IR 05000461/2010006

IR 05000461-10-006 (Drs), on 05/24/2010 - 06/25/2010; Clinton Power Station; Component Design Bases Inspection (CDBI)

ML102210169
Person / Time
Site:	Clinton
Issue date:	08/06/2010
From:	Ann Marie Stone NRC/RGN-III/DRS/EB2
To:	Pacilio M Exelon Generation Co, Exelon Nuclear
References
IR-10-006
Download: ML102210169 (33)

Text

gust 6, 2010

SUBJECT:

CLINTON POWER STATION - COMPONENT DESIGN BASES INSPECTION (CDBI) INSPECTION REPORT 05000461/2010006(DRS)

Dear Mr. Pacilio:

On June 25, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed a component design bases inspection at your Clinton Power Station. The enclosed report documents the inspection results, which were discussed on June 25, 2010, with Mr. F. 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, two NRC-identified findings of very low safety significance were identified. The findings involved a violation of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations (NCVs) in accordance with Section VI.A.1 of the NRC Enforcement Policy.

If you contest the subject or severity of a NCV, 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 20555-0001, 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 Inspectors Office at the Clinton Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspectors at the Clinton Power Station. 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 System (PARS)

component of NRC's Agencywide Documents Access and Management System (ADAMS),

accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-461 License No. NPF-62

Enclosure:

Inspection Report No. 05000461/2010-006 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-461 License No: NPF-62 Report No: 05000461/2010-006(DRS)

Licensee: Exelon Generation Company, LLC Facility: Clinton Power Station Location: Clinton, IL Dates: May 24 through June 25, 2010 Inspectors: A. Dunlop, Senior Reactor Engineer, Lead J. Gilliam, Reactor Engineer, Electrical Engineer T. Hartman, Reactor Engineer, Operations Inspector N. Della Greca, Electrical Contractor C. Baron, Mechanical Contractor C. Edwards, Mechanical Contractor Approved by: Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000461/2010006(DRS); 05/24/2010 - 06/25/2010; Clinton Power Station; Component

Design Bases Inspection (CDBI)

The inspection was a 3-week onsite baseline inspection that focused on the design of components that are risk-significant and have low design margin. The inspection was conducted by regional engineering inspectors and three consultants. Two Green findings were identified by the inspectors. The findings were considered Non-Cited Violations (NCVs) of NRC regulations. The significance of most findings are indicated by their color (Green, White, Yellow,

Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or may be assigned a severity level after NRC management review. 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.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI,

Test Control, having very low safety significance for the licensees failure to ensure adequate acceptance limits were incorporated into test procedures. Specifically, the licensee failed to properly consider instrument loop uncertainties and allowable emergency diesel generator frequency variance when determining the alert and required action values used in the inservice test procedure for testing of the residual heat removal pumps. Consequently, the acceptance criteria for the lower limits on degradation of pump head were non-conservative. This finding was entered into the licensees corrective action program and a preliminary calculation performed by the licensee concluded that the pumps were operable.

The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of the system to respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance (Green)because the licensee was able to demonstrate pump operability and therefore, there was no loss of safety function. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate operating experience that included similar issues relating to the failure to appropriately account for instrument uncertainties in design analysis. P.2(b) (Section 1R21.3.b.(1))

Green.

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI,

Test Control, having very low safety significance for the licensees failure to establish test conditions to assure that the 1B residual heat removal heat exchanger would perform satisfactorily in service under accident conditions. Specifically, the inspectors determined that the heat exchanger thermal performance test procedure did not assure adequate temperature differences to provide reliable test results. In addition, the most recent test was performed with lower temperature differences than those identified in plant calculations. This finding was entered into the licensees corrective action program and a preliminary analysis performed by the licensee concluded the test results were acceptable.

The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the residual heat removal heat exchanger performance test procedure did not establish appropriate test conditions to ensure that the component would perform its required function during an accident.

Also, the inspectors determined that the finding was similar to Examples 3.j and 3.k of IMC 612, Appendix E, in that there was a reasonable doubt of the operability of the component based on the most recent test conditions. The inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a cross-cutting aspect because it did not represent current performance. (Section 1R21.3.b.(2))

REPORT DETAILS

1R21 Component Design Bases Inspection

.1 Introduction

The objective of the component design bases inspection is to verify that design bases have been correctly implemented for the selected risk significant components and that operating procedures and operator actions are consistent with design and licensing bases. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The Probabilistic Risk-Assessment (PRA) model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for which there are no indicators to measure performance.

Specific documents reviewed during the inspection are listed in the Attachment to the report.

.2 Inspection Sample Selection Process

The inspectors selected risk-significant components and operator actions for review using information contained in the licensees PRA and the Clinton Power Station Standardized Plant Analysis Risk (SPAR) Model. In general, the selection was based upon the components and operator actions having a risk achievement worth of greater than 1.3 and/or a risk reduction worth greater than 1.005. The operator actions selected for review included actions taken by operators both inside and outside of the control room during postulated accident scenarios. In addition, the inspectors selected operating experience issues associated with the selected components.

The inspectors performed a margin assessment and a detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design reductions caused by design modification, or power uprates, or reductions due to degraded material condition. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as performance test results, significant corrective action, repeated maintenance activities, Maintenance Rule (a)(1) status, components requiring an operability evaluation, NRC resident inspectors input of problem areas/equipment, and system health reports. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. A summary of the reviews performed and the specific inspection findings identified are included in the following sections of this report.

This inspection constituted 30 samples as defined in Inspection Procedure 71111.21-05.

.3 Component Design

a. Inspection Scope

The inspectors reviewed the Updated Safety Analysis Report (USAR), Technical Specifications (TS), design basis documents, drawings, calculations and other available design basis information, to determine the performance requirements of the selected components. The inspectors used applicable industry standards, such as the American Society of Mechanical Engineers (ASME) Code, Institute of Electrical and Electronics Engineers (IEEE) Standards and the National Electric Code, to evaluate acceptability of the systems design. The NRC also evaluated licensee actions, if any, taken in response to NRC issued operating experience, such as Bulletins, Generic Letters (GLs),

Regulatory Issue Summaries (RISs), and Information Notices (INs). The review was to verify that the selected components would function as designed when required and support proper operation of the associated systems. The attributes that were needed for a component to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes to verify that the component condition and tested capability was consistent with the design bases and was appropriate may include installed configuration, system operation, detailed design, system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation.

For each of the components selected, the inspectors reviewed the maintenance history, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee corrective action program documents.

Field walkdowns were conducted for all accessible components to assess material condition and to verify that the as-built condition was consistent with the design. Other attributes reviewed are included as part of the scope for each individual component.

The following 19 components were reviewed:

• Residual Heat Removal (RHR) Pump 1B (1E12C002B): The inspectors reviewed the pump procurement specification and the design basis hydraulic analysis/calculations to verify that required total developed head (TDH), required net positive suction head (NPSH) and potential for vortex formation have been properly considered under all design basis accident/event conditions. The RHR pump inservice test (IST) procedures, recent test results, and trends in test data were reviewed to verify that component performance remains consistent with design basis requirements. The IST reference values (i.e., flow rate and developed head) were also reviewed to verify appropriate correlation to accident analyses conditions, taking into account set point tolerances and instrument inaccuracies. Documentation was reviewed to verify pump motor design was consistent with environmental qualification (EQ) basis for limiting conditions. The inspectors reviewed the pump room cooler and pump seal oil cooler differential pressure test and inspection procedures, including test/inspection results, to verify compliance with licensing commitments under GL 89-13, Service Water System Problems Affecting Safety-Related Equipment, program plan.

• RHR Heat Exchanger 1B (1E12B001B): The inspectors reviewed the design basis documentation, including procurement specifications and Tubular Heat Exchanger Manufacturers Association (TEMA) data sheet, and heat exchanger analysis to verify equipment heat removal capability under design basis conditions. The inspectors reviewed the heat exchanger inspection and thermal test procedures, including recent inspection/test results, and trending data to assess the licensees efforts to maintain the performance capability of this equipment. The licensees tube plugging analysis was reviewed to confirm that adequate margin on heat transfer capability had been maintained after recent maintenance activities to plug a significant number of tubes.

• RHR Pump 1B Low Pressure Coolant Injection Isolation Valve (1E12F042B):

The inspectors reviewed motor-operated valve (MOV) calculations and analysis to ensure the valve was capable of functioning under design conditions. These included calculations for required thrust, maximum differential pressure, and valve weak link analysis. Diagnostic testing and IST surveillance results, including stroke time and available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified.

Documentation was reviewed to verify valve motor design was consistent with EQ basis for limiting conditions. The inspectors also reviewed electrical calculations relating to actuator minimum terminal voltage under degraded voltage conditions and thermal overload sizing methodology.

• Standby Liquid Control (SLC) Pump 1B (1C42C001B): The inspectors reviewed the SLC system design basis hydraulic analysis/calculations to verify that required TDH, required NPSH, and potential for vortex formation have been properly considered under all design basis accident/event conditions. The SLC pump IST procedures, recent test results, and trends in test data were reviewed to verify that component performance remains consistent with design basis requirements. The IST reference values were also reviewed to verify appropriate correlation to accident analyses conditions, taking into account set point tolerances and instrument inaccuracies. Documentation was reviewed to verify pump motor design was consistent with EQ basis for limiting conditions.

• High Pressure Core Spray Suppression Pool Minimum Flow Bypass Valve (1E22F012): The inspectors reviewed MOV calculations and analysis to ensure the valve was capable of functioning under design conditions. These included calculations for required thrust, maximum differential pressure, and valve weak link analysis. Diagnostic testing and IST surveillance results, including stroke time and available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified. Documentation was reviewed to verify valve motor design was consistent with EQ basis for limiting conditions.

The inspectors also reviewed electrical calculations relating to actuator minimum terminal voltage under degraded voltage conditions and thermal overload sizing methodology.

• Emergency Diesel Generator (EDG) 1B (1DG01KB): The inspectors reviewed the EDG design, including seismic qualification, to confirm that it met the system design basis requirement. The design review also addressed the EDG starting and loading sequence. This was accomplished by evaluating logic and wiring diagrams, as well as the EDG voltage and frequency control circuits. The inspectors reviewed the EDG loading calculation and confirmed that the EDG vendor ratings conformed to the design basis load requirements. Additionally, the inspectors reviewed available instrumentation and alarms and verified that the EDG was adequately protected during normal, abnormal, and emergency conditions. The inspectors reviewed the EDG performance by evaluating the system health report and completed surveillance tests to confirm that the EDG reached speed and frequency within the time established by the accident analysis and the Technical Specification. Additionally, the inspectors confirmed that the EDG was capable of accepting, rejecting, and sharing loads in accordance with the guidelines of Regulatory Guide 1.9, Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants.

• EDG 1B Air Start System: The inspectors reviewed the EDG air start system for conformance with design basis requirements. This review included design basis calculations, test procedures, and test results to verify the capability of the system to start the EDG under limiting conditions. Specifically, the inspectors reviewed the capacity of the system, reviewed periodic leakage testing and acceptance criteria, and verified that adequate air pressure was maintained. The inspectors also verified that the system was adequately protected from internal flooding hazards.

• EDG Heating, Ventilation and Air-Conditioning (HVAC) Damper (VD01YB): The inspectors reviewed the EDG HVAC Damper and associated HVAC system for conformance with design basis requirements. This review included design basis calculations to verify adequate cooling during EDG operation. The inspectors reviewed the design of the damper and associated ductwork to withstand the pressure differentials associated with a postulated tornado without loss of function. The inspectors also reviewed the control system design associated with the damper and associated equipments. The inspectors also reviewed the effects of degraded voltage conditions on minimum power and voltage requirements. The inspectors also verified separation from other trains and divisions by reviewing electrical drawings.

• Shutdown Service Water (SX) Pump 1B (1SX01PB): The inspectors reviewed design basis calculations, test procedures, and test results to verify the capability of the SX pump to supply the required service water flow to components under limiting conditions. Specifically, the inspectors reviewed the bases of the pump test acceptance criteria, the calculated performance of the pump under post accident conditions, the performance of SX system vacuum breakers, and the basis for SX system valve throttle positions. The inspectors reviewed the performance of the SX strainers under test and accident conditions, reviewed the potential loss of SX flow due to valve leakage, and reviewed the licensees monitoring of the SX piping system conditions.

• SX Outlet RHR Heat Exchanger Room 1A/B Cooler Valve (1SX023B): The inspectors reviewed the air-operated valve associated with the RHR heat exchanger for conformance with design basis requirements. This review included design basis calculations and test results to verify the capability of the valve to perform its required function. Specifically, the inspectors reviewed valve thrust calculations and stroke test results, reviewed the required air pressure to close the valve, and reviewed the setpoints of the associated air pressure regulator and rupture disc to verify the capability of the valve to perform its function under the most limiting conditions. The inspectors also reviewed the effects of degraded voltage conditions on minimum power and voltage requirements. The inspectors also verified separation from other trains and divisions by reviewing electrical drawings.

• Screenhouse HVAC Fan (1VH01CB): The inspectors reviewed the fan and associated HVAC system for the safety-related portion of the screenhouse for conformance with design basis requirements. This review included design basis calculations to verify adequate cooling during post-accident operation. The inspectors reviewed the design to verify that the system would be protected in the event of a postulated tornado. The inspectors also reviewed the control system design associated with the fan and associated equipments. The inspectors also reviewed the effects of degraded voltage conditions on minimum power and voltage requirements. The inspectors also verified separation from other trains and divisions by reviewing electrical drawings.

• Instrument Air Supply for the Automatic Depressurization System (ADS): The inspectors reviewed the portions of the instrument air system associated with operation of the ADS for conformance with design basis requirements. The inspectors reviewed design basis calculations, leakage tests, and air quality to verify that the air supply would be capable of performing its function if the normal air supply was not available under accident conditions. Specifically, the inspectors reviewed the air system capacity, pressure, and leakage limits. The inspectors also reviewed the dew point of the makeup air supply to verify the required air system quality.

• Reserve Auxiliary Transformer (RAT) B, (1AP02EB): The inspectors reviewed modification EC 339047 that replaced the existing three-winding reserve auxiliary transformer, RAT 1, with three equivalent transformers, RATs A, B, and C, for potential impact on the design basis of the auxiliary power system. The inspectors confirmed that the three transformers were adequately sized to carry the existing loads. Additionally, the inspectors reviewed the one-line diagrams to verify that the revised design conformed to the system requirements. For RAT B, the inspectors reviewed transformer design data, including nameplate, sizing, current carrying capability, and input/output voltage rating. Also, the inspectors reviewed vendor test results and verified tap positions to confirm that the correct transformer impedance and tap settings were utilized in the voltage drop and short circuit calculations. The inspectors reviewed transformer loading and short circuit calculations. The review also confirmed the adequacy of the transformer protection, including lightning, over-current, differential, and ground fault protection. The inspectors reviewed protective relay setting calculations and surveillance testing of such relays to confirm that such settings conformed to the design system requirements. The review addressed the adequacy of the instrumentation and alarms available to the operators and the adequacy of the revised bus duct design to ensure that the bus duct rating was sufficient to support the transformer loading demands.

• 4.16 kV Auxiliary Power Bus 1B1 (1AP09E): The inspectors reviewed the one-line diagrams and the loading, short circuit and voltage drop calculations to evaluate the capability of the safety-related bus to supply adequate power to the associated loads in accordance with the design and licensing bases of the system. The review addressed minimum and maximum anticipated grid voltage, transformer impedance and tap settings, availability and performance of the static var compensator, and major pump load requirements to confirm that design variables were adequately included in the analyses. The review also evaluated the switchgear design, the relay protection provided, the circuit breaker interrupting capability, and the ability of the bus to withstand maximum loading and available symmetrical and asymmetrical short circuit. The inspectors reviewed control logics and wiring diagrams of the supply breakers to confirm that manual transfers between the normal an alternate sources and between these and the emergency source utilized synchronizing equipment and that automatic transfers operated as described in the USAR. The inspectors also reviewed the degraded grid voltage analysis to confirm that, under postulated minimum grid voltage, adequate voltage was available at all safety-related components. Protective relay coordination curves were also reviewed to assure that the electrical equipment was adequately protected and that selective breaker tripping was provided under overload and faulted conditions. The inspectors reviewed surveillance testing of voltage and over-current relays to verify conformance with design calculations assumptions and conclusions.

• 480 V Auxiliary Power Bus 1B - Auxiliary Building (1AP12E): The inspectors reviewed the one-line diagrams, bus loading, short circuit, and voltage drop calculations to ensure that the safety-related 480 V load center was capable of supplying adequate voltage to the auxiliary building loads. The review included load center and circuit breaker rating, circuit breaker interrupting capacity, and capability of the bus to withstand maximum loading and available symmetrical and asymmetrical short circuit. The coordination/protection calculation for the incoming line and feeder breaker was also reviewed to confirm adequacy of load protection and selective trip coordination between these breakers. The inspectors confirmed that adequate 125 Vdc was available to the circuit breakers spring charging motor and close and trip coils to ensure opening and closing of the breakers under all modes of operation.

• Nuclear System Protection System Bus B (1C71-PC01B): The inspectors reviewed seismic qualification, voltage drop and minimum voltage calculations.

The calculation review verified methodology, design inputs, assumptions, and results.

• 125 Vdc Motor Control Center (MCC) 1B (1DC14E): The inspectors reviewed seismic qualification and various electrical calculations associated with the 1B 125 Vdc MCC. These included voltage drop, minimum voltage, and short circuit calculations. The inspectors reviewed load flow and short circuit current calculations to determine the design basis for maximum load. The calculation review also verified methodology, design inputs, assumptions, and results. The inspectors reviewed bus surveillance and preventive maintenance testing for issues that affect reliability.

• 125 Vdc 1B Battery (1DC): The inspectors reviewed seismic qualification and various electrical calculations associated with the safety-related 1B 125 Vdc batteries. These included battery sizing, voltage drop, minimum voltage, and station blackout coping. The calculation review verified methodology, design inputs, assumptions, and results. The battery surveillance, corrective actions, system health report and performance history including cell voltage, charging, specific gravity, electrolyte level, and temperature correction were also reviewed to ensure acceptance criteria were met and performance degradation would be identified.

• Balance of Plant DC Bus E: The inspectors also reviewed seismic qualification, voltage drop, and minimum voltage calculation. The calculation review verified methodology, design inputs, assumptions, and results.

b. Findings

(1) Non-Conservative Acceptance Criteria for RHR Pump Performance Testing

Introduction:

A finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, was identified by the inspectors for licensees failure to include adequate acceptance limits in test procedures. Specifically, when determining the alert and required action values used in the IST procedure for testing of the RHR pumps, the licensee failed to properly consider instrument loop uncertainties and allowable EDG frequency variance. As a result, the acceptance criteria for the lower limits on degradation of TDH were non-conservative.

Description:

The General Electric (GE) design document 762E425AC, GE Process Diagram, RHR System, stated that each RHR pump shall have a minimum TDH of approximately 275 feet at a flowrate of 5050 gallons per minute (gpm). The design basis for this value was reconstituted as part of the extended power uprate and feedwater leakage control (FWLC) modification projects to verify that current IST acceptance criteria limits for the RHR pumps would still envelope any changes in system design requirements resulting from these modifications. The inspectors review of hydraulic calculation 01RH29, Development of RHR Pump Curves and Comparison with the System Resistance Curves for Concurrent Operation of FWLC with RHR Operating Modes A-1, A-2, B-1 and B-2, determined that the licensee had failed to correct this calculation for a previously determined change in assumed post-accident EDG frequency variance from 1 to 2 percent to correspond to the variance allowed by TS.

Further, the inspectors determined that the licensee had incorrectly concluded that the original pre-operational pump performance test curve used in this calculation did not have to be conservatively adjusted (downward) to account for instrument uncertainties associated with the test data used to generate the test curve. Correcting the hydraulic analysis for these errors, the inspectors determined that for the most limiting mode of operation (combined containment spray/FWLC mode) RHR pump 1A would not be able to supply the required flow of 4079 gpm if the pump were allowed to degrade more than approximately 4.3 percent below the current IST reference value. Thus, the current IST limits in procedure CPS 9053.7, RHR B/C Pumps & RHR B/C Water Leg Pump Operability, for alert (5 percent degradation) and required action (7 percent degradation)were both below the minimum value needed to support this safety-related design function of the RHR system.

The licensee initiated Action Request (AR) 1084176 and performed an immediate operability evaluation to address the issue. Based on the most recent IST test data, the performance curve for RHR pump 1A was 0.9 percent below the reference value curve and therefore, the licensee concluded the pump remained operable. The licensee implemented an administrative limit of 3.5 percent degradation for the next inservice test of the 1A RHR pump until completion of a comprehensive evaluation. The inspectors had no further concerns with the licensees evaluation of this issue.

Analysis:

The inspectors determined that the failure to properly account for instrument uncertainties and allowable EDG frequency variance in development of the acceptance criteria for inservice testing of the RHR pumps was a performance deficiency.

The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to consider instrument uncertainties and EDG frequency variance in the development of IST acceptance criteria resulted in the establishment of acceptance criteria values that did not ensure that the RHR pumps could meet their intended safety function.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 -

Initial Screening and Characterization of findings, Table 3b for the Mitigating System cornerstone. The finding screened as very low safety significance (Green) because the finding was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed an operability evaluation that concluded the RHR 1A pump actual degradation was only 0.9 percent, which was less than the preliminary allowable degradation limit of 4.3 percent. As such, sufficient margin existed to ensure the RHR system would be capable of successfully performing the combined containment spray/FWLC mode functions.

This finding has a cross-cutting aspect in the area of problem identification and resolution, because the licensee did not thoroughly evaluate IN 2008-02 Findings Identified during Component Design Bases Inspections, which included similar issues relating to the failure to appropriately account for instrument uncertainties in design analysis. P.2(b)

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, requires, in part, 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.

Contrary to the above, from 1999 until present, the licensee failed to incorporate adequate acceptance limits in IST test procedures. Specifically, instrument loop uncertainties and allowable EDG frequency variance were not adequately included when the new acceptance criteria were established in procedure CPS 9053.7 for the inservice testing of the RHR pumps after installing FWLC modifications. Because this violation was of very low safety significance and it was entered into the licensees corrective action program as AR 1084176, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000461/2010006-01, Non-Conservative Acceptance Criteria for RHR Pump Performance Testing)

(2) Inadequate Test Control of RHR Heat Exchangers

Introduction:

A finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, was identified by the inspectors for the licensees failure to establish test conditions to assure that the 1B RHR heat exchanger would perform satisfactorily in service under accident conditions.

Specifically, the inspectors determined that the heat exchanger thermal performance test procedure did not assure adequate temperature differences to provide reliable test results. In addition, the most recent test was performed with lower temperature differences than those identified in plant calculations.

Description:

The inspectors identified a performance deficiency related to the periodic thermal performance testing of the 1B RHR heat exchanger. This heat exchanger had a significant number of tubes plugged that exceeded the number assumed by the vendor in their determination of its heat removal capability. As a result, GE performed an analysis that exchanged allowed number of plugged tubes for lower tube fouling. The assumed fouling allowed by this analysis appeared unrealistically low based on the service water source being lake water and would be difficult to verify through testing.

The heat exchanger was tested every two years to verify its thermal performance in accordance with the licensees Generic Letter 89-13 program. The test was normally performed with the plant online, using the temperature difference between the lake (ultimate heat sink) and the suppression pool to provide a heat load for the test. The test included measuring both flows and temperatures under test conditions, then performing an analysis to verify that the heat exchangers performance would be adequate under accident conditions. The test was normally performed with significantly lower temperature differences than the design basis accident condition.

Test procedure CPS 2700.20, RHR A(B) Heat Exchanger Thermal Performance Test Covered by NRC Generic Letter 89-13, Revision 4, provided instructions for collecting RHR heat exchanger test data. The inspectors noted that this procedure did not include a required minimum temperature difference or a required minimum suppression pool temperature to perform this test. The procedure did include a note directing the suppression pool temperature to be increased to 93 degrees Fahrenheit (°F), as required, and stating that a lower temperature may be used at the discretion of the cognizant/test engineer. An additional note, added to revision 4 of the procedure (June 1, 2010), stated that if the suppression pool temperature is not increased then this temperature difference is recommended to be 25°F and optimally 30°F. The inspectors were concerned that this procedure would not ensure that reliable test data would be obtained with relatively small temperature differences and requested the basis of the minimum recommended temperature differences.

The licensee provided excerpts from calculations that included pre-test temperature measurement uncertainty. Calculations 065-017, Summary Report, Clinton GL-89-13 Program Support, Revision 3, and 065-019, RHR & DG Heat Exchanger Testing Specification & Acceptance Criteria, Revision 3, addressed the expected measurement uncertainties associated with this test; these calculations were based on an assumed minimum temperature difference of 30°F. The licensee stated that they had not performed formal analyses to support performance of this test with temperature differences of less than 30°F. Also, calculation 065-019 stated that if the difference is less than 30°F, the uncertainties start to substantially increase.

The inspectors determined that the most recent test conducted on December 11, 2008, had been performed with an average temperature difference of approximately 23°F. The licensee stated that the test condition had been evaluated prior to the test and found to be acceptable based on an informal evaluation. The post-test data evaluation for that test (EC 373382) concluded that the heat exchanger performance was acceptable with a margin of approximately 5 percent; however, discussions with engineering personnel indicated that the post-test evaluation had been based on less conservative temperature measurement uncertainties than the original calculations. An informal sensitivity analyses, performed by engineering personnel during the inspection, concluded that the corrected temperature measurement uncertainty would be slightly smaller than the predicted difference between a clean condition and a heat exchanger at the fouling limit. This analysis verified that the 2008 test results did demonstrate operability.

However, the 2008 post-test data evaluation did not include an appropriate temperature measurement uncertainty analysis and would not have reliably identified a heat exchanger in an inoperable condition. In response to the inspectors concern, the licensee initiated AR 1083290.

Analysis:

The inspectors determined that the failure to assure that the RHR heat exchanger would perform its design function was a performance deficiency that was reasonably within the licensees ability to foresee and prevent. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the RHR heat exchanger performance test procedure did not ensure that the component would perform its required function during an accident. Also, the inspectors determined that the finding was similar to Examples 3.j and 3.k of IMC 612, Appendix E, in that there was a reasonable doubt of the operability of the component based on the most recent test conditions. Additional analyses were required to verify that the component would be capable of performing its design function under limiting conditions.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of findings, Table 3b for the Mitigating System cornerstone. The finding screened as very low safety significance (Green)because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Additional informal analyses, performed during the inspection, demonstrated that the component was operable. This finding did not have a cross-cutting aspect because it did not represent current performance.

Enforcement:

Title 10 CFR 50, Appendix B, Criterion XI, Test Control, requires, in part, 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 performed in accordance with written test procedures and test results are documented and evaluated to assure that test requirements have been satisfied. Test procedures shall include provisions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions.

Contrary to the above, as of June 25, 2010, the licensees procedure CPS 2700.20 did not establish adequate test conditions to assure that the RHR heat exchanger would perform satisfactorily in service under accident conditions. Because this violation was of very low safety significance and it was entered into the licensees corrective action program as AR 1083290, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000461/2010006-02, Inadequate Test Control of RHR Heat Exchangers)

.4 Operating Experience

a. Inspection Scope

The inspectors reviewed 7 operating experience issues to ensure that NRC generic concerns had been adequately evaluated and addressed by the licensee. The operating experience issues listed below were reviewed as part of this inspection:

• IN 1992-27, Supplement 1, Thermally Induced Accelerated Aging and Failures of ITE/Gould AC Relays Used in Safety-Related Applications;

• IN 2005-30, Safe Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events and Inadequate Design;

• IN 2007-05, Vertical Deep Draft Pump Shaft and Coupling Failures;

• IN 2007-34, Operating Experience Regarding Electrical Circuit Breakers;

• IN 2008-02, Findings Identified During Component Design Bases Inspections;

• IN 2008-13, Main Feedwater System Issues and Related 2007 Reactor Trip Data; and

• IN 2009-14, Painting Activities and Cleaning Agents Render Emergency Diesel Generators and other Plant Equipment Inoperable.

b. Findings

No findings of significance were identified.

.5 Modifications

a. Inspection Scope

The inspectors reviewed 4 permanent plant modifications related to selected risk significant components to verify that the design bases, licensing bases, and performance capability of the components had not been degraded through modifications. The modifications listed below were reviewed as part of this inspection effort:

• ECN 28511, Drill 1/8 Inch Hole in 1E12F042C Disc;

• EC 339047, Reserve Auxiliary Transformer (RAT) Replacement Project;

• FECN 27062, SLC Tank Level Setpoint Change for Adequate NPSH; and

• WO 1124073-01, Replace 1DC02E Battery.

b. Findings

No findings of significance were identified.

.6 Risk Significant Operator Actions

a. Inspection Scope

The inspectors performed a margin assessment and detailed review of four risk significant, time critical operator actions. These actions were selected from the licensees PRA rankings of human action importance based on risk achievement worth values. Where possible, margins were determined by the review of the assumed design basis and USAR response times and performance times documented by job performance measures results. For the selected operator actions, the inspectors performed a detailed review and walk through of associated procedures, including observing the performance of some actions in the stations simulator and in the plant for other actions, with an appropriate plant operator to assess operator knowledge level, adequacy of procedures, and availability of special equipment where required.

The following operator actions were reviewed:

• Operator Fails to Align Division 3 EDG to Division 1 or 2;

• Operators Fail to Align SX Injection Through RHR B;

• Operator Fails to Both Lower Reactor Pressure Vessel Level and Control Level in an Anticipated Transient Without Scram; and

• Failure of Operator to Remove Q/S Relay to Prevent 1M-MCC Shunt Trip.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Review of Items Entered Into the Corrective Action Program

a. Inspection Scope

The inspectors reviewed a sample of the selected component problems that were identified by the licensee and entered into the corrective action program. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action program. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

4OA6 Meeting(s)

.1 Exit Meeting Summary

On June 25, 2010, the inspectors presented the inspection results to Mr. F. Kearney, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. Several documents reviewed by the inspectors were considered proprietary information and were either returned to the licensee or handled in accordance with NRC policy on proprietary information.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

F. Kearney, Site Vice President

M. Kanavos, Plant Manager

K. Baker, Senior Manager Design Engineering

S. Clary, Engineering Programs Manager

T. Chalmers, Operations Director

S. Clary, Engineering Programs Manager

J. Coombs, MOV Engineer

B. Corley, Operations

C. Culp, GL 89-13 Engineer

B. Davis, Senior Manager Plant Engineering

S. Fatora, Maintenance Director

R. Frantz, Regulatory Assurance

S. Gackstetter, Training Director

M. Gandi, Design Engineer

A. Hable, Probabilistic Risk Assessment Engineer

J. Hall, Plant Engineering

M. Heger, Mechanical/Structural Design Engineering Manager

S. Lakebrink, Design Engineering

D. Kemper, Regulatory Assurance Manager

M. Kimmich, Plant Engineering

S. Kowalski, Engineering Response Manager

J. Mosley, Electrical Design

F. Pournia, Site Engineering Director

D. Smith, Plant Engineering

C. VanDenburgh, Nuclear Oversight Manager

Nuclear Regulatory Commission

B. Kemker, Senior Resident Inspector

D. Lords, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened/Closed

000461/2010006-01 NCV Non-Conservative Acceptance Criteria for RHR Pump Performance Testing 000461/2010006-02 NCV Inadequate Test Control of RHR Heat Exchangers Attachment