IR 05000461/2017007

NRC Problem Identification and Resolution Inspection Report 05000461/2017007

ML17310A843
Person / Time
Site:	Clinton
Issue date:	11/06/2017
From:	Laura Kozak NRC/RGN-III/DRP/B1
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2017007
Download: ML17310A843 (30)

Text

B.

UNITED STATES ber 6, 2017

SUBJECT:

CLINTON POWER STATIONNRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000461/2017007

Dear Mr. Hanson:

On September 29, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a Problem Identification and Resolution (PI&R) inspection at your Clinton Power Station. The enclosed inspection report documents the inspection results, which were discussed at an exit meeting on September 29, 2017, with Mr. B. Kapellas, and other members of your staff.

This inspection was an examination of activities conducted under your license as they relate to PI&R and compliance with the Commissions rules and regulations and 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 Clinton Power Station was generally effective in identifying, evaluating and correcting issues. The team determined that station personnel generally 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. In addition, self-assessments and audits were found to be conducted at appropriate frequencies with sufficient depth for all departments. The assessments reviewed were thorough and effective in identifying site performance deficiencies, programmatic concerns, and improvement opportunities. On the basis of the interviews conducted, the inspectors did not identify any impediment to the establishment of a safety conscious work environment at Clinton Nuclear Power Station. Licensee staff was aware of and generally familiar with the CAP and other station processes, including the employee concerns program, through which concerns could be raised. The team determined that your stations performance in each of these areas supported nuclear safety.

Based on the results of this inspection, one NRC identified finding of very low safety significance (Green) was documented in this report. This finding involved a violation of NRC requirements.

The NRC is treating this violation as a Non-Cited Violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the subject or severity 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 Clinton Nuclear Power Station.

In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 Clinton Nuclear Power Station.

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/

Laura Kozak, Acting Chief Branch 1 Division of Reactor Projects Docket No. 50-461 License No. NPF-62 Enclosure:

Inspection Report 05000461/2017007 cc: Distribution via LISTSERV

SUMMARY OF FINDINGS

Inspection Report 05000461/2017007; 09/11/2017 - 09/29/2017; Clinton Power Station, Unit 1;

Biennial Problem Identification and Resolution.

This inspection was performed by four region-based inspectors and the Clinton Nuclear Power Station Resident Inspector. One Green finding was identified by the inspectors. This finding and violation was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red), and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated April 29, 2015.

Cross-cutting aspects are determined using IMC 0310, Aspects Within Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated 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.

Problem Identification and Resolution On the basis of the sample selected for review, the team concluded that implementation of the corrective action program (CAP) at Clinton Power Station was generally good. The licensee had a low threshold for identifying problems and entering them in the CAP. Items entered into the CAP were screened and prioritized in a timely manner using established criteria; were properly 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 determined to be performed at an appropriate level to identify deficiencies. On the basis of interviews conducted during the inspection, workers at the site expressed freedom to enter safety concerns into the CAP. There was one Green finding identified by the team during the inspection. The finding involved the failure to evaluate defeating reactor core isolation cooling system interlocks and trips before adding it to an emergency operating support procedure. This finding has a cross-cutting aspect in the area of Human Performance.

NRC-Identified

and Self-Revealed Findings Cornerstones: Mitigating Systems Green-Severity Level IV. The inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment.

Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the change to an emergency operating procedure (EOP) support procedure to incorporate three reactor core isolation cooling (RCIC) system interlock defeats and did not provide a basis for the determination that this change would not create a possibility for a malfunction of a structure, system or component (SSC) important to safety with a different result than any previously evaluated in the updated safety analysis report. The licensee entered this issue into the CAP as action request (AR) 04056394 and planned to perform a screening for the procedure change.

This performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B,

Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the change did not ensure RCIC system reliability and availability during and following design basis accidents because it introduced a new failure mode and added reliance on monitoring activities and manual actions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012,

Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV significance because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The team determined that this finding had a cross-cutting aspect of Resources in the area of Human Performance because the licensee did not ensure that procedures, and other resources were available and adequate to support nuclear safety. Specifically, the procedure which required a 50.59 screening for changes to EOP support procedures, was not explicit in requiring the screening. [H.1] (Section 4OA2.1.b.2)

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 Inspection Procedure 71152.

.1 Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The inspector reviewed the licensees Corrective Action program (CAP) implementing procedures and attended CAP meetings to assess the implementation of the CAP by site personnel.

The inspectors reviewed risk and safety significant issues in the licensees CAP since the last Nuclear Regulatory Commission (NRC) Problem Identification and Resolution (PI&R) inspection in September 2015. The selection of issues ensured an adequate review of issues across NRC cornerstones. The inspectors used issues identified through NRC generic communications, department self-assessment, licensee audits, operating experience reports, and NRC documented findings as sources to select issues. In addition, the inspectors reviewed Action Reports (ARs) and a selection of completed investigations from the licensees various investigation methods, which included root cause, apparent cause, equipment apparent cause, corrective action program evaluation and human performance investigations.

The inspectors selected two of high risk systems, which included the reactor core isolation cooling (RCIC) system and the switchgear ventilation (VX) system to review in detail. The inspectors review was to determine whether the licensee staff were properly monitoring and evaluating the performance of these systems through effective implementation of station monitoring programs. A five year review on the RCIC system and the VX system was also undertaken to assess the licensee staffs efforts in monitoring for system degradation due to aging aspects.

A supplemental inspection was performed in January 2016 in response to the White violation issued for the failure to verify the suitability of the replacement pump design for the Division 3 shutdown service water system. The same pump failed again in June 2017 and was being reviewed by the resident inspector office. As a part of their review, the resident inspector office would review the corrective actions that were not completed at the conclusion of the supplemental inspection. Therefore, the PI&R team did not complete the review of this item. (Inspection Report 05000461/2016008, ADAMS Accession Number ML16077A312)

During the reviews, the inspectors determined whether the licensee staffs actions were in compliance with the facilitys CAP and 10 CFR Part 50, Appendix B requirements.

Specifically, the inspectors determined if licensee personnel were identifying plant issues at the proper threshold, entering the plant issues into the stations CAP in a timely manner, and assigning the appropriate prioritization for resolution of the issues. The inspectors also determined whether the licensee staff assigned the appropriate 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 NRC findings, including non-cited violations.

b. Assessment

(1) Effectiveness of Problem Identification Based on the inspection results, the inspectors concluded that, in general, the station was effective in identifying issues at a low threshold and entering them into the CAP.

This was evident by the approximate 9000 ARs written yearly in each of the past 5 years. In interviews conducted by the inspectors, it was evident that station employees had no reservations about writing ARs. The inspectors determined that normally problems were identified and captured in a complete and accurate manner in the CAP. The inspectors also noted that deficiencies were identified by external organizations (including the NRC) that had not been previously identified by licensee personnel. These deficiencies were subsequently entered into the CAP for resolution.

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

The inspectors performed a 5-year review of the RCIC and VX system. As part of this review, the inspectors reviewed a sample of RCIC and VX system health reports, IRs, operating experience, and Maintenance Rule status. The inspectors reviewed licensees CAP and work management system procedures that provided guidance for trending. In addition, the inspectors walked down portions of the RCIC and VX system. The inspectors concluded that RCIC and VX 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 and Findings Condition Report not generated for Conditions Adverse to Quality The Inspectors identified one minor example of a condition adverse to quality not being documented in the corrective action program. The inspectors reviewed an equipment apparent cause evaluation performed under AR 1395861-02 where the licensee established an extent of condition review following the identification of internal damage to shutdown service water (SX) valve 1SX024A to determine the impact of Clinton raw water and corrosion rates on SX gate valves. The inspectors reviewed the work order associated with the internal inspection of the 1SX024B valve. The work order stated the valve was in the required open position but with degradation in several valve components, including the T-slot, stem, gland flange, and wedge guides. The inspectors determined the degraded condition was not entered into the corrective action program as required by PI-AA-120, Issue Identification and Screening Process.

Title 10 CFR Part 50, Appendix B, Criterion II, states, The applicant shall establish at the earliest practicable time, consistent with the schedule for accomplishing the activities, a Quality Assurance Program which complies with the requirements of this appendix. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout plant life in accordance with those policies, procedures, or instructions.

The licensees Quality Assurance Program was established in the Quality Assurance Topical Report (QATR). Corrective action procedure PI-AA-125 implemented a portion of the licensees CAP as described in the QATR and stated Specifically, at the direction of site management, Significant Conditions Adverse to quality and Conditions Adverse to Quality are resolved through direct action, the implementation of Corrective Actions to Prevent Recurrence (CAPRs) and Corrective Actions (CAs) and documented in the Computer Program.

Contrary to the above, the licensee failed to carry out the documentation portion of their Quality Assurance Program in accordance with procedure PI-AA-125. Specifically, the licensee failed to document the degraded condition identified during an extent of condition inspection of 1SX024B valve. The licensee captured this issue in AR 04056735, 2017 PIR Perform Past Operability for 1SX024B. However, because the valve was identified in the open position, did not challenge system operability, and was replaced under work order (WO) 04579646 this violation was determined to be minor.

(2) Effectiveness of Prioritization and Evaluation of Issues Based on the results of the inspection, the inspectors concluded that identified problems were generally prioritized and evaluated commensurate with their safety significance, including an appropriate consideration of risk. Higher level evaluations, such as root cause and apparent cause evaluations, were generally technically accurate; of sufficient depth to effectively identify the cause(s); and generally considered extent of condition, generic implications, and previous occurrences in an adequate manner.

The inspectors determined that the station ownership committee and management review committee meetings were generally thorough and meeting participants were actively engaged and well-prepared. Station ownership committee and management review committee meetings accurately prioritized issues.

The inspectors determined that, overall, Clinton Power Station personnel evaluated equipment operability and functionality requirements adequately after a degraded or non-conforming condition was identified, and appropriate actions were assigned to correct the degraded or non-conforming condition.

i) Observations and Findings Failure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support Procedure

Introduction.

The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),

Changes, Test, and Experiments, and an associated Green finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee changed an emergency operating procedures (EOP)-1 support procedure to incorporate three RCIC interlock defeats but did not provide a basis for the determination that this change would not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the updated safety analysis report.

Description.

The inspectors reviewed condition report AR 3984041, RCIC Low Suction Pressure Trip Needs a Defeat Action in EOPs, dated March 10, 2017, which was generated during a benchmarking effort at another Exelon facility that had incorporated the RCIC low suction pressure trip defeat and others, into an EOP support procedure.

The inspectors questioned what trip defeats Clinton had incorporated into their EOP support procedures and requested and reviewed procedure change records and documentation associated with these additional trip bypasses.

Inspectors reviewed the Clinton Safety Evaluation of the Upgraded Emergency Operating Procedure Program, dated December 1990 and noted that previously a list of RCIC interlock defeats were added to the EOPs and were reviewed by the NRC. The inspectors noted the following interlocks were not included in the safety evaluation:

(1) Defeating RCIC Exhaust Diaphragm High Pressure Isolation;

(2) Defeating RCIC High Exhaust Pressure Trip; and

(3) Defeating RCIC High Steam Flow Isolations.

Additionally, section A.1 of the Safety Evaluation acknowledged in part, The NRC stated that future changes to the [NRC accepted] procedure generation package (PGP) should be reviewed in accordance with 10 CFR 50.59.

Inspectors reviewed Exelon white paper, EXC-WP-10, RCIC Operation at Elevated Suppression Pool Temperatures Following an Extended Loss of AC Power (ELAP),dated February 5, 2015. Exelon Nuclear provided a list of RCIC Trip/Isolation Signals under Appendix A, which included the Interlocks:

(1) Defeating RCIC Exhaust Diaphragm High Pressure Isolation;

(2) Defeating RCIC High Exhaust Pressure Trip; and

(3) Defeating RCIC High Steam Flow. Section 4.3 stated in part that Each site should review [the list] to determine if trip/isolation signals would be expected. If the site determined trips were expected the site would also need to evaluate if the trip could be bypassed.

Inspectors reviewed EOP-1, RPV Control, and noted that both the path for RPV Water Level Control and the path for RPV Pressure, stated it is Ok to defeat RCIC Interlocks and referenced CPS 4410.00C001. CPS 4410.00C001 was updated on April 29, 2016 and May 26, 2016 to allow implementation of the following defeats when directed by a licensed Senior Reactor Operator, as required by CPS 4410.00, Defeating System Interlocks:

(1) defeating RCIC Exhaust Diaphragm High Pressure Isolation; (2)defeating RCIC High Exhaust Pressure Trip; and

(3) defeating RCIC High Steam Flow Isolations. The inspectors requested the documentation of the evaluation performed to add these interlocks to the EOP.

The licensee provided the AD-AA-101-F-01, Document Site Approval Form record for CPS 4410.00C001, Defeating RCIC Interlocks, dated April 28, 2015 which documented that 10 CFR 50.59 was not applicable for the following procedure changes that allowed

(1) Defeating RCIC Exhaust Diaphragm High Pressure Isolation;

(2) Defeating RCIC High Exhaust Pressure Trip; and

(3) Defeating RCIC High Steam Flow Isolations due to findings from the Fukushima event. The Procedure was also revised to incorporate an isolation reset section. Inspectors noted although the interlock defeats were added in response to actions recommended for beyond design basis events, they were incorporated into emergency procedures that are used for design basis events. This could introduce a new failure mechanism that was not screened or evaluated in accordance with 10 CFR 50.59 as required by licensee procedures and NRC regulations. The Procedure was also revised to incorporate an isolation reset section.

Inspectors reviewed AD-AA-101, Processing of Procedures and Training and Reference Materials (T&RMs) procedure which outlined the process for revising standard or site specific procedures. Inspectors noted that Section 1.2 stated in part, this procedure applies to all procedures and T&RMs except...Boiling water reactor (BWR) Emergency Operating Procedures (EOPs)... Therefore, this procedure was not applicable for the changes being made. Inspectors also reviewed CPS 1005.09, Emergency Operating Procedure (EOP) and Severe Accident Guideline (SAG)

Program. Step 8.9.4 of this procedure stated in part, A copy of EOP Revision Assessment letter shall be attached as supporting documentation to the 50.59 Screening Form prepared for the revision. However when asked about the 10 CFR 50.59 Screening Form, the licensee determined one was not completed for this procedure revision.

Inspectors noted although the interlock defeats were added in response to actions recommended for beyond design basis events, they were incorporated into emergency procedures that are used for design basis events. This could introduce a new failure mechanism that was not screened or evaluated in accordance with 10 CFR 50.59 as required by licensee procedures and NRC regulations. The inspectors also determined that adding manual operator actions to safety related components was not within the design basis of the facility; therefore the inspectors could not reasonably determine that this change would not have required prior NRC review and approval.

Analysis.

The inspectors determined that the failure to perform a written 10 CFR 50.59 evaluation to incorporate three additional defeats to RCIC system interlocks to EOP-1 support procedure, CPS 4410.00C001 Defeating RCIC Interlocks was contrary to 10 CFR 50.59(d)(1) and was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the change did not ensure RCIC system availability and reliability during and following design basis accidents because it introduced a new failure mode and added reliance on monitoring activities and manual actions.

Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the inspectors were able to answer all of the questions No.

The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Assessing the violation in accordance with Enforcement Policy, the inspectors determined it to be of Severity Level IV because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2).

The team determined that this finding had a cross-cutting aspect of Resources in the area of Human Performance because the licensee did not ensure that procedures and other resources were available and adequate to support nuclear safety. Specifically, licensee procedure CPS-1005.09, Emergency Operating Procedure (EOP) and Severe Accident Guideline (SAG) Program, requiring a 50.59 screening for changes to EOP-1 support procedures, was not explicit in requiring the screening. Procedure 1005.09 alluded to performance of a screening through step 8.9.9 which states a copy of EOP revision assessment letter shall be attached as a supporting documentation the 50.59 Screening Form prepared for the revision. [H.1]

Enforcement.

Title 10 CFR 50.59, Changes, Tests, and Experiments, Paragraph (d)(1)requires, in part, the licensee to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section.

Contrary to the above, from April 29, 2016 and May 26, 2016 until September 27, 2017, and ongoing, the licensee failed to maintain a record of a change in procedures made pursuant to 10 CFR 50.59(c) that included a written evaluation which provided the bases for the determination that the change did not require a license amendment pursuant to 10 CFR 50.59(c)(2). Specifically, the licensee made changes to the EOP-1, support procedure, CPS 4410.00C001, Defeating RCIC Interlocks, but failed to perform a written evaluation which provided the bases for the determination that the change did not require a license amendment pursuant to 10 CFR 50.59(c)(2)(vi). As a result, the licensee failed to provide a basis supporting that the change did not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the final safety analysis report (as updated).

The licensee documented this issue in their CAP as AR 04056394, 2017PIR 4410.00C001 Needs 50.59 Screening, and planned to perform a 50.59 screening to determine whether NRC approval was required prior to implementing the EOP changes.

Because the associated finding was of very low safety significance (Green) and because the licensee entered it into its corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy:

(NCV 05000461/2017007-01, Failure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support Procedure.)

(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 and licensee identified documented violations and determined that actions assigned were generally effective and timely.

i) Observations and Findings Corrective Action Assignments not Created when Required The inspectors identified two minor examples for the failure to document corrective action assignments for conditions adverse to quality. The inspectors reviewed AR 2614744 which was written to document the failure to perform an as found voiding inspection on the A residual heat removal system prior to performing a fill and vent of the system. This issue was classified as a level 3 condition in accordance with PI-AA-120, Issue Identification and Screening Process, which assigned a level between 1 and 5 to each identified condition. The level assigned to each issue was commensurate with the significance of the issue, level 1 being the most significant and level 5 being the least significant. The assigned level would indicate what the required follow-up actions should be. The inspectors identified that the action created to correct the issue was categorized as a procedure change request assignment rather than a corrective action assignment.

The inspectors also reviewed AR 3977720 which was written to document a discrepancy identified in the updated safety analysis report and the technical specification bases.

This issue was also classified as a level 3 condition. The inspectors identified an action item was created to correct this issue rather than a corrective action assignment.

Title 10 CFR Part 50, Appendix B, Criterion II, states, The applicant shall establish at the earliest practicable time, consistent with the schedule for accomplishing the activities, a Quality Assurance Program which complies with the requirements of this appendix. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout plant life in accordance with those policies, procedures, or instructions.

The licensees Quality Assurance Program was established in the QATR. Corrective action procedure PI-AA-125, Corrective Action Program Procedure, implemented a portion of the licensees CAP as described in the QATR and stated in part the following guidance would be used to determine if an action is a corrective action, Actions that correct a significance lever 1, 2, or 3 conditions.

Contrary to the above, the licensee failed to carry out the corrective action portion of their Quality Assurance Program in accordance with procedure PI-AA-125. Specifically, the licensee failed to assign corrective action assignments to significance level 3 conditions. The licensee captured this in AR 4056734, CA not created for Level 3 AR 2614744 and AR 4056951, AR 3977720 (SL3) does not Have CA as Required.

However, since the identified conditions were corrected under the created assignments the safety significance of this violation was determined to be minor.

Rigor Applied to the Generation of Effectiveness Reviews During the review of multiple causal evaluations, the inspectors identified a trend associated with the rigor applied to the generation of effectiveness reviews. Specifically, the inspectors identified that effectiveness reviews lacked rigor and failed to ensure that corrective actions were successful in addressing the identified issues. For example:

• Apparent cause evaluation 2605894, Procedure Adherence, was performed in response to a trend in failures to follow procedures. The identified cause was the inadequate awareness of topics contained in Information Use procedures. The corrective action developed to address the cause was to implement departmental Information Use procedure list/tool for reference during pre-job briefs. The effectiveness review criterion was that 80 percent of procedure adherence issue be self-identified. This criterion would not verify the licensee improved in the area of procedure adherence.

• Apparent cause evaluation 2659195, Missed Required Technical Specification Surveillance, was performed in response to the failure to perform a required technical specification surveillance that resulted in a licensee event report. The identified cause was that a senior reactor operator did not validate that a required surveillance was current. The corrective action developed to address the cause was to provide counsel to the individual responsible for this issue. The effectiveness review criterion was that the specific senior reactor operator involved in this event didnt cause a technical specification event that resulted in a licensee event report. This effectiveness review was narrowly focused and specific to one individual rather than ensuring the actions that led to this issues were addressed with all licensed operators and any failure to follow technical specification was considered in the review rather than only those that led to a licensee event report.

• Root cause 2689365, FLEX Inspection Performance, was performed in response to issues identified during the Temporary Instruction 191 inspection.

One of the issues was the timeliness and quality of responses to the NRC during the inspection. The identified cause was a lack of senior manager engagement to consistently drive resolution of NRC questions and issues. The corrective action developed to address the cause was to develop and implement a progressive accountability program for senior managers to ensure tracking resolution of NRC questions and issues. The effectiveness review criteria was to verify that progressive accountability actions had been implemented and documented for individuals missing due dates. The effectiveness review focused on the timeliness aspect of the issue but failed to address the quality of responses aspect.

• Corrective action program evaluation report 4009845, C1R17 [main steam isolation valve] MSIV [local leak rate test] LLRT Technical Specification [TS]

3.6.1.3 Limit Exceeded, was performed in response to the MSIV LLRT failure.

This test failure was a reportable condition because it represented a degraded barrier and a condition prohibited by TS. The identified cause was expected wear of MSIV internals. The corrective action developed to address the identified cause was repair/replace the main steam isolation valves that failed the LLRT.

The effectiveness review was to verify the work order for the as left test was acceptable. The corrective actions and effectiveness review did not address what led to the MSIVs failing the LLRT test.

The inspectors determined these examples did not constitute a violation of NRC requirements and communicated them to the licensee as observations.

.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 implementing operating experience program procedures, attended CAP meetings to observe the use of OE information, completed evaluations of OE issues and events, and selected monthly assessments of the OE composite performance indicators. The inspectors review was to determine whether the licensee was effectively integrating OE experience into the performance of daily activities, whether evaluations of issues were proper and conducted by qualified personnel, whether the licensees program was sufficient to prevent future occurrences of previous industry events, and whether the licensee effectively used the information in developing departmental assessments and facility audits. The inspectors also assessed if corrective actions, as a result of OE experience, were identified and effectively and timely implemented.

b. Assessment Based on the results of the inspection, the inspectors had no issues with the licensees review of operating experience.

i) Observations and Findings No observations or findings were identified.

.3 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 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 ARs as required by the CAP procedures.

i) Observations and Findings No observations or findings were identified.

.4 Assessment of Safety Conscious Work Environment

a. Inspection Scope

The inspectors interviewed selected Clinton Power Station personnel to determine if there were any indications that licensee personnel were reluctant to raise safety concerns to either their management or the NRC due to fear of retaliation. The inspectors also assessed the licensees safety conscious work environment through a review of Employee Concern Program (ECP) implementing procedures, discussions with an ECP manager, interviews with personnel from various departments, and reviews of ARs. The inspectors reviewed licensees self-assessments and assessments by external organizations of safety culture to determine if there were any organizational issues or trends that could impact the licensees safety performance.

b. Assessment The inspectors did not identify any issues that suggested conditions were not conducive to the establishment and existence of a safety conscious work environment at Clinton Power Station. Licensee staff members were aware of and generally familiar with the CAP and other station processes, including the ECP, through which concerns could be raised. In addition, a review of the types of issues in the ECP indicated that the licensee staff members were appropriately using the CAP and ECP to identify issues. The licensee staff also indicated that management had been supportive of the CAP by providing time and resources for employees to generate their IRs.

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

Since the beginning of 2015, various safety culture assessments had been performed by contractors, the licensees staff, and a nuclear plant owner/operators organization. The results indicated that there were no impediments to the identification of nuclear safety issues. The inspectors reviewed these surveys and did not identify any adverse trend.

i) Observations and Findings No observations or findings were identified.

4OA6 Management Meetings

.1 Exit Meeting

On September 29, 2017, the inspectors presented the inspection result to Mr. B. Kapellas 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

T. Stoner, Site Vice President

B. Kapellas, Plant Manager

K. Pointer, Regulatory Assurance

N. Santos, Regulatory Assurance

R. Bair, Work Management Director

J. Cunningham, Maintenance Director

T. Dean, Training Director

C. Dunn, Operations Director

M. Heger, Senior Manager Design Engineering

T. Krawyck, Engineering Director

W. Marsh, Organizational Effectiveness Manager

B. Golladay, CAP Manager

F. Paslaski, Radiation Protection Manager

D. Shelton, Regulatory Assurance Manager

R. Champley, Shift Operations Superintendent

D. Koons, Chemistry Manager

J. Wilson, Senior Manager Plant Engineering

A. Sigemund, Security Manager

U.S. Nuclear Regulatory Commission

L. Kozak, Acting Chief, Reactor Projects Branch 1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000461/2017007-01 NCV Failure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support Procedure (Section 4OA2.1.b.2)

Discussed

None.