IR 05000456/2014007

IR 050004562014007; 050004572014007, July 21, 2014 Through August 1, 2014, Braidwood, Units 1 and 2, Biennial Problem Identification and Resolution

ML14240A008
Person / Time
Site:	Braidwood
Issue date:	08/27/2014
From:	Eric Duncan Region 3 Branch 3
To:	Pacilio M Exelon Generation Co, Exelon Nuclear
References
IR-2014-007
Download: ML14240A008 (20)

Text

UNITED STATES ugust 27, 2014

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2, BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION (PI&R) INSPECTION REPORT 05000456/2014007; 05000457/2014007

Dear Mr. Pacilio:

On August 1, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution biennial inspection at your Braidwood Station, Units 1 and 2. The NRC inspection team discussed the inspection results at an interim exit meeting on August 1, 2014, with Ms. M. Marchionda and other members of your staff; and at the final exit meeting on August 7, 2014, with Mr. W. Spahr, and other members of your staff. The inspection team documented the results of this inspection in the enclosed inspection report.

Based on the inspection samples selected for review, the inspection team determined that your Braidwood Station staffs implementation of the corrective action program supported nuclear safety. In reviewing the corrective action program, the team assessed the Braidwood Station staffs ability to identify problems at a low threshold; to implement the stations process for prioritizing and evaluating these problems, and to implement effective corrective actions to resolve identified problems. In each of these areas, the team determined that performance was adequate to support nuclear safety.

The team also evaluated other processes your Braidwood Station staff used to identify issues for resolution. These included the use of audits and self-assessments to identify latent problems and incorporation of lessons learned from industry operating experience into station programs, processes, and procedures. The team determined that performance in each of these areas also supported nuclear safety.

Finally, based on the results of the interviews conducted, the inspection team did not identify any impediment to the establishment of a safety conscious work environment at Braidwood Station. Based on the inspection teams observations, employees expressed that they felt free to raise concerns related to nuclear safety without fear of retaliation.

The NRC inspectors did not identify any findings or violations of more than minor significance. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Eric R. Duncan, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77

Enclosure:

IR 05000456/2014007; 05000457/2014007 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 50-456; 50-457 License Nos: NPF-72; NPF-77 Report No: 05000456/2014007; 05000457/2014007 Licensee: Exelon Generation Company, LLC Facility: Braidwood Station, Units 1 and 2 Location: Braceville, IL Dates: July 21 through August 1, 2014 Inspectors: J. Lennartz, Project Engineer J. Benjamin, Senior Resident Inspector C. Brown, Reactor Engineer R. Winter, Reactor Engineer M. Perry, Resident Inspector Illinois Emergency Management Agency Approved by: E. Duncan, Chief Branch 3 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000456/2014007; 05000457/2014007; 07/21/2014-08/01/2014; Braidwood

Station, Units 1 and 2; Biennial Problem Identification and Resolution (PI&R) Inspection.

This inspection was performed by three NRC region-based inspectors, the Braidwood Senior Resident Inspector, and the Braidwood Illinois Emergency Management Agency (IEMA)

Resident Inspector. No findings of significance or violations of NRC requirements were identified during this inspection. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 2006.

Problem Identification and Resolution On the basis of the samples selected for review, the inspection team concluded that the corrective action program (CAP) at Braidwood Station was generally being implemented in an effective manner. Licensee personnel had a low threshold for identifying problems and entering them into the CAP. Issues entered into the CAP were found to be screened and prioritized in a timely manner using established criteria; were found to be properly evaluated commensurate with their safety significance; and corrective actions were found to be generally implemented in a timely manner, commensurate with their safety significance. The inspection team noted that the Braidwood Station staff reviewed operating experience (OE) for applicability to station activities and that, in general, OE was effectively utilized. Audits and self-assessments were generally thorough and intrusive and performed at an appropriate level to identify deficiencies.

Based on the interviews conducted during the inspection, the inspectors did not identify any impediment to the establishment of a safety conscious work environment (SCWE) at Braidwood Station. Workers at the site expressed freedom to raise concerns related to nuclear safety without fear of retaliation, and workers were aware of and generally familiar with the CAP process and other processes, including the Employee Concerns Program (ECP), which could be used to raise safety concerns.

NRC-Identified

and Self-Revealed Findings None.

Licensee-Identified Violations

None.

REPORT DETAILS

4OA2 Problem Identification and Resolution

This inspection constituted one biennial sample of PI&R as defined in Inspection Procedure 71152, Problem Identification and Resolution. Documents reviewed are listed in the Attachment to this report.

.1 Corrective Action Program Effectiveness

a. Inspection Scope

The inspector reviewed the licensees CAP implementing procedures and interviewed licensee personnel to assess CAP implementation. The inspectors also observed and assessed the effectiveness of CAP-related meetings, which included the station ownership committee meeting and the management screening committee meeting.

The inspectors reviewed risk-significant and safety-significant issue reports (IRs) entered into the licensees CAP since the last NRC PI&R inspection in June 2012. The IRs reviewed included a sampling of issues identified through NRC generic communications, licensee audits and self-assessments, operating experience reports, NRC documented findings, and licensee-identified violations. The inspectors also reviewed IRs for selected systems, structures, and components or functions classified as (a)(1) status in accordance with the maintenance rule (10 CFR 50.65). The IRs selected ensured an adequate review across NRC cornerstones and included completed root cause and apparent cause evaluations.

Based on input from the resident staff, the inspectors selected the auxiliary feedwater and essential service water systems for in-depth reviews of system performance over the past 5 years. These in-depth reviews were conducted to determine whether the licensee was properly evaluating and taking appropriate corrective actions for the problems documented in IRs related to these systems.

During these reviews, the inspectors determined whether licensee actions were in compliance with the CAP implementing procedures and 10 CFR Part 50, Appendix B requirements. Specifically, the inspectors assessed whether licensee personnel identified issues at a proper threshold, whether identified issues were being entered into the CAP in a timely manner with the appropriate significance characterization, and whether identified issues were appropriately prioritized for resolution. The inspectors determined whether licensee personnel assigned the appropriate evaluation method to ensure that the correct root, apparent, and contributing causes were determined; verified that issues were appropriately evaluated with respect to the maintenance rule and operability; and assessed the evaluations scope and depth. The inspectors also evaluated the timeliness and effectiveness of corrective actions. For significant conditions adverse to quality, the inspectors assessed the corrective actions to prevent recurrence. For less significant issues, the inspectors verified that the corrective actions were implemented in a timely manner commensurate with their safety significance.

b. Assessment

(1) Effectiveness of Problem Identification Based on the results of the inspection, the inspectors concluded that problem identification was generally effective. Based on the information reviewed, the inspectors determined that Braidwood Station personnel had a low threshold for initiating IRs; station personnel appropriately screened issues from both the NRC and industry operating experience at an appropriate level and entered them into the CAP when applicable; and identified problems were generally entered into the CAP in a complete, accurate, and timely manner.

Findings No findings were identified.

(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 adequately considered extent of condition, generic implications, and previous occurrences.

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, Braidwood 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.

Findings No findings were identified.

(3) Effectiveness of Corrective Actions Based on the results of the inspection, overall, the corrective actions reviewed were found to be appropriately focused to correct the identified problem and were implemented in a timely manner commensurate with the issues safety significance.

Problems identified through root or apparent cause evaluations were resolved in accordance with the CAP procedural and regulatory requirements. Corrective actions intended to prevent recurrence were generally comprehensive, thorough, and timely.

The corrective actions associated with selected NRC documented findings and violations, as well as licensee-identified violations, were generally appropriate to correct the problem and were implemented in a timely manner. However, the inspectors identified one unresolved item, as discussed below, concerning the corrective actions to address a degraded Unit 2 reactor coolant pump (RCP) thermal barrier.

Findings

Introduction:

The inspectors identified an Unresolved Item (URI) regarding the incorporation of Westinghouse Nuclear Safety Advisory Letter (NSAL) 99-05, Reactor Coolant Pump Operation During Loss of Seal Injection," into the current licensing basis (CLB). Specifically, Westinghouse issued NSAL 99-05 to inform plants with specific model Westinghouse RCPs that during a postulated loss of RCP seal injection, RCP seal package and/or lower bearing temperatures may rise more rapidly than previously assumed in the original design. The previous analysis assumed that an RCP could operate without seal injection for a relatively long period of time (e.g., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or longer)because the thermal barrier heat exchanger (TBHX) could provide adequate seal cooling if seal injection was lost. Westinghouse NSAL 99-05 notified licensees that this assumption may no longer be correct if RCP seal leakoff was less than 2.5 gallons per minute (gpm).

Description:

In 1999, Westinghouse notified Braidwood Station through NSAL 99-05, Reactor Coolant Pump Operation During Loss of Seal Injection, of a potential safety issue that had not been previously identified as part of the original design. Specifically, the NSAL described concerns that during a postulated loss of seal injection (LOSI)event, the RCP seal package and/or lower bearing temperatures may rise more rapidly than previously assumed in the original design. The original design considered the thermal barrier and associated heat exchanger as a fully functional backup to seal injection during a LOSI event. Consequently, following a loss of seal injection, the RCP thermal barrier and associated TBHX would cool the reactor coolant fluid that would flow up the RCP shaft and through the seals to maintain the lower bearing and the RCP seal temperatures within their normal temperature range for an extended period of time (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The NSAL notified the licensee of a non-conformance to this original design.

Specifically, during a postulated LOSI for RCPs with less than 2.5 gpm seal leakoff rate, it was determined that the RCP TBHX system would not be capable of maintaining the RCP seals within their nominal temperature range for the previously assumed extended period of time. Instead, the NSAL concluded that the RCP seal temperatures would rise above acceptable operating temperatures within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if the RCP had less than a 2.5 gpm seal leakoff rate. With low seal leakoff rates, even though the reactor coolant system (RCS) water is cooled in the TBHX, the water will be heated after it flows through the TBHX. Westinghouse postulated that there may not be enough mass flow up along the shaft to absorb the heat transferred from the RCS, potentially raising the seal and bearing temperatures above their operating limits. In addition to NSAL 99-05 applicability to all Braidwood RCPs, the Braidwood Unit 2, 2B RCP TBHX system was particularly adversely affected since the 2B RCP thermal barrier had also been identified to have degraded insulating properties. The licensee estimated that for a bounding set of plant conditions, Operations personnel would have approximately 27 minutes to respond after losing seal injection before the 2B RCP trip criteria would be reached.

This condition had existed since 1999 and as of the end of the inspection the licensee had not taken any action to correct the degraded thermal barrier.

The inspectors discussed the potential consequences that a LOSI event may have on the 2B RCP after reaching the RCP trip criteria. One consequence would be the need for operators to insert a manual reactor trip prior to the manual 2B RCP pump trip in accordance with plant procedures, training, and associated expectations. Also, the licensee informed the inspectors that the event was bounded by the total loss of RCP seal cooling analysis that concluded 21 gpm of controlled leakage could occur.

The licensee entered the original NSAL 99-05 operating experience issue into their CAP in 1999 and determined that the Updated Final Safety Analysis Report (UFSAR) was not required to be updated. Additionally, during the inspection the licensee affirmed that the decision to not update the UFSAR was correct because the UFSAR was still correct and the level of detail in NSAL 99-05 was not required to be discussed in the UFSAR.

Although a loss of RCP seal injection and/or RCP TBHX function was discussed in numerous instances in the UFSAR, the following excerpt in the UFSAR generally described the CLB discussed in other UFSAR sections. (REF: original Safety Analysis Report (SAR) and UFSAR Section 5.4.1.2).

High-Pressure seal injection water is introduced through a connection on the thermal barrier flange. A portion of this water flows through the radial bearing and the seals; the remainder flows down the shaft through the thermal barrier where it acts as a buffer to prevent system water from entering the radial bearing and seal section of the unit. The thermal barrier heat exchanger provides a means of cooling system water to an acceptable level in the event seal injection flow is lost.

The licensee informed the inspectors that the UFSAR was still correct because the RCP would be tripped upon reaching the pre-established temperature limits upon a LOSI event and that the worst case leakage through each RCP would be 21 gpm for a total of 84 gpm from the four RCPs. This amount of controlled leakage was within the capacity of a single high head injection charging pump. However, the inspectors questioned this response because the conditions described in NSAL 99-05 had not been identified during the time frame that the original SAR was approved. Consequently the inspectors questioned whether or not the assumptions in the original SAR that established an acceptable level of safety and the licensing basis for the RCP TBHX system were adversely affected.

Additionally, Westinghouse NSAL 99-05 recommended that all plants review their SAR relative to the loss of seal injection and ensure that the SAR was consistent with the NSAL, indicating a limited time frame for operation without seal injection (Ref: Westinghouse NSAL 99-05, Recommended Actions #1).

At the conclusion of the inspection, a detailed review of the CLB was in progress. This URI will remain open until that review is completed and the inspectors determine whether NSAL 99-05 was adequately incorporated into the Braidwood CLB and whether the licensee should have implemented additional corrective actions to address the degraded 2B RCP thermal barrier.

(URI 05000456/2014007-01; 05000457/2014007-01; Incorporation of Westinghouse NSAL 99-05,Reactor Coolant Pump Operation During Loss of Seal Injection, Into the Current Licensing Basis and Corrective Actions For 2B RCP Degraded Thermal Barrier)

.2 Use of Operating Experience

a. Inspection Scope

The inspectors reviewed the licensees OE program implementation. Specifically, the inspectors reviewed OE program implementing procedures, attended CAP meetings to observe the screening of OE information, reviewed completed evaluations of OE issues and events, and reviewed selected monthly assessments of the OE composite performance indicators. The inspectors performed this review to determine whether the licensee was effectively integrating OE into the performance of daily activities, whether evaluations of issues were proper and conducted by qualified personnel, whether the licensees program was sufficient to prevent future occurrences of previous industry events, and to determine whether NRC-identified and industry-identified OE were entered into the licensees OE system as prescribed by procedure and were properly evaluated for significance. The inspectors also assessed if corrective actions resulting from OE were identified and implemented in an effective and timely manner.

b. Assessment Based on the results of the inspection, the inspectors did not identify any issues of concern regarding Braidwood Stations use of OE and concluded that, in general, OE was effectively utilized at the station. Industry OE was effectively disseminated across the various plant departments and the inspectors did not identify any issues while reviewing OE evaluations. 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 Braidwood Station was thoroughly evaluated and actions were implemented in a timely manner to address any issues that resulted from the evaluations.

Findings No findings were identified.

.3 Self-Assessments and Audits

a. Inspection Scope

The inspectors reviewed selected self-assessments, including adverse trend assessments and performance assurance audits to assess the licensee staffs ability to identify and enter issues into the CAP with the appropriate characterization, to prioritize and evaluate issues commensurate with their safety significance, and to implement effective corrective actions in a timely manner. The inspectors also evaluated whether self-assessments and audits were effectively managed and adequately covered the subject areas and verified that assessments were conducted in accordance with plant procedures, including procedure LS-AA-126-1005, Self-Assessment Program.

b. Assessment Based on the results of the inspection, the inspectors did not identify any issues of concern regarding Braidwood Station staffs ability to conduct self-assessments and audits. Assessments were conducted in accordance with plant procedures, were generally thorough and intrusive, adequately covered the subject area, and were effective at identifying issues and enhancement opportunities at an appropriate threshold. Identified issues were entered into the CAP with an appropriate significance characterization and corrective actions were completed and/or scheduled to be completed in a timely manner commensurate with their safety significance.

Findings No findings were identified.

.4 Safety Conscious Work Environment

a. Inspection Scope

The inspectors assessed the licensees SCWE by reviewing the licensees ECP implementing procedures; through discussions with the ECP coordinators; by reviewing IRs; and by conducting interviews with licensee personnel from various departments on site including Operations, Maintenance, Security, Radiation Protection and Chemistry.

The inspectors also reviewed the results from a 2013 mid-cycle safety culture survey that was conducted in November 2013. The review was performed to ensure there was a free flow of information and to determine if individuals were willing to raise nuclear safety concerns without fear of retaliation.

b. Assessment Based on the results of the inspection, the inspectors did not identify any issues that suggested conditions were not conducive to the establishment and existence of a SCWE at Braidwood Station.

Information obtained during the interviews indicated that an environment was established where Braidwood Station employees felt free to raise nuclear safety issues without fear of retaliation; were aware of and generally familiar with the CAP and other processes, including the ECP and the NRC, through which concerns could be raised; and safety significant issues could be freely communicated to supervision.

Findings No findings were identified.

4OA5 Other Activities

(Closed) Notice of Violation 05000456/2012004-03; 05000457/2012004-03 Failure to Analyze Recycle Holdup Tank Inlet Piping Loads A non-cited violation (NCV) of Title 10 CFR 50, Appendix B, Criterion III, Design Control, was issued in February 2009, when licensee personnel failed to evaluate the effect of dynamic loads on inlet piping from Unit 1 and Unit 2 residual heat removal system suction relief valves that discharged to the recycle holdup tank (RHUT); and as a result, failed to verify if the RHUT design was adequate to withstand loads resulting from a discharge through residual heat removal system suction relief valves into the RHUT.

However, during a subsequent inspection, the NRC inspectors determined that the licensee had not restored compliance for this NCV within a time period commensurate with the significance of the issue. Consequently, the conditions for considering the violation as non-cited, in accordance with Section 2.3.2(a)(2) of the NRC Enforcement Policy, were not met. Therefore, a cited Notice of Violation was issued on November 8, 2012.

The licensee responded to the NRC regarding this Notice of Violation by letter dated December 7, 2012, which described corrective actions and when full compliance would be achieved. The licensees corrective actions included: 1) revising procedures to assure adequate quench volume was present whenever the RHUT was aligned to the residual heat removal system suction relief valves; 2) evaluating calculation analysis CN-CRA-09-29 to determine whether dynamic loads from potential over-pressurization were within the design limit of the RHUT provided the RHUT had an adequate volume of colder water to lower the temperature effect from the hotter residual heat removal system water; 3) evaluating calculation analysis BRW 10-0010 to assure acceptable large early release frequency dose conditions existed; and, 4) performing modifications to protect against separate water hammer concerns by providing adequate drain and venting capability and adequate structural support for the inlet piping. The inspectors reviewed the licensees written response and relevant documents to evaluate the adequacy of corrective actions and to verify that the corrective actions were completed.

The inspectors did not identify any issues of concern. Full compliance was achieved on May 30, 2014. This violation is closed.

4OA6 Management Meetings

.1 Interim Exit Meeting Summary

On August 1, 2014, the inspectors presented the preliminary inspection results to Ms. M. Marchionda, Braidwood Plant Manager, 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.

.2 Exit Meeting Summary

On August 7, 2014, the inspectors conducted a teleconference exit meeting to present the final inspection results to Mr. W. Spahr, Braidwood Maintenance Director, 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

M. Marchionda, Plant Manager

J. Bashor, Engineering Director

W. Spahr, Maintenance Director

T. Fisk, Nuclear Oversight

D. Jewell, Maintenance Corrective Action Program Coordinator

E. Johnston, Operations Corrective Action Program Coordinator

M. Morris, Radiation Protection Corrective Action Program Coordinator

D. Poi, Emergency Preparedness Manager

P. Raush, Regulatory Assurance Manager

A. Ronstadt, Engineering Corrective Action Program Coordinator

C. Tate, Corrective Action Program Manager

M. Abbas, NRC Coordinator

J. Zoeller, Nuclear Oversight

Nuclear Regulatory Commission

D. Betancourt, Resident Inspector

E. Duncan, Branch Chief, Division of Reactor Projects

LIST OF ITEMS

Opened

05000456/2014007-01; URI Incorporation of Westinghouse NSAL 99-05, Reactor

05000457/2014007-01 Coolant Pump Operation During Loss of Seal Injection, Into the Current Licensing Basis and Corrective Actions For 2B RCP Degraded Thermal Barrier

Closed

05000456/2012004-03; VIO Failure to Analyze Recycle Holdup Tank (RHUT) Inlet Piping

05000457/2012004-03 Loads

Discussed

None