IR 05000341/2017007

NRC Problem Identification and Resolution Inspection Report 05000341/2017007

ML17213A364
Person / Time
Site:	Fermi
Issue date:	08/01/2017
From:	Billy Dickson NRC/RGN-III
To:	Fessler P Detroit Edison, Co
References
IR 2017007
Download: ML17213A364 (27)

Text

UNITED STATES ust 1, 2017

SUBJECT:

FERMI POWER PLANT, UNIT 2NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000341/2017007

Dear Mr. Fessler:

On June 23, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a Problem Identification and Resolution (PI&R) inspection at your Fermi Power Plant, Unit 2 (Fermi-2).

The enclosed inspection report documents the inspection results, which were discussed at an exit meeting on June 23, 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 Fermi-2, was generally effective in identifying, evaluating and correcting issues. The licensee had a low threshold for identifying issues and entering them into the CAP.

A risk based approach was used to determine the significance of the issues and priority for issue evaluation and resolution. Corrective actions were generally implemented in a timely manner, commensurate with their safety significance. Operating experience was entered into the CAP when appropriate and evaluated according to procedure. The use of operating experience was integrated into daily activities and found to be effective in preventing similar issues at the plant. In addition, self-assessments and audits were conducted at appropriate frequencies with sufficient depth for all departments based on the documents the team reviewed. The self-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 Fermi-2. Your 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 one finding of very low safety significance (Green). The NRC has also determined that a violation is associated with this finding. Because you have initiated corrective actions to address the issue, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section 2.3.2.a 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 Fermi Power Plant.

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/

Billy Dickson, Chief Branch 2 Division of Reactor Projects Docket No. 50-341 License No. DPR-43 Enclosure:

Inspection Report 05000341/2017007 cc: Distribution via LISTSERV

SUMMARY

Inspection Report 05000341/2017007; 06/05/2017-06/23/2017; Fermi Power Plant, Unit 2;

Identification and Resolution of Problems.

This inspection was performed by three region-based inspectors and the resident inspector at Fermi-2. One Green finding, which had an associated Non-Cited Violation (NCV) of the U.S. Nuclear Regulatory Commission (NRC) regulations, was identified. 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, dated July 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 Fermi-2, was generally effective in identifying, evaluating and correcting issues. The licensee had a low threshold for identifying issues and entering them into the CAP.

A risk based approach was used to determine the significance of the issues and priority for issue evaluation and resolution. Corrective actions were generally implemented in a timely manner, commensurate with their safety significance. Operating experience was entered into the CAP when appropriate and evaluated according to procedure. The use of operating experience was integrated into daily activities and found to be effective in preventing similar issues at the plant. In addition, self-assessments and audits were conducted at appropriate frequencies with sufficient depth for all departments based on the documents the team reviewed. 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 Fermi-2. Your 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.

Although implementation of the CAP was determined to be effective overall, the inspectors identified several issues that represented potential weakness of the program.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50,

Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to correct a design deficiency that mis-quantified unidentified leakage from reactor coolant system (RCS) pressure boundary. Specifically, in April 2007, the licensee identified that the driver mount drain for the reactor recirculation pump could potentially drain leakage from nearby pipe cracks to the identified leakage collection point. However, the licensee had not corrected this design deficiency as of the start of this inspection. The licensee documented this issue into the CAP as Condition Assessment Resolution Document (CARD) 17-25489 and developed a night order to direct the operators how to calculate unidentified leakage. The licensee also planned to revise procedure 24.000.02 as an interim measure until the modification was implemented.

The inspectors determined that the licensees failure to correct the design deficiency that mis-quantified unidentified leakage is a performance deficiency that is reasonably within the licensees ability to foresee and correct. The inspectors determined that this issue is more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, leakage that would normally be collected and measured as unidentified leakage could be collected and measured as identified leakage, leading to a potential violation of the TS unidentified leakage rate. Because the finding did not represent a loss of system or function, or represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) Allowed Outage Time, or represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in the licensees Maintenance Rule Program, it was screened as very low safety significance. The inspectors did not identify a cross-cutting aspect since the issue originated more than three years ago. (Section 4OA2.1.b.3.ii)

REPORT DETAILS

4OA2 Problem Identification and Resolution

This inspection constituted one biennial sample of problem identification and resolution (PI&R) inspection 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 as CARDs.

.1 Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The inspectors reviewed the procedures and processes that described the CAP at Fermi-2 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 meetings and the CARD Ownership / Screening Committee meetings.

Selected licensee personnel were interviewed to assess their understanding of and their involvement in the CAP.

The inspectors reviewed selected CARDs 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 CARDs reviewed were issued since the last NRC biennial PI&R inspection completed in December of 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, common cause evaluations and direct cause 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 five-year review of the safety-related emergency equipment cooling water system based on input from the resident staff. The system provides cooling water to remove heat from essential equipment located in the auxiliary building and reactor buildings that are required to shutdown the reactor. Additionally, the system provides cooling to essential loads following a loss of reactor building closed cooling water system, a loss of coolant accident or a loss of offsite power. The primary purpose of this review was to determine whether the licensee was monitoring and addressing performance issues of the emergency equipment cooling water system. The inspectors performed walkdowns, as needed, to verify the resolution of issues.

A five-year review of the offgas system was undertaken to assess the licensee staffs efforts in monitoring system performance. Although this system is non-safety related, its function is to reduce the offsite exposures at the nearest site boundary to less than the established maximum limit. Therefore, its failure could adversely affect plant operation and require operator intervention. The inspectors review was to determine whether the licensee staff was properly monitoring and evaluating the performance of the system through effective implementation of station monitoring program, such as the system health report. The inspectors performed walkdowns, as needed, to verify the resolution of issues.

The inspectors examined the results of self-assessments of the CAP completed by the licensee 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 Fermi-2 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 licensee appropriately screened issues from both NRC and industry operating experience at an appropriate level and entered 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 five-year review on the emergency equipment cooling water system. As a part of this review, the inspectors interviewed the current system engineer, reviewed CARDs, 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 areas. The inspectors concluded that emergency equipment cooling water 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 Change in Rate of Identification During the last biennial problem identification and resolution inspection, the team identified a small decline in CARD generation rate over the preceding five years.

The licensee attributed the small decline partially to process changes in how low level conditions were captured. The licensee implemented actions to educate its staff on the CAP process. The team reviewed the CARD generation for the last two years and concluded that the decline in generation rate had been arrested. The licensee continued to monitor the generation rates and would implement additional corrective actions when necessary.

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 CARD Ownership/Screening Committee meetings were generally thorough and maintained a high standard for evaluation quality. Members of the committees were engaged and discussed selected issues in sufficient detail as well as 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 Weakness on the Engineering Product Update Process The inspectors reviewed the engineering backlog data provided by the licensee.

Licensee procedure MES21, Incorporation of Changes into Design Documentation, prescribes the methods and schedule for incorporation of the as-built design change documents. It requires incorporation within 180 days after accumulation of five as-built design change documents affecting QA Level 1 calculations, unless an exception was approved by the responsible supervisor. The inspectors reviewed a sample of calculations included on the backlog list and found more than five posted changes against the safety-related calculations with some of the posted changes dating back to the early 90s. The inspectors also noted that since the procedure specified time clock for incorporation started only after accumulation of five design changes, there may be additional calculations beyond the backlog list having multiple and very old postings.

Even though the procedure allowed exception based on supervisor approval, the inspectors were concerned that the need to review all the postings against the calculation while assessing a new design change affecting a calculation with numerous postings would impose additional burden on the engineers performing the review and increase the possibility of errors. The licensee entered this issue in the CAP for evaluation and possible process improvements.

ii) Findings No findings were identified.

(3) Effectiveness of Corrective Actions 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 five-year review of the offgas system. The inspectors interviewed program owner and reviewed corrective action documents, inspection reports, inspection procedures, as well the system health reports. The inspectors evaluated in-progress and planned actions and performed a partial system walkdown of visible parts of the system.

The inspectors determined that there were numerous offgas system chiller failures in the past five year due to hardware and logic issues indicating significant problems adversely affecting the system performance. At the time of the inspection, the system health rating was Red, indicating unacceptable performance requiring significant actions for improvement. However, the licensees recent actions including implementation of a modification to replace the logic controller and completion of a number of work orders is expected to return the performance to acceptable level within the next one or two quarters. The licensee also plans to replace the chiller tubing for further improvement.

The inspectors concluded that the licensee staff were properly monitoring the performance of the system and taking actions necessary for improvement.

i) Observations Outstanding Corrective Action Items During the last biennial PI&R inspection, the inspectors identified that there was a large population of outstanding correction action items. Specifically, there were over 3000 open corrective action items at the time of the inspection. More than 500 of these open corrective action items were Level 3 to Level 1 items, which the licensee considered conditions adverse to quality or significant conditions adverse to quality. There were 68 items that were greater than two years old. For those items that affected safety related equipment, the inspectors determined that the corrective actions were untimely and the issues were minor violations of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because the equipment affected were either operable or operable with appropriate compensatory actions in place. Since the majority of these outstanding corrective action items were design non-conformances, the inspectors were mainly concerned that these outstanding items could potentially affect the licensees understanding of the design basis of the plant and complicate future equipment issue resolution. The licensee acknowledged the inspectors concern and implemented a number of actions to reduce the backlog.

During this inspection, the inspectors reviewed the licensees outstanding corrective action items and determined that the licensee had made good progress on resolving these issues. Overall, the licensee had reduced this backlog by one-third. As of the start of this inspection, there were just under 2000 open CARDs with only 300 of them were Level 1 to 3. There were about 40 corrective action items that were greater than two years old. The inspectors sampled a number of these corrective actions in each significance level and determined that they were being tracked with appropriate level of attention to ensure their completion. The inspectors also sampled other action items and verified that the licensee did not systematically downplay the action items as non-corrective actions. Although good progress had been made, the licensee needs to remain focus on reducing the backlog.

ii) Findings Failure to Correct a Design Deficiency that Mis-Quantified Unidentified Leakage

Introduction:

The inspectors identified a finding of very low safety significance (Green)and an associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to correct a design deficiency that mis-quantified unidentified leakage from RCS pressure boundary. Specifically, in April 2007, the licensee identified that the driver mount drain for the reactor recirculation pump could potentially drain leakage from nearby pipe cracks to the identified leakage collection point. However, the licensee had not corrected this design deficiency as of the start of this inspection.

Description:

In February 2007, the drywell equipment sump temperature began a slow rise. Coupled with a slight increase in input into the sump, the licensee initialized a leak investigation as well as developed an Operational Decision Making Issue (ODMI) to address the leak. Due to the small quantity of the leak, the licensee chose to continue operation of the plant with a leak monitoring plan and implemented compensatory measures. During the review of potential sources of the leakage in support of the ODMI, the licensee identified that the driver mount drain for the reactor recirculation pump could potential drain leakage from nearby pipe cracks to the identified leakage collection point in April 2007. This configuration could mask the magnitude of an unidentified leak by directing it to the equipment drain instead of the floor drain. The licensee documented this issue in CARD 07-22140 and took credit for the administrative limits established by the ODMI to ensure that TS limits were satisfied. A plant modification was assigned to separate identified and unidentified leakage. An interim procedure change to the TS surveillance 24.000.02, Shiftly, Daily and Weekly Required Surveillance, to limit the sum of identified and unidentified leakage to five gallons per minutes (gpm) was evaluated but not implemented.

In September 2007, the licensee shutdown the reactor for a refueling outage. The drywell leak was located and repaired. Subsequently, the ODMI was closed when the unit returned to full power operation after refueling. However, the licensee did not implement any action to ensure that the TS limits were satisfied due to the design deficiency. Because of this design, the unidentified leakage surveillance acceptance criterion was no longer appropriate for the circumstance.

The inspectors verified that the total leakage, unidentified and identified, had not exceeded the five gallon per minute TS limits for unidentified leakage since 2013 and therefore no TS violation had occurred during this period. The due date for the modification to separate identified and unidentified leakage had been extended multiple times and was currently scheduled to be completed in October 2018.

Analysis:

The inspectors determined that the licensees failure to correct the design deficiency that mis-quantified unidentified leakage was contrary to the requirements in 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. This issue was reasonably within the licensees ability to foresee and correct and was therefore a performance deficiency. Consistent with the guidance in Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that this issue is more than minor because if left uncorrected, the performance deficiency have the potential to lead to a more significant safety concern. Specifically, leakage that would normally be collected and measured as unidentified leakage could be collected and measured as identified leakage, leading to a potential violation of the TS unidentified leakage rate. The inspectors also reviewed the examples of minor issues in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, dated August 11, 2009, and found no similar examples.

In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green), because the finding did not represent a loss of system or function, or represent an actual loss of function of at least a single Train for greater than its Tech Spec Allowed Outage Time, or represent an actual loss of function of one or more non-Tech Spec Trains of equipment designated as high safety-significant in the licensees Maintenance Rule Program.

The inspectors did not identify a cross-cutting aspect since the issue originated more than three years ago.

Enforcement:

Title 10 CFR, Part 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that measures be established to assure that conditions adverse to quality, such as deficiencies, are promptly corrected. Contrary to the above, from April 2007 to June 21, 2017, the licensee did not correct a condition adverse to quality. Specifically, the licensee identified in April 2007 a design deficiency that mis-quantified unidentified leakage from RCS pressure boundary. The licensee implemented compensatory actions to address this issue but they were closed out in September 2007 without taking any further actions. The licensee documented this issue into the CAP as CARD 17-25489 and developed a night order to direct the operators how to calculate unidentified leakage. The licensee also planned to revise procedure 24.000.02 as an interim measure until the modification was implemented.

Because this violation was not repetitive or willful, was of very low safety significance, and was entered into the licensees corrective action program, it is being treated as a Non-Cited Violation consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05000341/2017007-01, Failure to Correct a Design Deficiency that Mis-Quantified Unidentified Leakage).

.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 included NRC generic communications, significant industry issues, Part 21s, and General Electric Services Information Letters. Additional industry OE 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.

Generally, OE that was applicable to Fermi-2 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 Quality Assurance 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 CAP as required by the 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 condition reports. The inspectors also reviewed the results from a number of licensee initiated safety culture survey and pulses conducted in 2016.

The inspectors held scheduled interviews with 21 onsite staff members. The interviews included individual contributors and supervisors from both licensee and contractor organizations. During the interview, the inspectors assessed their willingness to raise nuclear safety issues. Additionally, the inspectors interviewed other personnel informally during plant walkdown to ascertain their views on the effectiveness of the CAP 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 12 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 function 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 personnel were generally aware of and 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 Meeting On June 23, 2017, the inspectors presented the inspection results to Mr. P. Fessler 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.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

P. Fessler, Senior VP & Chief Nuclear Officer

M. Caragher, Executive Director - Production

L. Bennett, Director - Operations

E. Kokosky, Director - Organizational Effectiveness

D. Noetzel, Director - Nuclear Engineering

W. Raymer, Director - Maintenance

P. Summers, Director - Nuclear Support

C. Harris, Manager - Performance Improvement

K. Hullum-Lawson, Manager - Plant Support Engineering

S. Maglio, Manager - Licensing

G. Strobel, Manager - Outage & Work Management

U.S Nuclear Regulatory Commission

B. Dickson, Branch Chief

B. Kemker, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Open

05000341/2017007-01 NCV Failure to Correct a Design Deficiency that Mis-Quantified Unidentified Leakage (Section 4OA2.1.b.3.ii)

Closed

05000341/2017007-01 NCV Failure to Correct a Design Deficiency that Mis-Quantified Unidentified Leakage (Section 4OA2.1.b.3.ii)

Discussed

05000341/2017007-01 NCV Failure to Correct a Design Deficiency that Mis-Quantified Unidentified Leakage (Section 4OA2.1.b.3.ii)