IR 05000498/2016009

NRC Problem Identification and Resolution Inspection Report 05000498/2016009 and 05000499/2016009

ML16323A297
Person / Time
Site:	South Texas
Issue date:	11/18/2016
From:	Thomas Hipschman Division of Reactor Safety IV
To:	Koehl D South Texas
References
IR 2016009
Download: ML16323A297 (28)

Text

ber 18, 2016

SUBJECT:

SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION, UNITS 1 AND 2, NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000498/2016009 AND 05000499/2016009

Dear Mr. Koehl:

On August 25, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution biennial inspection at the South Texas Project Electric Generating Station, Units 1 and 2, and the NRC inspection team discussed the results of this inspection with you and other members of your staff. The inspection team documented the results of this inspection in the enclosed inspection report.

Based on the inspection sample, the inspection team determined that the South Texas Project Electric Generating Stations, Units 1 and 2, corrective action program and your staffs implementation of this program were adequate to support nuclear safety.

In reviewing your corrective action program, the team assessed how well your staff identified problems at a low threshold, your staffs implementation of the stations process for prioritizing and evaluating these problems, and the effectiveness of corrective actions taken by the station to resolve these problems. The team also evaluated other processes your staff used to identify issues for resolution. These included your use of audits and self-assessments to identify latent problems and your incorporation of lessons learned from industry operating experience into station programs, processes, and procedures. The team determined that your stations performance in each of these areas supported nuclear safety.

Finally, the team determined that, overall, your stations management maintains a safety-conscious work environment in which your employees are willing to raise nuclear safety concerns through at least one of the several means available. However, the team received a number of statements in the area of security, which were contrary to a safety-conscious work environment. The NRC is concerned that a continued failure to address the underlying cause of Enclosure 2 transmitted herewith contains SUNSI. When separated from Enclosure 2 of this letter, this transmittal document and Enclosure 1 are decontrolled. a chilling effect within the security department could result in an environment in which security officers do not raise safety concerns. This indicated an area where additional management attention is warranted. The team discussed these potential problem areas with the Site Vice President and Chief Nuclear Officer. The team was informed of the actions that the plant has taken to address these areas of concern. However, it is too soon to evaluate the efficacy of these actions. As a result, the NRC will continue to evaluate your efforts, and will return at a later date to verify their effectiveness.

NRC inspectors documented three findings of very low safety significance (Green) in this report.

These findings involved violations of NRC requirements. Two of these findings contain examples of security-related issues which were determined to be of very low security significance (i.e., Green as determined by the physical protection significance determination process). All deficiencies were promptly corrected or compensated for and the plant was in compliance with applicable physical protection and security requirements within the scope of this inspection before the inspectors left the site.

If you contest the violations or significance of these NCVs, 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 20555-001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the South Texas Project.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the South Texas Project.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 Public Inspections, Exemptions, Requests for Withholding, a copy of this letter and Enclosure 1 to this letter will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system, ADAMS. ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). However, because of the security-related concerns discussed in Enclosure 2 to this letter, and in accordance with 10 CFR 2.390, a copy of Enclosure 2 will not be available for public inspection.

In accordance with 10 CFR 2.390(b)(1)(ii), the NRC is waiving the affidavit requirements for your response, if any. This practice will ensure that your response will not be made available electronically for public inspection in the NRC Public Document Room or from the NRCs document system, ADAMS. If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21. Otherwise, mark your entire response Security-Related Information-Withhold Under 10 CFR 2.390 and follow the instructions for withholding in 10 CFR 2.390(b)(1).

Sincerely,

/RA/

Thomas R. Hipschman, Team Leader Inspection Program and Assessment Team Division of Reactor Safety Docket Nos. 05000498 and 05000499 License Nos. NPF-76 and NPF-80 Enclosures:

1. NRC Inspection Report 05000498/2016009 and 05000499/2016009 w/Attachment:

Supplemental Information 2. Security-Related Findings Identified During South Texas Project 2016 PI&R Inspection Report 05000498/2016009, 05000499/2016009 Electronic Distribution for South Texas Project U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Dockets: 05000498, 05000499 License: NPF-76, NPF-80 Report: 05000498/2016009, 05000499/2016009 Licensee: STP Nuclear Operating Company Facility: South Texas Project Location: FM 521 - 8 miles west of Wadsworth, Texas Dates: August 8 through August 25, 2016 Team Lead: R. Azua, Senior Reactor Inspector Inspectors: G. George, Senior Reactor Inspector H. Freeman, Senior Reactor Inspector N. Hernandez, Resident Inspector J. Rollins, Senior Physical Security Inspector Approved By: T. Hipschman, Team Leader Inspection Program and Assessment Team (IPAT)

Division of Reactor Safety Enclosure 1

SUMMARY

IR 05000498/2016009, 05000499/2016009; 08/08/2016 - 08/25/2016; South Texas Project

Electric Generating Station, Units 1 and 2; Problem Identification and Resolution (Biennial)

The inspection activities described in this report were performed between August 8 and August 25, 2016 by five inspectors from the NRCs Region IV office and the resident inspector at the South Texas Project. The report documents three findings of very low safety significance (Green). These findings involved violations of NRC requirements. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Assessment of Problem Identification and Resolution Based on its inspection sample, the team concluded that the licensee maintained a corrective action program in which individuals generally identified issues at an appropriately low threshold.

Once entered into the corrective action program, the licensee generally evaluated and addressed these issues appropriately and timely, commensurate with their safety significance.

The licensees corrective actions were generally effective, addressing the causes and extents of condition of problems. However, in the area of security, it was noted that a number of items were identified during this inspection where the adequacy or the timeliness of the corrective actions were found to be in question.

The licensee appropriately evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. The licensee incorporated industry and internal operating experience in its root cause and apparent cause evaluations.

The licensee performed effective and self-critical nuclear oversight audits and self-assessments.

The licensee maintained an effective process to ensure significant findings from these audits and self-assessments were addressed.

Overall, the licensee maintained a safety-conscious work environment in which personnel were willing to raise nuclear safety concerns without fear of retaliation. However, the team received a number of statements in the area of security, which were contrary to a safety-conscious work environment. The NRC is concerned that a continued failure to address the underlying cause of a chilling effect within the security department could result in an environment in which security officers do not raise safety concerns. This indicated an area where additional management attention is warranted.

Cornerstone: Barrier Integrity

Green.

The team is documenting a self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Procedure 0PGP04-ZE-0309, Design Change Package,

Revision 2. Specifically, on January 16, 1996, the licensee failed to identify and correct changes to drawing and breaker overhaul procedures, which resulted from Design Change Package 93-3409-4, Circuit Breaker Replacement-Load Center 1W, in accordance with Step 4.2.2.5 of the procedure. This resulted in electrical maintenance personnel using an incorrect drawing and procedure during a technical support center diesel generator supply breaker overhaul, on July 16, 2014, which left in place internal jumper cables that prevented the supply breaker from automatically closing.

The inspectors determined that the failure to follow Procedure 0PGP04-ZE-0309, Design Change Package, Revision 2, was a performance deficiency. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, the performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, and component, and barrier performance - containment isolation, attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding adversely affected the Technical Support Center diesel generators capability to supply ac power to the containment hatch hoists in order to close that hatch in the event of a loss of offsite power during outage conditions. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding could be evaluated using the significance determination process. In accordance with Table 3, SDP Appendix Router, the inspectors determined that the subject finding would be processed through Appendix G,

Shutdown Operations Significance Determination Process. dated May 9, 2014. In accordance with Appendix G, Exhibit 4, Barrier Integrity Screening Questions,

Question B.6, directs the inspectors to Appendix H if the finding degrades the ability to close or isolate the containment. In accordance with Inspection Manual Chapter 0609,

Appendix H, Containment Integrity Significance Determination Process, Section 4.1,

Types of Findings, the finding was a Type B finding because it had potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. Appendix H, Section 6.2, Approach for Assessing Type B Findings at Shutdown,

Step 2.2.A directs the user to Table 6.3 with a containment status of intact. Table 6.3,

Phase 1 Screening - Type B Findings at Shutdown, requires a Phase 2 evaluation because South Texas Project has a large, dry containment and the finding affected containment isolation. Appendix H, Table 6.4, Phase 2 Risk Significance - Type B Findings at Shutdown, provided an estimated risk significance of White because South Texas Project has a large, dry containment; the leakage from containment was greater than 100 percent volume/day; South Texas Project had in-depth shutdown mitigation capability; and for part of the exposure period, the plant was in Plant Operational State 2E.

In accordance with Appendix H, Section 2.0, Limitations and Precautions, a more detailed assessment was performed in a Significant Determination Process Phase 3 evaluation.

The analyst performed a detailed risk evaluation of the subject performance deficiency.

During the exposure period, from July 16, 2014, through October 29, 2015, the failure of the Technical Support Center diesel generator affected risk of the unit, while at power, because of the failure to provide power to the positive displacement pump for reactor coolant pump seal cooling following a postulated loss of all alternating current event. Additionally, the Technical Support Center diesel would not have fulfilled its function to provide backup power to close the containment hatch during the outage period from October 18, 2015, to October 29, 2015. These two impacts on plant risk were evaluated. Because the combined risk of the at-power and shutdown risk evaluations were lower than the threshold, the analyst determined that this non-cited violation was of very low safety significance (Green).

This finding has no cross-cutting aspect assigned because the root cause of this issue occurred in 1996 and is not reflective of current licensee performance.

REPORT DETAILS

4OA2 Problem Identification and Resolution

The team based the following conclusions on a sample of corrective action documents that were open during the assessment period, which ranged from November 15, 2014, to the end of the on-site portion of this inspection on August 25, 2016.

.1 Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The team reviewed approximately 130 condition reports, including associated root cause analyses and apparent cause evaluations, from approximately 30,000 that the licensee had initiated or closed between November 15, 2014, and August 25, 2016. The majority of these were lower-level condition reports that did not require cause evaluations. The inspection sample focused on higher-significance condition reports for which the licensee evaluated and took actions to address the cause of the condition. In performing its review, the team evaluated whether the licensee had properly identified, characterized, and entered issues into the corrective action program, and whether the licensee had appropriately evaluated and resolved the issues in accordance with established programs, processes, and procedures. The team also reviewed these programs, processes, and procedures to determine if any issues existed that may impair their effectiveness.

The team reviewed a sample of audit reports, self-assessments, and various other documents related to the licensees corrective action program. The team evaluated the licensees efforts in determining the scope of problems by reviewing selected self-assessment results, audits, action plans, and results from surveillance tests and preventive maintenance tasks. The team reviewed daily condition reports and attended the licensees Management Performance Improvement Committee/Condition Review Group, Condition Report Screening Committee, and the Engineering Corrective Action Review Board meetings to assess the reporting threshold and prioritization efforts, and to observe the corrective action programs interfaces with the operability assessment and work control processes. The teams review included an evaluation of whether the licensee considered the full extent of cause and extent of condition for problems, as well as a review of how the licensee assessed generic implications and previous occurrences of issues. The team assessed the timeliness and effectiveness of corrective actions, completed or planned, and looked for additional examples of problems similar to those the licensee had previously addressed. The team conducted interviews with plant personnel to identify other processes that may exist where problems may be identified, and addressed, outside the corrective action program.

The team reviewed corrective action documents that addressed past NRC-identified violations to evaluate whether corrective actions addressed the issues described in the inspection reports. The team reviewed a sample of corrective actions closed to other corrective action documents to ensure that the ultimate corrective actions remained appropriate and timely. The team reviewed a sample of condition reports where the licensee had changed the significance level after initial classification to determine whether the level changes were in accordance with station procedure and that the conditions were appropriately addressed. No issues were identified.

The team considered risk insights from both the NRCs and the South Texas Projects risk models to focus the sample selection and plant tours on risk-significant systems and components. For this inspection, the team did not select a system for a focused review within the corrective action program, but selected a series of sequencer and power supply cards for safety-related equipment, including the qualified display processing system. These components were selected for review due to NRC concerns regarding the adequacy, effectiveness, and timeliness of corrective actions taken following repeated failures of these components over the last couple of years. The team performed an in-depth review of the licensees actions.

b. Assessments 1. Effectiveness of Problem Identification During the 24-month inspection period, licensee staff generated approximately 30,000 condition reports. The team determined that most conditions that required generation of a condition report by South Texas Project Procedure 0PGP03-ZX-0002, Condition Reporting Process, had been appropriately entered into the corrective action program.

Overall, the team concluded that the licensee generally maintained a low threshold for the formal identification of problems and entry into the corrective action program for evaluation. Licensee personnel initiated an average of 1,400 condition reports per month during the inspection period. Most of the personnel interviewed by the team understood the requirements for condition report initiation; most expressed a willingness to enter newly identified issues into the corrective action program at a very low threshold.

2. Effectiveness of Prioritization and Evaluation of Issues The sample of condition reports reviewed by the team focused primarily on issues screened by the licensee as having higher-level significance, including those that received cause evaluations, those classified as significant conditions adverse to quality, and those that required engineering evaluations. The team also reviewed a number of condition reports that included or should have included immediate operability determinations to assess the quality, timeliness, and prioritization of these determinations.

Prioritization and evaluation of issues in general was found to be good, especially with the more safety-significant condition reports. However, some examples were identified where the identified root cause or apparent cause were more in line with contributing causes, and the contributing causes listed should have been the actual root or apparent causes. Even though the corrective actions were adequate to address the concerns identified, the importance of determining the correct root or apparent cause is in the evaluation for the extent of cause. The licensee understood the teams concern with this issue and placed it in their corrective action program for review.

The team identified two examples of inadequate apparent cause evaluations in response to NRC-identified Green non-cited violation in the area of security. One of these examples was determined to be an NRC-identified Green non-cited violation that is documented in Enclosure 2. It must be noted that the team observed a lack of rigor and management oversight in the review and approval of the apparent cause and corrective actions for these examples.

Overall, the team determined that the licensees process for screening and prioritizing issues, that had been entered into the corrective action program, supported nuclear safety. The licensees operability determinations were consistent, accurately documented, and completed in accordance with procedures.

3. Effectiveness of Corrective Actions

Overall, the team concluded that the licensee generally identified effective corrective actions for the problems evaluated in the corrective action program. The licensee generally implemented these corrective actions in a timely manner, commensurate with their safety significance, and reviewed the effectiveness of the corrective actions appropriately. However, in the area of security, it was noted that a number of items were identified during this inspection where the adequacy or the timeliness of the corrective actions were found to be in question.

For this inspection, the team did not select a system for a focused review within the corrective action program, but selected a series of sequencer and power supply cards for safety-related equipment, including the qualified display processing system.

These components were selected due to NRC concerns regarding the adequacy, effectiveness and timeliness of corrective actions taken, following repeated failures of these components over the last couple of years. The team performed an in-depth review of the licensees actions. Based on this review, it was apparent that the licensee staff recognized that the South Texas Projects actions were inconsistent with industry practices in this area. With the help and support from industry organizations, the licensee set aside funding and started developing plans to replace these components. Due to the fact that this planned replacement schedule for these components was not fully developed by the end of this inspection, and because implementation will occur over a long period, the NRC will continue to monitor the licensees progress in this area.

.2 Assessment of the Use of Operating Experience

a. Inspection Scope

The team examined the licensees program for reviewing industry operating experience, including reviewing the governing procedures. The team reviewed a sample of industry operating experience communications and the associated site evaluations to assess whether the licensee had appropriately assessed the communications for relevance to the facility. The team also reviewed assigned actions to determine whether they were appropriate.

b. Assessment Overall, the team determined that the licensee appropriately evaluated industry operating experience for its relevance to the facility. Operating experience information was incorporated into plant procedures and processes as appropriate.

The team further determined that the licensee appropriately evaluated industry operating experience when performing root cause analysis and apparent cause evaluations. The licensee appropriately incorporated both internal and external operating experience into lessons learned for training and pre-job briefs.

.3 Assessment of Self-Assessments and Audits

a. Inspection Scope

The team reviewed a sample of licensee self-assessments and audits to assess whether the licensee was regularly identifying performance trends and effectively addressing them. The team also reviewed audit reports to assess the effectiveness of assessments in specific areas. The specific self-assessment documents and audits reviewed are listed in Attachment 1.

b. Assessment Overall, the team concluded that the licensee had an adequate departmental self-assessment and audit process. However, the team noted that previous NRC inspections had identified instances where audits and self-assessments of the South Texas Project programs had not been performed for a long time or not at all.

Some examples include the licensees 10 CFR 50.59 program and scaffolding program, the latter of which resulted in a recent NRC-identified violation because licensee departments were unaware of changes in industry requirements and standards.

.4 Assessment of Safety-Conscious Work Environment

a. Inspection Scope

The team interviewed approximately 60 individuals in six focus groups. The purpose of these interviews was

(1) to evaluate the willingness of licensee staff to raise nuclear safety issues, either by initiating a condition report or by another method,

(2) to evaluate the perceived effectiveness of the corrective action program at resolving identified problems, and

(3) to evaluate the licensees safety-conscious work environment. The focus group participants included personnel from the mechanical and electrical maintenance, instrumentation and controls, engineering (both plant staff and contract),and security departments. At the teams request, the licensees regulatory affairs staff selected the participants blindly from these work groups, based partially on availability.

To supplement these focus group discussions, the team interviewed the employee concerns program manager to assess her perception of the site employees willingness to raise nuclear safety concerns. The team reviewed the employee concerns program case log and select case files.

b. Assessment 1. Willingness to Raise Nuclear Safety Issues Based upon the interviews, all individuals indicated that they would raise nuclear safety concerns. All indicated that management was generally receptive to nuclear safety concerns and was willing to address them promptly. Individuals stated that management encourages them to raise and document safety concerns in the corrective action program and have publicly recognized some of the employees that had done so. Most expressed positive experiences after raising issues to their supervisors. All interviewees stated that they were able to and had initiated condition reports using the WIZARD electronic corrective action program system. They also stated that condition report initiation was straightforward and fairly simple.

However, even though it appeared that licensee management had established and promoted a safety-conscious work environment where individuals felt free to raise safety concerns without fear of retaliation, the team received a number of statements in the area of security, which were contrary to a positive environment for raising safety concerns. The main contributing factor to this environment appears to be due to the level of staffing. Continued understaffing of the security department has led to the use of mandatory overtime. The NRC is concerned that a continued failure to address the underlying cause of a chilling effect within the security department could result in an environment in which security officers do not raise safety concerns. This indicated an area where additional management attention continues to be warranted.

The team questioned focus group participants whether they were able to submit a condition report anonymously. Most individuals were aware that they could submit condition reports anonymously and were knowledgeable of the process.

2. Employee Concerns Program All interviewees were aware of the employee concerns program. Most explained that they had heard about the program through various means, such as posters, training, presentations, and discussion by supervisors or management at meetings. All interviewees stated that they would use Employee Concerns if they felt it was necessary. All expressed confidence that their confidentiality would be maintained if they brought issues to Employee Concerns.

3. Preventing or Mitigating Perceptions of Retaliation When asked if there have been any instances where individuals experienced retaliation or other negative reaction for raising issues, all individuals interviewed stated that they had neither experienced nor heard of an instance of retaliation, harassment, intimidation or discrimination at the site. The team determined that processes in place to mitigate these issues were being successfully implemented.

.5 Findings

a. Failure to Update Procedure Reference Leads to Non-functional Unit 1 Technical Support Center Diesel Generator

Introduction.

The inspectors are documenting a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Procedure 0PGP04-ZE-0309, Design Change Package, Revision 2. Specifically, on January 16, 1996, licensee personnel failed to identify changes that were needed to be made to a drawing and breaker overhaul procedure, in accordance with Step 4.2.2.5 of Procedure 0PGP04-ZE-0309, following the implementation of Design Change Package 93-3409-4, Circuit Breaker Replacement-Load Center 1W. This resulted in electrical maintenance personnel using an incorrect drawing and procedure during a technical support center diesel generator supply breaker overhaul, on July 16, 2014, which left in place internal jumper cables that prevented the supply breaker from automatically closing.

Description.

On October 29, 2015, the licensee performed a preventive maintenance test on the automatic start and load function of the Unit 1 technical support center diesel generator. The test simulated a loss of offsite power to the 480V load center 1W and automatic starting and loading of the technical support center diesel generator to the load center 1W. During this test, the technical support center diesel started; however, its supply breaker failed to close and load 480V load center 1W.

Following the failure, the licensee initiated Condition Report 15-24053 which investigated the cause of the supply breaker failure. The investigation determined that the supply breaker would not close because the supply breaker was installed with jumper cables installed in its control block. One of these jumper cables prevented the breaker from automatically closing on demand. Further investigation identified that these jumper cables were left in place during the supply breaker overhaul prior to installation of the breaker. The investigation determined that the supply breaker was installed in 480V load center 1W on July 17, 2014. Therefore, the technical support center diesel would not have fulfilled its function to provide backup power to the positive displacement pump for reactor coolant pump seal cooling during a loss of all alternating current, from July 17, 2014, to October 17, 2015. This loss of seal cooling could lead to a reactor cooling pump seal failure during a loss of all alternating current event, which would eventually lead to an increase in reactor coolant system leakage past the seals (i.e., intersystem loss of cooling accident). Additionally, the technical support center diesel, in its incorrect configuration, would not have fulfilled its function to provide backup power to close the containment hatch during the outage period from October 18, 2015, to October 29, 2015.

The licensee interviewed the electrical maintenance staff to determine why the electrical jumper cables were installed on the control block of the supply breaker. Electrical maintenance personnel explained to the investigators that they used Procedure 0PMP05-NA-0017, 480 Volt Type K Breaker Overhaul/Lubrication (Generic) for LC 1W/5B, work instructions PMI-EM-NA-0003, Gould 480 Volt Type K Breaker Test, and the procedures referenced drawings to disassemble and reassemble the supply breaker. The referenced drawing indicated that the jumper cables were integral to the control block of the circuit breaker. Therefore, the electrical maintenance staff reinstalled the jumper cables as indicated on the referenced drawing. Further investigation identified that the overhaul procedure referenced an incorrect drawing. The licensees investigation identified that the drawing, referenced in the procedure, was not updated to reflect a January 16, 1996, change to the supply breaker design and its cubicle.

On January 16, 1996, the licensee issued Design Change Package 93-3409-4, Circuit Breaker Replacement - Load Center 1W. The purpose of the change was to replace the normal supply breaker from an obsolete, less reliable, breaker to a more reliable ABB K-line breaker. The change also authorized retrofitting to 480V load center 1W to accept the K-line breaker. When these changes were instituted to the load center 1W and the K-line breakers, the Procedure 0PMP05-NA-0017, 480 Volt Type K Breaker Overhaul/Lubrication (Generic) for LC 1W/5B referenced the generic ABB K-line breaker control wiring diagram which included illustrations for K-line breakers with and without jumper cables installed. The licensees investigation determined that this procedure should have been updated to reference the retrofitted cubicle drawing which identified the correct configuration on the control block of the supply breaker without installed jumper cables.

Procedure 0PGP04-ZE-0309, Design Change Package, Revision 2, dated October 2, 1995, Step 4.2.2.5, states, Identify changes to engineering documents which will result from the change.

The licensees investigation determined that the 1996 design change package failed to identify changes to the overhaul procedures and drawings which led to electrical maintenance staff reinstalling jumper cables into the supply breakers control block, rendering the technical support diesel non-functional from July 17, 2014, to October 29, 2015.

Analysis.

The inspectors determined that the failure to follow Procedure 0PGP04-ZE-0309, Design Change Package, Revision 2 was a performance deficiency. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, the finding was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, and component, and barrier performance - containment isolation, attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency adversely affected the Technical Support Center diesel generators capability to supply ac power to the containment hatch hoists in order to close that hatch in the event of a loss of offsite power during outage conditions. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding could be evaluated using the significance determination process. In accordance with Table 3, SDP Appendix Router, the inspectors determined that the subject finding would be processed through Appendix G, Shutdown Operations Significance Determination Process. dated May 9, 2014. In accordance with Appendix G, Exhibit 4, Barrier Integrity Screening Questions, Question B.6, directs the inspectors to Appendix H if the finding degrades the ability to close or isolate the containment. In accordance with Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, Section 4.1, Types of Findings, the issue was a Type B finding because it had potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. Appendix H, Section 6.2, Approach for Assessing Type B Findings at Shutdown, Step 2.2.A, directs the user to Table 6.3 with a containment status of intact. Table 6.3, Phase 1 Screening - Type B Findings at Shutdown, required a Phase 2 evaluation because South Texas Project has a large, dry containment and the finding affected containment isolation. Appendix H, Table 6.4, Phase 2 Risk Significance - Type B Findings at Shutdown, provided an estimated risk significance of White because South Texas Project has a large, dry containment; the leakage from containment was greater than 100 percent volume/day; South Texas Project had in-depth shutdown mitigation capability; and for part of the exposure period, the plant was in Plant Operational State 2E. In accordance with Appendix H, Section 2.0, Limitations and Precautions, a more detailed assessment was performed in a Significant Determination Process Phase 3 evaluation.

The analyst performed a detailed risk evaluation of the subject performance deficiency.

During the exposure period, from July 16, 2014, through October 29, 2015, the failure of the Technical Support Center diesel generator affected risk of the unit, while at power, because of the failure to provide power to the positive displacement pump for reactor coolant pump seal cooling following a postulated loss of all alternating current event.

Additionally, the technical support center diesel would not have fulfilled its function to provide backup power to close the containment hatch during the outage period from October 18, 2015, to October 29, 2015. These two impacts on plant risk were evaluated as follows:

At Power Risk: Utilizing the Events and Conditions Assessment module of the site-specific Standardized Plant Analysis Risk Model for South Texas Project, Units 1 and 2, Version 8.12, the change in core damage frequency was quantified to be 1.43E-7. There were two dominant core damage sequences for this analysis as follows:

1. Loss of offsite power; failure of reactor coolant pump seal cooling; loss of a single stage of reactor coolant pump seals; failure of the residual heat removal system; failure of low pressure injection; and failure of containment fan coolers.

2. Loss of offsite power; failure of reactor coolant pump seal cooling; loss of a single stage of reactor coolant pump seals; failure to recover offsite power in two hours; and failure of containment fan coolers.

Based on inspector input, the analyst determined that the licensee could have recovered the failed condition by replacing the breaker with one of several spares that were available at the time. As a bounding assumption, the analyst considered that the breaker could be replaced prior to seal damage at least 50 percent of the time.

Therefore, the bounding change in core damage frequency was 7.2E-8.

Shutdown Risk: The analyst reviewed the subject finding related to the potential to close the containment equipment hatch with the reactor shut down, following a postulated loss of offsite power. A similar failure of the Technical Support Center diesel generator during the same outage period as discussed above was reviewed and closed as a Green finding in NRC Inspection Report Number 05000498/2016003 and 05000499/2016003. The analyst determined that the shutdown risk associated with the current performance deficiency was fully bounded by the evaluation performed in the referenced report.

Because the combined risk of the at-power and shutdown risk evaluations were lower than the threshold, the analyst determined that this finding was of very low safety significance (Green).

This finding has no cross-cutting aspect assigned because the root cause of this issue occurred in 1996 and is not reflective of current license performance.

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, states, in part, that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions. Contrary to the above, on January 16, 1996, licensee personnel failed to identify changes that were needed to be made to a drawing, and breaker overhaul procedure, in accordance with Step 4.2.2.5 of Procedure 0PGP04-ZE-0309, following the implementation of Design Change Package 93-3409-4, Circuit Breaker Replacement-Load Center 1W. This resulted in electrical maintenance personnel using an incorrect drawing and procedure during a technical support center diesel generator supply breaker overhaul, on July 16, 2014, which left in place internal jumper cables that prevented the supply breaker from automatically closing. The licensee documented this issue into the corrective action program as Condition Report 2015-24053 and has taken corrective actions to preclude repetition. This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy: NCV 05000498/2016009-01; 05000499/2016009-01, Failure to Update Procedure Reference Leads to Non-functional Unit 1 Technical Support Center Diesel Generator.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On August 25, 2016, the inspectors presented the inspection results to Mr. Dennis L. Koehl, President and Chief Executive Officer, STP Nuclear Operating Company, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On October 6, 2016, the inspectors re-exited with Mr. Jenewein, Manager, Performance Improvement. The purpose of this re-exit was to inform the licensee of the final risk significance of the inspection findings.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Koehl, President and Chief Executive Officer

J. Atkins, Manager, Systems Engineering

R. Barr, Supervisor, Performance Improvement

M. Berg, General Manager, Engineering (Acting)

J. Berrio, Manager, Operations - Production Support and Programs

A. Capristo, Executive Vice President and Chief Administrative Officer

J. Connolly, Site Vice President

L. Huerta, Manager, Security (Acting)

B. Jenewein, Manager, Performance Improvement

J. Milliff, Manager, Operations, Training Support

G. Powell, Chief Nuclear Officer

M. Reddix, Manager, Security Projects

D. Rencurrel, Senior Vice President, Operations

B. Scarborough, Manager, Quality Assurance

R. Stastny, Manager, Maintenance

R. Savage, Engineer Specialist, Licensing Consult

L. Sterling, Supervisor, Licensing

NRC Personnel

T. Hipschman, Team Leader, IPAT

S. Morris, Deputy Regional Administrator, Region IV

F. Sanchez, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000490; NCV Failure to Update Procedure Reference Leads to Non-functional 499/2016009-01 Unit 1 Technical Support Center Diesel Generator (Section 4OA2.5)