IR 05000456/2008009

IR 05000456-08-009 & 05000457-08-009 on 10/06/2008 - 10/24/2008 for Braidwood Station, Routine Biennial Problem Identification and Resolution Inspection

ML083400182
Person / Time
Site:	Braidwood
Issue date:	12/04/2008
From:	Richard Skokowski NRC/RGN-III/DRP/RPB3
To:	Pardee C AmerGen Energy Co, Exelon Nuclear
References
IR-08-009
Download: ML083400182 (27)

Text

December 4, 2008

SUBJECT:

BRAIDWOOD STATION UNITS 1 AND 2 NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000456/2008009 AND 05000457/2008009

Dear Mr. Pardee:

On October 24, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed a routine biennial Problem Identification and Resolution Inspection at your Braidwood Station, Units 1 and 2. The enclosed report documents the inspection results, which were discussed on October 24 with Mr. L. Coyle and members of his staff.

This inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, and compliance with the Commissions rules and regulations and the conditions of your operating license. Within these areas, the inspection involved examination selected procedures and representative records, observations of activities, and interviews with personnel.

On the basis of the sample selected for review, the team concluded that implementation of the Corrective Action Program at Braidwood was generally good. There was one Green finding identified by the team during this inspection, related to the failure to implement timely corrective actions for a previously identified Non-Cited Violation (NCV). The finding was determined to be a violation of NRC requirements. However, because the violation was of very low safety significance (Green) and because it was entered into your CAP, the NRC is treating this as a Non-Cited Violation in accordance with Section VI.A.1 of the NRC=s Enforcement Policy.

If you contest the subject or severity of this NCV, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC, 20555-0001, with copies to the Regional Administrator, Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001; and the NRC Resident Inspectors= Office at the Braidwood Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Richard A. Skokowski, Chief Projects Branch 3 Division of Reactor Projects Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77 Enclosure:

Inspection Report 05000456/2008009; 05000457/2008009 w/Attachment: Supplemental Information cc w/encl:

Site Vice President - Braidwood Station

Plant Manager - Braidwood Station

Regulatory Assurance Manager - Braidwood Station

Chief Operating Officer and Senior Vice President

Senior Vice President - Midwest Operations

Senior Vice President - Operations Support

Vice President - Licensing and Regulatory Affairs

Director - Licensing and Regulatory Affairs

Manager Licensing - Braidwood, Byron and LaSalle

Associate General Counsel

Document Control Desk - Licensing

Assistant Attorney General

J. Klinger, State Liaison Officer,

Illinois Emergency Management Agency

Chairman, Illinois Commerce Commission

SUMMARY OF FINDINGS

IR 05000456/2008009; 05000457/2008009; 10/06/2008 - 10/24/2008; Braidwood Station,

Routine Biennial Problem Identification and Resolution Inspection.

This inspection was conducted by two resident inspectors and two regional inspectors, with the assistance of the Illinois Emergency Management Agency (IEMA) resident inspector. One finding of very low safety significance (Green) was identified during this inspection. The finding was classified as a Non-Cited Violation (NCV). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using NRC Inspection Manual Chapter (IMC) 0609,

ASignificance Determination Process@ (SDP). 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.

Identification and Resolution of Problems The team concluded that the implementation of the Corrective Action Program (CAP) at Braidwood 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 their safety significance. The team noted that the licensee was adequate at reviewing and applying industry operating experience lesson learned. Audits and self-assessments were also noted to be acceptable. On the basis of interviews conducted during the inspection, workers at the site expressed freedom to enter safety concerns into the CAP, exhibiting a good safety conscience work environment.

There was one Green NCV identified by the team during this inspection. The finding was related to the licensees failure to perform timely corrective actions for a previously identified violation.

a.

Cornerstone: Mitigating Systems

• Green: The inspectors identified a Green NCV of 10 CFR 50, Appendix B,

Criterion XVI, for failure to take timely corrective actions to address a previously issued NCV regarding the substitution of manual actions for automatic actions on the A train auxiliary feedwater pumps. Specifically, the licensee did not perform a full evaluation in accordance with 10 CFR 50.59 for the addition of new Step, 2.c, in Revision 101 of Abnormal Operating Procedure 1/2BwOA-ELEC-4, Loss of Offsite Power, which instructed operators to place the A train auxiliary feedwater pumps in pull-out position. This violation was originally identified by NRC inspectors in January 2007. The inspection team identified that the licensee had not taken timely actions to correct the violation.

This finding was considered to be more than minor because it impacted the procedure quality attribute of the mitigating systems cornerstone. As a result, the inspectors completed a Phase 1 Significance Determination Process Screening in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor

Inspection Findings for At-Power Situations. The inspectors answered no to all of the Mitigating Systems Cornerstone questions in Table 4a of IMC 0609,

Attachment 4, and determined the issue to be of very low safety significance,

Green.

This issue of untimely corrective actions was entered into the licensees corrective action program, and the licensee took immediate corrective actions by issuing Revision 104 to 1/2BwOA-ELEC-4, which removed Step 2.c until the full 50.59 evaluation was completed. (Section 4OA2.a.4).

b.

Licensee-Identified Violations

None.

REPORT DETAILS

4OA2 Problem Identification and Resolution

Completion of sections a. through d. constitutes one biennial sample of problem identification and resolution as defined by Inspection Procedure 71152.

a.

Assessment of the Corrective Action Program Effectiveness

(1) Inspection Scope The inspectors reviewed the licensees CAP implementing procedures and attended CAP meetings for both the Station Ownership Committee (SOC) and the Management Review Committee (MRC) to assess the implementation of the CAP by site personnel.

The inspectors reviewed risk and safety significant issues in the licensees program since the last NRC Problem Identification and Resolution (PI&R) Inspection in May 2007. The selection of issues ensured an adequate review of issues across NRC cornerstones. The inspectors reviewed issue reports (IRs) generated as a result of daily plant activities. In addition, the inspectors reviewed IRs and a selection of completed investigations from the licensees various investigation methods, which included root cause, apparent cause, common cause, and trending performance investigations. The inspectors also reviewed issues identified through NRC generic communications, department self assessments, licensee audits, operating experience reports, and NRC documented findings.

The inspectors performed a five year review of the residual heat removal system, safety injection system, and the auxiliary building ventilation system to assess the licensees efforts in monitoring for any system degradation due to aging. The inspectors performed partial system walkdowns and a detailed CAP document review for each of the above mentioned systems.

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 level of priority and 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.

(2) Assessment - Effectiveness of Problem Identification Based on the information reviewed, the inspectors concluded that the threshold for initiating issue reports was good and well below the plant procedural requirements. The inspectors observed one potential vulnerability in the security organization which is discussed below under observations.

The inspectors concluded that the program was effective at identifying issues. However, a review of the IRs attributed to NRC and other outside agency identification in the last two years illustrated that improvements could be made. The control of scaffolding constructed adjacent to safety related and fire protection equipment was an example of an area where identification and resolution were prompted by the NRC.

Observations:

Potential Corrective Action Program Vulnerability The inspectors identified a low-level issue tracking system utilized by the security organization, referred to informally as the parking lot. A review of the items on the parking lot list indicated primarily low level issues such as minor housekeeping items.

However, the use of an informal tracking list posed a potential vulnerability in that issues may not be widely known or reviewed. In addition members of the security organization expressed concern regarding the eventual resolution of issues binned into the parking lot with no formal means of documenting resolution. The licensee reviewed all parking lot issues for incorporation into the corrective action program. No violations of NRC requirements were identified.

No findings of significance were identified.

(3) Assessment - Effectiveness of Prioritization and Evaluation of Issues The inspectors reviewed the classification of issue reports for resolution ranging from 1 for the most significant to 5, the least significant. In addition, the inspectors verified the licensee response to issue reports of various significance levels was appropriate (root cause reviews, apparent cause evaluations, common cause assessments, and divisional assignments). The inspectors also reviewed the daily SOC and MRC reports and attended meetings to observe the disposition and management review of IR classification. All IRs reviewed were assigned appropriate prioritization and evaluation levels.

The inspectors determined that the evaluations in root cause reports and apparent cause reports that were reviewed were mostly adequate. The corrective actions addressed the identified problems and the timeliness of corrective actions was appropriate to the safety significance. However, the inspectors did identify one equipment apparent cause evaluation (EACE) (IR 803300) where vendor concerns regarding site maintenance practices were not adequately addressed. This issue is recorded in greater detail in the observations section. The inspectors noted that evaluations involving equipment failures also received maintenance rule screenings. A select number of these screenings reviewed by the inspectors indicated that maintenance rule attributes were being properly assigned.

Observations:

Equipment Apparent Cause Evaluation Lacked Sufficient Detail On August 3, 2008, equipment operators noted that they were adding oil to the safety related 1A component cooling water (CC) pumps inboard pump seal oil bulb on a more frequent basis (weekly) than normal. Subsequent inspection by the system engineer identified an oil leak along the pump shaft and the ratcheting of the stationary portion of the inboard pump bearing oil seal. The 1A CC pump oil seals had been recently replaced with a new style seal in June 2008. A detailed internal inspection of the seal identified that the stationary seal was rotating due to the seal o-ring having the incorrect dimensions. The o-ring was determined to have an outside diameter of 3.119 inches versus the required 3.127 inches necessary to create the interference fit needed to hold the stator in place. Licensee-drawn oil samples and inspection of the o-ring did not indicate that wear had occurred during the short duration the seal had been installed, and therefore it was determined that the vendor had supplied an incorrectly sized part.

The inspectors noted that the original EACE referenced a vendor concern that incorrect tooling had been used by the licensee when attempting to establish the interference fit.

No explanation of the validity of this concern was addressed in the EACE. In addition, the licensee did not discuss whether there had been a deficiency in the skill of the craft work performed during seal installation. The inspectors noted that the corrective actions generated by the EACE called for maintenance procedure changes, requiring in-field measurements of the components creating the seal interference fit. The licensee entered the inspectors concerns into the CAP under IR 832347. The EACE was reopened and more detail of the vendors original concerns and how they were addressed was added that showed that licensee technicians installing the seal followed procedures and vendor guidance and that appropriate tooling was used.

No findings of significance were identified.

(4) Assessment - Effectiveness of Corrective Actions The inspectors reviewed licensee generated effectiveness reviews for corrective actions to prevent recurrence from selected root cause reports. The inspectors determined the reviews to be of sufficient detail and timed appropriately to effectively determine whether the corrective actions led to eventual issue resolution. In addition, the inspectors reviewed licensee corrective actions for NRC identified violations identified since the last PI&R inspection. The inspectors identified one previously identified NCV which the licensee did not adequately resolve and is described below.

Introduction:

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, for failure to take timely corrective actions to address a previously issued NCV regarding the substitution of manual actions for automatic actions on the A train AF pumps.

Description:

NCV 05000456/2007002;05000457/2007002-02 was issued for the failure to document, in accordance with 10 CFR 50.59, an evaluation that provided the basis for the determination that a change, test, or experiment did not require a license amendment for a change that the licensee made to their loss of Power abnormal operating procedure. The licensee added a new step, Step 2.c, in Revision 101 of Abnormal Operating Procedure 1/2BwOA-ELEC-4, Loss of Offsite Power, which instructed operators to place the A train auxiliary feedwater (AF) pumps in pull-out position. By procedurally directing operators to place the A train AF pumps in pull-out position, the pumps could no longer respond to accidents as described in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). The licensee performed a screening in accordance with 10 CFR 50.59 but did not perform a full 50.59 evaluation.

By performing only the screening, the licensee did not evaluate if replacing an automatic pump start function with a manual action represented more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR.

This issue was originally captured in IR 583152 with an assigned action (Assignment #3)to perform a full evaluation in accordance with 10 CFR 50.59. The full evaluation would address whether a license amendment was required to implement Step 2.c in 1/2BwOA-ELEC-4. The original due date of the corrective action was April 27, 2007. The evaluation was completed and brought for final approval to the Plant Operations Review Committee (PORC) on February 8, 2007. The PORC did not approve the evaluation and assigned Engineering representatives to provide comments to the originator for incorporation. Additional entries were added to IR 583152 on April 25, July 26, and September 27, 2007, that documented extensions to the evaluation due date because the Engineering comments had not been provided to the originator. At the time the PI&R inspection team was on-site the due date for the evaluation had been extended to October 24, 2008. The inspectors also identified that two subsequent revisions to procedure 1/2BwOA-ELEC-4 had not addressed the original NRC concerns with Step 2.c.

This issue of untimely corrective actions was entered into the licensees corrective action program as IR 831223. The licensee took immediate corrective actions by issuing Revision 104 to 1/2BwOA-ELEC-4, which removed step 2.c until the full 50.59 evaluation was completed.

Analysis:

The failure to complete timely corrective actions for a previous NCV is a performance deficiency. The inspectors evaluated the issue in accordance with IMC 0612, Appendix B, Issue Screening. The traditional enforcement questions were reviewed and the inspectors determined this issue does not warrant review under traditional enforcement. This finding was considered to be more than minor because it impacted the procedure quality attribute of the mitigating systems cornerstone. As a result, the inspectors completed a Phase 1 Significance Determination Process Screening in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors answered no to all of the Mitigating Systems Cornerstone questions in Table 4a of IMC 0609, 4, and determined the issue to be Green or of very low safety significance.

No cross-cutting issues were identified.

Enforcement:

10 CFR 50, Appendix B, Criterion XVI, requires in part that conditions adverse to quality be promptly identified and corrected. Contrary to this requirement, the licensee failed to take timely corrective actions to address the condition adverse to quality identified in NCV 2007002-02. Because this finding was entered into the licensees corrective action program as IR 831223, this violation is being treated as a non-cited violation (NCV) in accordance with Section VI.A of the NRC Enforcement Policy. (NCV 05000456/2008009-01; 05000457/2008009-01)b.

Assessment of the Use of Operating Experience

(1) Inspection Scope The inspectors reviewed the licensees implementation of the facilitys Operating Experience (OPEX) program. Specifically, the inspectors reviewed implementing operating experience program procedure and completed evaluations of a sample of OPEX issues and events.

The inspectors review was to determine 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 OPEX experience, were identified and effectively implemented in a timely manner.

(2) Assessment The inspectors noted that screening of OPEX was performed frequently via teleconferencing between the site and company headquarters. The inspectors determined that operating experience was adequately reviewed at the site. The inspectors noted that root cause reports and apparent cause evaluations included discussions of OPEX.

Observations:

Corrective Actions for Manual Actions Used in Plant Procedures The inspectors identified a weakness with regard to the disposition of the timing of manual operator actions used while performing off-normal procedures and emergency operating procedures (EOP). The licensee identified in training OPEX IR 518546 that the time to complete actions to provide centrifugal charging pump (CV) cooling via an alternate min-flow path was never field tested. The IR stated that the probabilistic risk assessment (PRA) assumed the actions could be completed in 10 minutes but noted the valves were located 29 feet in the air. An action to create a formal critical response time was closed to IR 624518-11.

IR 624518 questioned the operators ability to manually cycle certain air operated valves credited for manual operation in the EOPs. Assignment #3 was generated to create procedures to validate that operators could physically operate the equipment in the time frames expected in the EOPs for the equipment listed in Assignment #2, extent of condition. Assignment #8 was to validate the manual operations when plant conditions could be established. This assignment was closed to actions already taken to validate existing time critical actions under IR 829955.

IR 829955 had one assignment: Operations Crew 3 Shift Manager needs to discuss with Operations Manager and present this request to the Ops Director Peer Group for approval if desired. The action was still open and had a due date of December 11, 2008.

During the inspectors verification that the licensee had adequately addressed the concern originally identified in IR 518546, the licensee PRA expert indicated that the original concern was addressed in the latest revision to the PRA model. The new model did not credit the manual lineup to the radioactive waste hold up tank requiring the manipulation of remotely located valves. In addition, the new PRA model allotted the operators 30 minutes to provide adequate cooling to the CV pumps following a loss of component cooling water. The new model did take credit for control room operators realigning the CV pump suction to the refueling water storage tank (RWST) with an assumed time of 1 minute. The inspectors review of the new PRA model, the inspectors ascertained that the licensee was unable to provide the technical basis for the 30 minute recovery time, how the 1 minute pump suction realignment was derived, and how changing the suction source from the volume control tank to the RWST would protect the pump from overheating with no CC flow.

In IR 834951, the licensee captured that the stations PRA expert was not included in the PRA update plan and that actions to address the review of operator actions were not specifically addressed in the plan. Also in IR 834951, the licensee documented the need to verify the adequacy of rigor associated with the PRA model to ensure the technical basis for all changes were captured.

Actions to Address CV Shaft Failure Vulnerability The inspectors reviewed actions for a prior NRC concern regarding the licensees response to industry and plant events involving shaft failures in CV pumps, which are the licensees high-head safety injection pumps. Industry experience has shown that most of the failures of original manufactured shafts with significant run-time occurred with no indications of problems immediately prior to failure. Westinghouse has recommended replacing the shafts. The licensee initiated a study to determine the best course of action, and assessed the feasibility and cost of several options. The option chosen by Braidwood and Corporate engineering was to replace the shafts if they failed, as documented in IR 443080. This issue was to be brought back to the Plant Health Committee with an action due date of November 13, 2008. The licensee presently maintains a newer generation pump shaft in stores in the event of a shaft failure. The licensee based their decision not to preemptively replace the pump shafts on recently implemented operating procedures that maintain sufficient pump flow and system valve configuration that minimize stress on the CV pumps, and that they could replace the pump shaft without exceeding more than half of the limiting condition for operation allowed outage time of seven days for the CV pumps. No violations of NRC requirements were identified.

No findings of significance were identified.

c.

Assessment of Self-Assessments and Audits

(1) Inspection Scope The inspectors reviewed samples of the governing procedures, schedules, plans, reports, and resulting IRs for licensee self assessments and quality assurance (QA)audits. A sample of corrective actions generated for issues was also reviewed.

(2) Assessment The licensee used numerous corporate and station performance measures to monitor station activities. Departmental assessments were performed and rolled up into station wide assessments. QA audits were effective in identifying a number of findings pertaining to radiation protection, the corrective action program, and maintenance activities. The self assessment and audit functions appeared to be well established by procedures and functioning effectively. A self assessment of the CAP had been performed. This assessment was thorough and well organized. The issues that were identified had assigned corrective actions. The assessment findings and conclusions generally matched those of the inspection team.

No findings of significance were identified.

d.

Assessment of Safety Conscious Work Environment

(1) Inspection Scope The inspectors assessed the licensees safety conscious work environment through the reviews of the facilitys employee concern program (ECP) implementing procedures, discussions with the ECP manager, interviews with personnel from various departments, and reviews of issue reports. The inspectors interviewed approximately twenty individuals from various departments about their willingness to raise nuclear safety issues and reviewed selected corrective action program records to assess safety-conscious work environment.

(2) Assessment Based on interviews with plant personnel and reviews of licensee generated safety culture surveys performed across the organization, the inspectors determined that the site possessed a healthy safety conscious work environment. Two potential vulnerabilities were identified through the interview process. In the first instance, multiple members of the licensees maintenance and security staff reported that they were not able to determine how IRs that they generated were vetted for final closure.

This left the members feeling that their issues were not being adequately reviewed or addressed. The licensee entered this concern into the CAP as IR 831248 with action to create a CAP users guide for the workforce.

The second issue concerned the inactivation of licensee staff computer access due to prolonged periods of inactivity. In this instance licensee staff would need to rely on other personnel to log them into the CAP database in order to generate issues. The inspectors determined that this workaround could potentially lead to low level issues not being entered into the CAP out of inconvenience. This issue was entered into the CAP as IR 834912.

No findings of significance were identified.

4OA6 Management Meetings

Exit Meeting Summary

The inspectors presented the inspection results to Mr. L. Coyle and other members of the Braidwood staff at an exit meeting on October 24, 2008. The licensee acknowledged the issues presented. No proprietary information was identified in the possession of the team.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Hanson, Site Vice President

L. Coyle, Plant Manager

S. Butler, Emergency Preparedness Manager

G. Dudek, Site Training Director

R. Gadbois, Maintenance Director

G. Golwitzer, Acting Regulatory Assurance Manager

D. Gullott, Regulatory Assurance Manager

J. Knight, Nuclear Oversight Manager

T. McCool, Operations Director

J. Moser, Radiation Protection Manager

B. Schipiour, Work Management Director

M. Smith, Engineering Director

T. Schuster, Chemistry, Environmental, and Radioactive Waste Manager

Nuclear Regulatory Commission

R. Skokowski, Chief, Project Branch 3, Division of Reactor Projects, Region III

B. Dickson, Senior Resident Inspector, Braidwood

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000456/2008009-01;

05000457/2008009-01 NCV Failure to take timely corrective action for a previously identified NRC violation (Section 4OA2.a.4)

Closed

05000456/2008009-01;

05000457/2008009-01 NCV Failure to take timely corrective action for a previously identified NRC violation (Section 4OA2.a.4)