification to add a second valve in the system to allow testing without having to cross-connect the systems. The modification had a projected completion date of October 2012; however, the inspectors found that the modification had not been completed in October 2012 and when the next work order came due in 2013 the licensee canceled it and extended the testing from a five year interval to a nine year interval. The inspectors review found that the licensee planned to extend the testing beyond their implementation of National Fire Protection Association Standard 805, "Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants," which then would not require opening CC-722A for cold shutdown. After implementation of the standard the licensee planned to cancel the testing without performing the valve cycling.

The inspectors also questioned why the testing was not shown as overdue in the licensees work order system, since the valve hadnt been tested since 1997. The inspectors review found that even though the testing in 2010 was canceled, the work order system credited the canceled work as if it had been performed. The licensee informed the inspectors that a human performance error, by the individual canceling the work order in 2010, inappropriately left the box checked that credited completion. After reviewing the inspectors questions and concerns, the licensee concluded that the safe shutdown functionality of the valve could not be substantiated and commenced four hour fire rounds for the non-functional valve.

Analysis:

The inspectors determined that failure to demonstrate the capabilities of systems needed to perform a design function for Appendix R safe shutdown was contrary to the licensees Fire Protection Evaluation Report (FPER) and was a performance deficiency. The finding was determined to be more than minor because the failure to demonstrate the capabilities of systems needed to perform a design function for Appendix R safe shutdown was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage).

In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding affected the ability to reach and maintain cold shutdown, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1-A, because the inspectors determined that the finding would not prevent the reactor from reaching and maintaining hot shutdown.

This finding has a cross-cutting aspect of Resolution (P.3), in the area of problem identification and resolution, because the licensee did not take effective corrective actions to address the issue in a timely manner. Specifically, in 2007, the licensee identified that they had not been testing the valve as specified in their FPER and as of July 2015 had still not corrected it. Additionally, the licensee had an opportunity to identify and correct the situation in 2013 during their review and subsequent cancellation of the work order to add a second crosstie valve.

Enforcement:

License condition 4.F for both Unit 1 and Unit 2 required the licensee to implement and maintain in effect all provisions of the approved fire protection program as described in the final safety analysis report (FSAR) and Safety Evaluation Report dated August 2, 1979, (and Supplements dated October 21, 1980, January 22, 1981, and July 27, 1988) and the Safety Evaluation Report issued January 8, 1997, for technical specifications (TS) Amendment No. 170 and No. 174. Section 9.10 of the FSAR stated that the FPER was incorporated into the FSAR by reference. Fire Protection Evaluation Report, Section 9.0, Control of Appendix R Safe Shutdown Equipment, states in part that existing PBNP tests, inspections, and maintenance activities demonstrate the capabilities of such systems to perform their design basis functions.

Contrary to the above, from March 4, 1997 until July 9, 2015, the licensee failed to include the cycling of CC-722A, in existing PBNP tests, inspections, and maintenance activities to demonstrate the valve was capable of performing its design basis function to open. Because this violation was of very low safety significance and it was entered into the licensees CA program as CR 02059145, CR 02059446, and CR 02060486, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. The licensee corrective actions included declaring CC-722A non-functional and commencing four hour fire rounds (NCV 05000266/2015007-01; 05000301/2015007-01, Failure to Demonstrate the Functionality of a Credited Safe Shutdown Component).

(3) Effectiveness of Corrective Actions On the basis of the corrective action documents reviewed, the inspectors concluded that the CAs appeared generally appropriate for the identified issues. Those CAs addressing selected NRC documented violations were also determined to be generally effective and timely. The inspectors review of the previous five years of the licensees efforts to address issues with the 120 volt AC inverters did not identify any recent negative trends or inability by the licensee to address long-term issues. The inspectors questioned the validity of effectiveness reviews that just verified that other specified corrective actions were done and did not check if the actions were effective.

Observations The inspectors evaluated the effectiveness review (CA 01896156-47) completed for the corrective actions for the RCE associated with CR 01896156, Common Cause Degraded Cornerstone-Mitigating Systems, Two White Findings. The inspectors noted that the completed review only verified that the specified corrective action of conducting training was completed and documented; the review did not determine or try to determine whether the training had accomplished its purpose.

In the Site Management Review Meeting document, dated April 24, 2015, the licensee stated that CA program oversight does not meet expectations and that guidance is not provided for effectiveness reviews other than CRs classified as Significance Level (SL)1, the most significant category; guidance does not exist for SL2 and SL3 CRs, the lower categories available for safety issues.

Findings No findings were identified.

(4) Corrective Actions Associated with Root Cause Evaluations for the Degraded Cornerstone Resulting From Potential Flooding Issues In February 2015, the NRC completed a supplemental inspection (NRC Inspection Report 05000266/2015009; 05000301/2015009 (ADAMS Accession Number ML15077A007)) in accordance with Inspection Procedure (IP) 95002, Inspection for One Degraded Cornerstone or Any Three White Inputs in a Strategic Performance Area, to assess the licensees evaluation of one White inspection finding that affected the Mitigating Systems Cornerstone. The NRC, during that inspection, reviewed completed corrective actions from licensees root cause evaluations 01883633, White Finding-Flooding, and 01896156, Common Cause Degraded Cornerstone-Mitigating Systems, Two White Findings. During this inspection the NRC reviewed corrective actions from those root causes that were closed subsequent to the February 2015 inspection.

For RCE 01883633, the NRC during this inspection reviewed corrective actions number 39, 46, 51, 60, 61, 63, and 71. Corrective actions that were not closed at the time of this inspection were number 48, 49, 53, 58, 62, and 70.

For RCE 01896156, the NRC during this inspection reviewed corrective actions number 43, 44, 45, 47, 50, 52, 59, 60 62, 66, and 67. Corrective actions that were not closed at the time of this inspections were number 49, 53, 54, 63, 64, 65, 68, 69, and 70.

.1 Assessment of the Use of Operating Experience

a. Inspection Scope

The inspectors reviewed the licensees implementation of the facilitys operating experience (OE) program. Specifically, the inspectors reviewed OE program implementing procedures, attended CA program meetings, reviewed completed evaluations of OE issues and events, and selected assessments of the OE composite performance indicators. The inspectors review was to determine whether the licensee was effectively integrating OE into the performance of daily activities, whether evaluations of issues were proper and conducted by qualified personnel, and whether the licensees program was sufficient to prevent future occurrences of previous industry events. The inspectors also assessed whether CAs, as a result of OE, were identified and implemented in an effective and timely manner.

Assessment Overall, the inspectors determined that the licensee was generally effective at evaluating NRC and industry OE for relevance to the facility. The inspectors also verified that the use of OE in formal CA program products such as root cause evaluations and apparent cause evaluations was appropriate and adequately considered. Generally, OE that was applicable to the facility was appropriately evaluated and actions were implemented in a timely manner to address any issues that resulted from the evaluations.

Observations The inspectors identified some examples and trends that indicated that the licensee could improve some areas in their OE Program. Specifically:

• A review of licensee OE performance indicators found that the system engineering group had high average closure time for OE actions;

• The licensees evaluation associated with the higher than expected radiation dose rates during the 2014 fall outage concluded that the licensee failed to incorporate existing OE on proper primary chemistry control after their power uprate; and

• Multiple examples were identified by the licensee of failures to incorporate vendor technical information into procedures including two NRC violations, one in the 2014 NRC fourth quarter integrated report and one in the 2015 first quarter report.

Also, after reviewing a licensee evaluation of industry OE related to vibration induced copper piping failures, which concluded that the station adequately identifies and addresses copper piping vibration, the inspectors toured the turbine building to look for unidentified examples. The inspectors chose an instrument air pipe with the highest vibration and asked the licensee if it was acceptable. The licensee evaluated the piping and concluded that it was above desired levels. The licensee documented the issue in CR 02058263, Vibration in IA Line to AOV, and initiated a work request to further evaluate and correct the vibration.

The inspectors reviewed the licensees evaluation of NRC operating experience related to a 10 CFR Part 21 report regarding deviations utilizing computer software program STAAD.Pro in design of structures or components. Specifically, the NRC received notification pertaining to identification of deviations associated with calculations that were supplied to various licensees (AECOM letters to NRC dated May 18, 2015 (ADAMS Accession No. ML15147A547), and June 17, 2015 (ADAMS Accession No.

ML15168B075)). The Part 21 reports indicated that the noted deviations in STAAD.Pro can result in non-conservative evaluations of structural members and can potentially result in the inadequate design of structural components.

In discussion with the licensee, the inspectors determined that NextEra Energy had not utilized STAAD.Pro software for structural design and PBNP also verified that a supplier of design calculations had not utilized STAAD.Pro software for structural design at PBNP. However, the licensee had not documented this evaluation. The licensee initiated condition report AR 02056361 to address the inspectors concern that the licensees Part 21 evaluation was not documented.

The inspectors inquired if any vendor had performed structural calculations on behalf of PBNP that utilized the STAAD.Pro software. The licensee initiated condition report CR 02056363 to determine whether vendors had used STAAD.Pro in support of PBNP and any potential impact. As a result, the licensee determined that STAAD.Pro was utilized in PBNP calculation S-11165-116-02. The licensee initiated condition report CR 02056821, Vendor Notify Of Calc Impact/Staad Part 21, and determined the effect of STAAD.Pro deviations to be insignificant. The inspectors concluded the safety concern was minor since calculation S-11165-116-02 conclusions remained valid.

b. Findings

No findings were identified.

.2 Assessment of Self-Assessments and Audits

a. Inspection Scope

The inspectors assessed the licensee staffs ability to identify and enter issues into the CA program, prioritize and evaluate issues, and implement effective corrective actions, through efforts from departmental assessments and audits.

b. Assessment The inspectors concluded that self-assessments and audits were typically accurate, thorough, and effective at identifying most issues and enhancement opportunities at an appropriate threshold with some exceptions when Quick-Hit assessments were used.

The inspectors concluded that personnel involved in audits and self-assessments were knowledgeable in the subject area they audited or assessed. In many cases, self-assessments and audits identified issues that were not previously recognized by the licensee.

Observations The inspectors reviewed the licensees assessment of their self-assessment and benchmarking program. The provided assessment appeared to review and assess the compliance of their assessment program with existing procedural requirements and did not look at the effectiveness of the program. However, the inspectors overall assessment of the self-assessment process indicated that the program appeared to meet licensee intended requirements for identifying issues.

The inspectors observed that the licensee had many Quick Hit and Benchmarking items, which were small in scope, and fewer of the focused self-assessments, which would be a larger or more in depth review. The inspectors review of the licensees Nuclear Oversight audits found them to be thorough and effective at identifying issues. Some of the reviewed Quick Hit assessments had very narrow scopes, which was consistent with licensee expectations, but appeared to not add much value to the overall assessment process.

Findings No findings were identified.

.3 Assessment of Safety Conscious Work Environment

a. Inspection Scope

The inspectors assessed the licensees safety conscious work environment (SCWE)through the reviews of the facilitys employee concerns program (ECP) implementing procedures, discussions with the coordinator of the ECP, interviews with personnel from various departments, and reviews of issue reports. The inspectors also reviewed the results from a 2010 and a 2013 organization effectiveness survey, the results from a 2014 Employee Engagement Survey, and recent results from licensee-initiated Health Checks surveys.

As part of the overall inspection effort, inspectors discussed department and station programs with a variety of people. In addition, the inspectors held scheduled interviews with 43 non-supervisory individuals, in groups of four to seven people, from various departments to assess their willingness to raise nuclear safety issues. Additionally other personnel were randomly asked their views of the effectiveness of the CA program.

The individuals for the scheduled interviews were randomly selected to provide a distribution across the various departments at the site. In addition to assessing individuals willingness to raise nuclear safety issues, the interviews also included discussion on any changes in the plant environment over the last 12 months. Other items discussed included:

• knowledge and understanding of the CA program;

• effectiveness and efficiency of the CA program;

• willingness to use the CA program; and

• knowledge and understanding of ECP.

The inspectors also discussed the functioning of the ECP with the program coordinator; reviewed program logs from 2013, 2014, and 2015; and reviewed three case files.

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 majority of licensee personnel felt free to raise nuclear safety issues without fear of retaliation. Licensee personnel were aware of and generally familiar with the CA program and other processes, including the ECP and the NRCs allegation process, through which concerns could be raised; safety significant issues could be freely communicated to supervision. 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 even with survey issues identifying conditions that might act to inhibit discussion of items including mistrust of management and lack of effective communications among groups and departments.

Observations The licensees March 2015 Health Check survey results, compiled from 103 respondents, showed that less than 50 percent of survey participants answered agree or strongly agree to questions on effective communications and coordination existing among work groups and downward flow of information from management. While 88 percent of survey participants agreed with the statement I trust my immediate supervisor, only 54 percent agreed that there was strong evidence of effective station leadership. The March survey, however, showed positive trends from monthly surveys going back to October 2014.

Non-supervisory personnel in the interviewed arranged groups, when presented with the March 2015 survey results, stated that at their level there were no issues with working with and communicating with workers in other groups. The majority of the groups expressed their opinion that communications and coordination issues, if they existed, were at levels above them.

All interviewees indicated that they could and would bring up safety issues with supervision, management, and through the CA program. None of the interviewed personnel stated that there was intimidation or retaliation when they brought up issues.

Those same interviewees predominantly said they would use the ECP but saw no need to have to resort to the ECP for issue reporting.

Findings No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On July 10, 2015, the inspectors presented the inspection results to the Plant Manager, Mr. DeBoer, 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 and that all material considered proprietary by the licensee was returned to the licensee.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. DeBoer, Plant General Manager

S. Aerts, Performance Improvement Manager

P. Baranowski, Engineering

L. Hawki, Engineering ERRT Manager

K. Lemmens, Maintenance

J. Olson, Operations CAPCO

M. Omillian, Training

S. Pfaff, Performance Improvement Supervisor

T. Schneider, Senior Engineer

B. Scherwinski, Licensing

J. Schultz, Training

Nuclear Regulatory Commission

J. Cameron, Chief, Reactor Projects Branch 4

D. Oliver, Senior Resident Inspector

LIST OF ITEMS

Opened

05000266/2015007-01 NCV Failure to Demonstrate the Functionality of a Credited Safe

05000301/2015002-02 Shutdown Component (Section 4OA2.2b.(2))

Closed

05000266/2015007-01 NCV Failure to Demonstrate the Functionality of a Credited Safe

05000301/2015007-01 Shutdown Component (Section 4OA2.2b.(2))