IR 05000331/2017007

NRC Problem Identification and Resolution Inspection Report 05000331/2017007

ML17123A087
Person / Time
Site:	Duane Arnold
Issue date:	05/03/2017
From:	Stoedert K NRC/RGN-III/DRP/B1
To:	Dean Curtland NextEra Energy Duane Arnold
References
IR 2017007
Download: ML17123A087 (28)

Text

May 3, 2017

SUBJECT:

DUANE ARNOLD ENERGY CENTERNRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000331/2017007

Dear Mr. Curtland:

On April 25, 2017, 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 interim exit meeting on March 24, 2017, and an exit teleconference on April 25, 2017, with you 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.

The licensee had a low threshold for identifying issues and entering them into the CAP.

Through consideration of risk and consequence, the significance of the issues and priority for issue evaluation and resolution were determined. 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 based on the samples we reviewed. In addition, self-assessments and audits were conducted at appropriate frequencies with sufficient depth and details for all departments. The assessments 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, the NRC has identified an issue that was evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that a violation is associated with this issue. Because the licensee initiated condition reports (CRs) to address the issue, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section 2.3.2a of the Enforcement Policy. The NCV is described in the subject inspection report.

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; and the NRC Resident Inspector at the DAEC.

If you disagree with the 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III, and the NRC Resident Inspector at the DAEC.

This letter, its enclosure, and your response, (if any), will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Karla Stoedter, Chief Branch 1 Division of Reactor Projects Docket No. 50-331 License No. DPR-49

Enclosure:

Inspection Report 05000331/2017007 cc: Distribution via LISTSERV

SUMMARY OF FINDINGS

Inspection Report 05000331/2017007; 03/06/2017 - 04/25/2017; Duane Arnold Energy Center;

Identification and Resolution of Problems.

This report covers an 8-week period of inspection by four regional inspectors. A Green finding was identified by the inspectors. The findings involved a non-cited violation (NCV) of the U.S. Nuclear Regulatory Commission (NRC) requirements. 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, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated November 1, 2016.

The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6, dated February 2016.

Identification and Resolution of Problems On the basis of the samples selected for review, the team concluded that the Corrective Action Program (CAP) at Duane Arnold Energy Center (DAEC) was generally effective in identifying, evaluating and correcting issues. The licensee had a low threshold for identifying issues and entering them into the CAP. Through consideration of risk and consequence, the significance of the issues and priority for issue evaluation and resolution were determined. 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 based on the samples we reviewed. In addition, self-assessments and audits were conducted at appropriate frequencies with sufficient depth and details for all departments. The assessments 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 licensees 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 represented potential weakness of the program.

Cornerstone: Barrier Integrity

Green.

The inspectors identified a finding and an associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(f)(1) for the licensees failure to scope in multiple check valves of the main steam isolation valve leakage treatment system (LTS) into the Inservice Testing (IST) Program. Specifically, these valves were credited to mitigate the consequences of the main steam isolation valve leakage following a loss of coolant accident but they were not scoped into the IST program. Since the licensee made a commitment to the NRC to put these valves into the IST program as part of License Amendment 207, this issue is also a Deviation in accordance with the NRC Enforcement Policy. The licensee put this issue into the CAP as Action Requests (ARs) 2193481 and 2193482 and planned to include these valves in the full IST program.

This performance deficiency was more than minor because if left uncorrected, there was a potential to lead to a more significant safety concern. Specifically, these valves that were credited to mitigate the consequence of an accident were not tested in accordance with the IST program. The finding screened as very low safety significance (Green)because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components, nor did it involve an actual reduction in function of hydrogen igniters in the reactor containment.

The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of evaluation because the licensee justified that the valves be put into the augmented IST program since they were non-code components. In addition, the licensee did not re-scope these components into the IST program when 10 CFR 50.55(f)(1) was changed in 1999. This misconception continued when the licensee discovered several valves of the LTS were not in the IST program scope in 2015. [P.2] (Section 4OA2.1.b.2.ii)

REPORT DETAILS

4OA2 Problem Identification and Resolution

This inspection constituted one biennial sample of problem identification and resolution (PI&R) as defined by Inspection Procedure 71152, Problem Identification and Resolution. Documents reviewed are listed in the Attachment to this report. Note that the licensees computer program tracks condition reports (CRs) as action requests (ARs).

.1 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 Department 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 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 May 2015. The inspectors also reviewed selected issues that were more than five 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, apparent cause evaluations and condition evaluations. The inspectors assessed the scope and depth of the licensees evaluations. For issues that were characterized as significant conditions adverse to quality, the inspectors evaluated the licensees corrective actions to prevent recurrence and for issues that were less significant, the inspectors reviewed the corrective actions to determine if they were implemented in a timely manner commensurate with their safety significance.

The inspectors performed a 5-year review of the safety-related control building chiller system based on input from the resident staff. The system is part of the control building heating, ventilation and air conditioning system and its function is to provide chilled water for temperature control in the building. This ensures operability of plant equipment and maintains accessibility and habitability of the building, including the control room. The primary purpose of this review was to determine whether the licensee was monitoring and addressing performance issues of the control building chiller system. The inspectors performed walkdowns, as needed, to verify the resolution of issues.

A 5-year review of the instrument air system was also performed to assess the licensees efforts in monitoring the effectiveness of maintenance. Although this system is non safety related, its failure would adversely affect plant operation and require operator intervention. The system is currently a Maintenance Rule (a)(1) system that is in the monitoring phase of the (a)(1) action plan. 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 also performed walkdowns, as necessary, to verify the resolution of the issues.

b.

Assessment

(1) Identification of Issues Based on the results of the inspection, the inspectors concluded that DAEC was generally effective in identifying issues at a low threshold and entering them into the CAP. The inspectors determined that problems were normally 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 licensee was generally effective at trending low level issues to prevent larger issues from developing. The licensee used the CAP to document instances where previous corrective actions were ineffective or were inappropriately closed.

The inspectors performed a 5-year review on the control building chiller systems. As part of this review, the inspectors interviewed the current system engineer, reviewed CRs, critical equipment failure evaluations and condition evaluations. In addition, the inspectors performed a system walkdown to assess the material condition of the system and surrounding area. The inspectors concluded that control building chiller system 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)

Observation

Declining Rate of Identification

The inspectors review the CR generation rate for the last five years and noted a steady decline over this period. Specifically, there were over 9,100 CRs generated in 2012 compared with only about 6,600 generated in 2016, a 28 percent drop. The most significant decline was from 2014 to 2016 when a 27 percent drop was observed.

Considering the impact of an outage year to a non-outage year and the change in how the licensee classified routine work activities as non-corrective action items, the inspectors still observed an 11 percent drop in identification rate.

As documented in the pre-inspection self-assessments, the licensee had also recognized this declining trend and had taken steps to address this issue. The inspectors recognized that there may be multiple reasons for this issue such as:

organizational change, staff reduction, backlog reduction, etc. Based on the samples reviewed, both low and high safety significance issues were in the CAP. Through interviews with the licensees staff, the inspectors concluded that the staff were willing to bring up safety issue and write CRs. Therefore, this declining trend had not affected plant operations but the licensee needs to be cognizant of this trend before it affects the problem identification process.

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

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.

i)

Observations Corrective Action Program Process Issues During this inspection, the inspectors identified a number of CAP process issues at DAEC. For example, AR 1776321, LPCI Manual Realignment from S/D Cooling in Mode Three, documented an issue related to the low pressure cooling injection operation. The licensee performed an apparent cause evaluation but did not assign corrective action to the apparent cause identified as required by the licensees CAP procedure. A number of actions were assigned but were not corrective actions such that they could be changed or cancelled without as much oversight as corrective actions would receive. The licensee initiated AR 2192557, NRC PI&R - Corrective Actions for ACE 1776321-05, to address this issue.

In another example, AR 1599839, Replacement of 1VAC015A/B Cooling Coils Has Not Been Timely, documented timeliness issues with safety-related room cooler cooling coils replacement. One of the corrective action assignments to replace the coil was canceled without proper justification and approval from the Management Review Committee as required by the licensees CAP procedure. The licensee initiated AR 2192698, NRC PI&R - LTCA 1599839-04 Cancelled Inappropriately, to address this issue.

Although these issues were minor procedure violations, the licensee needs to be vigilant and adhere to procedures in order to maintain the overall CAP effectiveness.

ii)

Findings Failure to Include Valves in the Inservice Testing Program

Introduction:

The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.55a for the licensees failure to scope multiple check valves into the IST Program. Since the licensee made a commitment to the NRC to put these valves into the IST program as part of License Amendment 207, this issue is also a Deviation in accordance with the NRC Enforcement Policy.

Description:

On August 15, 1994, as supplemented on December 21, 1994, and January 20, 1995, the licensee submitted a request for a license amendment to increase the allowable main steam isolation valve (MSIV) leakage and delete the Technical Specification requirements applicable to the MSIV leakage control system (LCS). MSIV leakage would be directed to the main steam drain lines and the main condenser instead of the LCS. The licensee proposed to use non-safety-related components to fulfill the safety related leakage control function through an alternate treatment path. The licensee committed in their submittal, among implementing modifications, that all valves within the seismic verification boundary that were required to reposition to establish the boundary or treatment path would be included in the American Society of Mechanical Engineers (ASME),Section XI, IST program. License Amendment 207 was approved and issued by the NRC on February 22, 1995. The ASME Section XI Code is now the ASME Code for Operations and Maintenance (OM Code).

Despite the commitment, the licensee did not include all the valves within the seismic verification boundary into the scope of the ASME IST program when Amendment 207 was implemented in 1995. This new leakage treatment system (LTS) contained valves that were scoped into the IST program because of meeting other scoping criteria.

However, five LTS valves were put in to the augmented IST program and four other LTS valves, due to an oversight, were not included in the IST program at the time. During an outage scope review in 2015, the licensee identified that four valves for the LTS were not in the IST program and proceeded to put one of them into the augmented IST program. Actions to put the other three valves into the augmented IST program were assigned but not completed.

The licensee uses the augmented IST program for components that are not ASME Class 1, 2 or 3 components but have a safety function or have a commitment to be included in the IST program. Testing of the valves in the augmented IST program would be performed in accordance with the ASME Code only to the extent practical. However, relief requests would not be submitted for a valve if the code requirement could not be met. Therefore, components in the augmented IST program are subjected to less stringent requirements than those in the IST program.

Prior to September 1999, 10 CFR 50.55a(f)(1) required, in part, that safety-related pressure vessels, piping, pumps and valves must meet the requirements applicable to components which are classified as ASME Code Class 2 or Class 3 for a boiling water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971. On September 22, 1999, 10 CFR 50.55a(f)(1) was revised and requires, in part, that pumps and valves that perform a function to shut down the reactor or maintain the reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for safety related systems (in meeting the requirements of the 1986 Edition, or later, of the Boiler and Pressure Vessel or OM Code) must meet the test requirements applicable to components which are classified as ASME Code Class 2 or Class 3 for a boiling water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971. DAEC received their construction permit on June 22, 1970. Therefore, all the valves in the LTS were required to be scoped in the full IST program per 10 CFR 50.55a(f)(1) as they were credited to mitigate the consequences of an accident.

On December 8, 2015, during a refueling outage scoping review, the licensee identified that the four valves were not included in the IST program. Actions were created to scope these valves into the augmented IST program. Upon the inspectors questioning, the licensee reviewed the work history and determined only corrective maintenance had been performed on some of these valves. The licensee completed a functional assessment and determined the valves were functional but non-conforming.

Analysis:

The inspectors determined that the failure to include the nine valves into the IST program in accordance with 10 CFR 50.55a(f)(1) as well as the commitment for License Amendment 207 was within the licensees ability to foresee and correct. This issue was therefore a performance deficiency and was more than minor because if left uncorrected, there was a potential to lead to a more significant safety concern.

Specifically, these valves that were credited to mitigate the consequence of an accident were not tested in accordance with the IST program and may not function appropriately when needed. The inspectors evaluated the finding using the Significance Determination Process in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, dated June 19, 2012. The finding screened as very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components, nor did it involve an actual reduction in function of hydrogen igniters in the reactor containment.

The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of evaluation to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.

Specifically, through License Amendment 207, the licensee applied and the NRC approved the use of non-code components in lieu of a safety-related system provided that the licensee included the valves in the IST program to provide a high degree of confidence that these valves would remain functional. Instead the licensee mistakenly concluded that the valves could be put into the augmented IST program since they were non-code components. In addition, the licensee did not re-scope these components into the IST program when 10 CFR 50.55(f)(1) was changed in 1999. This misconception continued when the licensee discovered several valves of the LTS were not in the IST program scope in 2015. [P.2]

Enforcement:

Title 10 of the CFR, Section 50.55a(f)(1) requires, in part, that pumps and valves that perform a function to mitigate the consequences of an accident must meet the test requirements applicable to components which are classified as ASME Code Class 2 or Class 3 for a boiling water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971.

Contrary to the above, DAEC, which received their construction permit on June 22, 1970, failed to ensure that nine MSIV leakage treatment system valves which perform a function to mitigate the consequences of an accident met the test requirements applicable to components which are classified as ASME Code Class 2 or Class 3 for a boiling water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971. Specifically, the licensee put these nine valves in the augmented IST program and therefore, did not meet the all the requirements applicable to components which are classified as ASME Code Class 2 or Class 3.

The inspectors determined that the failure of the licensee to meet the commitments for License Amendment 207 was also a Deviation from February 22, 1995 when the amendment was issued, to September 22, 1999, when the requirements were codified in 10 CFR 50.55a.

The licensee put this issue into the CAP and planned to include these valves in the full IST program. Because this violation was of very low safety significance and the issue was entered in into the licensees CAP as ARs 2193481 and 2193482, this violation is being treated as a NCV, consistent with Section 2.3.2.a of the Enforcement Policy.

(NCV 0500031/2017007-01, Failure to Include Valves in the Inservice Testing Program)

(3) Effectiveness of Corrective Action Based on the results of the inspection, the inspectors concluded that the licensee was generally effective 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 performed a 5-year review of the instrument air system. As part of this review, the inspectors interviewed the current system engineer, reviewed the instrument air system health report, CRs, operating experience, and Maintenance Rule (a)(1) action plan. The system action plan was initiated in May 2016 to address a number of critical component failures, which exceeded the systems condition monitoring performance criterion. The failures had been appropriately addressed and the system was in the monitoring phase of its (a)(1) action plan. In addition, the inspectors walked down the instrument air system to assess material condition. The inspectors concluded that instrument air system 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)

Observations Inadequate Implementation of Technical Specification Bases Change Process The inspectors assessed the corrective actions associated with non-cited violation (NCV)05000331/2015007-02, Failure to Correctly Update the Updated Final Safety Analysis Report. The corrective actions associated with this NCV also included addressing errors in the Technical Specification Bases. The inspectors identified that one of the proposed corrections to Technical Specification Bases page B3.5-28 had been made but, due to an oversight, was changed back to read incorrectly. The inspectors considered this a minor violation of TS 5.5.10 a, Technical Specification Bases Change Control Program. Technical Specification 5.5.10.a., stated, Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews. Licensee procedure ACP 102.24, Preparation, Review and Processing of Bases Changes, Revision 10, Step 3.3.2.(3) stated that the Licensing Engineer determines the content of the proposed changes. This shall include consideration of the effect(s) on outstanding change requests. The licensee was working on two change packages with the same page in both packages simultaneously and failed to ensure that both packages contained the correct wording on the effected page. The licensee wrote AR 2192712, 2017 NRC PIR TS Bases Page Incorrectly Updated to correct this error.

Corrective Action Program Vulnerability The inspectors identified a vulnerability in the licensees CAP process. Currently, the licensees process allows the owner of a Significance Level 2 or 3 condition adverse to quality CR to approve intent changes and due dates extensions for assignments unless these actions were designated by the MRC as requiring MRC approval for due date extensions and intent changes. As such, corrective actions can be cancelled without scrutiny. For example, in AR 2063651, Allowable Stress Higher Than CLBs Listed in Block Wall Calc, the licensee documented an issue with a number of calculations that listed an incorrect allowable limit. This CR was initially screened by the licensee as a condition not adverse to quality. However, the MRC changed that to a condition adverse to quality and noted that corrective action was required. The MRC did not specifically require MRC approval for due date extensions and intent changes. However, the corrective action was later changed by the owner to a routine work assignment item and was currently pending for completion. Per procedure PI-AA-104-1000, Condition Reporting, a routine work assignment, when tied to a CR, does not meet the corrective action definition and are not part of the CAP. Therefore, the owners action that changed a corrective action into a routine work assignment directly conflicted with the intent of the MRC. The licensee entered this issue into the CAP as AR 2200047, DAEC NRC PI&R:

Vulnerability of the CAP to evaluate this issue.

ii) Findings No findings were identified.

.2 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 the OE program implementing procedures, attended CAP meetings to observe the use of OE information, and reviewed licensee evaluations of OE issues and events. The objective of the review was to determine whether the licensee was effectively integrating OE into the performance of daily activities, whether evaluations of issues were appropriate, 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 OE, were identified and implemented in an effective and timely manner.

b.

Assessment The inspectors observed that operating experience was discussed as part of the daily and pre-job briefings. Operating experience evaluations were limited to certain types; for example, NRC generic communications, significant industry issues, Part 21s, and General Electric Service Information Letters. Additional industry operating experience was disseminated across plant departments for their review and use, if needed. Specific equipment related issues were distributed to appropriate engineers for evaluating and screening into the CAP. The inspectors also verified that the use of OE in formal CAP products such as root cause evaluations and equipment apparent cause evaluations was appropriate and adequately considered. In addition, operating experience was used when developing the instrument air system Maintenance Rule (a)(1) action plan.

Generally, OE that was applicable to DAEC was thoroughly evaluated and actions were implemented in a timely manner to address any issues that resulted from the evaluations.

Based on the results of the inspection, the inspectors concluded that operating experience was effectively utilized at the station. No significant 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 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. The inspectors also determined that findings from the CAP self-assessment were consistent with the inspectors assessment.

c. Findings

No findings were identified.

.4 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 CRs. The inspectors also reviewed the results from a 2015 safety culture survey and meeting minutes of the Safety Culture Monitoring Panel.

The inspectors held scheduled interviews with 21 individuals in various group and individual settings to assess their willingness to raise nuclear safety issues. These individuals included supervisory and non-supervisory licensee and contractor personnel.

Additionally, the inspectors interviewed other licensee staff informally during plant walkdowns to ascertain their views on the effectiveness of the CA program and their willingness and freedom to raise issues.

The individuals in the scheduled interviews were randomly selected to provide a distribution across 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 six months. 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.

The inspectors also discussed the functioning of the ECP with the program coordinator; reviewed program logs from 2015 through 2017; and reviewed selected case files to identify any emergent issues or potential trends.

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 licensee personnel felt free to raise nuclear safety issues without fear of retaliation. Licensee and contractor personnel were aware of and generally familiar with the CAP and other processes, including the ECP and the NRCs allegation process, 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 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.

c. Findings

No findings were identified.

4OA6 Management Meetings

Exit Meetings On March 24, 2017, the inspectors presented the inspection results to Mr. D. Curtland and other members of the licensee staff. The licensee acknowledged the issues presented. One item had remained open pending licensees evaluation. This open item was discussed and closed during a teleconference exit on April 25, 2017. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Casey, Chemistry Manager

D. Curtland, Site Director

M. Davis, Licensing Manager

J. Debois, Performance Improvement Manager

M. Durbin, Maintenance Director

P. Hanson, Engineering Director

C. Hill, Training Manager

D. Hobson, Manager of Projects

D. Morgan, Radiation Protection Manager

M. Strope, Operations Manager

J. Schwertfeger, Security Manager

M. Foritz, Emergency Preparedness Manager

B. Simmons, Performance Assessment Manager

NRC

K. Stoedter, Branch Chief

B. Bergeon, Acting Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Open and

Closed

05000331/2017007-01 NCV Failure to Include Valves in the Inservice Testing (IST)

Program (Section 4OA2.1.b.2.ii)

Discussed

None