IR 05000461/2007005

IR 05000461-07-005 on 10/01/07 - 12/31/07 for Clinton, Fire Protection, Permanent Plant Modifications

ML080360611
Person / Time
Site:	Clinton
Issue date:	02/04/2008
From:	Ring M NRC/RGN-III/DRP/B1
To:	Pardee C AmerGen Energy Co
References
IR-07-005
Download: ML080360611 (44)

Text

ary 4, 2008

SUBJECT:

CLINTON POWER STATION NRC INTEGRATED INSPECTION REPORT 05000461/2007005

Dear Mr. Pardee:

On December 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Clinton Power Station. The enclosed report documents the inspection results, which were discussed on January 10, 2008, with Mr. F.A. Kearney and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and to compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, NRC identified two findings of very low safety significance (Green). Both of these issues involved violations of NRC requirements. However, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating these violations as Non-Cited Violations consistent with Section VI.A.1 of the NRC Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Clinton Power Station.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark A. Ring, Chief Branch 1 Division of Reactor Projects Docket No. 50-461 License No. NPF-62 Enclosure: Inspection Report No. 05000461/2007005 w/Attachment: Supplemental Information cc w/encl: Site Vice President - Clinton Power Station Plant Manager - Clinton Power Station Regulatory Assurance Manager - Clinton Power Station Chief Operating Officer and Senior Vice President Senior Vice President - Midwest Operations Senior Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Director - Licensing and Regulatory Affairs Manager Licensing - Clinton, Dresden and Quad Cities Associate General Counsel Document Control Desk - Licensing Assistant Attorney General State Liaison Officer Illinois Emergency Management Agency Chairman, Illinois Commerce Commission Illinois Emergency Management Agency

SUMMARY OF FINDINGS

IR 05000461/2007005, AmerGen Energy Company LLC, on 10/01/07 - 12/31/07 Clinton Power

Station. Fire Protection, Permanent Plant Modifications.

This report covers a three-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. Two green findings, both involving non-cited violations, were identified by the inspectors. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process (SDP). The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 2006.

Cornerstone: Initiating Events

Green.

The inspectors identified a performance deficiency involving a Non-Cited Violation (NCV) of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement the fire protection program in accordance with program requirements. The inspectors identified multiple instances of the licensees failure to follow approved fire protection program procedures concerning control of Transient Combustible Material and Fire Protection Impairment Reporting. Corrective actions for this issue included removing the unattended combustible material and repairing the non-functional latches on the fire doors.

The inspectors determined that this issue was more than minor because it could be reasonably viewed as a precursor to a significant event. Specifically, the inspectors identified issues where fire had a potential of impacting safety related equipment used for safe shutdown purposes. This finding was of very low safety significance because the transient combustible materials identified by the inspectors were not combustibles of significance, and the licensee maintained non-degraded fire suppression systems in the areas where the fire door latches were not functional. The inspectors determined that this finding was cross-cutting in the area of Problem Identification and Resolution.

Specifically, the licensee implements a corrective action program with a low threshold for identifying issues. The licensee identifies such issues completely, accurately, and in a timely manner commensurate with their safety significance P.1(a). (Section 1R05)

Cornerstone: Mitigating Systems.

Green.

The inspectors identified a performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, having very low safety significance for the licensees failure to properly test a recently installed permanent plant modification to both the Division 1 and Division 2 shutdown service water (SX) systems. This issue resulted in two of four newly installed vacuum breakers installed by the licensee during the modification that did not pass the minimal design specification during subsequent testing.

The inspectors also determined that the issue was more than minor because it could be reasonably viewed as a precursor to a significant event. Failure to perform appropriate modification testing could lead to components being installed within safety-related systems that do not work as designed. This finding was of very low safety significance because the vacuum breaker not meeting the minimal design specification would not make the shutdown service water system inoperable. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance. Specifically, the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. H.4(c) (Section 1R17)

REPORT DETAILS

Summary of Plant Status

The plant was operated at approximately 96 to 97 percent rated thermal power (maintaining 100 percent electrical output) throughout the inspection period with several de-rates of approximately two to three percent at the requests of the grid operator. The grid operator made those requests due to concerns regarding grid stability.

On December 2, 2007, the licensee entered coast down operations and maintained that mode of operation throughout the end of the inspection period.

REACTOR SAFETY

Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness [R]

1R01 Adverse Weather Protection

.1 Winter Seasonal Readiness Preparations

a. Inspection Scope

The inspectors conducted a review of the licensees preparations for winter conditions to verify that the plants design features and implementation of procedures were sufficient to protect mitigating systems from the effects of adverse weather. Documentation for selected risk-significant systems was reviewed to ensure that these systems would remain functional when challenged by inclement weather. During the inspection, the inspectors focused on plant specific design features and the licensees procedures used to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Safety Analysis Report (USAR) and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant specific procedures. Cold weather protection, such as heat tracing and area heaters, was verified to be in operation where applicable. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures.

Specific documents reviewed during this inspection are listed in the Attachment. The inspectors reviews focused specifically on the following plant systems due to their risk significance or susceptibility to cold weather issues:

• Reactor Core Isolation Cooling System;

• Circulating Water System; and

• Service Water System.

This inspection constitutes one winter seasonal readiness preparations sample as defined in Inspection Procedure 71111.01.

b. Findings

No findings of significance were identified.

.2 Readiness For Impending Adverse Weather Condition - Severe Thunderstorm Watch

a. Inspection Scope

Since thunderstorms with potential tornados and high winds were forecast in the vicinity of the facility for October 17, 2007, the inspectors reviewed the licensees overall preparations/protection for the expected weather conditions. On October 17, 2007, the inspectors walked down the reactor core isolation cooling system, in addition to the licensees emergency alternating current (AC) power systems, because their safety related functions could be affected or required as a result of high winds or tornado-generated missiles or the loss of offsite power. The inspectors evaluated the licensee staffs preparations against the sites procedures and determined that the staffs actions were adequate. During the inspection, the inspectors focused on plant specific design features and the licensees procedures used to respond to specified adverse weather conditions. The inspectors also toured the plant grounds for loose debris, which could become missiles during a tornado, and ascertained operator staffing and if they could access controls and indications for those systems required to control the plant.

Additionally, the inspectors reviewed the USAR and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant specific procedures. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. Specific documents reviewed during this inspection are listed in the Attachment.

This inspection constitutes one readiness for impending adverse weather condition sample as defined in Inspection Procedure 71111.01.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

• Residual Heat Removal Subsystem C with Residual Heat Removal Subsystem A out of service;

• Reactor Core Isolation Cooling System;

• Division 2 Diesel Generator;

• Control Rod Drive System.

The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstone at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, USAR, Technical Specification (TS) requirements, Administrative TS, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Documents reviewed are listed in the Attachment.

These activities constituted four partial system walkdown samples as defined in Inspection Procedure 71111.04.

b. Findings

No findings of significance were identified.

.2 Semi-Annual Complete System Walkdown

a. Inspection Scope

On November 10, 2007, the inspectors performed a complete system alignment inspection of the High Pressure Core Spray system to verify the functional capability of the system. This system was selected because it was considered both safety-significant and risk-significant in the licensees probabilistic risk assessment. The inspectors walked down the system to review mechanical and electrical equipment line-ups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of past and outstanding work orders (WOs) was performed to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the corrective action program (CAP) database to ensure that system equipment alignment problems were being identified and appropriately resolved. The documents used for the walkdown and issue review are listed in the attached List of Documents Reviewed.

These activities constituted one complete system walkdown sample as defined in Inspection Procedure 71111.04.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Routine Resident Inspector Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of fire fighting equipment. The control of transient combustibles, ignition sources, and the condition of installed fire barriers was also reviewed. The inspector selected fire areas for inspection based on overall contribution to internal fire risk and the potential to impact equipment which could cause a plant transient. The inspector also verified the following:

• fire hoses and extinguishers were in their designated locations and ready for immediate use;

• fire detectors and sprinklers were free from obstructions;

• fire doors functioned properly; and

• dampers and penetration seals were in a satisfactory condition.

The inspectors reviewed portions of the licensees fire protection evaluation report and the USAR to verify consistency in the documented analysis with installed fire protection equipment at the station.

The inspectors completed six samples by inspection of the following areas:

• Fire Zone A-2d, Auxiliary Building 737 Personnel Hatch Area and Fire Area C-2, Elevations 737 and 755 Containment;

• Fire Zone FB-1b, High Pressure Core Spray system pump room;

• Fire Area C-2, Elevations 803 and 828 Containment;

• Fire Zone FB-1a, Fuel Building Elevation 712 General Access Area;

• Fire Area M-1, Division 1 Shutdown Service Water pump room, Fire Zone M-2a, Division 2 Shutdown Service Water pump room, Fire Zone M-2b, Division 3 Shutdown Service Water pump room, and Fire Area M-4, A Turbine Driven Fire pump room; and

• Fire Area CB-2 Division 2 Essential Switchgear Room and Cable Spreading Room.

b. Findings

Introduction:

The inspectors identified a performance deficiency involving a Non-Cited Violation (NCV) of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement the fire protection program in accordance with program requirements. The inspectors identified multiplied instances of the licensees failure to follow approved fire protection program procedures concerning control of transient combustible material and fire protection impairment reporting.

Description:

Throughout the inspection period the inspectors identified multiple instances where the licensee failed to follow fire protection program procedures. For example, on November 29, 2007, the inspectors observed transient combustible items (cardboard boxes containing fluorescent light bulbs) stored on the Control Building 751 elevation mezzanine (east). The licensee had posted this area as a Transient Combustible Free Zone. Licensee Procedure OP-AA-201-009, Control of Transient Combustibles, defines a Transient Combustible Free Zone as an area in the plant in which transient combustible material is strictly controlled. Therefore, authorization in the form of a transient combustible permit is required prior to staging or storing any transient combustibles in the area. Attachment 5 of OP-AA-201-009 states that placement of transient combustible material in specified area without prior approval and additional compensatory measures is prohibited in Modes 1, 2, and 3. Control Building 751 was an area specified in the procedure. Soon after the inspectors notified the licensee regarding this issue, the licensee removed the combustible material from this area.

Another example included a plant walkdown, conducted on December 3, 2007, where the inspectors noted transient combustible items (rubber hoses and plastic bucket) in the transient combustible free zone on the Control Building 762 elevation and rubber hoses on the Control Building 751 elevation mezzanine (west). Both areas are specified in OP-AA-201-009 Attachment 5, as transient combustible free zones. The area in which the transient combustible items were located on elevation 762 contained highly visible red stripes on the floor and markings indicating Combustible Free Area. A note in OP-AA-201-009 notes that Striped Red Floor areas and areas posted by signage at Clinton Power Station are provided for the purpose of separating redundant Safe Shutdown Equipment. The inspectors immediately notified the licensee regarding this issue; however, the transient combustibles were not removed until December 4, 2007, following identification of the same concern by the licensees nuclear oversight organization. The Nuclear Oversight organization concern was documented in issue report 707303. Both licensee procedure OP-AA-201-009 and Clinton Power Station (CPS) procedure 1893.01, Fire Protection Impairment Reporting, require that compensatory measures be established when combustible material is staged in a combustible free zone.

This issue is a continuation of issues identified by the inspectors regarding combustible material being found on transient combustible free zones since May 3, 2007, when the inspectors identified combustible material in Fire Zone D-12, 762 Diesel Building.

Also, the inspectors identified a number of instances where fire door latches were not functioning properly, leaving the door improperly secured. Clinton Power Station procedure 1893.01, Fire Protection Impairment Reporting, Appendix A, Section 8, states that, All fire barrier assemblies (walls, floors and ceilings) and fire barrier components (penetration seals, fires, fire dampers and envelope systems) comprising rated fire boundaries protecting safety-related areas shall be functional at all times when the barrier is relied upon to provide separation of safe shutdown functions in the event of a fire. For example, during a walkdown of the Division 2 switchgear cable tunnel inspectors noted that Fire Door 1DR1-462 was able to open without turning the door knob. Upon close inspection, the inspector identified that the latch would not engage.

The inspectors also identified the latches for fire doors 1DR1-367, Diesel Generator Division 3 day tank room, and 1DR1-369, Diesel Generator Division 1 Room to Diesel Generator Division 2 rooms, were not functioning during a walkdown on the Division 2 Diesel generator room. After notification to the Shift Manager one-hour fire watches were established for each of these areas.

Analysis:

The inspectors determined that the licensees failure to follow the procedural requirements of Clinton Power Stations fire protection program was a performance deficiency warranting a significance evaluation in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on November 2, 2006. Specifically, the licensee failed to appropriately implement procedures governing both the control of transient combustibles and fire barrier integrity. The inspectors determined that this issue could be reasonably viewed as a precursor to a significant event. Specifically, the inspectors identified issues where fire had a potential of impacting safety related equipment used for safe shutdown purposes.

Using IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that this issue involved two finding categories. The first category involved Fire Prevention and Administrative Controls. The second category involved Fire Confinement. Referencing IMC 0609, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the inspectors assigned a Low Degradation rating to the issues involving the failure to comply with the licensee transient combustible program. The inspectors conclusions were based on the fact that none of the items found in the combustible free zone could be considered transient combustibles of significance, as described in IMC 0609, Appendix F,

2. This attachment defines transient combustibles of significance as low

flash point liquids (below 200oF) and self igniting combustibles (oily rags). Because this issue was assigned a low degradation rating this issue was of very low safety significance (Green) in accordance with IMC 0609, Appendix F, Task 1.3.1.

In accordance with IMC 0609, Appendix F, Attachment 2, Table A2.2, the second finding category involving Fire Confinement was assigned a Moderate B degradation rating because the door latches were not functional. Referencing IMC 0609, Appendix F, Task 1.3.1, Supplemental Screening for Fire Confinement Findings, the inspectors answered no to both questions 1 and 2, and because the finding category was Fire Confinement, continued as directed to Task 1.3.2, Supplemental Screening for Fire Confinement Findings. The inspectors determined that this issue was of very low safety significance (Green). This decision was based on the fact that the exposing areas where fire door latches were not functional had either a non-degraded auto gaseous room flooding fire suppression system (Division 2 Diesel Generator Room) or a non-degraded automatic full area water-based fire suppression system (Division 1 Switchgear Room).

Enforcement:

Operating license NPF-62, Section 2.F states: The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the final safety analysis report (FSAR) as amended, for the CPS and as approved in the Safety Evaluation Report (SER) (NUREG-0853) dated February 1982 and Supplement Numbers 1 through 8...

Updated Safety Analysis Report, Appendix E, Section 4.0, Fire Protection Evaluation Report Compliance with BTP APC9.5-1, Appendix A, Plant Under Construction and Operating Plant Program contains program requirements of the licensee fire protection program. Fire Protection Evaluation Report, Section C.2, Instructions, Procedures, and Drawings states the inspections, tests, administrative controls, fire drills and training that govern the fire protection program should be prescribed by documented instructions, procedures, or drawings and should be accomplished in accordance with these documents.

Contrary to the above, the inspectors identified multiple instances where the licensee failed to follow fire protection procedures OP-AA-201-009 (Transient Combustible Control) and CPS 1983.01 (Fire Protection Impairment Reporting). This was a violation of the licensee's operating license NPF-62, section 2.F, relating to the fire protection program. The licensees corrective actions for this issue included removing combustible material out of combustible free zones and repairing the non-functioning latches on the fire doors. Because this issue was of very low safety significance and has been entered into the licensees corrective action program (Issue Report 00721120), this violation is being treated as a Non-Cited Violation, consistent with Section VI.A, of the NRC Enforcement Policy. (NCV 05000461/2007-005-01)

Throughout the inspection period, the inspectors identified multiple instances of the licensee failing to identify compliance with the transient combustible material control program. Additionally, the inspectors noted that various licensee personnel have had opportunities to identify fire doors where the latch was not functional. Because the licensee personnel had not self-identified and corrected these issues through the corrective action program prior to the inspectors identification of these issues, the inspectors concluded that the primary cause of this finding was related to the cross-cutting aspect of Problem Identification and Resolution. Specifically, the licensee implements a corrective action program with a low threshold for identifying issues. The licensee identifies such issues completely, accurately, and in a timely manner commensurate with their safety significance. P.1(a)

.2 Annual Fire Protection Drill Observation

a. Inspection Scope

On December 17, 2007, the inspectors observed a fire brigade activation in response to a simulated fire outside the residual heat removal system A heat exchanger room on the 737 elevation of the auxiliary building. The observation was used to determine the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies; openly discussed them in a self-critical manner at the drill debrief, and took appropriate corrective actions. Specific attributes evaluated were:

(1) proper wearing of turnout gear and self-contained breathing apparatus;

(2) proper use and layout of fire hoses;

(3) employment of appropriate fire fighting techniques;

(4) sufficient firefighting equipment brought to the scene;

(5) effectiveness of fire brigade leader communications, command, and control;

(6) search for victims and propagation of the fire into other plant areas;

(7) smoke removal operations;

(8) utilization of pre-planned strategies;

(9) adherence to the pre planned drill scenario; and

(10) drill objectives.

These activities constituted one annual fire protection inspection sample as defined in Inspection Procedure 71111.05.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

a. Inspection Scope

The inspectors observed flood protection equipment in the reactor core isolation cooling pump room, low pressure core spray pump room, residual heat removal A pump room, and the high pressure core spray pump room. The inspectors verified that flooding mitigation plans and equipment were consistent with the design requirements and risk analysis assumptions. The inspector reviewed USAR Section 3.4.1 for internal flood protection measures, reviewed the licensees flooding mitigation procedures, and reviewed issue reports related to possible flood protection issues. Additionally, a plant walkdown was performed to verify design barriers were properly maintained. These barriers included penetrations between rooms containing safety related equipment, watertight doors, room drains, and electrical conduit seals and covers. These items were inspected to verify their material condition would meet design assumptions. The inspector performed a review of the stations maintenance database to verify preventative maintenance was current and equipment deficiencies were being appropriately reported and resolved. Additionally, the corrective action program was reviewed for the past 12 months for issues related to internal flood protection. The inspection of internal flooding for the reactor core isolation cooling, low pressure core spray, residual heat removal A, and high pressure core spray pump rooms constituted one inspection sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

.1 Facility Operating History

a. Inspection Scope

The inspectors reviewed the plant=s operating history from December 2005 through October 2007 to identify operating experience that was expected to be addressed by the Licensed Operator Requalification Training (LORT) program. It was then verified that the identified operating experience had been addressed by the facility licensee in accordance with the station=s approved Systems Approach to Training (SAT) program to satisfy the requirements of 10 CFR 55.59(c), ARequalification program requirements.@

b. Findings

No findings of significance were identified.

.2 Licensee Requalification Examinations

a. Inspection Scope

The inspectors performed a biennial inspection of the licensee=s LORT test/examination program for compliance with the station=s SAT program which would satisfy the requirements of 10 CFR 55.59(c)(4), AEvaluation.@ The operating examination material reviewed consisted of six operating tests, each containing two dynamic simulator scenarios and five or six job performance measures (JPMs). The written examinations reviewed consisted of six written examinations, each containing approximately 35 questions. The inspectors reviewed the annual requalification operating test and biennial written examination material to evaluate general quality, construction, and difficulty level. The inspectors assessed the level of examination material duplication from week-to-week during the current year operating test. The examiners assessed the amount of written examination material duplication from week-to-week for the written examination administered this year. The inspectors reviewed the methodology for developing the examinations, including the LORT program 2-year sample plan, probabilistic risk assessment insights, previously identified operator performance deficiencies, and plant modifications.

b. Findings

No findings of significance were identified.

.3 Licensee Administration of Requalification Examinations

a. Inspection Scope

The inspectors observed the administration of a requalification operating test to assess the licensee=s effectiveness in conducting the test to ensure compliance with 10 CFR 55.59(c)(4), AEvaluation.@ The inspectors evaluated the performance of two crews in parallel with the facility evaluators during two dynamic simulator scenarios and evaluated various licensed crew members concurrently with facility evaluators during the administration of several JPMs. The inspectors assessed the facility evaluators= ability to determine adequate crew and individual performance using objective, measurable standards. The inspectors observed the training staff personnel administer the operating test, including conducting pre-examination briefings, evaluations of operator performance, and individual and crew evaluations upon completion of the operating test.

The inspectors evaluated the ability of the simulator to support the examinations. A specific evaluation of simulator performance was conducted and documented under Section 1R11.8, AConformance with Simulator Requirements Specified in 10 CFR 55.46,@ of this report.

b. Findings

No findings of significance were identified.

.4 Examination Security

a. Inspection Scope

The inspectors observed and reviewed the licensee=s overall licensed operator requalification examination security program related to examination physical security (e.g., access restrictions and simulator considerations) and integrity (e.g., predictability and bias) to verify compliance with 10 CFR 55.49, AIntegrity of examinations and tests.@

The inspectors also reviewed the facility licensee=s examination security procedure, any corrective actions related to past or present examination security problems at the facility, and the implementation of security and integrity measures (e.g., security agreements, sampling criteria, bank use, and test item repetition) throughout the examination process.

b. Findings

No findings of significance were identified.

.5 Licensee Training Feedback System

a. Inspection Scope

The inspectors assessed the methods and effectiveness of the licensee=s processes for revising and maintaining its LORT Program up to date, including the use of feedback from plant events and industry experience information. The inspectors reviewed the licensee=s quality assurance oversight activities, including licensee training department self-assessment reports. The inspectors evaluated the licensee=s ability to assess the effectiveness of its LORT program and their ability to implement appropriate corrective actions. This evaluation was performed to verify compliance with 10 CFR 55.59

(c) ARequalification program requirements@ and the licensee=s SAT program.

b. Findings

No findings of significance were identified.

.6 Licensee Remedial Training Program

a. Inspection Scope

The inspectors assessed the adequacy and effectiveness of the remedial training conducted since the previous biennial requalification examinations and the training from the current examination cycle to ensure that the training addressed weaknesses in licensed operator or crew performance identified during training and plant operations.

The inspectors reviewed remedial training procedures and individual remedial training plans. This evaluation was performed in accordance with 10 CFR 55.59

(c) ARequalification program requirements@ and with respect to the licensee=s SAT program.

b. Findings

No findings of significance were identified.

.7 Conformance with Operator License Conditions

a. Inspection Scope

The inspectors reviewed the facility and individual operator licensees' conformance with the requirements of 10 CFR Part 55. The inspectors reviewed the facility licensee's program for maintaining active operator licenses and to assess compliance with 10 CFR 55.53

(e) and (f). The inspectors reviewed the procedural guidance and the process for tracking on-shift hours for licensed operators and which control room positions were granted watch-standing credit for maintaining active operator licenses.

The inspectors reviewed the facility licensee's LORT program to assess compliance with the requalification program requirements as described by 10 CFR 55.59 (c).

Additionally, medical records for licensed operators were reviewed for compliance with 10 CFR 55.53 (i).

b. Findings

No findings of significance were identified.

.8 Conformance with Simulator Requirements Specified in 10 CFR 55.46

a. Inspection Scope

The inspectors assessed the adequacy of the licensee=s simulation facility (simulator) for use in operator licensing examinations and for satisfying experience requirements as prescribed in 10 CFR 55.46, ASimulation Facilities.@ The inspectors also reviewed a sample of simulator performance test records (i.e., transient tests, malfunction tests, steady state tests, and core performance tests), simulator discrepancies, and the process for ensuring continued assurance of simulator fidelity in accordance with 10 CFR 55.46. The inspectors reviewed and evaluated the discrepancy process to ensure that simulator fidelity was maintained. Open simulator discrepancies were reviewed for importance relative to the impact on 10 CFR 55.45 and 55.59 operator actions as well as on nuclear and thermal hydraulic operating characteristics. The inspectors conducted interviews with members of the licensee=s simulator staff about the configuration control process and completed the IP 71111.11, Appendix C, checklist to evaluate whether or not the licensee=s plant-referenced simulator was operating adequately as required by 10 CFR 55.46

(c) and (d).

b. Findings

No findings of significance were identified.

.9 Annual Operating Test Results and Biennial Written Examination Results

a. Inspection Scope

The inspectors reviewed the pass/fail results of the individual biennial written tests administered by the licensee during calendar year 2007. The inspectors also reviewed the results for the operating and simulator tests (required to be given annually per 10 CFR 55.59(a)(2)) administered by the licensee during calendar years 2006 and 2007.

The overall written examination and operating test results were compared with the significance determination process in accordance with NRC Manual Chapter 0609, Appendix I, AOperator Requalification Human Performance Significance Determination Process.@

b. Findings

No findings of significance were identified.

.10 Resident Inspector Quarterly Review

a. Inspection Scope

On October 10, 2007, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator re-qualification examinations to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance;

• Crews clarity and formality of communications;

• Ability to take timely actions in the conservative direction;

• Prioritization, interpretation, and verification of annuciator alarms;

• Correct use and implementation of abnormal and emergency procedures;

• Control board manipulations;

• Oversight and direction from supervisors; and

• Ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. The following scenarios were observed:

• EGS-LOR-02, Revision 01; Loss of feedwater, Small Break LOCA;

• EGS-LOR-40, Revision 03; Small Leak into secondary Containment.

This inspection constitutes one quarterly licensed operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 Routine Quarterly Evaluations

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the Residual Heat Removal System.

The inspectors reviewed events where ineffective equipment maintenance has resulted in invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• implementing appropriate work practices;

• identifying and addressing common cause failures;

• scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

• characterizing system reliability issues for performance;

• charging unavailability for performance;

• trending key parameters for condition monitoring;

• ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

• verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Documents reviewed are listed in the Attachment.

This inspection constitutes one quarterly maintenance effectiveness sample as defined in Inspection Procedure 71111.12.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• Reviewed work associated with WO 912637-03, Replacement and Testing of 1A1 Main Feed 1AP07EK per CPS 3501.01, WO 912427 and WO 1044764, 95 Relay Test for Division 1 Shutdown Service Water Valve Operability Test.

• Reviewed work associated with WO 861827 and 861828, Condensate Booster Pump, WO# 1049059, SX Operability Test, WO# 106128-02, Division 2 DG Operability, WO# 814214 9080.19, DG 1B Overcrank, Delay timer Test, Diffferential Overcurent, Trip Test, and Trip Bypass, and WO# 863080, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run and Hot restart.

• Reactor Core Isolation Cooling System Outage Window.

• Reviewed licensee risk assessment associated with work implementing plant process computer modifications, concurrent with Main Steam Line area temperature calibrations and the Manual Scram Channel Functional.

These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstone. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.

These activities constituted four samples as defined in Inspection Procedure 71111.13.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• AR 00695330, 1SX346A Shutdown Service Water System Vacuum Breaker Modification/Vendor Testing;

• AR00687545, Water hammer on Shutdown Service Water Piping Potential Transient, High Pressure Core Spray Room Coolers;

• OE 691277, Replenishment Diesel Fuel Oil Lubricity Values Out of Specification;

• IR 695404, Reactor Core Isolation Cooling Local Leak Rate Test High Flow Rate and

• AR 00715429, Suspected Unreliable 138 KV Indication South Bloomington and AR 00715427, South Bus Voltage Exceeded 362.2 KV Due to Low Grid Load.

The inspectors selected these potential operability issues based on the risk-significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and the USAR to the licensees evaluations, to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the

.

These activities constituted five inspection samples as defined in Inspection Procedure 71111.15.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

.1 Annual Resident Review

a. Inspection Scope

Engineering Change (EC) Package 358900 was reviewed and selected aspects were discussed with engineering personnel. This document and related documentation were reviewed for adequacy of the associated 10 CFR 50.59 safety evaluation screening, consideration of design parameters, implementation of the modification, post-modification testing, and relevant procedures, design, and licensing documents were properly updated. The inspectors reviewed completed work activities to verify that installation was consistent with the design control documents. The modification installed a vacuum breaker in the upper portion of the shutdown service water system to lessen the severity of a hydraulic transient on the system following a system startup in response to a design basis accident.

This inspection constitutes one sample as defined in Inspection Procedure 71111.17.

b. Findings

Introduction:

The inspectors identified a performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, having very low safety significance for licensees failure to properly test a recently installed permanent plant modification to both the Division 1 and Division 2 shutdown service water (SX)system.

Description:

As part of an effort to improve the reliability of the SX system, the licensee performed an engineering evaluation to address the impact of water hammer (hydraulic transients) loads in the SX system. The results of the evaluation were based on Clinton Power Station calculation IP-M-734. Calculation IP-M-734 showed that the existing SX vacuum breakers on the Division 1 and Division 2 SX system were required to reduce the transient loads in Division 1 and Division 2 to within code allowable and that additional vacuum breakers were required on both divisions of the SX piping system in the SX piping just upstream and downstream of the essential switchgear chillers. Water hammer is a concern in the system because following a loss of offsite power (LOOP)event coincident with a loss of coolant accident (LOOP/LOCA), the SX system is not available for operation for 30 seconds per design, due to the LOOP. During this loss, SX system pressure becomes degraded. Because of this degraded pressure condition, the upper portions of the SX piping system have a very high probability of voiding due to the effect of gravity (drain-down). Studies performed by the licensee have shown this voiding/column separation causes sections of the piping system to experience vacuum conditions. This vacuum condition has a tendency to amplify the hydraulic transient experience when the SX pumps are started following a LOOP/LOCA.

The engineering evaluation also concluded that while the SX system has been shown to be very robust in design and the only noted problem on either Division 1 or Division 2 has been the damage to the control room ventilation chiller divider plate, the analysis does show that higher than normal transient pressures could occur if the new vacuum breakers were not installed and placed in service. Finally, the analysis concluded that at least one of the two new vacuum breakers must be in service to support SX operability or an engineering evaluation will be required to support continued operation of that division of SX.

Engineering Change (EC) package 358900 was developed to install the vacuum breakers to relieve/lessen the effect of a hydraulic transient. In September 2007, the licensee installed the new vacuum breakers (two per division) on the portion of the SX system that supports the Division 1 and Division 2 essential switchgear system. The vacuum breakers were installed under work orders 908458 and 908461. During a review of the EC, the inspectors noted that the licensee did not specify any pre or post installation testing to ensure the vacuum breakers would perform their intended safety function during a water hammer condition, nor did the licensee address why dynamic post-modification testing was not required. Licensee Procedure CC-AA-107-101, Post Modification Acceptance Testing, section 4.1 stated that all configuration changes (modifications) require testing. The procedure stated that this testing verifies that the installed configuration corresponds to the design configuration and that the quality of the workmanship is acceptable. These typically consist of continuity checks, non-destructive examination of welds, leak checks, pressure tests, instrument calibration and bump check of motors to ensure they rotate in the correct direction.

According to the general requirements contained in the bill of material list contained in EC package 358900, the valves were to meet all the technical and operating requirements of K-2873A, ASME Section III, Division 3, Revision 4. Specifically, the documents stated that valves shall be designed such that the vacuum breakers fully open under a 0.5 PSID (vacuum), close at 0 psid and close bubble tight at 0.2 psid (position pressure). The inspectors noted that the only post modification testing required was a leak check at normal operating temperature and pressure.

A licensee investigation into the inspectors concern confirmed that post modification test was indeed required to validate that the vacuum breakers performed their intended safety function of relieving shutdown service water system piping vacuum conditions following a LOOP/LOCA event. The inspectors question and a subsequent issue report prompted additional testing by the licensee under WO 1078819. During the tests, two

(2) of the four vacuum breakers (one in each division) failed the test acceptance criteria of opening at 0.5 psid (Vacuum) or </= 3 pounds-force. The testing failures were documented in IR 696764 and 697157, respectively. Worst case differential pressure recorded during the licensee testing indicated 15 pounds-force to open the vacuum breaker. The licensee completed a formal operability evaluation using RELAP 5 (a thermal-hydraulic modeling computer code) that showed that the increase in transient loads caused by the vacuum breaker not opening until 3.0 psid (17 pounds-force) were within the ASME code-allowable.

Analysis:

The inspectors determined that the licensees failure to specify appropriate post-modification testing in accordance with the licensees testing program was a performance deficiency warranting a significance evaluation in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on November 2, 2006. The inspectors also determined that the issue was more than minor because this finding can be reasonably viewed as a precursor to a significant event. Failure to perform appropriate modification testing could lead to components being installed within a safety-related system that do not work as designed. As discussed above, two of the required four vacuum breakers did not meet the minimal required acceptance testing criteria, which represents a reduction in the reliability of the shutdown service water system in response to a LOOP/LOCA event. Additionally, the inspectors determined that the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, dated March 23, 2007, because the finding was associated with the operability and reliability of a system or train in a mitigating system. The finding also affected the Mitigating System Cornerstone objective of ensuring the reliability and capability of a system that responds to an initiating event to prevent undesirable consequences. Additionally, the inspectors concluded that this finding affected the Mitigating System Cornerstone for Core Decay Heat Removal Degraded (Long-Term Cooling). For the Phase 1 Screening Worksheet, Mitigation Systems Cornerstone the inspectors answered Yes to Question 1: Is the finding a design or qualification deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment? As a result, the issue was screened to be of very low safety significance, Green.

A licensee investigation into this issue concluded that the apparent cause of this issue was personal error and lack of attention to the details. The licensees investigation also noted that multiple engineers had opportunities to find and identify this issue prior to installation of the modification in the field. Using this information, the inspector concluded that the primary cause of this finding was related to the cross-cutting aspect of Human Performance. Specifically, the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c)).

Enforcement:

10 CFR Part 50 Appendix B, Criteria XI Test Control states that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall include, as appropriate, proof test prior to installation, preoperational tests, and operational tests during nuclear power plant operation.

Contrary to the above, on September 21, 2007, the licensee failed to perform appropriate post modification testing following the installation of vacuum breakers in the shutdown service water system. This issue resulted in vacuum breakers that were required for system operability being installed on December 16, 2007, that would not have functioned at the minimal operating requirement specifications as described in licensee design documentation.

The licensees corrective actions for this issue included prompt testing of the shutdown service water vacuum breakers and accelerating the testing frequency of these vacuum breaker tests to every two months. Because this issue is of very low safety significance and has been entered into the licensees corrective action program (Issue Report 00721065 and 00706713), this violation is being treated as a Non-Cited Violation, consistent with Section VI.A, of the NRC Enforcement Policy.

(NCV 05000461/2007-005-02)

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance (PM) activities for review to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• WO# 1053166, 1E12F085A Repair RHR Line Fill Pump Check Valve;

• WO# 1022326, SX piping replacement (Freeze Seal) 1VY04S, Emergency Core Cooling System Reactor Core Isolation Cooling Pump Room Coil Cabinet.

WO# 950311-01, Replacement of TDR Temperature Indicator steam line flow indication (Self-test system), WO# 48995, Reactor Core Isolation Cooling Room cooler temperature element replacement 1TEVY004, WO# 963382 Replacement of RCIC Storage Tank Level Transmitter 1E51-N501 PMT'd using CPS 9432.49 RCIC Storage Tank Level indicator;

• WO# 706233-03 and 925536 Division 2 Essential Switchgear Heat Removal System;

• Division 3 Diesel Generator B air start motors and

• WO# 1229877 Division 2 Control Room Ventilation Hydromotor replacement.

These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion), and test documentation was properly evaluated. The inspectors evaluated the activities against TS, the USAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the Attachment.

These inspection activities constitute five samples as defined in Inspection Procedure 71111.19.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

.1 Routine Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

• CPS 9080.13, Rev. 38b; Diesel Generator 1B 24 Hour Run and Hot Restart -Operability

• CPS 9041.01, Rev. 36b; Jet Pump Operability Test.

The inspectors observed in plant activities and reviewed procedures and associated records to determine whether: preconditioning occurred; effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing; acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis; plant equipment calibration was correct, accurate, and properly documented; as left setpoints were within required ranges; and the calibration frequency were in accordance with TSs, the USAR, procedures, and applicable commitments; measuring and test equipment calibration was current; test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied; test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used; test data and results were accurate, complete, within limits, and valid; test equipment was removed after testing; where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable; where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure; where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished; prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test; equipment was returned to a position or status required to support the performance of its safety functions; and all problems identified during the testing were appropriately documented and dispositioned in the corrective action program. Documents reviewed are listed in the Attachment.

This inspection constitutes two routine surveillance testing samples as defined in Inspection Procedure 71111.22.

b. Findings

No findings of significance were identified.

.2 In-service Testing

a. Inspection Scope

The inspectors reviewed the test results for the C Residual Heat Removal Pump performed on December 4, 2007, to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements.

During this effort, the inspectors observed in plant activities and reviewed procedures and associated records to determine whether: preconditioning occurred; effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing; acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis; plant equipment calibration was correct, accurate, and properly documented; as left setpoints were within required ranges; and the calibration frequency were in accordance with TSs, the USAR, procedures, and applicable commitments; measuring and test equipment calibration was current; test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied; test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used; test data and results were accurate, complete, within limits, and valid; test equipment was removed after testing; where applicable for in-service testing activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers Code, and reference values were consistent with the system design basis; where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable; where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure; where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished; prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test; equipment was returned to a position or status required to support the performance of its safety functions; and all problems identified during the testing were appropriately documented and dispositioned in the corrective action program.

Documents reviewed are listed in the Attachment.

This inspection constitutes one in-service inspection sample as defined in Inspection Procedure 71111.22.

b. Findings

No findings of significance were identified.

.3 Containment Isolation Valve Testing

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

• CPS 9053.04, 1B Residual Heat Removal System Quarterly Isolation Valve Testing.

The inspectors observed in plant activities and reviewed procedures and associated records to determine whether: preconditioning occurred; effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing; acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis; plant equipment calibration was correct, accurate, and properly documented; as left setpoints were within required ranges; and the calibration frequency were in accordance with TSs, the USAR, procedures, and applicable commitments; measuring and test equipment calibration was current; test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied; test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used; test data and results were accurate, complete, within limits, and valid; test equipment was removed after testing; where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable; where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure; where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished; prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test; equipment was returned to a position or status required to support the performance of its safety functions; and all problems identified during the testing were appropriately documented and dispositioned in the corrective action program. Documents reviewed are listed in the Attachment.

This inspection constitutes one containment isolation valve inspection sample as defined in Inspection Procedure 71111.22.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety2OS2 As-Low-As-Is-

Reasonably-Achievable (ALARA) Planning And Controls (71121.02)

.1 Radiological Work Planning

a. Inspection Scope

The inspectors evaluated the licensees list of outage planning work activities ranked by estimated exposure and reviewed the following six work activities of highest exposure significance:

• Drywell Permanent Shielding;

• Drywell Temporary Shielding;

• Drywell Bioshield In-service Inspection and Support Activities;

• Drywell Flex-hose Repair;

• Drywell Scaffold; and

• Refuel Floor Cavity Work. For these six activities, the inspectors reviewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements in order to determine if the licensee had established procedures and engineering and work controls that were based on sound radiation protection principles in order to achieve occupational exposures that were ALARA. This also involved determining if the licensee had reasonably grouped the radiological work into work activities based on historical precedence, industry norms, and/or special circumstances.

The inspectors evaluated the licensees interfaces between operations, radiation protection, maintenance, maintenance planning, scheduling and engineering groups for interface problems or missing program elements.

The inspectors reviewed work activity planning to determine if there was consideration of the benefits of dose rate reduction activities such as shielding provided by water filled components and piping, job scheduling, along with shielding and scaffolding installation and removal activities.

b. Findings

No findings of significance were identified.

.2 Source-Term Reduction and Control

a. Inspection Scope

The inspectors reviewed licensee records to determine the historical trends and current status of tracked plant source terms and to determine if the licensee was making allowances and developing contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety

2PS3 Radiological Environmental Monitoring Program (REMP) and Radioactive Material Control Program (71122.03)

.1 Inspection Planning

a. Inspection Scope

The inspectors reviewed the most current annual Radiological Environmental Operating Report dated April 27, 2007, and licensee assessment results to evaluate if the REMP was implemented as required by the Radiological Effluent Technical Specifications (RETS) and the Offsite Dose Calculation Manual (ODCM). The inspectors reviewed the report for changes to the ODCM with respect to environmental monitoring and commitments in terms of sampling locations, monitoring and measurement frequencies, land use census, inter-laboratory comparison program, and data analysis. The inspectors reviewed the ODCM to identify environmental monitoring stations and evaluated licensee self-assessments, audits, licensee event reports, and inter-laboratory comparison program results. The inspectors reviewed the FSAR for information regarding the environmental monitoring program and meteorological monitoring instrumentation. The inspectors also reviewed the scope of the licensees audit program to determine if it met the requirements of 10 CFR 20.1101(c). This review represented one sample.

b. Findings

No findings of significance were identified.

.2 Onsite Inspection

a. Inspection Scope

The inspectors walked down selected air sampling stations (greater than 20 percent)and approximately 10 percent of the thermoluminescent dosimeter monitoring stations to determine whether they were located as described in the ODCM and to determine the equipment material condition.

The inspectors observed the collection and preparation of environmental samples including surface water and air. The environmental sampling program was evaluated to determine if it was representative of the release pathways as specified in the ODCM and if sampling techniques were performed in accordance with station procedures.

The inspectors evaluated the condition of the meteorological instruments using observations and record reviews, and assessed whether the equipment was operable, calibrated, and maintained in accordance with guidance contained in the Final Safety Analysis Report, NRC Safety Guide 23, and licensee procedures. The inspectors assessed whether the meteorological data readout and recording instruments, including computer interfaces and data loggers that measure and record wind speed, wind direction, delta temperature, and atmospheric stability measurements, were available on the licensees computer system and whether this information was available in the control room.

The inspectors reviewed each event documented in the Radiological Environmental Operating Report which involved missed samples, inoperable samplers, lost thermoluminescent dosimeters, or anomalous measurements for the cause and corrective actions.

The inspectors reviewed the ODCM for significant changes that resulted from land use census modifications, or sampling station changes made since the last inspection.

This included a review of technical justifications for changed sampling locations. The inspectors assessed whether the licensee performed reviews required to ensure that the changes did not affect their ability to monitor the impacts of radioactive effluent releases on the environment.

The inspectors reviewed the calibration and maintenance records for nine air samplers to evaluate operating parameters. The inspectors reviewed results of the vendors inter-laboratory comparison program and quality assurance programs to assess the adequacy of environmental sample analyses performed by the licensee.

The inspectors reviewed quality assurance audit results of the REMP to determine whether the licensee met the Technical Specification/ODCM requirements.

These reviews represent six samples.

b. Findings

No findings of significance were identified.

.3 Unrestricted Release of Material From the Radiologically Restricted Area

a. Inspection Scope

The inspectors observed the access control locations where the licensee monitored potentially contaminated material leaving the radiologically controlled area and inspected the methods used for control, survey, and release of material from this area.

The inspectors observed the performance of personnel surveying and releasing material for unrestricted use to verify that the work was performed in accordance with plant procedures.

The inspectors evaluated whether the radiation monitoring instrumentation was appropriate for the radiation types present and was calibrated with appropriate radiation sources that represented the expected isotopic mix. The inspectors reviewed the licensees criteria for the survey and release of potentially contaminated material and verified that there was guidance on how to respond to an alarm indicating the presence of licensed radioactive material. The inspectors evaluated the licensees equipment to determine if radiation detection sensitivities were consistent with the NRC guidance contained in Circular 81-07 and Information Notice 85-92 for surface contamination and Health Physics Position-221 for volumetrically contaminated material.

The inspectors reviewed the licensees procedures and records to verify that the radiation detection instrumentation was used at its typical sensitivity level based on appropriate counting parameters, such as counting times and background radiation levels. The inspectors assessed whether the licensee had established a release limit by altering the instruments typical sensitivity through such methods as raising the energy discriminator level or locating the instrument in a high radiation background area.

These reviews represent two samples.

b. Findings

No findings of significance were identified.

.4 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed the licensees self-assessments, audits, condition reports, and special reports related to the radiological environmental monitoring program since the last REMP inspection to determine if identified problems were entered into the corrective action program for resolution. The inspectors also assessed whether the licensee's self-assessment program was capable of identifying and addressing repetitive deficiencies or significant individual deficiencies that were identified by the problem identification and resolution process.

The inspectors also reviewed selected corrective action documents related to the REMP that affected environmental sampling and analysis and meteorological monitoring instrumentation. Staff members were interviewed and documents were reviewed to determine if the following activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk:

• Initial problem identification, characterization, and tracking;

• Disposition of operability/reportability issues;

• Evaluation of safety significance/risk and priority for resolution;

• Identification of repetitive problems;

• Identification of contributing causes;

• Identification and implementation of effective corrective actions;

• Resolution of NCVs tracked in the corrective action system; and

• Implementation/consideration of risk-significant operational experience feedback.

This review represented one sample.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors performed a screening review of Revisions 9, 10, and 11 of the Clinton Annex to the Standardized Emergency Plan to determine whether changes identified in Revisions 9, 10, and 11 decreased the effectiveness of the licensees emergency planning for the Clinton Power Station. This review did not constitute an approval of the changes, and as such, the changes are subject to future NRC inspection to ensure that the emergency plan continues to meet NRC regulations.

These activities completed one inspection sample.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

Other Activities [OA]

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the data submitted by the licensee for the 3rd Quarter 2007 performance indicators for any obvious inconsistencies prior to its public release in accordance with IMC 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings of significance were identified.

.2 Reactor Coolant System Leakage

a. Inspection Scope

The inspectors sampled licensee submittals for the Reactor Coolant System Leakage performance indicator first quarter 2007 through third quarter 2007. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in Revision 5 of the Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline, were used. The inspectors reviewed the licensees operator logs, RCS leakage tracking data, issue reports, event reports and NRC Integrated Inspection reports for the period of January 1, 2007, through September 30, 2007, to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the Attachment this report.

This inspection constitutes one reactor coolant system leakage sample as defined in Inspection Procedure 71151.

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution

.1 Semi Annual Review

a. Inspection Scope

The inspectors searched the licensees corrective action program database using key words failed, failure, broke, and broken for items identified between January 1, 2007, and September 30, 2007, to identify any possible trends in equipment or component malfunctions or failures. Through this search, the inspectors identified multiple issues with back-up power supplies to area radiation monitors required by technical specifications. The inspectors selected several of these issue reports and the related work orders and corresponding procedures for a more detailed review. A list of the documents reviewed is included in the Attachment. This review completed one semi-annual trend review inspection sample.

b. Observations The inspectors reviewed issue reports and work orders related to battery back-up failures on the Main Control Room Air Intake Radiation Monitors 1RIXPR009C, 1RIPR009D, and the Spent Fuel Storage Pool Area Radiation Monitor, 1RIXAR016. In addition, the inspectors reviewed the technical specifications, operational requirements manual, and USAR requirements for these monitors and their back-up power supplies.

The back-up battery test is performed as part of the channel functional testing required by technical specification surveillance requirement 3.3.7.1.2. Step 8.1 35 of CPS 9437.60, Main control room air intake radiation 1RIX-PR009A(B,C,D) channel calibration, and step 8.1.25 of CPS 9437.67, Area radiation monitors 1RIX-AR016(19, 35, 52) channel calibration, are performed to verify the back-up batteries are functioning properly. Clinton Power Station maintenance personnel identified failures to the back-up power supplies for the 1RIX-PR009C on April 3, 2007, 1RIX-009D on April 18, 2007, and 1RIX-AR016 on April 26, 2007. These batteries are normally replaced every three years, and had last been replaced in July of 2004. The licensee determined that the back-up power supply is not required by technical specifications for detector operability.

The inspectors reviewed the operability requirements related to power supplies for these monitors and agreed that the battery back-ups are not required by technical specifications. The licensee has replaced the failed batteries and continues to monitor the performance of these detectors through its preventative maintenance program.

Although the inspectors identified a trend in unexpected failures of these back-up power supplies, because they are not required by technical specifications for monitor operability, and because the failures were captured in the licensees corrective action program and necessary corrective actions were completed, no findings or violations of regulatory requirements were identified.

.2 Annual Sample - Reserve Auxiliary Transformer Static VAR Compensator Tripped Twice

a. Inspection Scope

The inspectors reviewed corrective action documents related to trips of the Reserve Auxiliary Transformer (RAT) Static VAR Compensator (SVC) that occurred on April 29, 2007, and July 7, 2007. The inspectors selected these events based on their risk significance, which placed the station in an unplanned 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO and changed on-line risk level to Red. The inspectors reviewed issue reports, prompt investigation reports, and apparent cause evaluation reports to determine if the licensee correctly identified the cause of each event, if appropriate corrective actions were taken in each case, and if there was any relationship between the two events. A complete list of documents reviewed is included in the Attachment at the end of this report. This review represents the completion of one annual inspection sample.

b. Observations On April 29, 2007, at 1355, the RAT SVC automatically shutdown and RAT voltage increased from 4.2kV to 4.4 kV, outside its technical specification allowed limit of 4.3 kV.

The licensee declared the RAT inoperable, entered a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO, and elevated plant risk level to Red. The licensee transferred vital electrical loads from the RAT to the emergency reserve auxiliary transformer (ERAT) and commenced troubleshooting on the RAT SVC. Troubleshooting revealed the trip was due to aging or overheating of components on the control power module in the Powerlogic Circuit Monitor. The licensee replaced the Powerlogic Circuit Monitor and restored the RAT SVC to service on April 30, 2007. Corrective actions taken by the licensee included preventive maintenance items to monitor these components more frequently and an extent of condition action to replace all of the remaining original control power monitors in the RAT and ERAT SVCs.

On July 7, 2007, at 2146, the RAT SVC automatically shutdown after a phase voltage unbalance relay sensed a voltage imbalance on the 4160 volt bus. The licensee entered an unplanned 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO. The licensee with the assistance of the SVC vendor performed troubleshooting and determined the cause of the trip to be a failure of the PV C514 branch logic circuit card on Thyristor Switched Capacitor (TSC) V1 (A-B phase)to properly control the TSC V1. On July 9, 2007, the licensee replaced the PV C514 branch logic card on TSC V1, and successfully returned the RAT SVC to service.

Neither the licensee nor the vendor was able to determine the cause of the PV C514 branch logic card. Follow up corrective actions included visual inspection of all branch logic circuit cards and card connections in both the RAT and ERAT SVCs and to evaluate the branch logic circuit card for reverse engineering or an equivalent replacement card.

The inspectors reviewed the apparent cause determination reports, the extent of condition reviews, and corrective action assignments for both events and determined that the licensee appropriately identified and corrected the problems that caused the RAT SVC to trip, adequately evaluated the extent of condition, and appropriately addressed any possible common causes. The inspectors determined that the two events were not related and both were unexpected equipment failures that could not have been predicted. No findings of significance were identified.

4OA4 Cross-Cutting Aspects of Findings

.1 A finding described in section 1R05.1 of this report had, as its primary cause, an aspect

of Problem Identification and Resolution, in that, licensee personnel on multiple occasions failed to identify and correct non-compliances with transient combustible material control procedures and non-functional fire door latches. (P.1(a))

.2 A finding described in section 1R17 of this report had, as its primary cause, a Human

Performance deficiency, in that, the licensee failed to ensure supervisory oversight of work activities such that nuclear safety is supported. Specifically, several engineers had opportunity, but failed, to review and identify the lack of adequate post maintenance testing prior to installation of a modification in the field. (H.4(c))

4OA6 MANAGEMENT MEETINGS

.1 Exit Meeting Summary

On January 10, 2008, the inspectors presented the inspection results to Mr. B. C. Hanson, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

.2 Interim Exit Meetings

Interim exit meetings were conducted for:

• Biennial Operator Requalification Program Inspection with Mr. B. C. Hanson, Site Vice President, on November 9, 2007.

• Biennial Operator Requalification Examination Results Inspection with Mr. M. Otten, Operations Training Manager, on November 19, 2007, via telephone.

• Radiological Environmental Monitoring Program (REMP) and Radioactive Material Control Program Inspection with Mr. J. Domitrovich, Maintenance Director on November 9, 2007.

• As-Low-As-Is-Reasonably-Achievable (ALARA) Planning and Controls with Mr. B. Hanson, Site Vice President on November 30, 2007.

• Emergency Preparedness Inspection with Mr. M. Friedman, Emergency Preparedness Manager, on December 18, 2007.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Hanson, Site Vice President

R. Kearney, Plant Manager

B. Campbell, Chemistry Supervisor

R. Schenck, Work Management Director

G. Vickers, Radiation Protection Manager

J. Gackstetter, Regulatory Assurance Manager

R. Frantz, Regulatory Assurance Representative

M. Hiter, Access Control Supervisor

M. Friedmann, Emergency Preparedness Manager

C. VanDenburgh, Nuclear Oversight Manager

J. Domitrovich, Maintenance Director

D. Schavey, Operations Director

J. Rappeport, Chemistry Manager

J. Lindsay, Training Director

C. Williamson, Security Manager

R. Peak, Site Engineering Director

T. Chalmers, Shift Operations Superintendent

M. Otten, Operations Training Manager

LIST OF ITEMS

Opened

05000461/2007-005-01 NCV Failure to implement fire protection program in accordance with program requirements.

05000461/2007-005-02 NCV Failure to properly test a recently installed permanent plant modification to both the Division 1 and Division 2 shutdown service water system.

Closed

05000461/2007-005-01 NCV Failure to implement fire protection program in accordance with program requirements.

05000461/2007-005-02 NCV Failure to properly test a recently installed permanent plant modification to both the Division 1 and Division 2 shutdown service water system.

Attachment