IR 05000348/2008002
| ML081220298 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/30/2008 |
| From: | Scott Shaeffer NRC/RGN-II/DRP/RPB2 |
| To: | Jerrica Johnson Southern Nuclear Operating Co |
| References | |
| IR-08-002 | |
| Download: ML081220298 (30) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
ril 30, 2008
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000348/2008002 AND 05000364/2008002
Dear Mr. Johnson:
On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Joseph M. Farley Nuclear Plant, Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on April 3, 2008, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The NRC reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents two NRC-identified findings and one self-revealing finding of very low safety significance (Green). These findings were determined to be violations of NRC requirements. Also, two licensee-identified violations, which were determined to be of very low safety significance, are listed in this report. The NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy because of the very low safety significance of the violations and because they are entered into your corrective action program (CAP). If you contest any NCV in this report, 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 20555-0001; 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 Resident Inspector at Joseph M. Farley Nuclear Plant.
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
NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket No.: 50-348, 50-364 License No.: NPF-2, NPF-8
Enclosure:
Inspection Report 05000348/2008-002 and 05000364/2008-002 w/Attachment: Supplemental Information
REGION II==
Docket Nos.: 05000348, 05000364 License Nos.: NPF-2, NPF-8 Report No.: 05000348/2008-002 and 05000364/2008-002 Licensee: Southern Nuclear Operating Company, Inc.
Facility: Joseph M. Farley Nuclear Plant, Units 1 and 2 Location: Columbia, AL Dates: January 1, 2008 through March 31, 2008 Inspectors: Eddy L. Crowe, Senior Resident Inspector Shane R. Sandal, Resident Inspector Mark A. Bates, Senior Operations Engineer (Section 1R11)
Bruno L. Caballero, Operations Engineer (Section 1R11)
Approved by: Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects Enclosure
SUMMARY OF FINDINGS
IR 05000348/2008-002 and 05000364/2008-002; 01/01/2008 - 03/31/2008; Joseph M. Farley
Nuclear Plant, Units 1 and 2; Operability Evaluations, Event Follow-up, and Other Activities.
The report covered a three-month period of inspection by resident inspectors and two operations inspectors. Three Green non-cited violations were identified. The significance of most findings is indicated by its color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, ASignificance Determination Process@ (SDP). Findings for which the SDP does not apply may be Green or assigned a severity level after management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, AReactor Oversight Process.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
The NRC identified a Green NCV of 10 CFR 50 Appendix B, Criterion III for failing to implement measures to verify design adequacy resulting in the installation of a maintenance jumper on the cell switch for the Unit 2 2C Component Cooling Water (CCW) pump. This resulted in a condition unknown to the licensee at the time of installation, allowing simultaneous start of both the 2C and 2B CCW pumps in response to a loss of offsite power (LOSP) or safety injection (SI) sequencer signal. This finding has been entered into the licensees CAP as Condition Report (CR) 2007112315.
Failure to verify design adequacy for safety-related components is a performance deficiency. This finding is more than minor because inadequate design evaluations challenged the operability of the A train of CCW. Subsequently, the A CCW train was shown to be operable following additional engineering evaluations. The finding affects the design control attribute of the Mitigating Systems cornerstone. The cornerstone objective of ensuring the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences was not met. The Phase 1 screening performed by the NRC concluded the finding is of very low safety significance (Green). (Section 1R15)
- Green.
A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI was identified for inadequate corrective actions which resulted in the 1C CCW Pumps circuit breaker failing to operate when required. The combination of inadequate tolerances, manipulation of the breaker foot pedal, and the interlock plunger being bound in the interlock bar resulted in the circuit breaker experiencing a trip free operation during its demanded closing operation. During the time the 1C CCW Pump was inoperable, the 1A CCW Pump would not have restarted during LOSP or SI conditions due to a latent failure of its circuit breaker. Thus, a loss of safety function existed for approximately seven hours and fifteen minutes. Because the latent failure of the 1A CCW pump was not a trendable or foreseeable failure, no performance deficiency was identified. The NRC reviewed both breaker failures for a common performance deficiency and none was identified. This finding has been entered into the licensees CAP as CR 2007108601.
The licensees failure to ensure the interlock plunger was correctly aligned to allow proper operation of the 4160 volt 1C CCW pump circuit breaker is a performance deficiency. This finding is more than minor because it affected the equipment reliability attribute of the Mitigating Systems cornerstone. The cornerstone objective of ensuring the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences was not met. The NRC performed a Phase 3 Significance Determination of the performance deficiency and concluded the finding was of very low safety significance. (Section 4OA3)
- Green.
The NRC identified a Green NCV of 10 CFR 50 Appendix B, Criterion XV for failing to properly control nonconforming components resulting in the installation of a 4160 volt breaker for the Unit 1 1C CCW pump with a stop bolt gap dimension not meeting vendor and station maintenance acceptance criteria. This finding has been entered into the licensees CAP as CRs 2007108654 and 2008101720.
Failure to control components not conforming to requirements in order to prevent their inadvertent use or installation in safety-related applications is a performance deficiency.
The NRC determined this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective. Specifically, installation of a breaker not meeting vendor or station acceptance criteria challenged the reliability of the 1C CCW pump. Because the finding did not result in a loss of operability or safety function and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event, the NRC concluded the finding was of very low safety significance (Green). A human performance cross-cutting aspect was identified regarding effectively communicating expectations for procedural compliance and personnel following procedures (H4(b)). (Section 4OA5)
Licensee-Identified Violations
Violations of very low safety significance, which were identified by the licensee, have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensee=s CAP. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near 100 percent Rated Thermal Power (RTP) during this inspection period.
Unit 2 operated at or near 100 percent RTP during this inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection
a. Inspection Scope
Impending Adverse Conditions Review. The NRC evaluated implementation of adverse weather preparation procedures and compensatory measures for the following weather condition. The NRC walked-down portions of the Main Steam Systems, Condensate Storage Systems, Refueling Water Storage (RWS) Systems, and the Emergency Diesel Generators (EDGs). These systems were selected because their safety-related functions could be affected by freezing weather. The NRC verified the applicable portions of procedure FNP-0-SOP-0.12, Cold Weather Contingencies, were performed.
Documents reviewed are listed in the Attachment.
- Projected freezing temperatures from January 3-4 Seasonal Readiness Review. The NRC evaluated implementation of the licensees Cold Weather Contingency procedure, FNP-0-SOP-0.12, and conditions for entry into the procedure. The NRC examined protective coverings of the grating on the Main Steam Valve Rooms, circulating water piping, heat tracing lines on the condensate storage tanks, reactor makeup water storage tanks, and RWS tanks to verify these protections for cold weather conditions were functional. The EDG building was also evaluated to ensure that provisions were implemented to compensate for any known deficiencies.
Documents reviewed are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
a. Inspection Scope
Partial System Walkdowns. The NRC performed partial walk-downs of the following three systems to verify the operability of redundant or diverse trains and components when safety equipment was inoperable. The NRC attempted to identify discrepancies impacting the function of the system, and therefore, potentially increasing risk. The walk-downs were performed using the criteria in licensee procedures FNP-0-AP-16, Conduct of Operations - Operations Group, and FNP-0-SOP-0, General Instructions to Operations Personnel. The walk-downs included reviewing the Updated Final Safety Analysis Report (UFSAR), plant procedures, drawings, checks of control room and plant valves, switches, components, electrical power, support equipment, and instrumentation.
Documents reviewed are listed in the Attachment.
- Unit 2 A Train Residual Heat Removal (RHR) System during planned maintenance to 2B RHR Pump Complete Walk-down. The NRC conducted a complete walk-down of the accessible portions of the following system. The NRC used licensee procedures FNP-0-SOP-56.0, Control Room System, and Functional System Description (FSD) A181006, Control Room Ventilation System, to verify the system alignment of on-service equipment. The NRC also interviewed personnel, reviewed control room logs, Maintenance Rule (MR)monthly reports, CRs, quarterly system health reports, outstanding work orders (WO),and industry operating experience to verify alignment and equipment discrepancies were being identified and appropriately resolved. Documents reviewed are listed in the
.
- Unit 1/Unit 2 Control Room Ventilation System
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
Fire Area Tours The NRC conducted a tour of the six fire areas listed below to assess the material condition and operation status of fire protection features. The NRC verified that combustibles and ignition sources were controlled in accordance with the licensee's administrative procedures; fire detection and suppression equipment was available for use; passive fire barriers were maintained in good material condition, and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with the requirements of licensee procedures FNP-0-AP-36, Fire Surveillance and Inspection; FNP-0-AP-38, Use of Open Flame; FNP-0-AP-39, Fire Patrols and Watches; and the associated Fire Zone Data sheets. Documents reviewed are listed in the Attachment.
- Unit 1/Unit 2 (shared) Security Computer Inverter Room, Fire Zone 13
- Unit 1 CCW Room, Fire Zone 6
- Unit 1 Service Water Intake Structure (SWIS), Fire Zone 72
- Unit 2 Cable Spreading Room, Fire Zone 40
- Unit 2 SW Intake Structure, Fire Zone 72
- Unit 2 SFP Pump Room, Fire Zone 4 Fire Drill. On March 21, 2008, the NRC observed a fire drill for a simulated fire in the Unit 2 A Train Motor Driven Auxiliary Feedwater Pump (MDAFWP) Room. The NRC observed the licensee response in the fire equipment staging area, main control room, and entry into the simulated fire area was in accordance with plant procedures. The NRC verified station personnel utilized proper fire fighting techniques and equipment was properly restored to operating status following the fire drill. The NRC reviewed procedures FNP-0-AOP-29.0, Plant Fire, FNP-0-EIP-13.0, Fire Emergencies, and FNP-0-FVP-14.0, Auxiliary Building Smoke and CO2 /Halon Removable (Portable Equipment)to verify these procedures were properly implemented.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
a. Inspection Scope
Internal Flooding The NRC reviewed selected risk-important plant design features and licensee procedures intended to protect the plant and its safety-related equipment from internal flooding events. The NRC reviewed flood analysis and design documents, including the UFSAR, engineering calculations and abnormal operating procedures for licensee commitments. The NRC walked-down the area listed below to verify plant design features and plant procedures for flood mitigation were consistent with design requirements and internal flooding analysis assumptions. The NRC reviewed flood protection barriers, which included plant floor drains, condition of room penetrations, condition of the sumps in the rooms, and condition of water-tight doors. The NRC also reviewed CRs to verify the licensee was identifying and resolving problems. Documents reviewed are listed in the Attachment.
- Unit 1 CCW Room 185
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
a. Inspection Scope
Resident Inspector Quarterly Review. On January 14, 2008, the NRC observed portions of the licensed operator training and testing program to verify implementation of procedures FNP-0-AP-45, Farley Nuclear Plant Training Program, FNP-0-TCP-17.6, Simulator Training Evaluation/Documentation, and FNP-0-TCP-17.3, Licensed Retraining Program Administration (Classroom). The NRC observed operations simulator exam scenario 6C, conducted in the licensee's simulator for failed steam generator pressure instrument, a broken RCP drive shaft, an inadvertent SI with main steam isolation valve (MSIV) closure, and a steam generator tube rupture (SGTR) with an ALERT emergency declaration. The NRC observed high risk operator actions, overall performance, self-critiques, training feedback, and management oversight to verify operator performance was evaluated against the performance standards of the licensee's scenario. Documents reviewed are listed in the Attachment.
Biennial Requalification Program Review. The NRC reviewed documentation, interviewed licensee personnel, and observed administration of operating tests associated with the licensees operator requalification program. Each of the activities performed by the NRC was to assess the licensees effectiveness in implementing requalification requirements identified in 10 CFR Part 55, Operators Licenses. The evaluations were also performed to determine if the licensee effectively implemented operator requalification guidelines established in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and Inspection Procedure 71111.11, Licensed Operator Requalification Program. The NRC also evaluated the adequacy of the licensees simulation facility for use in operator licensing examinations using ANSI/ANS-3.5-1985, American National Standard for Nuclear Power Plant Simulators for use in Operator Training and Examination. The NRC observed two crews during the performance of the operating tests. Documentation reviewed included written examinations, Job Performance Measures (JPMs), simulator scenarios, licensee procedures, on-shift records, simulator modification request and performance test records, the feedback process, licensed operator qualification records, remediation plans, watchstanding, and medical records. The records were inspected using the criteria listed in Inspection Procedure 71111.11. Documents reviewed are listed in the
.
b. Findings
No findings of significance were identified.
1R12 Maintenance Rule Effectiveness
a. Inspection Scope
The NRC reviewed the two samples listed below for items such as:
- (1) appropriate work practices;
- (2) identifying and addressing common cause failures;
- (3) scoping in accordance with 10 CFR 50.65(b) of the MR;
- (4) characterizing reliability issues for performance;
- (5) trending key parameters for condition monitoring;
- (6) charging unavailability for performance;
- (7) classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2); and
- (8) appropriateness of performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified as (a)(1). In addition, the NRC specifically reviewed events where ineffective equipment maintenance resulted in invalid automatic actuations of Engineered Safeguards Systems affecting the operating units. Documents reviewed are listed in the
.
- Unit 1/Unit 2 1-2A EDG
- Unit 1 1A SW Pump replacement
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
The NRC reviewed the following seven activities to verify appropriate risk assessments were performed prior to removing equipment for work. The NRC verified risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the NRC verified the appropriate use of the licensees risk assessment and risk categories in accordance with the requirements in licensee procedures FNP-0-ACP-52.3, Mode 1, 2, & 3 Risk Assessment; NMP-GM-006, Work Management; and FNP-0-AP-16, Conduct of Operations -
Operations Group.
- Unit 1, January 4, 2008 - YELLOW risk condition due to scheduled B train solid state protection system (SSPS) surveillance testing concurrent with planned work in the high voltage switch yard (HVSY) associated with the #2 auto transformer
- Unit 2, January 17, 2008 - GREEN risk condition due to rescheduled A train SSPS surveillance testing concurrent with main steam valve room entry, surveillance testing of the 2B auxiliary building battery charger, and 2B DG automatic fuel transfer pump equipment outage
- Unit 2, January 22, 2008 - GREEN risk condition due to main steam flow instrumentation normalization concurrent with 2A auxiliary building battery charger load test and HVSY semi-weekly inspections
- Unit 1, February 4, 2008 - YELLOW risk condition due to scheduled equipment outage for 1B DG concurrent with semi-weekly inspection of HVSY
- Unit 1, February 6, 2008 - GREEN risk condition due to inoperable CCW isolation valve (Q1P17HV3096B) concurrent with 1B DG equipment outage and tornado watch in effect for Houston County
- Unit 1, February 13, 2008 - GREEN risk condition due to corrective maintenance on B train SW lubrication and cooling strainer
- Unit 2, February 13, 2008 - YELLOW risk condition due to 2B DG equipment outage concurrent with equipment outages for the 2B SW pump and 2C charging pump
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
The NRC reviewed the following five operability evaluations to verify they met the requirements of licensee procedures FNP-0-AP-16, Conduct of Operations - Operations Group and FNP-0-ACP-9.2, Operability Determination (OD) for technical adequacy, consideration of degraded conditions, and identification of compensatory measures. The NRC reviewed the evaluations against the design bases, as stated in the UFSAR and FSDs to verify system operability was not affected.
- CR 2007112315, Unit 2, Installation of jumper on cell switch to allow auto-start of 2B CCW pump
- CR 2007112748, Unit 2, Foreign material in bearing housing of 2A MDAFWP
- CR 2008101055, Unit 1/2, CCW non-essential isolation valve (HV3096A & B) not fully closing
- CR 2008101148, Unit 2, 2B SW Pump discharge piping leak
b. Findings
Introduction.
The NRC identified a Green NCV for failure to implement measures verifying the adequacy of system design for a maintenance jumper installed on the cell switch for the 2C CCW pump in accordance with the requirements of 10 CFR 50 Appendix B, Criterion III.
Description.
On December 5, 2007, following a scheduled replacement of the breaker for the 2C CCW pump, the associated mechanized operated cell (MOC) switch for the breaker was declared inoperable due to dimensional acceptance criteria for the MOC operator being exceeded. In an effort to allow the 2C CCW pump breaker to be returned to service while allowing additional troubleshooting of the MOC switch, the licensee authorized a temporary alteration to install a jumper for the 2C CCW pump cubicle cell switch allowing automatic start of the 2B CCW swing pump from the sequencer and returning the 2C CCW pump breaker to service. The residents noted the installation of the maintenance jumper during a review of CRs and control room logs on December 6, 2007. The residents questioned the licensee regarding the configuration control and testing utilized to support the intended function of the installed jumper. The licensee determined the installed jumper did not bypass the automatic start feature of the 2C CCW pump and resulted in a condition allowing the simultaneous start of both the 2C and 2B CCW pumps on the A train in response to a LOSP or SI sequencer signal with both breakers in the connect position. This condition was then documented in the licensees CAP as CR 2007112315. The licensee subsequently bypassed the automatic start feature of the 2C CCW pump for the remaining duration of the jumper installation.
Because the simultaneous pump start condition challenged the hydraulic and electrical response of the system under emergency conditions, the licensee performed an evaluation ultimately supporting system operability through additional engineering analysis. Licensee administrative procedure FNP-0-AP-13, Control of Temporary Alterations requires that a sufficient review be performed prior to implementing the alteration to determine not only the effect of the alteration being considered on the circuit or system being worked, but also on associated/inter-tiered circuits or systems. The NRC determined the licensees review of the alteration was inadequate because it failed to identify the alteration resulted in a condition that challenged the hydraulic and electrical response of the mitigating system during accident conditions.
Analysis.
Failure to verify the adequacy of design for the maintenance jumper installed on the cell switch for the 2C CCW pump was a performance deficiency. This finding was more than minor because it affected the design control attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective because a condition allowing simultaneous start of 2 CCW pumps challenged the CCW systems reliability. The NRC performed a Phase 1 screening worksheet because the finding was associated with the reliability and function of a mitigating system. Because the finding did not result in a loss of operability or safety function and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event, the NRC concluded the finding was of very low safety significance (Green).
Enforcement.
10 CFR 50 Appendix B, Criterion III, requires design control measures be established for verifying the adequacy of system design. Contrary to the above, on December 5, 2007, the licensee reviewed, authorized, and installed a temporary alteration for the cell switch of the 2C CCW pump breaker, which, unknown to the licensee at the time of installation, resulted in a condition allowing the simultaneous start of both the 2C and 2B CCW pumps in response to an LOSP or SI signal with both breakers in the connect position. Because this failure to verify the adequacy of design was of very low safety significance (Green) and has been entered into the CAP as CR 2007112315, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000364/2008002-01, Installation of a Maintenance Jumper for the 2C CCW Pump Cell Switch.
1R18 Plant Modifications
a. Inspection Scope
Temporary Modifications. The NRC reviewed the following temporary modification (TM)and associated 10CFR50.59 screening criteria against the system design bases documentation and the licensee's TM procedure FNP-0-AP-8, Design Modification Control. The NRC reviewed implementation, configuration control, post-installation test activities, drawing and procedure updates, and operator awareness for this TM.
Documents reviewed are listed in the Attachment.
- CR 2007105294, Unit 1 Temporary Capping of Tubing to Differential Pressure Cylinder for Containment Emergency Hatch
b. Findings
No findings of significance were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
The NRC reviewed the criteria contained in licensee procedures FNP-0-PMT-0.0, Post-Maintenance Test Program, to verify post-maintenance test procedures and test activities for the following five systems/components were adequate to verify system operability and functional capability. The NRC also witnessed the test or reviewed the test data to verify test results adequately demonstrated restoration of the affected safety function(s). Documents reviewed are listed in the Attachment.
- FNP-1-STP-33.0B, SSPS Train B Operability Test following corrective maintenance on Unit 1 B train SSPS
- FNP-2-STP-4.3, Charging Pump 2C Inservice Test following corrective maintenance on Unit 2 C charging pump discharge check valve (Q2E21V0122C)
- FNP-1-STP-24.16, Containment Cooler and RCP Motor Air Cooler SW Valves Inservice Test following corrective maintenance on Unit 1 RCP motor air cooler SW return valve (Q1P16MOV3134)
- FNP-2-STP-33.0B, SSPS Train B Operability Test following corrective maintenance on Unit 2 B train SSPS
- FNP-1-STP-80.1, DG 1B Operability Test following failure of the diesel speed signal generator
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The NRC reviewed surveillance test procedures and either witnessed the test or reviewed test records for the following surveillance tests to determine if the tests adequately demonstrated equipment operability and met the TS requirements. The NRC reviewed the activities to assess for preconditioning of equipment, procedure adherence, and valve alignment following completion of the surveillance. The NRC reviewed licensee procedures FNP-0-AP-24, Test Control; FNP-0-M-050, Master List of Surveillance Requirements; and FNP-0-AP-16, Conduct of Operations - Operations Group; and attended selected briefings to determine if procedure requirements were met.
Documents reviewed are listed in the Attachment.
Surveillance Tests
- FNP-2-STP-24.1, 2A, 2B and 2C SW Pump Quarterly Inservice Test
- FNP-1-STP-11.6, RHR Valves Inservice Test
- FNP-2-STP-41.3, Power Range Functional Test Channels (N-41)(N-42)(N-43)
- FNP-1-STP-11.1, 1A RHR Pump Quarterly Inservice Test Reactor Coolant System (RCS) Leak Detection
- FNP-1-STP-9.0, RCS Leakage Test
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
The NRC evaluated the conduct of routine licensee emergency drills on the following dates to identify any weaknesses and deficiencies in classification, notification, and protection action recommendation (PAR) development activities. The NRC observed emergency response operation in the simulated control room to verify event classification and notifications were performed in accordance with FNP-0-EIP-9.0, Emergency Classification and Actions. The NRC used procedure FNP-0-EIP-15.0, Emergency Drills, as the inspection criteria. The NRC also attended the licensee critique of the drill to compare any inspector-observed weaknesses with those identified by the licensee in order to verify whether the licensee was properly identifying failures.
- February 12, dual unit LOSP/reactor trip; failure of the 1B DG to start with subsequent trip of 1A MDAFWP, faulted steam generator (SG) inside containment, large break loss of coolant accident (LBLOCA) with breach of containment and failed fuel
- March 12, annual drill, failure of reactor to automatically trip following three dropped rods, trip of 1A RHR pump, small break RCS leak with incomplete containment isolation, followed by LBLOCA with breach of containment and failed fuel
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a. Inspection Scope
The NRC sampled licensee data for the PIs listed below to verify the accuracy of the PI data reported during the period listed. Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Rev. 5, was used to verify the basis in reporting for each data element. Documents reviewed are listed in the Attachment.
Initiating Events Cornerstone
- Unplanned Scrams per 7,000 Critical Hours
- Unplanned Power Changes per 7,000 Critical Hours
- Unplanned Scrams with Complications The NRC reviewed samples of raw PI data, Licensee Event Reports (LERs), and Monthly Operating Reports for the period covering January 2007 through December 2007. The data reviewed from the LERs and Monthly Operating Reports was compared to graphical representations from the most recent PI report. The NRC also examined a sampling of operations logs and procedures to verify the PI data was appropriately captured for inclusion into the PI report, as well as ensuring the individual PIs were calculated correctly.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
a. Inspection Scope
Daily Condition Report Reviews. As required by Inspection Procedure 71152, "Identification and Resolution of Problems," and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the NRC performed a daily screening of items entered into the licensee=s CAP. This review was accomplished by reviewing hard copies of each CR, attending daily screening meetings and accessing the licensee=s computerized database.
Selected Issue Follow-up Inspection. The NRC performed a detailed review of the failures of the Unit 2 B Train SSSPS failures on February 20, 2008 and December 21, 2007. The NRC interviewed station personnel and reviewed CR 2007112795 and CR 2008101637. The NRC also performed a review of the condition report database to evaluate the existence of additional failures. The NRC reviewed the root cause performed for CR 2007112795 to verify safety concerns were properly classified and prioritized for resolution; technical issues were evaluated and dispositioned to address operability and reportability. The NRC evaluated the thoroughness of this root cause determination to ensure a sufficiently thorough evaluation was performed; the extent of condition, generic implications, common causes and previous history was adequately considered; and appropriate corrective actions were implemented or planned in a manner consistent with safety and compliance. The NRC also evaluated the documents against the requirements of the licensees corrective action program as defined in NMP-GM-003, Corrective Action Program.
b. Findings
No findings of significance were identified.
4OA3 Event Follow-up
.1 (Closed) LER 05000348/2007003-00. Component Cooling Water Pump Breaker Failures
(Closed) URI 05000348, 364/2007010-01. Unavailability of CCW System to Automatically Actuate Due to Breaker Failures
a. Inspection Scope
On September 4, 2007, the 1C CCW Pump failed to start from the main control board due to a combination of the interlock plunger misadjustment and manipulation of the foot pedal by the system operator. At the time of this failure, the opposite train CCW Pumps (1A) circuit breaker was is a condition such that an automatic start of this pump would not have occurred. These conditions existed for approximately seven hours and fifteen minutes. Corrective actions for the 1C CCW Pump included replacing the affected breaker, verification of proper adjustment of interlock plungers on installed Cutler Hammer 4160 volt circuit breakers, and changes to station procedures. This event is also further discussed in NRC Augmented Inspection Team report 05000348, 364-2007-010. Unresolved Item (URI) 05000348(364)/2007010-01 was opened pending evaluation of the operational effects of the CCW system being unable to automatically actuate. The NRC reviewed the maintenance history and circumstances surrounding the misadjusted interlock plunger. The NRC reviewed licensee procedures associated with the control of circuit breakers and interviewed licensee personnel regarding the implementation of those procedures. The NRC reviewed root cause evaluations for both circuit breakers and evaluated licensee corrective actions. This LER is closed. The NRC also performed the required evaluation required by URI 05000348(364)/2007010-01, Unavailability of CCW System to Automatically Actuate Due to Breaker Failures, and this URI is closed.
b. Findings
Introduction.
A self-revealing Green finding was identified for inadequate corrective actions resulting in the 1C CCW Pumps circuit breaker failing to operate when demanded from the Control Room.
Description.
On September 4, 2007, the licensee attempted to start 1C CCW Pump to facilitate post maintenance testing of the 1B CCW Pump A function. The licensee dispatched a system operator to the pumps 4160 volt circuit breaker cubicle to perform pre-start checks per station procedure FNP-0-SOP-0.0, Figure 6, ESF Equipment 4160-volt Breaker Pre-Start Check Sheet. The system operator decided to depress the breaker foot pedal to verify full insertion of the interlock plunger. This technique for verifying the plunger was fully inserted was consistent with requalification training given to the system operator the previous week, on a recent revision to station procedure FNP-0-SOP-36.6 Circuit Breaker Racking. However, the licensee did not intend for the system operator to perform this check during breaker pre-start checks. During the unintended check of the breaker foot pedal, the system operator encountered resistance.
The system operator called for assistance of electrical maintenance personnel. Electrical maintenance confirmed the interlock plunger was fully inserted, while the breaker foot pedal was bound. With the assistance of electrical maintenance, the system operator reported to the control room the circuit breaker was ready for pump start. Shortly thereafter, a control room reactor operator attempted to start the 1C CCW Pump, but the pump failed to start.
The licensee quarantined the circuit breaker and performed troubleshooting activities.
The licensees troubleshooting activities indicated the interlock plunger was bound due to the breaker walking forward in its cubicle during breaker closing evolutions. The licensee also discovered inadequate tolerances of the interlock plunger linkage, foot pedal, and trip latch. Over the past several years, the licensee has experienced multiple demand failures of 4160 volt circuit breakers, some of which included the breaker interlock plunger not being located correctly in the breaker interlock bar. The licensee implemented corrective actions which included multiple verifications of plunger height in relation to the interlock bar. However, these corrective actions did not establish the tolerances necessary for proper interaction between the interlock plunger linkage and the trip latch. The licensees troubleshooting activities also revealed these inadequate tolerances were exacerbated by the system operator manipulating the foot pedal during breaker pre-start checks. The combination of the inadequate tolerances, the manipulation of the breaker foot pedal, and the interlock plunger being bound in the interlock bar resulted in the circuit breaker experiencing a trip free operation during its demanded closing operation. This resulted in the failure of the pump to start.
On September 5, 2007, the licensee restored 1C CCW Pump to an operable status by placing another 4160 volt circuit breaker into the electrical bus cubicle. The licensee started the 1C CCW Pump and completed post maintenance testing of the 1B CCW Pump A function. The licensee then focused maintenance activities of the 1B CCW Pump to its B function. The licensee attempted to start 1A CCW Pump to facilitate post maintenance testing of the 1B CCW Pump, but this pump failed to start when demanded from the Control Room. Troubleshooting subsequently identified a bent roll pin in the closing latch mechanism of this pumps 4160 volt circuit breaker which prevented the breaker from completing its closing evolution. The licensee installed a new Cutler hammer breaker to replace the Allis Chalmer 4160 breaker. The Allis Chalmer breakers had a latent failure existing at the time the 1C CCW Pump failed to start resulting in the loss of CCW function for approximately seven hours and fifteen minutes. Because, the latent failure of the 1A CCW pump was not a trendable or foreseeable failure, no performance deficiency was identified. The NRC reviewed both breaker failures for a common performance deficiency and none was identified.
Analysis.
The licensees failure to ensure the interlock plunger was properly aligned to allow proper operation of the 4160 volt circuit breaker is a performance deficiency. This finding is more than minor because it affected the equipment reliability attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences. The Phase 1 screening worksheet of Manual Chapter 0609, SDP, determined a Phase 2 analysis was required because the finding represented a loss of system safety function. The Senior Reactor Analyst performed a Phase 3 Significance Determination of the performance deficiency and concluded it was of very low safety significance (Green). A major contributor to this conclusion was the short duration the component was out of service. The SDP review also adjusted for actual increased breaker failure rates for the 4160 breaker failures at Farley. The dominant accident sequence involved the failure of the entire CCW function followed by a RCP seal LOCA without the ability to inject ECCS to maintain the core covered. The seal failure and the failure of ECCS were a direct result of the loss of Component Cooling Water. Due to the nature of the performance deficiency, neither a common cause failure nor a recovery of the failed component prior to the end of the exposure time was assumed credible.
Enforcement.
10 CFR Part 50, Appendix B, Criterion XVI, requires that measures be established to correct conditions adverse to quality. Contrary to the above, the licensee failed to establish sufficient corrective actions to ensure 4160 circuit breaker interlock plungers were properly adjusted and that station personnel operated circuit breakers in a manner which preserved the mechanical interlock. Because of the very low safety significance and because the licensee included this condition in their CAP as CR 2008100108, this violation is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000348/2008002-02, Failure of CCW 4160 Circuit Breaker to Operate When Demanded Due to Ineffective Corrective Actions.
.2 (Closed) LER 05000364/2007001-00. Technical Specification 3.8.1 Violation Due to
failure of Breaker/Mechanism-Operated Cell (MOC) Switch The NRC documented the closure of this LER as it related to Unit 1 in NRC inspection reports 05000348, 364/2007005 and 05000348, 364/2007008. The inspectors reviewed this LER as it related to Unit 2 and did not identified any additional findings of significance.
4OA5 Other
.1 (Closed) URI 05000348, 364/2007010-002. Use of Non-Conforming Components in
Safety Related Applications
a. Inspection Scope
On September 5, 2007, the licensee installed a breaker into the cubicle for the Unit 1 A train CCW pump that had a trip latch to latch roll clearance (stop bolt gap) which exceeded station maintenance and vendor acceptance criteria. Corrective actions included replacing the affected breaker. This event is discussed in NRC Augmented Inspection Team report 05000348, 364-2007-010. URI 05000348(364)/2007010-002 was opened pending determination of the reason a nonconforming breaker was installed into a safety related application. The NRC reviewed the maintenance history and circumstances surrounding the installation of the nonconforming breaker. The NRC also reviewed licensee procedures associated with the control of nonconforming components and interviewed licensee personnel regarding the implementation of those procedures.
During tours of the plant, the NRC verified administrative controls associated with nonconforming components were being appropriately implemented. Documents reviewed are listed in the Attachment.
b. Findings
Introduction.
The NRC identified a Green NCV for failure to implement measures to prevent installation of a nonconforming breaker into the cubicle for the 1C CCW pump in accordance with the requirements of 10 CFR 50 Appendix B, Criterion XV.
Description.
On September 5, 2007, the licensee installed a breaker into the cubicle for the 1C CCW pump which had a stop bolt gap clearance of 0.062. Licensee maintenance procedure FNP-0-EMP-1313.03, Maintenance of Siemens-Allis 4.16KV Circuit Breakers Type MA-350 and associated vendor manuals specify the stop bolt gap should be set between 0.015 and 0.047. Station maintenance procedures also state, in part, breakers with stop bolt gap clearances found in the high range have experienced trip-free conditions and that, when making these adjustments, clearances should be set to the lower end of the acceptance criteria. The breaker out-of-tolerance stop bolt gap measurement was previously identified during maintenance checks by the licensee on January 4, 2007 when the breaker was then moved to a non-safety related application.
The licensee discovered the previously identified out of tolerance condition during a review of the paperwork after the breaker had been moved back to the 1C CCW pump and had been returned to service on September 5, 2007. Station procedure FNP-0-AP-22, Nonconformance Control/Deficiency Reporting, requires nonconforming items be tagged with a nonconforming material tag. The stated purpose of this procedure is to establish administrative mechanisms to preclude the inadvertent use or installation of nonconforming material. The NRC noted a nonconforming material tag was not placed on the breaker on January 4, 2007 to prevent its subsequent use in a safety related application. The NRC also noted two other examples where nonconforming components were not tagged. Specifically, the original breaker occupying the cubicle for the 1C CCW pump, which had failed to close on September 4, 2007, was not tagged with a nonconforming material tag. Additionally, on October 16, 2007, the breaker for the 1B RHR pump failed to close and following that event, the inspectors did not observe the use of a nonconforming material tag on that breaker to prevent the potential for its inadvertent use. The original 1C CCW pump breaker was refurbished and the 1B RHR pump breaker was never reused in the plant.
Analysis.
Failure to implement measures to prevent the installation and use of a nonconforming breaker into the cubicle for the 1C CCW pump was a performance deficiency. This finding was more than minor because it affected the equipment reliability attribute of the Mitigating Systems cornerstone objective and adversely impacted the cornerstone objective in that breakers with high stop bolt gap measurements have been attributed to breaker fail-to-close events as documented in the licensees CAP and thus challenged the systems reliability. The NRC performed a Phase 1 screening worksheet because the finding was associated with the reliability and function of a mitigating system. Because the finding did not result in a loss of operability or loss of safety function and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event, the NRC concluded the finding was of very low safety significance (Green). Because the breaker was not indentified with a nonconformance tag as required by FNP-0-AP-22, the NRC determined this finding involved a human performance cross-cutting aspect in that the licensee defines and effectively communicates expectations regarding procedural compliance and that personnel follow procedures (H.4(b)).
Enforcement.
10 CFR 50 Appendix B, Criterion XV, requires measures be established to control materials, parts, or components not conforming to requirements in order to prevent their inadvertent use or installation. Contrary to the above, on September 5, 2007, a breaker, previously known by the licensee not to conform to vendor and station maintenance criteria, was inadvertently installed in the cubicle for the Unit 1 C CCW pump and the pump was returned to service. Because this failure to implement measures to prevent the installation of the nonconforming breaker into the cubicle of the 1C CCW pump was of very low safety significance (Green) and has been entered into the CAP as CRs 2007108654 and 2008101720, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000348/2008002-03, Failure to Prevent the Installation of a Nonconforming Breaker Into the Cubicle for the 1C CCW Pump.
4OA6 Meetings, Including Exit
On April 3, 2008, the NRC presented the inspection results to Mr. Randy Johnson and the other members of his staff who acknowledged the findings. The NRC confirmed that proprietary information was not provided or examined during the inspection.
4OA7 Licensee Identified Violations
The following violations of very low safety significance were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI.A.1 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
- 10 CFR 50.55a (f), Inservice Testing Requirements, in paragraph
- (4) requires, in part, that throughout the service life of a nuclear facility, pumps and valves which are classified as American Society of Mechanical Engineers (ASME) Code Class 1, Class 2, and Class 3 must meet the inservice test requirements set forth in the ASME OM Code. Contrary to this, on August 15, 2007, the licensee determined Pressurizer (PZR) Safety Valves were not being tested at the 60 month test frequency required by the Code. The licensee was measuring the 60 month test frequency from the time the valve was installed to next test date, as opposed to the Code-required frequency of test date-to-test date. This finding is determined to be of very low safety significance because the test history and reliability of the PZR safety valves shows these valves routinely meet the test acceptance criteria and the licensees risk analysis performed on the extended surveillance showed the Incremental Conditional Core Damage Probability (ICCDP) is less than 1.0E-06 and the Incremental Conditional Large Early Release Probability (ICLERP) is less than 1.0E-07. The licensee entered the finding into their CAP as CR 2007108020. On Unit 1, PZR safety valves outside of the 60 month testing frequency were removed and forwarded to vendor for testing. On Unit 2, one valve is currently installed but will be removed during the next refueling outage scheduled for October 2008.
- 10 CFR 50, Appendix B, Criterion III requires in part that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization performing the original design, unless the applicant designates another responsible organization.
Contrary to the above, the licensee made modifications to the design operating pressure of the Unit 1 CCW System Valves HV3096A and HV3096B which isolate CCW flow to and from the non-safety related evaporator packages and hydrogen recombiners. These modifications were made using Calculation SM-99-2189-017, which had been revised with errors in regulator settings for the above valves. The licensee identified the errors in the calculation following changes to regulator settings and entered the error into their CAP. Station personnel performed an operability evaluation of these valves and determined compensatory measures were necessary until their regulator settings could be readjusted to their correct design setpoints.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- R. Bayne, Performance Improvement Manager
- S. Chesnut, Engineering Support Manager
- C. Collins, Plant Manager
- P. Hayes, Engineering Director
- L. Hogg, Security Manager
- J. Jerkins, Performance Improvement Senior Engineer
- J.R. Johnson, Site Vice President
- T. Livingston, Chemistry Manager
- H. Mahan, Licensing Engineer
- B.D. McKinney, Acting Licensing Manager
- B.L. Moore, Site Support Manager
- W. Oldfield, Fleet Oversight Supervisor
- J. Swartzwelder, Work Control Superintendent
- R. Wells, Operations Manager
NRC personnel
Scott
- M. Shaeffer, Branch Chief, Region II, Division of Reactor Projects
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None
Opened and Closed
- 05000364/2008002-01 NCV Installation of a Maintenance Jumper for the 2C CCW Pump Cell Switch (1R15)
- 05000348/2008002-02 NCV Failure of CCW 4160 Circuit Breaker to Operate When Demanded Due to Ineffective Corrective Actions (4OA3.1)
- 05000348/2008002-03 NCV Failure to Prevent the Installation of a Nonconforming Breaker into the Cubicle for the 1C CCW Pump (4OA5.1)
Closed
- 05000348(364)/2007-010-001 URI Unavailability of CCW System to Automatically Actuate Due to Breaker Failures (4OA3.1)
- 05000348(364)/2007-010-002 URI Use of Non-Conforming Components in Safety Related Applications (4OA5.1)
- 05000348/2007003-00 LER Cooling Water Pump Breaker Failures (4OA3.1)
- 05000364/2007001-00 LER T.S. 3.8.1 Violation Due to Failure of Breaker/MOC (4OA3.2)
Discussed
None.