IR 05000348/2009007

Download: ML093650019

Text

December 31, 2009

Mr. J. Randy Johnson Vice President - Farley Southern Nuclear Operating Company, Inc. 7388 North State Highway 95 Columbia, AL 36319

SUBJECT: JOSEPH M. FARLEY NUCLEAR PLANT - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000348/2009007 AND 05000364/2009007

Dear Mr. Johnson:

On November 20, 2009, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at the Farley Nuclear Station. The enclosed inspection report documents the inspection findings, which were discussed on November 20, 2009, with you and other members of your staff during an exit meeting.

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

On the basis of the samples selected for review, the team concluded that in general, problems were properly identified, evaluated, and corrected. There was one Green self-revealing finding identified during this inspection associated with an inadequate surveillance test procedure. Specifically, the finding was related to the licensee's failure to maintain adequate procedures for testing the 'B' train solid state protection system. Additionally, a licensee-identified violation (LIV), which was determined to be of very low safety significance, is listed in this report. These findings were determined to be violations of NRC requirements. However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you wish to contest these non-cited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-

0001; and the NRC Senior Resident Inspector at the Farley Nuclear Station.

SNC 2 In addition, examples of minor problems were identified by the team, such as equipment issues that were not entered into the corrective action program, and corrective action item closures that did not implement the actions required to be performed.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response, if any, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document 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/ Daniel Merzke, Acting Chief Reactor Projects Branch 7 Division of Reactor Projects Docket Nos.: 50-348 and 50-364 License Nos.: NPF-2 and NPF-8

Enclosure:

Inspection Report 05000348/2009007 and 05000364/2009007

w/Attachment:

Supplemental Information cc w/encl: (See page 3)

SNC 2 In addition, examples of minor problems were identified by the team, such as equipment issues that were not entered into the corrective action program, and corrective action item closures that did not implement the actions required to be performed.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response, if any, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document 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/ Daniel Merzke, Acting Chief Reactor Projects Branch 7 Division of Reactor Projects Docket Nos.: 50-348 and 50-364 License Nos.: NPF-2 and NPF-8

Enclosure:

Inspection Report 05000348/2009007 and 05000364/2009007

w/Attachment:

Supplemental Information cc w/encl: (See page 3) X PUBLICLY AVAILABLE G NON-PUBLICLY AVAILABLE G SENSITIVE X NON-SENSITIVE ADAMS: G Yes ACCESSION NUMBER:_________________________ X SUNSI REVIEW COMPLETE DJM OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP SIGNATURE Via telecom SON /RA/ TXL /RA/ Via email DJM /RA/ NAME RTaylor SNinh TLighty SSandal DMerzke DATE 12/29/2009 12/22/2009 12/30/2009 12/22/2009 12/31/2009 E-MAIL COPY? YES NO YES NO YES NO YES NO YES YES NO YES OFFICIAL RECORD COPY DOCUMENT NAME: S:\DRP\RPB7\PI&R\PI&R\INSPECTION REPORTS\FARLEY PI&R INSPECTION REPORT 2009007 (DRAFT).DOC SNC 3 cc w/encl: B. Doug McKinney Licensing Manager B-031 Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Angela Thornhill Managing Attorney and Compliance Officer Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Jeffrey T. Gasser Executive Vice President Southern Nuclear Operating Company, Inc. Electronic Mail Distribution William D. Oldfield Fleet Support Supervisor Southern Nuclear Operating Company Electronic Mail Distribution L. Mike Stinson Vice President Fleet Operations Support Southern Nuclear Operating Company, Inc.

Electronic Mail Distribution Paula Marino Vice President Engineering Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Moanica Caston Vice President and General Counsel Southern Nuclear Operating Company, Inc. Electronic Mail Distribution John G. Horn Site Support Manager Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc.

Electronic Mail Distribution Aimee J. Gray Performance Improvement Supervisor Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc.

Electronic Mail Distribution C. Howard Mahan Principal Licensing Engineer Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Senior Resident Inspector U.S. Nuclear Regulatory Commission Joseph M. Farley Nuclear Plant U.S. NRC 7388 N. State Highway 95 Columbia, AL 36319 Mr. Mark Culver Chairman Houston County Commission P. O. Box 6406 Dothan, AL 36302 Jim Sommerville (Acting) Chief Environmental Protection Division Department of Natural Resources Electronic Mail Distribution Dr. D. E. Williamson State Health Officer Alabama Dept. of Public Health Electronic Mail Distribution

SNC 4 Letter to J. Randy Johnson from Daniel Merzke dated December 31, 2009

SUBJECT: JOSEPH M. FARLEY NUCLEAR PLANT - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000348/2009007 AND 05000364/2009007 Distribution w/encl: C. Evans, RII L. Slack, RII OE Mail RIDSNRRDIRS PUBLIC R. Martin, NRR Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.: 50-348, 50-364 License Nos.: NPF-2, NPF-8 Report No.: 05000348/2009007, 05000364/2009007 Licensee: Southern Nuclear Operating Company Inc.

Facility : Joseph M. Farley Nuclear Plant Location: Columbia, AL Dates: November 2 - 6, 2008, and November 16 - 20, 2009 Inspectors: R. Taylor, Senior Project Inspector S. Ninh, Senior Project Engineer T. Lighty, Project Engineer S. Sandal, Resident Inspector, Farley Accompanying W. Deschaine, Project Engineer (in training) Personnel: Approved by: Daniel Merzke, Acting Chief Reactor Projects Branch 7 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000348/2009007, 05000364/2009007; 11/02/2009 - 11/20/2009; Farley Nuclear Plant, Units 1 and 2; biennial inspection of the identification and resolution of problems. The inspection was conducted by a senior project inspector, senior project engineer, project engineer, and a resident inspector. One Green self-revealing finding was identified. The significance of most findings is indicated by its color (Green, White, Yellow, Red) using the Significance Determination Process in Inspection Manual Chapter (IMC) 0609, Significance

Determination Process (SDP). The cross-cutting aspect was determined using IMC 0305, Operating Reactor Assessment Program. Findings for which the Significance Determination Process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's 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, in general, problems were properly identified, evaluated, prioritized, and corrected. Generally, the threshold for initiating condition reports (CRs) was appropriately low, as evidenced by the types of problems identified and the large number of CRs entered annually into the Corrective Action Program (CAP). Employees were encouraged by management to initiate CRs. However, the team did identify some examples where plant issues were not appropriately entered into the CAP. Generally, prioritization and evaluation of issues were consistent with the licensee's CAP guidance, formal root cause evaluations for significant problems were adequate, and corrective actions specified for problems were acceptable. Overall, corrective actions developed and implemented for issues were generally timely, effective, and commensurate with the safety significance of the issues. However, the team did identify some examples where plant issues were not appropriately evaluated consistent with the licensee's CAP guidance.

The team determined that, overall, audits and self-assessments were adequate in identifying deficiencies and areas for improvement in the CAP, and appropriate corrective actions were developed to address the issues identified. The licensee's operating experience (OE) usage was found to be generally acceptable and integrated into the licensee's processes for performing and managing work, and plant operations. Based on discussions and interviews conducted with plant employees from various departments, the inspectors determined that personnel at the site felt free to raise safety concerns to management and use the CAP to resolve those concerns.

Cornerstone: Mitigating Systems

Green.

A self-revealing violation of 10 CFR 50 Appendix B, Criterion V "Instructions, Procedures, and Drawings," was identified for an inadequate procedure for testing the 'B' train solid state protection system (SSPS) (FNP-1-STP-33.3). While Unit 1 was in Mode 5 and N31 source range instrumentation was tagged out and unavailable, the licensee performed step 5.6.1.1B of FNP-1-STP-33.3, resulting in N32 source range instrument being de-energized.

Procedure FNP-1-STP-33.3 inadvertently de-energized the only operable source range instrument for Unit 1. TS 3.3.1 required a minimum of one source range neutron flux monitor with the plant in this condition. When the licensee recognized this condition, they immediately restored power to N32 and exited the TS action. The licensee initiated condition report (CR) 2009105672 to address this issue. The issue was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the SSPS procedure resulted in a total loss of source range instrumentation during testing. This affected the safety function of source range neutron flux indication in Mode 5. This finding was assessed using the Phase 1 and 2 Shutdown Operations screening worksheet of the SDP and was determined to require a Phase 3 analysis because the finding involved a loss of source range monitors while shutdown. This finding was determined to be of very low safety significance because the dominant sequence is a boron dilution event and de-energizing the source range for less than a minute would not prevent an operator from taking the necessary actions to address potential boron dilution. No cross-cutting issue was identified. (Section 4OA2.a.(3))

B. Licensee Identified Violations

A violation of very low safety significance, identified by the licensee, has been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensee's CAP. The violation and corrective action tracking number is listed in Section 4OA7 of this report.

REPORT DETAILS

4OA2 Problem Identification and Resolution

a. Assessment of the Corrective Action Program (1) Inspection Scope The inspectors reviewed the licensee's CAP procedures which described the administrative process for initiating and resolving problems primarily through the use of condition reports (CRs). To verify that problems were being properly identified, appropriately characterized, and entered into the CAP, the inspectors reviewed CRs that had been issued between August 2008 and November 2009, including a detailed review of selected CRs associated with four risk-significant systems: Auxiliary Feedwater (AFW), Service Water (SW), Diesel Generators (DGs), and Radiation Monitors. Where possible, the inspectors independently verified that the corrective actions were implemented as intended. The inspectors also reviewed selected common causes and generic concerns associated with root cause evaluations to determine if they had been appropriately addressed. To help ensure that samples were reviewed across all cornerstones of safety identified in the NRC's Reactor Oversight Process (ROP), the team selected a representative number of CRs that were identified and assigned to the major plant departments, including operations, maintenance, engineering, emergency preparedness, health physics, chemistry, and security. These CRs were reviewed to assess each department's threshold for identifying and documenting plant problems, thoroughness of evaluations, and adequacy of corrective actions. The inspectors reviewed selected CRs, verified corrective actions were implemented, and attended meetings where CRs were screened for significance to determine whether the licensee was identifying, accurately characterizing, and entering problems into the CAP at an appropriate threshold.

The inspectors conducted plant walkdowns of equipment associated with selected systems and other plant areas to assess the material condition and to look for any deficiencies that had not been previously entered into the CAP. The inspectors reviewed CRs, maintenance history, completed work orders (WOs) for the systems, and reviewed associated system health reports. These reviews were performed to verify that problems were being properly identified, appropriately characterized, and entered into the CAP. Items reviewed generally covered a 16month period of time; however, in accordance with the inspection procedure, a five-year review was performed for selected systems for age-dependent issues. Control Room walkdowns were also performed to assess the main control room (MCR) deficiency list and to ascertain if deficiencies were entered into the CAP. Operator Workarounds and Operator Burden screenings were reviewed, and the inspectors verified compensatory measures for deficient equipment were being implemented in the field.

The team conducted a detailed review of selected CRs to assess the adequacy of the root-cause and apparent-cause evaluations of the problems identified. The inspectors reviewed these evaluations against the descriptions of the problem described in the CRs and the guidance in licensee procedures. The inspectors assessed if the licensee had adequately determined the cause(s) of identified problems, and had adequately addressed operability, reportability, common cause, generic concerns, extent-of-condition, and extent-of-cause. The review also assessed if the licensee had appropriately identified and prioritized corrective actions to prevent recurrence.

The team reviewed selected industry operating experience items, including NRC generic communications, to verify that they had been appropriately evaluated for applicability and that issues identified through these reviews had been entered into the CAP. The team reviewed site trend reports, to determine if the licensee effectively trended identified issues and initiated appropriate corrective actions when adverse trends were identified. The inspectors attended various plant meetings to observe management oversight functions of the corrective action process.

Documents reviewed are listed in the Attachment.

(2) Assessment Identification of Issues The team determined that the licensee was generally effective in identifying problems and entering them into the CAP and that there was a low threshold for entering issues into the CAP. This conclusion was based on a review of the requirements for initiating CRs as described in licensee procedure NMP-GM-002-001, "Corrective Action Program Instructions," management expectation that employees were encouraged to initiate CRs for any reason, and a review of system health reports. Generally, the threshold for initiating (CRs) was appropriately low, as evidenced by the types of problems identified and the large number of CRs entered annually into the CAP. Trending was generally effective in monitoring equipment performance. Site management was actively involved in the CAP and focused appropriate attention on significant plant issues.

However, the team identified deficiencies which were not entered into the licensee's CAP. Specifically:

• During a walk down of the EDG system, the team identified residue on the 1C EDG under the fuel compartment.

Additionally, during a walk down of the SW system, the team identified minor leaks, and other housekeeping issues that were not entered into the CAP. No CRs or WOs had been initiated in the CAP to correct or evaluate the observed conditions. The licensee entered CRs 2009113334, 2009113337, and 2009113318 to address these conditions. The licensee's failure to enter conditions adverse to quality into the corrective action program was a performance deficiency. This performance deficiency was assessed using IMC 0612 Appendix B and was screened as Minor because the observed conditions did not challenge system or component operability and no safety consequences to the plant had occurred as a result. The failure to comply with the requirements of NMP-GM-002-001, "Corrective Action Program Instructions," constituted a violation of minor significance that was not subject to enforcement action in accordance with the NRC's Enforcement Policy. Prioritization and Evaluation of Issues Based on the review of audits conducted by the licensee and the assessment conducted by the inspection team during the onsite period, the team concluded that the licensee was generally effective in the prioritization and evaluation of identified problems. Problems were generally prioritized and evaluated in accordance with the licensee's CAP procedures as described in the CR severity level determination guidance in NMP-GM-002, "Corrective Action Program." Each CR written was assigned a severity level at the CAP coordinator meeting, and adequate consideration was given to system or component operability and associated plant risk.

The team determined that the licensee had conducted root cause and apparent cause analyses in compliance with the site CAP procedures, and assigned cause determinations were appropriate considering the significance of the issues being evaluated. A variety of causal-analysis techniques were used depending on the type and complexity of the issue consistent with licensee procedure NMP-GM-002-GL03, "Cause Determination Guideline." The licensee had performed evaluations that were technically accurate and of sufficient depth. The team further determined that operability, reportability, and degraded or non-conforming condition determinations had been completed consistent with the guidance contained in NMP-AD-012, "Operability Determinations and Functionality Assessments for Resolution of Degraded and Nonconforming Conditions." However, the team did make the following observations in the area of prioritization and evaluation of issues:

• CR 2009104096 identified that the motor-operated valve (MOV) torque switch actuation point for Q2P16MOV3134 (service water from reactor coolant pump motor air coolers) appeared to be set at a spring pack displacement of 0.146 inches and that the target displacement should be 0.242 inches to ensure that there is adequate margin to prevent dual-indication issues during the performance of required in-service surveillance tests. The condition report spawned action item (AI) 2009203977 to modify reptasks associated with FNP-0-EMP-1501.17 to set an appropriate limit to minimize potential for dual indication-type closing stroke results. CR 2009104096 and AI 2009203977 were coded as severity level 5 issues. The team concluded that CR 2009104096 described a condition adverse to quality as defined by NMP-GM-002, Corrective Action Program. Additionally, the team concluded that the CR and associated action item should have, at minimum, been assigned a severity level 4 as specified by NMP-GM-002 section 6.4.

The team also noted that during performance of a time stroke test of Q2P16MOV3134 on November 6, 2009, the MOV failed to meet test acceptance criteria due to dual position indication. The team also noted that this was a repeat issue for this valve and that corrective actions specified by AI 2009203977 had not been implemented prior to the November failure. This issue was documented in the CAP as CR 2009113404 and 2009113410.

• CR 2009107346 identified that WO 2053045701 to repair inoperable core exit thermocouple (CETC) Q2B14TI2302-B was voided in error. Because all previous corrective actions to repair the CETC had been closed or voided in the CAP, the CR spawned elective WO 2091590701 to effect repairs. CR 2009107346 was coded as a severity level 5 issue. The team concluded that CR 2009107346 described a condition related to TS LCO 3.3.3 for Post Accident Monitoring Instrumentation and therefore instrument inoperability would constitute a condition adverse to quality as defined by NMP-GM-002, Corrective Action Program. Additionally, the team concluded that the CR and associated action item should have been assigned a severity level 4 as specified by NMP-GM-002 section 6.4. The team noted that the CETC was first identified to be inoperable on March 1, 2000 and is being tracked by open LCO 2-2000-006. The team also noted that this condition may represent an operable but degraded nonconforming (OBDN) condition for Unit 2 CETC channel B as specified by NMP-AD-012, Operability Determinations and Functionality Assessments, Section 4.14.3. The team did not find evidence that this issue is currently being tracked as OBDN for Unit 2 CETC channel B.

• The team noted that the maintenance preventable functional failure (MPFF) evaluation for the 1-2L 600V load center failure documented in CR 2008103720 concluded that the bus bar failure was not an MPFF. The team reviewed both the apparent cause determination and root cause evaluation and concluded that the event should have been classified as an MPFF. This conclusion was based on the lack of preventive maintenance (pre-event) that allowed degradation to go undetected and the substantial amount of industry OE that existed indicating the need to perform these preventive maintenance inspections. The team concluded that maintenance rule failure determination was contrary to FNP-0-M-89, Maintenance Rule Site Implementation Manual, Appendix 1, MPFF Determination Guidelines. This issue was entered into the CAP as CR 2009113291.

• CR 2009112319 identified air leaking from the diaphragm for Q2N12HV3226 (steam admission valve to the U2 TDAFW pump). An initial determination of operability was completed that concluded that the valve was OBDN and compensatory actions were identified to monitor the valve for continued degradation that may challenge pump operability. The team reviewed the condition report and current list of OBDN equipment and noted that this valve was not included on the OBDN list. The team also noted that the equipment field listed Q2N12HV3226 as "Operable." The team concluded that designating the valve as Operable in the equipment field instead of OBDN as required by NMP-AD-012, Operability Determinations and Functionality Assessments, Section 6.3.3.1 prevented the capture of this issue in the OBDN list for periodic review by operations. The team acknowledged that an administrative tracking item was created to monitor the valve condition. This issue was entered into the CAP as 2009113795 and corrective actions were taken to include Q2N12HV3226 on the current list of OBDN equipment. The team also noted that 2 WOs written to address the leaking actuator (2092556701 and 2092556701) were coded as elective maintenance. The team concluded that coding these WOs as elective maintenance was not consistent with NMP-AD-012, Operability Determinations and Functionality Assessments, Section 6.3.3.4 which states that all OBDN work orders should be coded as corrective maintenance.

• CR 2009112150 and CR 2009112166 identified leak-by through valve Q2P17HV3096B (component cooling water (CCW) to waste evaporator packages and H2 recombiners). An initial determination of operability (IDO) was completed that concluded that the aligned CCW train would be OBDN and compensatory actions were identified to maintain a basis for reasonable assurance that the train of CCW aligned to the miscellaneous header could be established in the event of a line break downstream of the valve. The team reviewed the condition reports and noted that elective maintenance WO 2092528101 was spawned from CR 2009112150 to effect repairs to Q2P17HV3096B. The team concluded that coding this WO as elective maintenance was contrary to the NMP-AD-012, Operability Determinations and Functionality Assessments, Section 6.3.3.4 which states that all OBDN work orders should be coded as corrective maintenance. The team acknowledged that an administrative tracking item had been established to brief operating crews on procedures used for filling the CCW expansion tank in the event of a downstream line break. However, the team also noted that no compensatory actions were established to ensure that the make-up water sources to CCW or expansion lank level indication would not be removed from service without prior evaluation given the existing degraded condition of Q2P17HV3096B. Effectiveness of Corrective Actions Based on a review of corrective action documents, interviews with licensee staff, and verification of completed corrective actions, the team determined that overall, corrective actions were timely, commensurate with the safety significance of the issues, and effective, in that conditions adverse to quality were corrected and non-recurring. For significant conditions adverse to quality, the corrective actions directly addressed the cause and effectively prevented recurrence in that a review of performance indicators, all CRs, and effectiveness reviews demonstrated that the significant conditions adverse to quality had not recurred. Effectiveness reviews for corrective actions to prevent recurrence were sufficient to ensure corrective actions were properly implemented and were effective.

(3) Findings

Introduction.

A Green self-revealing NCV was identified for failure to comply with 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The procedure used for testing the 'B' train solid state protection system (SSPS), FNP-1-STP-33.3, resulted in all source range instrumentation being de-energized, which was not appropriate for plant conditions.

Description.

On April 29, 2009, while Unit 1 was in Mode 5 and the N31 source range instrument was tagged out and unavailable, N32 was the only operating source range instrument and was required by TS 3.3.1 to be operable. When performing procedure FNP-1-STP-33.3, step 5.6.1.1 B, the operating N32 source range instrument was de-energized, resulting in Unit 1 entering TS 3.3.1 condition L, "required source range neutron flux channel inoperable." The licensee immediately recognized this condition and returned the input error inhibit switch to its normal position which restored power to N32 and stopped the surveillance test. The licensee exited TS 3.3.1 Condition L once power was restored to N32. In 2001 the licensee revised FNP-1-STP-33.3 to allow testing of the 'B' SSPS system with or without inhibits in effect. The team determined that the procedure change was inadequate to deal with existing plant conditions on April 29, 2009. The use of inhibits allowed the licensee to test the SSPS system in a test mode configuration without initiating safety system actuations. However, placing the SSPS system in inhibit de-energized a source range monitor.

Analysis.

The failure to include adequate guidance in procedure FNP-1-STP-33.3 which resulted in all source range instruments being inadvertently de-energized was a performance deficiency. The issue was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the SSPS procedure inadvertently de-energized the source range monitor and did not verify the availability of source range instrumentation during testing. This affected the safety function of source range neutron flux indication in Mode 5. This finding was assessed using the Phase 1 and 2 Shutdown Operations screening worksheet of the SDP and was determined to require a Phase 3 analysis because the finding involved a loss of source range monitors while in Mode 5. This finding was determined to be of very low safety significance because the dominant sequence is a boron dilution event, and de-energizing the source range for less than a minute would not prevent an operator from taking the necessary actions to address potential boron dilution. No cross-cutting issue was identified.

Enforcement.

10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, states in part that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances. Contrary to the above, since 2001, the licensee failed to provide adequate guidance in Procedure FNP-1-STP-33.3, resulting in loss of all source range instrumentation during testing of the 'B' train SSPS system. The shift supervisor immediately directed the operator to return the switch to its normal position and power was restored to N32. Because of the very low safety significance and because this finding has been entered into the licensee's CAP as CR 2009105672, this violation is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy and is identified as NCV 05000348/2009007-01, Inadequate Procedure for the 'B' Train SSPS System Testing.

b. Assessment of the Use of Operating Experience (OE) (1) Inspection Scope The team examined licensee programs for reviewing industry operating experience and reviewed licensee procedure NMP-GM-008, "Operating Experience Program," to assess the effectiveness of how external and internal operating experience data was handled at the plant. In addition, the team selected operating experience documents (e.g., NRC generic communications, 10 CFR Part 21 reports, licensee event reports, vendor notifications, and plant internal operating experience items, etc.), which had been issued since August 2008 to verify whether the licensee had appropriately evaluated each notification for applicability to the Farley plant, and whether issues identified through these reviews were entered into the CAP. Documents reviewed are listed in the Attachment.

(2) Assessment Based on a review of documentation related to the review of OE issues, the team determined that the licensee was generally effective in screening OE for applicability to the plant. The inspectors verified for selected issues that industry OE was evaluated at either the corporate or plant level depending on the source and type of document. Relevant information was then forwarded to the applicable department for further action or informational purposes. OE issues requiring action were entered into the CAP for tracking and closure. In addition, operating experience was included in each root cause evaluation reviewed by the inspectors in accordance with licensee procedure NMP-GM-002-GL03, "Cause Determination Guideline." (3) Findings No findings of significance were identified.

c. Assessment of Self-Assessments and Audits (1) Inspection Scope The team reviewed audit reports and self-assessment reports, including those which focused on problem identification and resolution, to assess the thoroughness and self-criticism of the licensee's audits and self-assessments, and to verify that problems identified through those activities were appropriately prioritized and entered into the CAP for resolution in accordance with licensee procedure NMP-GM-003, "Self Assessment."

(2) Assessment The team determined that the scopes of assessments and audits were adequate. Self-assessments were generally detailed and critical, as evidenced by findings consistent with the team's independent review. The team verified that CRs were created to document all areas for improvement and findings resulting from the self-assessments, and verified that actions had been completed consistent with those recommendations. Generally, the licensee performed evaluations that were technically accurate. Site trend reports were thorough and a low threshold was established for evaluation of potential trends, as evidenced by the CRs reviewed that were initiated as a result of adverse trends.

(3) Findings No findings of significance were identified.

d. Assessment of Safety-Conscious Work Environment (1) Inspection Scope During normal interactions with plant employees during the course of this inspection, the inspectors informally interviewed plant personnel regarding their knowledge of the CAP at Farley and their willingness to write CRs or raise safety concerns. The inspectors conducted interviews to develop a general perspective of the safety-conscious work environment at the site to determine if any conditions existed that would cause employees to be reluctant to raise safety concerns. The inspectors reviewed the licensee's Concerns Program Procedure and interviewed the Concerns Coordinator. Additionally, the inspectors reviewed a sample of employee concern issues which had been entered into the CAP to verify concerns were being properly reviewed and deficiencies were being resolved.

(2) Assessment Based on the interviews conducted and the CRs reviewed, the team determined that licensee management emphasized the need for all employees to identify and report problems using the appropriate methods established within the administrative programs, including the CAP and concerns program. These methods were readily accessible to all employees. Based on discussions conducted with a sample of plant employees from various departments, the inspectors concluded that employees felt free to raise issues, and that management encouraged employees to place issues into the CAP for resolution. The inspectors did not identify any reluctance on the part of the licensee staff to report safety concerns.

(3) Findings No findings of significance were identified.

4OA3 Event Follow-up

.1 (Closed) LER 05000364/2008-001-00, Condition Prohibited by TS 3.8.6 Battery Cell Parameters

On February 11, 2008, the licensee incorrectly approved a quarterly surveillance test on the Unit 2 Auxiliary Building B-Train Battery for Cell number 33 with 2.06 V as satisfactory. This quarterly surveillance test was repeated on May 7, 2008, and Cell number 33 of B-Train Battery was determined to have a voltage of 2.06 V which was below the minimum allowed cell voltage of 2.08 V per TS Table 3.8.6-1. If one or more required batteries with one or more battery cell parameters are not within Category B, TS 3.8.6 Action statements were required to verify battery cell parameters meet Table 3.8.6-1 Category C limits (> 2.02V) once per 7 days thereafter and to restore battery cell parameters to Category B limits (> 2.08 V) within 31 days. This error resulted in the B-Train Battery being inoperable between February 11 and May 7, 2008, due to not completing the required TS Action statements. The licensee determined that the apparent cause was the journeymen taking the voltage reading failed to follow the procedure and compare the voltage to the correct acceptance criteria, and the supervisor also failed to ensure the procedure was correctly completed. Corrective actions included replacing Cell Number 33 on June 3, 2008, providing additional supervisor personnel reviews of TS battery surveillance test packages, and coaching personnel involved in the event on the importance of attention-to-detail when using procedures and legibility of recorded test data. The inspectors determined that this finding was more than minor because it affected the objective of the equipment performance attribute under the Mitigating Systems cornerstone in that the Unit 2 Auxiliary Building B-Train Battery was degraded. The inspectors evaluated this finding against NRC SDP Phase I screening worksheets and determined it to be of very low safety significance (Green) because the battery cell was well above the 2.02 V Category C limits during the entire period and therefore, the battery remained available to perform its safety function. This licensee-identified finding involved a violation of TS 3.8.6, Train A and B Auxiliary Building and Service Water Intake Structure (SWIS) Batteries. The enforcement aspects of the violation are discussed in Section 4OA7. This LER is closed.

.2 (Closed) LER 05000348/2008-004-00, Reactor Trip Due to Loss of RCP Breaker

Position On November 19, 2008, Unit 1 was operating at 100 percent power. Due to a gradual drop in 4160 V safety-related bus voltage over several hours, Farley contacted Alabama Power - Alabama Control Center (ACC) to discuss removing the shunt reactor from service in the Farley High Voltage Switch Yard (HVSY) to raise the 230 kV grid voltage. After receiving the request, the ACC operator used an air-break disconnect switch to isolate the shunt reactor, instead of using the breakers normally used for this purpose. Opening the disconnect switch, which is not designed to break load, caused a significant electrical arc and phase-to-phase fault. The transmission protection system tripped the appropriate devices and cleared the fault in approximately 3 cycles. However, the instantaneous HVSY voltage drop resulted in loss of breaker position indication for the 1B Reactor Coolant Pump (RCP) which initiated an automatic Solid State Protection System (SSPS) reactor trip. Flow to the reactor core was never lost. The reactor tripped without complications and all safety systems performed normally. Unit 1 returned to power operation on November 20, 2008.

The licensee determined that the event was caused when the air-break disconnect switch was opened while still under load. The ACC operator utilized a control display that was inadequate to perform the requested operation. The man-machine interface, together with the ACC operator's experience and training were insufficient to prevent the inappropriate action by the operator. The licensee also determined that the reactor trip would not have occurred from the close-in transmission fault, had the RCP Breaker Position relay power supply been aligned to the inverters instead of the AC constant voltage source transformers. Corrective actions included alerting the ACC operator of the requirement to isolate the shunt reactor prior to operating the air-break disconnect and the requirement to use switching orders for operations at the Farley HVSY, briefing the event with Farley Operations shift personnel and ACC personnel, aligning Reactor Trip relay power supplies to the inverters for all RCPs on both Units 1 and 2, issuing OE on the event to all Southern Nuclear sites and externally to the nuclear industry, and implementing modifications to eliminate the RCP breaker position single point vulnerability on Units 1 and 2 during refueling outages 1R23 and 2R20, respectively.

The inspectors reviewed the root cause and noted that Farley had experienced numerous reactor trips prior to 2003 and two additional reactor trips since. The root cause evaluation for the 2003 event initiated an action item (AI) to issue design change requests (DCRs) to eliminate the RCP breaker open trips for Units 1 and 2. However, these DCRs were later cancelled based on the costs outweighing the benefit without addressing the AI and no formal evaluation was performed to determine the risks or other actions to mitigate the risks. The inspectors verified that the CAP and engineering process allowed the deletion of the relay design control packages without evaluating the risk and without revising the affected root cause. The inspectors determined that the RCP breaker open trips were not considered safety-related equipment and the current CAP had been improved to require risk evaluation and documentation before deferring corrective actions associated with equipment. No findings of significance were identified and no violation of NRC requirements occurred. The licensee documented the issue in CR 2008112610. This LER is closed.

4OA7 Licensee Identified Violations

The following violation of very low safety significance was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI.A.1 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.

• TS 3.8.6 requires that battery cell parameters for Train A and Train B Auxiliary Building and Service Water Intake Structure (SWIS) batteries shall be within the limits of Table 3.8.6-1. TS SR 3.8.6.2 requires battery cell parameters to be verified to meet TS Table 3.8.6-1. If one or more required batteries with one or more battery cell parameters are not within Category B, TS 3.8.6 Action statements are required to verify battery cell parameters meet Table 3.8.6-1 Category C limits (> 2.02 V) once per 7 days thereafter and to restore battery cell parameters to Category B limits (> 2.08 V) within 31 days.

Contrary to the above, between February 11 and May 7, 2008, Unit 2 Auxiliary Building B-Train Battery Cell number 33 was 2.06 V which was below TS Table 3.8.6-1 limits and the licensee failed to restore Cell number 33 float voltage to Category B limits within 31 days as required per TS 3.8.6 Action statements. The licensee entered this condition into their CAP as CR 2009105768. This finding was evaluated against NRC SDP Phase 1 screening worksheets and determined to be of very low safety significance (Green) because the battery cell was above the 2.02 V Category C limits during the entire period and therefore, the battery remained available to perform its safety function during the affected period.

4OA6 Meetings, Including Exit

On November 20, 2009, the inspectors presented the inspection results to Randy Johnson and other members of the site staff. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

ATTACHMENT: SUPPPLEMENTAL INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

P. Autrey, Buried Piping Program Engineer

S. Brumfield, CAP Supervisor

M. Byrd, Design Engineering Supervisor

M. Caldwell, CCW System Engineer

C. Collins, Plant Manager

B. Deloach, Scaffolding Program Engineer

A. Gray, Performance Improvement Supervisor

B. Griner, Engineering Support Manager

J.R. Johnson, Site Vice President

B.D. McKinney, Licensing Supervisor

H. Mahan, Principal Licensing Engineer

R. Martin, Technical Services Manager

J. Morris, LLRT/ILRT Program Engineer

D. Morrow, Engineering Supervisor ILRT/ LLRT Program

K. Haynes, SW System Engineer

B. Smith, System Engineering Supervisor

V. Wilherson, Corporate Engineer

NRC

D. Merzke, Branch Chief, Reactor Projects Branch 7

S. Shaeffer, Branch Chief, Reactor Projects Branch 2

E. Crowe, Senior Resident Inspector, Farley Nuclear Plant

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000348/2009007-01 NCV Inadequate Procedure for the B train SSPS System Testing (Section 4OA2)

Closed

05000364/LER-2008-001-00 LER Condition Prohibited by TS 3.8.6 Battery Cell Parameters (Section 4OA3.1)

05000348/LER-2008-004-00 LER Reactor Trip Due to Loss of RCP Breaker Position (Section 4OA3.2)

Discussed

None

LIST OF DOCUMENTS REVIEWED

Procedures

NMP-AD-012-F02, Formal Functionality Assessment, Version 1

NMP-AD-012-GL02, Functionality Assessment Guideline, Versions 1 and 2

NMP-GM-002, Corrective Action Program, Versions 9, 8, 6, 5, 4, 3 and 1

NMP-GM-002-001, Corrective Action Program Instructions, Versions 14 and 13

NMP-GM-002-F02, Apparent Cause Determination Report Format, Versions 4 and 3

NMP-GM-002-F03, Root Cause Quality Review Sheet, Version 6

NMP-GM-002-F03, Root Cause Grading Sheet, Version 4

NMP-GM-002-F04, Standard Apparent Cause Quality Review Sheet, Version 5

NMP-GM-002-F07, Effectiveness Review Form, Version 2

NMP-GM-002-F33, Enhanced Apparent Cause Quality Review Sheet, Version 1

NMP-GM-003, Self-Assessment Procedure, Versions 13 and 14

NMP-GM-003-GL01, Self-Assessment Guideline, Versions 9 and 8

NMP-GM-002-GL03, Cause Determination Guideline, Version 12

NMP-GM-002-002, Effectiveness Review Instructions, Version 1

NMP-GM-008, Operating Experience Program, Versions 8, 7, and 6

NMP-AD-012, Operability Determination and Functionality Assessments, Versions 5 and 6

NMP-AD-012-GL-01, Prompt Determination of Operability Preparation Guideline, Version 1

NMP-ES-001, Equipment Reliability Process Description, Version 7.0

NMP-ES-006, Preventative Maintenance Implementation and Continuing Equipment Reliability Improvement, Version 6.0

NMP-ES-008, Component Health Monitoring, Version 7.0

FNP-0-M-87, Maintenance Rule Scoping Manual, Version 22

FNP-0-SYP-17.0, Maintenance Rule Monitoring and Reporting, Versions 13 and 15

FNP-0-M-89, Maintenance Rule Site Implementation Manual, Version 13.0

FNP-0-SYP-19, Performance Criteria for Systems under the Scope of the Maintenance Rule,

Version 10

FNP-1-SOP-24.0, Service Water System, Version 71 A-181001, Service Water System - Functional System Description, Revision 25 A-181000, Component Cooling Water System - Functional System Description, Revision 12

A-181005 Diesel Generator System - Functional System Description, Version 36.0

FNP-0-GMP-60.0, General Guidelines and Precautions for Erecting, Modifying, and

Disassembling Scaffolding

FNP-0-M-89, Maintenance Rule Site Implementation Manual, Version 13.0

FNP-0-SYP-17.0, Maintenance Rule Monitoring and Reporting, Version 15.0

FNP-0-SYP-19.0, Performance Criteria for Systems under the Scope of the Maintenance Rule, Version 11.0

FNP-1-RCP-252, Radiation Monitoring System Setpoints, Version 45.0

FNP-1-SOP-45.0, Radiation Monitoring System, Version 35.0

FNP-2-STP-80.6 Diesel Generator 2B 24 Hour Load Test, Version 16.0, 20.0, 23.0

FNP-0-STP-80.7 Diesel Generator 1B 24 Hour Load Test, Version 19.0, 21.0

FNP-1-STP-33.3 Verification of Steam Dump and Generator Trip SSPS K631 and K635 Slave Relays, Versions 6.0, 7.0, 9.0

Attachment Condition Reports

2006108584

2006200922

2009108662

2008112610

2003003089

2006102474

2008111946

2008112610

2009101650

2009102616

2009109700

2009110325

2003002747

2007108613

2007108807

2008106447

2008110125

2008110647

2008110647

2008113602

2009100519

2009101944

2009102695

2009102812

2009103207

2009106304

2009107391

2009107473

2009108827

2009108985

2009109848

2009110151

2009111158

2009111997

2009112205

2009112198

2009112000

2008112219

2008112665

2008113705

2008113922

2009100720

2009101637

2009101848

2009101946

2009103473

2009106295

2009106465

2009107631

2009108748

2009108927

2009109094

2009109353

2009109493

2009110673

2009110823

2009110894

2009111068

2009111109

2009111180

2009111212

2009111214

2009111286

2009111356

2009110658

2009111882

2008109980

2008110156

2008110330

2008112475

2008113333

2008113470

2008113922

2009100863

2009100922

2009101696

2009102450

2009102502

2009104575

2009109027

2009109036

2009109043

2009109248

2009111010

2009111097

2009111121

2009112092

2007103277

2007108600

2007108601

2007112284

2008103720

2008107415

2008111262

2008111606

2008111826

2008112165

2008112297

2008112497

2008112593

2008112595

2008113828

2009100037

2009100319

2009100326

2009100570

2009100772

2009100971

2009101172

2009101360

2009101710

2009101938

2009102141

2009103131

2009104096

2009104144

2009104363

2009104955

2009105782

2009105918

2009106304

2009106856

2009106857

2009107127

2009107346

2009107437

2009107906

2009108664

2009108991

2009109281

2009110308

2009110732

2009110955

2009111614

2009111652

2009113410

2009203263

2009203977

2009205235

2008110880

2008109275

Attachment

2009107314

2007112145

2008106337

2008113476

2009100602

2009107457

2009101710

2009112805

2008102490

2008100108

2007108601

2007100142

2007200550

2006101160

2006108584

2009103028

2008111262

2009100517

2009100458

2009101467

2009100768

2009105620

2009107856

2009105539

2009107823

2009107490

2009109848

2009109849

2009110033

2009109626

2009109219

2009109258

2009112391

2009112319

2009105601

2009100820

2008109171

2009113361

2009105671

2009111047

2008113021

2009111469

2009110907

2009107314

2009105149

2009102271

2008111870

2009112601

2009111374

2009104259

2008113352

2009112601

2009111017

2009105539

2009102876

2009100385

2008112353

2008110475

2009100030

2009100025

2008112253

2009106719

2009108119

2009111469

2009111392

2009110580

2009110474

2009109678

2009103542

2009102352

2009101518

2009100083

2008113842

2008112737

2008112746

2008111870

2008110439

2008110049

2009112601

2009111129

2009108439

2009106725

2009106614

2008113840

2008113818

2009112601

2009112097

2009111820

2009111130

2009109219

2009106265

2009104261

2008110348

2008110344

2008111273

2009108063

2009108061

2009110519

2008111019

2009100031

2009101940

2009400032

2009108231

2009108223

2009108167

2009108017

2009108018

2009107959

2009110601

2009108014

2009108247

2008111585

2009107956

2009104546

2009111469

2009110974

2009110775

2008107589

2009101917

2009101714

2009101672

2009100577

2009100577

2009100602

2009100603

2009100604

2009100607

2008113170

2008111870

2008111739

2008110101

2009112070

2009111486

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2009111353

2009110038

2009109714

2009109715

2009109716

2008113476

2008113245

2008112053

2009112573

2009111469

Attachment

2009109808

2009109760

2008113476

2008113245

2008113081

2009112601

2009111671

2009108633

2009102954

2009101039

2008113529

2009112601

2009107785

2009103753

2009112601

2009110723

2009108682

2009107067

2009101941

2009100020

2008113061

2009112601

2009111618

2009110824

2009110724

2009106615

2009105462

2009100021

2008112429

2008111604

2009112601

2008112196

2009112601

2009111155

2009108009

2009109713

2009105702

2009109687

2009111128

2009102813

2009100794

2008111747

2009111659

2009108450

2009101026

2009111885

2009101942

2009109711

2009103991

2009100033

2009100034

2009105460

2009102225

2009100035

2009101939

2009100052

2009100714

2009100717

2009100716

2009100734

2009100768

2009105916

2009108366

2008111943

2009110977

2009100330

2009107584

2009105095

2009102796

2009101914

2009100777

2009100643

2009107457

2009102152

2009105259

2009111155

2009111417

2009103906

2009109553

2009112290

2009101721

2008112817

2009111137

2008113574

2008109946

2009200895

2009205521

2009106867

2006108639 Work Orders

107224401

1070145701

1070145801

1090212001

1080111801

2008112001

2008111901

2007146001

2008112301

2008112201 S081236201

S081196901

2060449301

2092457801 S092425301

2092427901

Attachment Self-Assessments and Audits

CRP-PA-I-09-006, Nuclear Power Group's Self-Assessment and Benchmarking Program Assessment

NA-SQ-09-003, Assessment of

PIDP-6 Safety Culture Evaluation Process

SQN-PI-S-09-09, Snapshot Self-Assessment Report, Corrective Action Program

SQN-OPS-09-028, Focused/Snapshot Self-Assessment Report, Effectiveness of Actions Associated with

PER 112718

SQN-M&M-S-09-025, Snapshot Self-Assessment Report, Understanding, Proficiency, and Implementation for the Corrective Action Program within MSB

SQN-SCH-F-09-02, Focused Self-Assessment Report, Plant Health and Work Management Inter-Relations SSA903, Corrective Action Program (CAP) Audit SSA0806, Maintenance Functional Area Audit SSA0807, Systems Engineering Functional Area Audit

Other Documents

Farley System Health Report, Service Water, 2nd and 3rd Quarters 2009 Farley System Health Report, Component Cooling Water, 2nd and 3rd Quarters 2009 Farley System Health Report, Emergency Diesel Generators, 3rd Quarter 2009 Farley Nuclear Plant Safety Culture Assessment, August 2008 Farley Buried Pipe and Tanks Monitoring Program Health Report, July 13, 2009

Farley Technical Specifications (TS) Units 1 and 2, Amendment No. 146 (Unit 1) and

Amendment No. 137 (Unit 2) Farley's Implementation of NRC Generic Letter 89-13 Timeline of Events Preceding the SWP Motor Replacement Stainless Steel Cooling Coil,

11/12/09

Farley Quarterly CAP Trend Report, 8/7/09 Farley FSAR Farley TS 1D11/2D11, System Health Report, Radiation Monitors, 1st, 2nd, and 3rd Quarter 2009 A181015, Radiation Monitoring System Functional System Description, Version 11.0 Operating Experience Program Improvement Project Plan

NRC Identified

CRs 200911334 /

2009113337 /200911339, SW System Walkdown

CRs 2009113318, EDG System Walkdown CRs

2009113291, 1-2L LCC Buss Bar Not Classified as MPFF

2009113378 /

2009113380, Control of Equipment That is OBDN per

NMP-AD-012

2009113795, OBDN Report Inaccuracies

2009113479, IDO PDO Items to Have Corrective Maintenance WOs Written

2009113868, CRs Assigned Severity Level 5