IR 05000461/2005002

IR 05000461-05-002(DRS); Clinton Power Station; on 11/14/2005 Through 01/20/2006; Safety System Design and Performance Capability Inspection

ML060670370
Person / Time
Site:	Clinton
Issue date:	03/06/2006
From:	Ann Marie Stone NRC/RGN-III/DRS/EB2
To:	Crane C Exelon Generation Co
References
IR-05-002
Download: ML060670370 (31)

Text

rch 6, 2006

SUBJECT:

CLINTON POWER STATION, NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY, INSPECTION REPORT 05000461/2005002(DRS)

Dear Mr. Crane:

On January 20, 2006, the U. S. Nuclear Regulatory Commission (NRC) completed a safety system design and performance capability inspection at your Clinton Power Station. The enclosed inspection report documents the inspection results, which were discussed at an interim exit meeting held on December 2, 2005, and during an exit meeting held by telephone on January 20, 2006, with Mr. R. Bement and other members of your staff.

The 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. Specifically, this inspection focused on the design and performance capability of the high pressure core spray system and its support systems to ensure that they were capable of performing their required safety related functions.

Based on the results of this inspection, two NRC identified findings of very low safety significance, all of which involved violations of NRC requirements were identified. However, because these violations were of very low safety significance, and because the findings were entered into the licensees corrective action program, the NRC is treating these findings as Non-Cited Violations in accordance with Section VI.A.1 of the NRCs 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 20555-0001, 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 Inspector Office at the Clinton Power Station facility.

In accordance with 10 CFR 2.390 of the NRC's "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 Publically Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-461 License No. NPF-62 Enclosure: Inspection Report 05000461/2005002(DRS)

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 Senior Vice President - Nuclear Services Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Manager Licensing - Clinton Power Station Senior Counsel, Nuclear, Mid-West Regional Operating Group Document Control Desk - Licensing

SUMMARY OF FINDINGS

IR 05000461/2005002(DRS); 11/14/2005 - 01/20/2006; Clinton Power Station; Safety System

Design and Performance Capability Inspection.

This report covers a 3 week period of announced baseline inspection on the design and performance capability of the high pressure core spray (HPCS) system and support systems.

The inspection was conducted by Region III inspectors, the resident inspector and a mechanical engineering consultant. Two Green findings associated with two non-cited violations were identified. The significance of most findings is 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 be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A. Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" requirements.

Specifically, the licensee failed to incorporate the most restrictive hydraulic conditions into the calculation which established the acceptance criteria for a technical specification surveillance test. This resulted in a HPCS system hydraulic calculation that was non-conservative when determining the pumps minimum acceptance criteria. Once identified, the licensee evaluated operability and entered the finding into their corrective action program to revise the affected documents.

The finding was more than minor because the failure to account for all modes of HPCS system operation in the surveillance tests acceptance criteria could result in unacceptable degradation and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin existed for the HPCS system and did not represent an actual loss of a safety function. (Section 1R21.2b)

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" requirements.

Specifically, in 2000, 2002 and 2003, the licensee failed to recognize that the calculated value for the diesel generator (DG) jacket-water (JW) flow rate, as determined from test data obtained during thermal performance testing of the division III DG JW cooler heat exchanger (HX), was significantly higher than the flow rate that could be attained by the engine-driven water pump. Once identified, the licensee entered the finding into their corrective action program as Condition Report (CR) 426459, NRC SSD&PC Is the Calculated Process Flow Rate Reasonable, dated November 21, 2005, and CR429726,

Discrepancies Not Identified in Corrective Action Process, dated December 2, 2005, to evaluate and/or revise the affected test procedures.

The finding was more than minor because the failure to account for flow rates that were significantly greater than that identified by the equipments design specification produced equipment performance data that did not accurately demonstrate the HXs availability and reliability. The finding was of very low safety significance because the licensees evaluation showed that the Division III DGs JW Cooler HX would have performed its safety function and did not represent an actual loss of a safety function. A contributing cause of the finding was related to the cross-cutting element of problem identification and resolution. Specifically, a similar issue was identified during another NRC inspection in 2001; however, the licensee did not properly evaluate and take actions. As a result, testing done in 2002 and 2003 showed the same discrepant flow rates. (Section 1R21.3b)

Licensee-Identified Violations

None.

REPORT DETAILS

1R21 Safety System Design and Performance Capability

Introduction Inspection of safety system design and performance verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected systems to perform design bases functions. As plants age, the design bases may be lost and important design features may be altered or disabled. The plants risk assessment model was based on the capability of the as-built safety system to perform the intended safety functions successfully. This inspectable area verifies aspects of the mitigating systems cornerstone for which there are no indicators to measure performance.

The objective of the safety system design and performance capability inspection was to assess the adequacy of calculations, analyses, other engineering documents, and operational and testing practices that were used to support the performance of the selected systems during normal, abnormal, and accident conditions.

The system and components selected were from the high pressure core spray (HPCS)system. This system was selected for review based upon:

• having a high probabilistic risk analysis ranking;

• having had recent significant issues;

• not having received recent NRC review; and

• being interacting systems.

The criteria used to determine the acceptability of the systems performance was found in documents such as:

• applicable technical specifications (TS);

• applicable updated safety analysis report (USAR) sections; and

• the systems' design documents.

.1 System Requirements

a. Inspection Scope

The inspectors reviewed the USAR, TS, system descriptions, drawings and available design basis information to determine the performance requirements of the HPCS system. The reviewed system attributes included process medium, energy sources, control systems, operator actions and heat removal. The rationale for reviewing each of the attributes was:

Process Medium: This attribute required review to ensure that the HPCS flow paths would be available and unimpeded during and/or following design basis events. To achieve this function, the inspectors verified that the system(s) would be aligned and maintained in an operable condition as described in the plants USAR, TS and design bases. In addition, the inspectors reviewed and verified for adequacy the:

• design basis calculations for flow rates, levels, pressures and temperatures,

• total dynamic head and net positive suction head (NPSH),

• alternate water source(s) capacity and

• pipe stress analysis results.

Energy Sources: This attribute required review to ensure that the HPCS motive/electrical source would be available/adequate and unimpeded during/following design basis events, that appropriate valves and system control functions would have sufficient power to change state when required. To achieve this function, the inspectors verified that interactions between HPCS and its support system(s) were appropriate, such that, all components would operate properly when required. To complete this attribute, the inspectors reviewed and verified for adequacy the:

• 125Vdc battery capacity,

• 4.16kV system capacity,

• breaker coordination,

• fuse coordination,

• voltage drop calculations,

• undervoltage (UV) calculations,

• degraded voltage calculations,

• air reservoir capacity (for air operated equipment),

• instrument air availability was as needed and that

• power was available to support operation of the HPCS system and support system(s).

Controls: This attribute required review to ensure that the automatic controls for operating HPCS and associated systems were properly established and maintained.

Additionally, review of alarms and indicators were necessary to ensure that operator actions would be accomplished in accordance with design requirements. To complete this attribute, the inspectors reviewed and verified for adequacy:

• setpoints established to ensure sufficient water inventory and prevent loss of required NPSH,

• instrument uncertainty & loop error calculations,

• relay setting calculations,

• setpoint calculations and

• controls functionality.

Operations: This attribute was reviewed because the operators perform a number of actions during normal, abnormal and emergency operating conditions that have the potential to affect HPCS operation. In addition, the emergency operating procedures (EOPs) require the operators to manually realign the systems flow paths during and following design basis events. Therefore, operator actions play an important role in the ability of the selected systems to achieve their safety related functions. To complete this attribute, the inspectors reviewed and verified for adequacy the following:

• Operating procedures (normal, abnormal, or emergency) to ensure they were consistent with operator actions for accident and/or event conditions.

• Operating procedure timing for manual actions were initiated within the assumed time periods and that testing was performed to validate the procedures consistent with design basis assumptions.

• Instrumentation and alarms were available to operators for making necessary decisions.

• Alarms and level instrumentation provided operators with sufficient information to perform the task and operability determinations supported calculations.

Heat Removal: This attribute was reviewed to ensure that there was adequate and sufficient heat removal capability for HPCS. To complete this attribute, the inspectors reviewed and verified:

• heat exchanger (HX) heat removal design calculations (e.g. lube oil cooler, room cooler) and

• HVAC calculations.

b. Finding Vortex Analysis Methodology Not Appropriate

Introduction:

The inspectors identified an unresolved item (URI) concerning the reactor core isolation cooling (RCIC) water storage tank volumes design analysis. Specifically, the inspectors identified that the licensee did not select an appropriate method for calculating the onset of vortexing at the intake of the HPCS and RCIC pumps suction lines from the RCIC water storage tank.

Description:

The inspectors reviewed Calculation IP-M-0384, Evaluation of Vortex in the RCIC [Water] Storage Tank, Revision 1. The purpose of the calculation was to determine the appropriate analytical level (i.e., elevation of water) where vortexing would occur above the HPCS and RCIC pumps suction lines. The analytical level was then used as a design input to calculate the automatic RCIC water storage tank to suppression pool low level switchover setpoint for the HPCS and RCIC pumps.

The inspectors noted that the methodology used in Calculation IP-M-0384 to determine the minimum height of water above the HPCS and RCIC pumps intake lines to preclude vortex formation was not appropriate. The calculations methodology did not account for the actual fluid configuration where air ingestion into the HPCS and RCIC pumps suction lines would potentially occur. The onset of vortexing was calculated using a methodology extrapolated from test data contained in NUREG/CR-2772, Hydraulic Performance of Pump Suction Inlets for Emergency Core Cooling Systems in Boiling Water Reactors, June, 1982. The extrapolated test data used in Calculation IP-M-0384 was that of a straight line drawn between two points on a graph of void fraction versus Froude Number selected from the subject NUREG. The graph was generated from test data where the minimum water submergence from the centerline of the suction pipe was at least 2-feet and with uniform approach flow (i.e., no water swirl at the suction line).

The inspectors requested the licensee to provide justification for their use of the test data from the subject NUREG to predict the onset of vortexing. In particular, the inspectors requested the licensee to justify why the minimum submergence of 2-feet of water and with no water swirling at the pump inlet, as evaluated in the subject NUREG, would be similar to the piping configuration in the licensees RCIC water storage tank.

The RCIC water storage tank had no design feature to prevent swirling of the water and the calculated submergence from Calculation IP-M-0384 was 9.36-inches from the centerline of the HPCS suction line (i.e., 1.93-inches from the top of the HPCS suction line), compared to at least 2-feet as described in the subject NUREG.

The licensee was unable to provide adequate technical justification for the methodology used and stated they would consider other methods applicable to this configuration that were more readily accepted by the industry. The licensee entered the finding into their corrective action program as Condition Report (CR) 429583, NRC SSD&PC RCIC

[Water] Tank Vortex Issue, dated December 1, 2005, to evaluate (i.e., perform an operability evaluation) and revise the affected documentation.

On December 1, 2005, the licensee shifted the HPCS and RCIC inventory source to the suppression pool as a conservative measure since the inspectors concern was specifically link to the RCIC water storage tank. The use of the suppression pool as a qualified inventory source was allowed per Clintons USAR and TS. Vortexing from the suppression pool should not occur due to the depth of the HPCS and RCIC suction lines.

Subsequent to the NRCs Interim Exit on December 2, 2005, the licensee completed Minor Revision 1/A to Calculation IP-M-0384 dated December 9, 2005, which used a different approach to determine the onset of vortexing for the RCIC pump. The revised calculation indicated that the RCIC pump suction line would have adequate submergence with the current low level switchover setpoint. However, a preliminary calculation for the HPCS pump indicated that an additional 9-inches of water over the top of the suction line would be required. The inspectors questioned the preliminary calculations value used for HPCS pump runout flow (5010 gpm versus 5650 gpm analyzed runout flow) and why there was no allowance for tank level change in the calculation for stroking of suction valves during realignment of pump suction sources.

On December 19, 2005, the licensee completed Minor Revision 1/B to Calculation IP-M-0384, which used a similar approach to determine the onset of vortexing for the HPCS pump. The results of this calculation determined that air entrainment was possible with the plants existing vortex limit (i.e., historical low level switchover setpoint). As a result, the licensee developed a RELAP5 Mod 3.3 model of the HPCS suction piping from the RCIC water storage tank to the suppression pool to evaluate the introduction and transport of air in the HPCS suction piping. A number of scenarios were analyzed to evaluate the affects of air entrainment on the HPCS pumps performance. Based on the RELAP5 model, the licensee concluded that the HPCS system would have been operable with the historical low level switchover setpoint.

However, the model did not support the historical low level switchover setpoint as an acceptable design setpoint for future operation. As a result, the licensee entered this finding into their corrective action program as CR 435174, Need to Recover RCIC and HPCS Vortex Margin, dated December 19, 2005.

The inspectors had not completed a review of the licensees re-analysis by the end of the inspection. Therefore, this issue is considered an unresolved item (URI)05000461/2005002-01(DRS) pending completion of this review.

.2 System Condition and Capability

a. Inspection Scope

The inspectors reviewed design basis documents and plant drawings, abnormal and EOP, requirements, and commitments identified in the USAR and TS. The inspectors compared the information in these documents to applicable electrical, instrumentation and control, and mechanical calculations, setpoint changes and plant modifications.

The inspectors also reviewed operational procedures to verify that instructions to operators were consistent with design assumptions.

The inspectors reviewed information to verify that the actual system condition and tested capability was consistent with the identified design bases. Specifically, the inspectors reviewed the installed configuration, the system operation, the detailed design, and the system testing, as described below.

Installed Configuration: To complete this attribute, the inspectors reviewed and verified that the installed configuration of the HPCS system met the design basis by performing detailed system walkdowns. The walkdowns focused on the installation and configuration of piping, components, and instruments; the placement of protective barriers and systems; the susceptibility to flooding, fire, or other environmental concerns; physical separation; provisions for seismic and other pressure transient concerns; and the conformance of the currently installed configuration of the systems with the design and licensing bases.

Operation: To complete this attribute, the inspectors performed procedure walk-throughs of selected manual operator actions to confirm that the operators had the knowledge and tools necessary to accomplish actions credited in the design basis; operation and system alignments were consistent with design and licensing basis assumptions and; emergency operating procedure changes had not impacted design assumptions and requirements.

Design: To complete this attribute, the inspectors reviewed the mechanical, electrical and instrumentation design of the HPCS system to verify that the systems and subsystems would function as required under accident conditions. The review included a review of the design basis, design changes, design assumptions, calculations, boundary conditions, and models as well as a review of selected modification packages.

Instrumentation was reviewed to verify appropriateness of applications and setpoints based on the required equipment function. In addition, the inspectors performed limited analyses in several areas to verify the appropriateness of the design values used for:

• pressure transient/water hammer evaluations,

• relief valve sizing calculations,

• tank sizing calculations,

• tank over-pressurization calculations and

• motor operated valve (MOV) - air operated valve calculations.

Testing: To complete this attribute, the inspectors reviewed records of selected periodic testing and calibration procedures and results to verify that the design requirements of calculations, drawings, and procedures were incorporated in the system and were adequately demonstrated by test results. Test results were also reviewed to ensure automatic initiations occurred within required times and that testing was consistent with design basis information.

b. Finding Non-Conservative Acceptance Criteria

Introduction:

The inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" having very low safety significance (Green)involving the HPCS systems hydraulic design analysis. Specifically, the inspectors identified that the licensee failed to correctly specify the minimum pump operability limits to be used in HPCS system surveillance testing.

Description:

The inspectors reviewed Calculation 01HP09, TS Surveillance Requirement for HPCS Pump Differential Pressure at Rated Flow (EC336808),

Revision 6. The purpose of this calculation was to develop HPCS pump curves to be used in IST procedures when testing the HPCS pump. The inspectors also reviewed Calculation 01HP15, Development of HPCS Pump Curves (1E22C001) & Comparison with System Resistance Curves for Operating Modes A, B, C, CC, E, F, G, & H, Revision 2. The purpose of Calculation 01HP15 was to develop pump curves for the HPCS pump and compare the pump curves to the system resistance curves for Operating Modes A, B, C, CC, E, F, G, and H. The inspectors review of Calculation 01HP15, identified that Calculation 01HP09, which established the HPCS pumps minimum acceptance criteria, to be used during testing, did not evaluate the most limiting hydraulic system resistance in which the HPCS pump was required to operate.

In particular, the HPCS system had the following modes of operation:

Mode Description (1 Accident / 2 System Test)

A Reactor at High Pressure - Suction from the RCIC Water Storage Tank1 B Reactor at High Pressure - Suction from the Suppression Pool1 Mode Description (1 Accident / 2 System Test)

C System Injection at Rated Core Spray Flow - Suction from the Suppression Pool1 CC Reactor at High Pressure, Split Flow - Suction from the Suppression Pool1 E System Injection at Rated Core Spray Flow - Suction from the RCIC Water Storage Tank1 F System at Runout - Suction from the Suppression Pool to the Reactor Vessel1 G Suction from the Suppression Pool - Discharging Back to the Suppression Pool2 H Suction from the RCIC Water Storage Tank - Discharging Back to the RCIC Water Storage Tank2 The inspectors noted that the results of Calculation 01HP15 indicated that the hydraulic requirements of Modes F, G and H were less restrictive than the test basis. However, the hydraulic requirements for Modes A, B and CC were more restrictive than the test basis. Because of this, the inspectors concluded that it was possible for pump degradation to be acceptable using the test basis, but may not be acceptable in Modes A, B, and CC. By not accounting for the higher head and lower flow requirements of Modes A, B and CC, Calculation 01HP09 was non-conservative when calculating the allowable degradation of the pump curve.

The licensee agreed that the pumps minimum acceptance criteria for the test basis based on Modes C and E was non-conservative when compared to the requirements based on Modes A, B, and CC. The inspectors reviewed the most recent pump tests and determined that adequate design margin remained between the higher minimum test points and current operating points. As a result, the inspectors concluded the HPCS system was operable.

The licensee determined that Calculation 01HP09 required revision to include the hydraulic evaluation of all modes of HPCS system operation that were evaluated in Calculation 01HP15. In addition, because Calculation 01HP09 determined the minimum acceptance criteria for HPCS system surveillance testing, the associated procedures would require revision if the acceptance criteria changed.

Analysis:

The inspectors determined that failure to correctly specify the minimum pump operability limits to be used in HPCS system surveillance testing was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on September 30, 2005. The finding involved the attribute of design control, where failure to account for all modes of HPCS system operation in the surveillance tests acceptance criteria could result in not identifying unacceptable degradation and could have affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that despite the loss of design margin in the HPCS flow delivery for the high head and low flow mode of operation, the HPCS system would have performed its safety function. Therefore, the inspectors concluded that the finding was a design deficiency that did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of December 2, 2005, the licensee failed to assure that the minimum pump operability limits as defined by design calculations were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the hydraulic requirements for the HPCS pump under high head and low flow conditions, as was determined in Calculation 01HP15, were not translated into Calculation 01HP09, TS Surveillance Requirement for HPCS Pump Differential Pressure at Rated Flow (EC336808), Revision 6, or subsequently into the routine testing surveillance. Once identified, the licensee entered the finding into their corrective action program as CR429366, SSD&PC - HPCS Pump Surveillance Acceptance Criteria Concern, dated December 1, 2005, to revise the affected documents. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000461/2005002-02(DRS)).

.3 Components

a. Inspection Scope

The inspectors examined the HPCS system and support systems associated pumps, HXs and instrumentation to ensure that component level attributes were satisfied.

Component Degradation: To complete this attribute, the inspectors reviewed and verified that potential degradation was monitored or prevented and component replacement was consistent with inservice and/or equipment qualification life. The inspectors examined existing system programs to ensure that components were adequately maintained.

Equipment/Environmental Qualification: To complete this attribute, the inspectors reviewed and verified that equipment was qualified to operate under the environment in which it was expected to be subjected to under normal and accident conditions. The inspectors reviewed design information, specifications, and documentation to ensure that the HPCS system and support systems were qualified to operate within the environmental conditions specified in the environmental qualification documentation.

Equipment Protection: To complete this attribute, the inspectors reviewed and verified that the HPCS system and subsystems were adequately protected from natural phenomenon and other hazards, such as HELBs, floods or missiles. The inspectors reviewed design information, specifications, and documentation to ensure that the systems were adequately protected from those hazards identified in the USAR, which could impact the systems ability to perform their safety function.

Component Inputs/Outputs: To complete this attribute, the inspectors reviewed and verified that the HPCS system and subsystems component inputs and outputs were suitable for the application and would be acceptable under accident and/or event conditions; that required inputs to components, such as coolant flow, electrical voltage, and control air necessary for proper component operation were provided and; that components (e.g., valve, circuit breakers, etc.) failed in the safe configuration.

Operating Experience: To complete this attribute, the inspectors reviewed and verified that the licensee was appropriately tracking and applying operating experience.

b. Finding Inadequate Heat Exchanger Thermal Performance Testing

Introduction:

The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" having very low safety significance (Green) involving the division III diesel generator (DG) jacket-water (JW) cooler HXs thermal performance testing. Specifically, the licensee failed to recognize that the calculated value for the DG JW flow rate, as determined from test data obtained during thermal performance testing, was significantly higher than the flow rate that could be attained by the engine-driven water pump.

Description:

The inspectors reviewed test procedure CPS 2700.19, Div III DG (16 Cyl)

JW Cooler (1DG13A) HX Performance Covered by GL89-13, Revision 3A. The purpose of the test procedure was to confirm the heat removal capability of the division III DG JW cooler HX. The DG rejects engine heat to the service water via the JW cooler HX. The licensee committed to perform HX thermal performance testing in response to Generic Letter (GL) 89-13, as described in procedure CPS 1003.10, CPS Program for NRC GL89-13 (SW Problems Affecting Safety-Related Equipment),

Revision 5A. The testing consisted of measuring service water flow rate, inlet and outlet service water temperatures, as well as measurements of JW inlet and outlet temperatures across the HX. By using a heat balance method (i.e., JW heat rejection equals heat added to service water), the mass flow rate of the JW was calculated.

The inspectors noted that there was a bypass around the HX for heat up of the JW.

The licensee stated that there would be no bypass of cooling water around the JW cooler HX because thermal equilibrium of the engine was maintained during the testing and that the engine was fully loaded. Therefore, the JW flow rate should be fully developed through the HX with a flow rate close to the value stated in the vendors design specification.

On November 18, 2005, the inspectors reviewed the 2002, 2003 and 2004 thermal performance tests for the division III DG JW cooler HX. The inspectors questioned the licensees test results because reduction of the test data indicated that the JW mass flow rate was significantly greater than the mass flow rate specified by the vendors design specification. In particular, the calculated JW flow rates were 1866 gallons per minute (gpm), 1471 gpm and 1165 gpm for the 2002, 2003 and 2004 year tests, respectively. The maximum JW flow rate specified by the vendors design specification was 850 gpm.

The inspectors questioned how a fixed speed JW pump could provide such a wide variation in cooling water flow rates, especially during the 2002 test that calculated a flow rate of 1866 gpm, which was over twice as much as the 850 gpm specified by the vendor. The licensee subsequently evaluated the 2002, 2003 and 2004 test results and determined that the 2002 and 2003 test results were invalid. The licensee determined that there was no impact on operability of the DG because the 2004 test results indicated a JW flow rate of 1165 gpm, which was considered valid by the licensee. The inspectors were not convinced that the 2004 tests were valid, but concluded that with the licensees HX inspection and cleaning effort performed in 2005, the HX would remove the heat generated by the engine. The licensee determined that test procedure CPS 2700.19 needed to be evaluated and/or revised. Condition Report 426459, NRC SSD&PC Is the Calculated Process Flow Rate Reasonable, and CR 429726, Discrepancies Not Identified in Corrective Action Process, dated November 21 and December 2, 2005, respectively, were issued.

The inspectors noted that during the NRC Heat Sink Inspection in 2001, questions regarding the 2000 test results were raised. The licensee initiated three condition reports to address the issue. In particular condition report, CR-2-01-03-180, Unstable Testing Conditions Invalidated Div III DG HX Test Performed in November 2000," dated March 21, 2001, reported that due to anomalies in the test data, the performance tests were invalid. Therefore, the SSDPC inspectors concluded that the licensee was aware of the testing discrepancies but did not properly evaluate and correct the concerns.

Analysis:

The inspectors determined that the licensees failure to recognize that the calculated value for the DG JW flow rate was significantly greater than the capability of the vendors design specification and that the test results did not represent actual HX performance was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on September 30, 2005. The finding involved the attribute of equipment performance, where the licensees failure to obtain accurate and reliable test data did not provide the information needed to demonstrate the functional capability of the HX and could have affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the failure to adequately test the heat transfer capability of the HX, the division III DGs JW cooler HX would have performed its safety function. Therefore, the inspectors concluded that the finding was a test control deficiency that did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green. A contributing cause of the finding was related to the cross-cutting element of problem identification and resolution. Concern with this performance test was previously identified in 2000; however, the licensee did not properly evaluate the adverse condition. In addition, the licensee did not recognize the condition during testing in 2002 and 2003.

Enforcement:

10 CFR Part 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 identified and that test results shall be evaluated to assure that test requirements have been satisfied.

Contrary to the above, for 2000, 2002 and 2003, the licensee failed to assure that the division III DGs JW cooler HX thermal performance test results were adequately evaluated to assure that test requirements had been satisfied. This resulted in HX test results that did not represent actual HX thermal performance. Once identified, the licensee entered the finding into their corrective action program as CR 426459 and CR 429726 to evaluate and/or revise the affected test procedures. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000461/2005002-03(DRS)).

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed a sample of problems associated with the HPCS system that were identified and entered into the corrective action program by the licensee. 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, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. R. Bement and other members of licensee management at the conclusion of the inspection on January 20, 2006. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

.2 Interim Exit Meetings

An interim exit was conducted for the safety system design and performance capability inspection with Mr. R. Bement on December 2, 2005.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

A. Bailey, Operations Training Manager

R. Bement, Site Vice President

W. Carsky, Shift Operations Superintendent

B. Corley, Reactor Operator

J. Cunningham, Work Management Director

T. Danley, Design Engineering Response Support

R. Davis, Radiation Protection Manager

R. Frantz, Regulatory Assurance

M. Gandhi, Mechanical/Structural Design Support

G. Hughes, Design Engineering

J. Hunsicker, Electrical/Instrumentation and Control Design Support

W. Iliff, Regulatory Assurance Manager (Response Team Lead)

B. Kerestes, Design Engineering

S. Lakebrink, Mechanical/Structural Design Manager

D. Lillyman, Balance of Plant Support

J. Lindsey, Training Director

T. Marini, Nuclear Oversight Manager

M. McDowell, Plant Manager

T. Parrent, Balance of Plant Support

C. Patel, High Pressure Core Spray System Manager

R. Peak, Site Engineering Director

D. Schavey, Operations Director

E. Schweitzer, Design Engineering

K. Scott, Senior Manager Plant Engineering

D. Smith, Diesel Generator System Manager

M. Smith, Electrical Systems Support

E. Tiedemann, Regulatory Assurance

D. Tucker, Electrical/Instrumentation and Control Design Support

C. Williamson, Security Manager

Nuclear Regulatory Commission

B. Dickson, Senior Resident Inspector

J. Lara, Chief, Engineering Branch 3

A.M. Stone, Chief, Engineering Branch 2

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000416/2005002-01(DRS) URI Vortex Analysis Methodology Not Appropriate (Section 1R21.1b)

05000461/2005002-02(DRS) NCV Non-Conservative Acceptance Criteria (Section 1R21.2b)

05000461/2005002-03(DRS) NCV Inadequate Heat Exchanger Thermal Performance Testing (Section 1R21.3b)

Closed

05000461/2005002-02(DRS) NCV Non-Conservative Acceptance Criteria (Section 1R21.2b)

05000461/2005002-03(DRS) NCV Inadequate Heat Exchanger Thermal Performance Testing (Section 1R21.3b)

Discussed

NONE Attachment