IR 05000244/2010009

IR 05000244-10-009, on 10/18/2010 - 11/11/2010, R. E. Ginna Nuclear Power Plant, LLC (Ginna), Component Design Bases Inspection

ML103560009
Person / Time
Site:	Ginna
Issue date:	12/22/2010
From:	Doerflein L Engineering Region 1 Branch 2
To:	John Carlin Constellation Energy Nuclear Group
References
2010-009, IR-10-009
Download: ML103560009 (33)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD KlNG OF PRUSSlA. PA 19406-1415 December 22, 2010 Mr. John T. Carlin, Vice President R. E. Ginna Nuclear Power Plant, LLC Constellation Energy Nuclear Group, LLC 1503 Lake Road Ontario, New York 14519 DESIGN R. E. GINNA NUCLEAR POWER PLANT - NRC COMPONENT OOO9 BASES INSPECTION REPORT O5OOO244I2O1

Dear Mr. Carlin:

an on November 11, 2010, the u.S. Nuclear Regulatory_commission (NRC) completed report documents power-Plant. The enclosed inspection inspection at your n.g. binna Nuclear you and other the inspection results, which were discussed on November 11,2O1O,with members of Your staff.

relate to safety and The inspection examined activities conducted under your license as they and with the conditions of your license'

compliance with the commission's rules and regulatibns and .

of selected components In conducting the inspection, the team examinel the adequacy and design basis accidents'

operator actions to mitigate postulated transients, initiating events, walkdowni, eiamination of Jelected procedures, calculations and The inspection invotveo-nero records, and interviews with station personnel'

low safety significance This report documents two NRC-identified findings that were of very (Green). These nnOinls were determined to inv6Ne violations of NRC requirements' However, because they were entered into because of the u"w ioiu-trfety significance of the violations and violations (Ncv)

your corrective action program, tne NnC li treating these findings as non-cited poricy. rf yo-u.contest any NCV in this.

consistent with sectio n 2.3.2of the NRC Enforcerient inspection report' with report, you should provide a response *itnin 30 days of the date of this the basis for your Oeniat, to the U.S. Nuclear Regulatory Commission, ATTN: Document Regional Administrator, Region control Desk, wasn-ing6n, o.c. 20555-0001, wTth copies to the l; the Director, Office olf Enfor."1nent, U.S. Nuclear Regulatory Commission, Washington' D'C'

Power Plant'

20555-0001; and the NRC Resident Inspector at the R.E. Ginna Nuclear 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 the public inspection in the NRC Public Docket Room or from the Publicly Available Records component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.sov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely, id^,*^*fr Lawrence T. Doerflein, Chief, Engineering Branch 2 Division of Reactor Safety Docket No. 50-244 License No. DPR-18 Enclosure: lnspection Report 0500024412010009 w/Attachment: Supplemental Information cc w/encl: Distribution via ListServ

SUMMARY OF FINDINGS

f R 05000244/201OO09i 10t1812010 - 11t1112010; R. E. Ginna Nuclear Power Plant, LLC (Ginna); Component Design Bases Inspection The report covers the Component Design Bases Inspection conducted by a team of four NRC inspectors and two NRC contractors. Two findings of very low risk significance (Green) were identified, both of which were considered to be non-cited violations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using NRC lnspection Manual Chapter (lMC) 0609, "significance Determination Process" (SDP). Cross-cutting aspects associated with findings are determined using IMC 0310, "Components Within the Cross-Cutting Areas." 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 4, dated December 2006.

NRC-ldentified Findinqs

Cornerstone: Mitigating Systems

. Green: The team identified a finding of very low safety significance involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion lll, Design Control. Specifically,

Constellation had not verified the adequacy of their design with respect to the impact of the installed Amptector type LSG trip unit discriminator feature on breaker coordination.

The discriminator circuit design had not been evaluated to ensure the 480V load center bus motor control center (MCC) feeder breakers would maintain coordination and be capable of maintaining power to downstream safety-related components in response to design basis events such as seismic or steam line break transients. Constellation entered the issue into their corrective action program to evaluate the adequacy of their design and ensure the feeder breakers remained operable.

The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the 480V busses to respond to initiating events to prevent undesirable consequences.

The team evaluated the finding in accordance with Inspection Manual Chapter (lMC)0609, Significance Determination Process, Attachment 0609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4afor the Mitigating Systems Cornerstone. The team determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. The team did not identify a cross-cutting aspect with this finding because this was an old design issue and therefore was not reflective of current performance. (Section 1R21.2.1.1)

Green: The team identified a finding of very low safety significance involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion lll, Design Control. Specifically,

Constellation had not correctly translated residual heat removal (RHR) pump net positive suction head (NPSH) operating limits into emergency operating procedures. Emergency operating procedure ES-1.3, Transfer to Cold Leg Recirculation, included criteria for aligning the discharge of the RHR pump to the suction of the safety injection pump under post-accident sump recirculation conditions which had not been adequately analyzed for RHR pump NPSH. Constellation entered the issue into their corrective action program to address the inconsistency between the design analysis and procedure and performed a review to ensure the RHR pump remained operable with respect to NPSH margin.

The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, design control measures had not ensured consistency between the design analysis assumptions and the operating procedure to ensure adequate RHR pump NPSH margin when aligned to the safety injection (Sl) pump during sump recirculation. The team evaluated the finding in accordance with IMC 0609, Significance Determination Process,

Attachment 0609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4afor the Mitigating Systems Cornerstone. The team determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. The team did not identify a cross-cutting aspect with this finding because it did not represent current performance. The discrepancy between the design analysis and procedure occurred outside of the timeframe which reflects current performance. (Section 1R21.2.1.2)iii

REPORT DETAILS

REACTOR SAFETY

Cornerstones: lnitiating Events, Mitigating Systems, Barrier I ntegrity

1R21 Component Desiqn Bases Inspection (lP 71111.21)

.1 Inspection Sample Selection Process

The team selected risk significant components and operator actions for review using information contained in the R. E. Ginna Nuclear Power Plant Probabilistic Safety Assessment and the U.S. Nuclear Regulatory Commission's (NRC) Standardized Plant Analysis Risk (SPAR) model. Additionally, the R.E. Ginna Significance Determination Process (SDP) Phase 2 Notebook (Revision 2.1a) was referenced in the selection of potential components and operator actions for review. In general, the selection process focused on components and operator actions that had a Risk Achievement Worth (RAW)factor greater than 1.3 or a Risk Reduction Worth (RRW) factor greater than 1.005. The components selected were located within both safety-related and non-safety related systems and included a variety of components such as pumps, breakers, transformers, and valves.

The team initially compiled a list of components and operator actions based on the risk factors previously mentioned. Additionally, the team reviewed the previous component design bases inspection report (05000244/2007006) and excluded those components previously inspected. The team then performed a margin assessment to narrow the focus of the inspection to 13 components, 4 operator actions, and 4 operating experience items. The team's evaluation of possible low design margin included consideration of original desigrt issues, margin reductions due to modifications, or margin reductions identified as a result of material condition/equipment reliability issues.

The assessment also included items such as failed performance test results, corrective action history, repeated maintenance, maintenance rule (aX1) status, operability reviews for degr:aded conditions, NRC resident inspector insights, system health reports, and industry operating experience. Finally, consideration was also given to the uniqueness and complexity of the design and the available defense-in-depth margins. The margin review of operator actions included complexity of the action, time to complete the action, and extent-oftraining on the action.

The inspection performed by the team was conducted as outlined in NRC Inspection Procedure (lP) 71 111.21. This inspection effort included walkdowns of selected components, interviews with operators, system engineers and design engineers, and reviews of associated design documents and calculations to assess the adequacy of the components to meet design basis, licensing basis, and risk-informed beyond design basis requirements. Summaries of the reviews performed for each component, operator action, and operating experience sample, and the specific inspection findings identified are discussed in the subsequent sections of this report. Documents reviewed for this inspection are listed in the Attachment.

.2.1.1 Station Service Transformer

16 (PXABSS016)

a. Inspection Scope

The team inspected station service transformer (SST) 16 to verify that it was capable of meeting its design basis requirements. The station service transformer is designed to provide the preferred power source to safety-related 480V Bus 16. The team reviewed load flow and short circuit current calculations to determine the design basis for maximum load and breaker interrupting duty, and the Bus 16 load center equipment vendor ratings for conformance with the design basis. The team also reviewed the coordination/protection calculation for the bus incoming line and motor control center (MCC) feeder breakers for design basis load flow conditions and breaker coordination.

The team performed walkdowns to assess the material condition and to identify potential seismic llil issues. The team reviewed SST 16 transformer cooling fan requirements and verified fan operation was in accordance with design requirements. The team also reviewed surveillance tests on the incoming line and MCC feeder breaker Amptector trip units to ensure test results were in accordance with design requirements. Finally, corrective action documents and system health reports were reviewed to verify deficiencies were appropriately identified and resolved, and that the SST was properly maintained.

b.

Findinqs

Introduction:

The team identified a finding of very low safety significance (Green)involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion lll, Design Control. Specifically, Constellation had not verified the adequacy of their design with respect to the impact of the installed Amptector type LSG (long, short & ground) trip unit discriminator feature on breaker coordination.

Description:

The team determined that an Amptector discriminator trip feature was enabled on the startup service transformer Bus 16 and Bus 14 incoming lines and MCC feeder breakers. These 480V breakers with an LSG Amptector overcurrent protective device have an associated safety feature called a discriminator. The discriminator is designed such that when a breaker is carrying a minimal current load (nominally less than 3 percent of its current sensor tap), and a fault (overload condition) occurs of significant magnitude, the protective device will instantaneously trip the breaker open.

However, when the current load is above the minimum value, the instantaneous trip is bypassed and the breaker will trip in accordance with its short time delay setpoint. The team noted that NRC Information Notice 92-29, Potential Breaker Miscoordination Caused By Instantaneous Trip Circuitry, had been issued to alert licensees to potential breaker miscoordination involving instantaneous trip circuitry. The licensee had reviewed this notice and determined that their design was acceptable. This was based on the determination that the applicable circuits had sufficient current load above that required to bypass the discriminator instantaneous trip feature, thereby assuring coordination would not be affected.

The team noted that the circuit breaker Amptector type LSG discriminator trip unit function had not been evaluated within design calculation DA-EE-104-07, 480V Coordination and Circuit Protection Study. The team was concerned that a lack of breaker coordination for instantaneous trip conditions could exist during an event such as a seismic or steam line break (SLB)where non-safety related equipment could fault with a concurrent postulated loss-of-offsite-power (LOOP) condition. For SLB events, UFSAR section 3.6.2.3.2.4 assumptions are offsite power to be unavailable if a trip of the turbine-generator system or reactor trip system is a direct consequence of the postulated piping failure. The LOOP would result in the loss of the minimum current flow which was relied on to bypass the discriminator circuit. The team noted that if non-safety related equipment became faulted due to the event, the MCC feeder breakers may trip when the 16 and 14 safety busses would be re-energized by their emergency diesel generators.

The team determined that the 480V breaker coordination study evaluated the circuit breaker with an Amptector type LSG trip unit (with a discriminator feature) on the feeder circuit to safety-related Class 1E MCC 'D'. The team noted that the breaker trip unit was required to provide coordination with downstream non-Class 1E MCC circuit breakers during fault conditions. In response to the team's concerns, Constellation reviewed the short circuit study and determined that the Bus 16 MCC feeder breaker Amptector trip unit was susceptible to instantaneously tripping after a LOOP condition, because sufficient short circuit current could exist for a fault on specific non-Class 1E MCC circuits. The team noted that a feeder breaker trip would complicate operator recovery actions because safety related loads would be lost while they would be attempting to respond to the initiating event.

Constellation entered the issue into their corrective action program (CAP) and performed an operability review. Constellation reviewed non-Class 1E circuits where sufficient fault current could exist to challenge breaker coordination with safety-related equipment.

Their initial review determined that there was reasonable assurance that circuit failure in the applicable non-class 1E equipment would not occur due to postulated events such as seismic, steam line break, or loss-of-coolant accidents. This was due in part to the location of the equipment and existing circuit configurations. Constellation concluded that a fault of sufficient magnitude would not be present when the emergency diesel generator (EDG) breaker would close to re-energize the safety bus at the time when the discriminator circuits would not be bypassed due to the 10 second interim loss of power (current load). This review was performed for both busses (16 and 14). The team reviewed Constellation's evaluation and found their initial assessment to be reasonable.

Analvsis: The team determined that the licensee's failure to adequately evaluate the Amptector's discriminator circuit function design for all postulated design basis conditions was a performance deficiency. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the 480V busses (16 and 14) to respond to initiating events to prevent undesirable consequences.

Specifically, the discriminator circuit design had not been evaluated to ensure the 480V load center bus MCC feeder breakers would maintain coordination and be capable of maintaining power to downstream safety-related components in response to design basis events such as seismic, steam line break transients, or loss-of-coolant accidents (LOCAs). The team evaluated the finding in accordance with Inspection Manual Chapter (lMC) 0609, Significance Determination Process (SDP), Attachment 0609.04, Phase 1-lnitial Screening and Characterization of Findings, Table 4afor the Mitigating Systems Cornerstone. The tdam determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. The team did not identify a cross-cutting aspect with this finding because this was an old design issue and therefore was not reflective of current performance.

Enforcement:

10 CFR Part 50, Appendix B, Criterion lll, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, as of November 9, 2010, the protective device coordination design for Amptector trip units with the discriminator instantaneous trip circuits had not been adequately verified under all postulated design conditions. Specifically, the design which included the discriminator circuit had not been evaluated to ensure the 480V load center bus MCC feeder breakers would maintain coordination and be capable of maintaining power to downstream safety-related components in response to design basis events such as seismic, steam line break conditions, or LOCAS. Because this finding was of very low safety significance, and it was entered into Constellation's CAP as CR 2010-7062, this violation is being treated as a non-cited violation (NCV)consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 0500024/+12010009-01, Inadequate Evaluation of Breaker Coordination for Amptector Type LSG Trip Unit Discriminator Feature)

.2.1.2 Residual Heat Removal Pump (PAC01A)

a.

Insoection Scope The team inspected the 'A' residual heat removal (RHR) pump to verify that it was capable of meeting its design basis requirements. The team reviewed applicable portions of the UFSAR and drawings to identify the design basis requirements for the pump. The team reviewed calculations and surveillance test procedures to verify that the pump was capable of achieving design basis head/flow requirements during limiting design basis conditions and that test acceptance criteria were consistent with these requirements. The team reviewed the hydraulic calculations associated with system flowrate and pressure as well as net positive suction head (NPSH) margin for the pump to ensure that the required performance could be achieved.

The team interviewed design and system engineers to review the design and system functional requirements as well as historicaltest performance results. In addition, the team reviewed work orders and corrective action documents to identify failures or nonconforming issues, and to determine if deficiencies were being appropriately identified, evaluated, and corrected. Finally, the team performed a review of the emergency operating procedures (EOPs) associated with post-accident pump operation to ensure the capability of the component to perform as required under actual accident conditions.

Findinos lntroduction: The team identified a finding of very low safety significance (Green)involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion lll, Design Control. Specifically, Constellation had not correctly translated RHR pump NPSH operating limits into the EOPs. The EOP ES-1.3, Transfer to Cold Leg Recirculation, Revision 44, included criteria for aligning the discharge of the RHR pump to the suction of the safety injection pump under post-accident sump recirculation conditions which had not been adequately analyzed for RHR pump NPSH.

Description:

The team reviewed design analysis DA-ME-2005-085, NPSH for the Emergency Core Cooling System (ECCS) Pumps during lnjection and Sump Recirculation, Revision 2. This design calculation included an evaluation of the minimum NPSH that would be available to an RHR pump while taking suction from the containment sump during post-accident operation. The conditions evaluated included the operating RHR pump being aligned to the suction of the operating safety injection pump(s). For this mode of operation the calculation concluded that the RHR pump NPSH margin would be acceptable based on the reactor coolant system pressure being at least 57 psig greater than the containment building pressure. The calculation stated that this design input was based on the EOP ES-1.3 criterion for entering this alignment.

However, the team observed that ES-1.3 did not directly include this pressure criterion.

The procedure referred to EOP Figure 19.0, High Head Safety Injection (Sl) Required, to determine if RHR pump alignment to the Sl pump was required. This figure was based on measured core exit temperature. The team questioned how Figure 19.0 related to the minimum reactor coolant system pressure criterion used in the calculation.

During their review of the team's concern, Constellation confirmed that Figure 19.0 was not correctly applied or consistent with engineering analysis assumptions for the determination of RHR NPSH margin when aligned in series with an Sl pump. The measured core exit temperature values included in the figure would not ensure adequate available NPSH for the operating residual heat removal pump, assuming conservative design saturated conditions in the containment sump. Constellation personnel stated that the values included in Figure 19.0 were non-conservative by approximately 25 degrees Fahrenheit (oF), and determined that the reactor pressure corresponding to the Figure 19.0 values in the design calculation would result in a RHR pump NPSH deficit of approximately 1.8 feet. Constellation personnel stated that the RHR NPSH design basis analysis had been previously based on using a different EOP Figure, (Figure 5, RHR Injection), than the one that had been translated into the existing procedure ES 1.3, during an October 2006 revision.

The team was also concerned that the reactor coolant pressure could decrease below the 57 psid assumption with respect to containment pressure over the course of post accident pump operation, resulting in a reduction of available NPSH as the RHR pump flow increased. The team noted the current EOPs did not include any criterion for stopping the safety injection pumps once their suction supply was aligned to an operating RHR pump.

Constellation initiated Condition Report 2010-7084 on November 10, 2010, to evaluate this issue. The associated operability evaluation verified that the RHR pump would still be operable if the system was aligned as allowed by Figure 19.0. Constellation's technical evaluation analyzed several different break sizes and the corresponding reactor pressures and temperatures. The evaluation considered that containment accident pressure would reasonably exist above and beyond the required containment pressure necessary to ensure adequate NPSH margin for the RHR pump. The evaluation took credit for less than 1 psig of containment accident pressure under post accident conditions and considered that the sump temperature would be expected to decrease over the course of the event. The team reviewed the operability evaluation and determined Constellation's conclusion was reasonable.

Analvsis: The team determined that the failure to correctly translate RHR pump NPSH operating limits into the EOPs was a performance deficiency. The finding was determined to be more than minor because it was similar to example 3.j. of NRC IMC 0612, Appendix E, Examples of Minor lssues, in that based on design (saturated)conditions the team had a reasonable doubt of operability with respect to the NPSH margin for the RHR pumps until additional analysis was performed. Additionally, the finding was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliabitity, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, design control measures had not ensured consistency belween the design analysis assumptions and the operating procedure to ensure adequate RHR NPSH when aligned to the Sl pump during sump recirculation.

The team evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1- lnitial Screening and Characterization of Findings, Table 4afor the Mitigating Systems Cornerstone. The team determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability.

The team did not identify a cross-cutting aspect with this finding because it did not represent current performance. The discrepancy between the design analysis and procedure occurred outside of the timeframe which reflects current performance.

Enforcement:

10 CFR Part 50, Appendix B, Criterion lll, Design Control, requires, in part, that measures be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from October 27,2006, to November 9, 2010, the design conditions assumed within calculation DA-ME-2005-085 to evaluate the adequacy of RHR NPSH during sump recirculation, had not been correctly translated into.procedure ES-1

.3. Because this finding was of very low safety significance, and it

was entered into Constellation's CAP as Condition Report 2010-708{ this violation is being trbated as a non-cited violation (NCV) consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 0500024/,12010009-02, Inadequate Translation of NPSH Design Limits into EOPs)

.2.1.3 Component Coolinq Water Pump (PAC02A)

a. Inspection Scope

The team inspected the 'A'component cooling water (CCW) pump to verify that it was capable of meeting its design basis requirements. The CCW system is designed to provide cooling water to essential components under normal, transient, and accident conditions. The team reviewed the UFSAR, drawings, and procedures to identify the most limiting requirements for the pump. The team reviewed a sample of surveillance test results to verify that pump performance met the acceptance criteria and that the criteria were consistent with the design basis. The team also reviewed calculations for NPSH to ensure that the pump could successfully operate under the most limiting conditions. The team discussed the design, operation, and corrective maintenance of the pump with engineering staff to gain an understanding of the performance history and overall component health. Additionally, the team reviewed corrective action documents and performed a walkdown to assess the material condition of the pump.

b.

Findinqs No findings were identified.

.2.1.4 Charqinq Pump (PCH01A)

a. Inspection Scope

The team inspected the 'A' charging pump to verify that it was capable of meeting its design basis requirements. The charging system is designed to provide injection to the reactor coolant system under normal, transient, and accident conditions. The team reviewed the UFSAR, drawings, and procedures to identify the most limiting requirements for the pump. The team reviewed a sample of surveillance test results to verify that pump performance met the acceptance criteria and that the criteria were consistent with the design basis. The team reviewed calculations for NPSH to ensure that the pump could successfully operate under the most limiting conditions, including a loss of component cooling water to the non-regenerative heat exchanger. The team discussed the design, operation, and corrective maintenance of the pump with engineering staff to gain an understanding of the performance history and overall component health. The team also reviewed corrective action documents and performed a walkdown to assess the material condition of the pump. In addition, the team reviewed the primary and back-up sources of electrical power and instrument air required to operate the pump.

b.

Findinqs No findings were identified.

.2.1.5 Component Coolinq Water Motor Operated Valve (MOV-738A)

a. Inspection Scope

The team inspected the CCW to RHR heat exchanger motor operated valve (MOV),

MOV-738A, to verify that it was capable of performing its design function. The team reviewed the UFSAR, calculations, and procedures to identify the design basis requirements of the valve. The team also reviewed accident system alignments to determine if component operation would be consistent with the design and licensing bases assumptions. Valve testing procedures and valve specifications were also reviewed to ensure consistency with design basis requirements. The team reviewed periodic verification diagnostic test results and stroke test documentation to verify acceptance criteria were met and consistent with the design basis. Additionally, the team verified the valve safety function was maintained in accordance with Generic Letter (GL) 89-10 guidance by reviewing torque switch settings, performance capability, and design margins. The team reviewed degraded voltage conditions and voltage drop calculations to confirm that the MOV would have sufficient voltage and power available to perform its safety function at worst case degraded voltage conditions.

The team interviewed the MOV program engineer to gain an understanding of maintenance issues and overall reliability of the valve. The team conducted a walkdown to assess the material condition of the valve, and to verify the installed valve configuration was consistent with design basis assumptions and plant drawings. Finally, corrective action documents were reviewed to verify that deficiencies were appropriately identified and resolved, and that the valve was properly maintained.

b.

Findinqs No findings were identified.

.2.1.6 Safetv Iniection Motor Operated Valve (MOV-857C)

a. Inspection Scope

The team inspected the safety injection (Sl) pump suction valve from RHR, MOV-857C, to verify that it was capable of performing its design function. The team reviewed the UFSAR, calculations, and procedures to identify the design basis requirements of the valve. The team also reviewed accident system alignments to determine if component operation would be consistent with the design and licensing bases assumptions. Valve testing procedures and valve specifications were also reviewed to ensure consistency with design basis requirements. The team reviewed periodic verification diagnostic test results and stroke test documentation to verify acceptance criteria were met and consistent with the design basis. Additionally, the team verified the valve safety function was maintained in accordance with GL 89-10 guidance by reviewing torque switch I

settings, performance capability, and design margins. The team also reviewed degraded voltage conditions and voltage drop calculations to confirm that the MOV would have sufficient voltage and power available to perform its safety function at worst case degraded voltage conditions.

The team interviewed the MOV program engineer to gain an understanding of maintenance issues and overall reliability of the valve. The team conducted a walkdown to assess the material condition of the valve, and to verify the installed valve configuration was consistent with design basis assumptions and plant drawings.

Corrective action documents were reviewed to verify that deficiencies were appropriately identified and resolved and that the valve was properly maintained. In addition, the team performed a review of the valve interlock design and testing to ensure that the valve and other associated emergency core cooling system (ECCS) valves would function as designed under the most limiting design basis condition, including a single failure of a valve or power supply.

b.

Findinqs No findings were identified.

.2.1.7 Emeroencv Diesel Generator (KDG01B)

a. Inspection Scope

The team inspected the 'B' emergency diesel generator to verify that it was capable of meeting its design basis requirements. The design function of the 'B' EDG is to provide standby power to safety-related 480V busses 16 and 17 when the preferred power supply is not available. The team reviewed the EDG loading study to ensure consistency with actual loading expected in response to a design basis accident. The team reviewed the break horsepower basis for selected pump motors to ensure loads were adequately considered in the loading study at conservative motor conditions.

The team reviewed completed Technical Specification (TS) performance tests to ensure the EDG met all applicable test acceptance criteria. The team reviewed applicable procedures associated with the use of ultra-low sulfur diesel (ULSD) and bio-dieselfuels to ensure that the correct fuel was being used. The EDG fuel consumption and unusable volume calculations were reviewed to assess the capacity of the fuel oil storage and day tanks and to verify the capability of the EDG to operate for the required mission time. The team reviewed calculations to assess the fuel oil storage tank protection against external events such as a postulated tornado event. The fuel oil monitoring limits were reviewed to assess fuel oil quality to ensure test results were consistent with design specifications. The team reviewed the design and supporting calculations of the EDG air start system and the jacket water and lube oil cooling systems to ensure the EDG was capable of performing in accordance with its design basis.

In addition, the team reviewed engineering change package (ECP) 2008-0040, that involved switching from continuous service water (SW) flow through the heat exchangers to isolation of the SW flow using two normally closed, parallel configured, air operated valves (AOV) that open on an EDG start signal. The team verified that the AOV's fail in the open position to ensure EDG cooling capability was maintained on loss of power or air to the AOVs. In addition, the team performed interviews with the EDG system engineer, reviewed applicable corrective action documents, and performed an extensive walk-down of the 'B' EDG and associated support equipment to assess the material condition and potential vulnerability to hazards such as flooding.

b.

Findinos No findings were identified.

.2.1.8 Service Water Pump Discharqe Check Valve (CV-4602)

a.

lnspection Scope The team inspected the 'B' service water pump discharge nozzle check valve, CV-4602, to verify that it was capable of meeting its design basis requirements. The check valve was designed to minimize SW coolant loss from the system as a result of an idle or out of service pump to ensure safety related loads are cooled. The team reviewed the UFSAR, drawings, and procedures to identify the design basis requirements of the check valve. The check valve testing procedures and SW system hydraulic analyses were reviewed to verify the design basis requirements were appropriately incorporated into the test acceptance criteria. The team reviewed a sample of test results to verify the acceptance criteria were met. The team reviewed the corrective and preventive maintenance of the check valve to gain an understanding of the performance history and overall component health. ln addition, the team reviewed maintenance pictures of the check valve to assess material condition. Finally, corrective action documents and system health reports were reviewed to verify deficiencies were appropriately identified and resolved, and that the check valve was properly maintained.

b.

Findinos No findings were identified.

.2.1.9 Pressurizer Relief Valve (RV-434)

a. Inspection Scope

The team inspected pressurizer safety relief valve, RV-434, to verify it was capable of performing its design basis function. The team reviewed the UFSAR, TSs, drawings, and procedures to identify the design basis requirements of the valve. The team verified that the valve setpoint was in accordance with TS requirements and the American Society of Mechanical Engineers (ASME) operating and maintenance code. The team reviewed design documentation for sizing and the lift setpoint, and the analysis for overpressure protection capability of the valve to determine if the valve would meet design requirements. The team also discussed valve performance and trending with the system engineer, and reviewed condition reports and system health reports to assess the material condition of the valve.

b.

Findinos No findings were identified.

.2.1.1 0 Motor Driven Auxiliarv Feedwater (MDAFW) Flow Control Valve (MOV-4007)

a. Inspection Scope

The team inspected the 'A' MDAFW pump flow control valve, MOV-4007, to verify that the valve was capable of supporting the pump design basis flow requirements to the steam generator. The team reviewed the UFSAR, drawings, and procedures to identify the design basis requirements of the valve. Design calculations and system operating parameters were reviewed to verify that the design basis had been appropriately translated into specifications and procedures. The team reviewed test procedures to verify that acceptance criteria for the tested parameters were appropriately supported by calculations to ensure the design and liiensing bases were satisfied. The team verified instrument control loop settings to ensure the design function of the MDAFW pump was supported. The team verified that the thermal overload bypass circuitry was appropriately tested to ensure MOV operation during a design basis event. The team interviewed the MOV program engineer to review maintenance issues and assess overall reliability of the valve. The team also conducted a walkdown to assess the material condition of the valve and to verify the installed valve configuration was consistent with design basis assumptions and plant drawings. Finally, corrective action documents, preventive maintenance, and system health reports were reviewed to verify that deficiencies were appropriately identified and resolved.

b.

Findinos No findings were identified.

.2.1.1 1 Station Auxiliarv Transformer (1 2-PXYDO1

24)

a. Inspection Scope

The team inspected the 12A station auxiliary transformer (SAT) to verify that it was capable of meeting its design basis requirements. The 12A SAT was designed to provide offsite power to 4160V busses 12A and 128. The team reviewed one line diagrams, the transformer nameplate, and vendor test results for impedance data to confirm that correct transformer impedances were used in electrical analyses. The team confirmed the adequacy of the overcurrent relay settings for design basis loading requirements. Additionally, the team reviewed transformer dissolved gas analysis results, transformer bushing condition monitoring, and the transformer and auxiliary's preventive maintenance condition monitoring for adverse conditions that could affect reliability. The team performed a walkdown of the 12,A SAT to assess the observable material condition. Finally, corrective action documents and system health reports were reviewed to verify deficiencies were appropriately identified and resolved and the SAT was properly maintained.

b.

Findinqs No findings were identified.

.2.1.1 2 Tie Breaker for Bus 14 to Bus 13 (52lBT14-13)

a.

lnspection Scope The team inspected the Bus14 to Bus 13 tie-breaker to verify that it was capable of meeting its design basis requirements. The breaker was designed to tie Bus 14 to Bus 13 when allowed by plant conditions. The team reviewed one line diagrams and vendor equipment data to confirm the breaker ratings were sufficient to meet design basis conditions. The team reviewed the electrical analyses for load flow, short circuit, and breaker trip unit coordination requirements to confirm the adequacy of the settings for bus tie operation. The team reviewed operating and preventive maintenance procedures for conformance with design basis load conditions and breaker trip unit setting requirements. Finally, condition reports and system health reports were reviewed to verify deficiencies were appropriately identified and resolved.

b.

Findinss No findings were identified.

.2.1.1 3 4160 Volt Switchqear (Bus 12A)

a. Inspection Scope

The team inspected the 4kV switchgear Bus 12A to verify that it was capable of meeting its design basis requirements. Bus 12A was designed to distribute preferred power to safety-related 480V busses 14 and 18. The team reviewed load flow and short circuit current calculations for maximum load, momentary and interrupting duty, and bus bracing requirement to ensure conformance with the design basis. The team confirmed the use of maximum switchyard voltage for short circuit calculations and reviewed vendor equipment data for adequate margin in breaker momentary and interrupting duty.

The team confirmed the calculated minimum voltage (for degraded grid conditions) and short circuit current (for maximum switchyard voltage) were based on switchyard operating limits. The team reviewed preventive maintenance for selected breakers, component replacements, and the results of inspections/tests to confirm the reliability of the equipment. The team performed a walkdown of the 4kV switchgear to assess the observable material condition and to identify potential seismic llil issues. Finally, condition reports and system health reports were reviewed to verify deficiencies were appropriately identified and resolved.

b.

Findinqs No findings were identified.

.2.2 Review of Low Marqin Operator Actions (4 samples)

The team assessed manual operator actions and selected a sample of fouroperator actions for detailed review based upon risk significance, time urgency, and factors iffecting the likelihood of human error. The operator actions were selected from a probabi[stic risk assessment (PRA) ranking of operator action importance based on RAW and RRW values. The non-PM conliderations in the selection process included the following factors:

o Margin between the time needbd to complete the actions and the time available prior to adverse reactor consequences;

.

Complexity of the actions; o Reliability and/or redundancy of components associated with the actions; o Extent of actions to be performed outside of the control room;

. Procedural guidance to the operators; and o Amount of relevant operator training conducted.

.2.2.1 Letdown lsolation Followinq a Loss of Component Coolinq Water

a. Inspection Scope

The team evaluated the manual operator actions to isolate reactor letdown flow within 10 minutes following a loss of CCW to preclude a common mode failure of the charging pumps due to elevated volume controltank (VCT) temperature. The loss of charging 1ow to the reactor coolant pump (RCP) seals, concurrent with the loss of CCW cooling to the RCP thermal barriers, increases the likelihood of a RCP seal loss-of-coolant accident (LOCA). Operator critical tasks included:

o Recognize loss of CCW, enter abnormal procedure o Trip the reactor o Trip both RCPs o lsolate letdown by closing AOV-427 The team interviewed licensed operators and operator simulator instructors and reviewed associated operating procedures and operator training, including associated Operations Night Orders, to evaluate the operators' ability to perform the required actions. The tbam walked down applicable control and indicating panels in the simulator and in the main control room to assess the likelihood of cognitive or execution errors.

The team evaluated the available time margins to perform the actions to verify the reasonableness of Constellation's operating procedures and risk assumptions. The team also reviewed equipment deficiency reports, and performed independent infield observations, to assess the material condition of the CCW pumps, motors, heat exchangers, and support systems. In addition, the team reviewed the VCT heat-up analysis for a loss of CCW, including design and operating assumptions, to ensure that it used appropriate and conservative inputs. The team evaluated the available process margins, based on fluid flow rates, temperatures, and heat transfer capacities, and performed independent calculations to verify the reasonableness of engineering analysis supporting the prescribed operator actions.

b.

Findinqs No findings were identified.

.2.2.2 lsolate Break in Service Water Common Discharqe Pipinq

in the Auxiliarv Buildinq

a. Inspection Scope

The team evaluated operator actions to recognize and mitigate a service water (SW)pipe break in the common discharge line within the auxiliary building. Specifically, operator critical tasks included:

r Recognize condition r Direct response in accordance with alarm response procedure

.

Determine cause

.

Confirm flooding o lsolate source The team interviewed licensed and non-licensed operators, reviewed associated alarm response procedures and operator training, and conducted a detailed walkdown of accessible portions of the auxiliary building with an auxiliary operator (AO) to assess the operators' ability to perform the required actions and the likelihood of cognitive or execution errors. The team evaluated the available time margins to perform the actions to verify the reasonableness of Constellation's alarm response procedures and risk assumptions. The team reviewed equipment deficiency reports, maintenance history, internalflood analyses, and inspection results and performed independent in-field observations to assess potential internalflood vulnerabilities and to ensure that Constellation maintained appropriate configuration control of critical design features. In addition, the team independently walked down accessible portions of the auxiliary building to assess the material condition of the associated structures, systems and components (SSCs) with particular focus on potential high volume internal flood sources.

b.

Findinqs No findings were identified.

.2.2.3 Aliqn and Start Standbv Auxiliarv Feedwater Pumps

a.

lnspection Scope The team evaluated the manual operator actions to align and start the standby auxiliary feedwater (SAFW) pumps given a failure of the main and auxiliary feedwater (AFW)sources. Specifically, operator critical tasks included:

o Recognize loss of feedwater flow, enter abnormal procedure (FR-H.1)o Transition to EOP Attachment 5.1 (SAFW alignment)

.

Ensure safety injection (Sl) reset

.

Ensure normally open valves are open

.

Open SAFW pump C(D) suction valve MOV-9629A(B)o Verify at least one SW pump running r Align discharge valves as directed to feed desired steam generator A(B)

.

Start SAFW pumps as directed by FR-H.1 The team interviewed licensed operators and operator simulator instructors, reviewed associated alarm response procedures and operator training, and observed a licensed operator respond to a simulated demand to align and start the SAFW pumps from the main control room to independently assess the likelihood of cognitive or execution errors. The team evaluated the available time margins to perform the actions to verify the reasonableness of Constellation's alarm response procedures and risk assumptions.

The team reviewed SAFW valve and breaker verification surveillances, pump testing results, and equipment deficiency reports to assess the SAFW system availability and reliability. The team also walked down accessible portions of the SAFW and AFW systems to independently assess Constellation's configuration control and the material condition of these risk significant SSCs.

b.

Findinqs No findings were identified.

.2.2.4 Aliqn the Technical Support Center Batterv Charqer to

DC Train A or B

a. Inspection Scope

The team evaluated the manual operator actions to align the technical support center (TSC) battery charger to DC train A or B, given a loss of a single train of 480 VAC power which would eventually fail the associated battery chargers. Specifically, operator critical tasks included:

.

Enter abnormal procedure in response to low voltage condition on DC train A/B o Remove TSC battery from equalizing charge

.

Open AC input breaker to TSC charger, verify TSC battery voltage, close AC input breaker o Proceed to A(B) battery room, unlock and close disconnect panel switch o Proceed to TSC battery room and ensure fuse disconnect switch is closed r Proceed to turbine building basement, unlock and close manual throw-over switch The team interviewed licensed and non-licensed operators, reviewed associated operating procedures and operator training, and observed an AO perform a simulated transfer of DC train B to the TSC battery charger to independently assess the AO's ability to perform the required actions and the likelihood of cognitive or execution errors.

The team evaluated the available time margins to perform the actions to verify the reasonableness of Constellation's operating procedures and risk assumptions. The team also walked down the associated battery rooms, battery chargers, switching panels, and essential main control room instrumentation to independently assess Constellation's configuration control and the material condition of the associated SSCs.

b.

Findinqs No findings were identified.

.2.3.1 Operatinq Experience Smart Sample FY 2007-02: Floodinq Vulnerabilities Due to

Inadeouate Desiqn and Conduit/Hvdrostatic Seal Barrier Concerns

a. Inspection Scope

NRC Operating Experience Smart Sample (OpESS) FY 2007-02 is directly related to NRC Information Notice (lN) 2005-30, "Safe Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events and Inadequate Design," and issues associated with conduiVhydrostatic seal issues. The team evaluated internal and externalflood protection measures for the EDG rooms, battery rooms, turbine building basement, auxiliary building, and SW screenhouse. The team walked down the areas to assess operational readiness of various features in place to protect redundant safety-related components and vital electric power systems from flooding. These features included equipment drains, door seals, backflow check valves, flood detection and alarms, flood barriers, circulating water (CW) pump trip sensors, and wall penetration seals.

The team conducted several detailed walkdowns of the turbine, EDG, screenhouse, and auxiliary buildings to assess potentialflood vulnerabilities. In addition, the team conducted a step-by-step walkthrough of two time-critical flood mitigation strategies with an AO to independently assess procedure quality, flood barrier material condition, and the operators'ability to perform the required actions. The team also reviewed engineering evaluations, calculations, alarm response procedures, preventive and corrective maintenance history, operator training, and correct action condition reports associated with flood protection equipment and measures. Finally, the team interviewed Constellation personnel regarding their knowledge of indications, procedures, and required actions associated with several postulated internal and externalflood scenarios.

b.

Findinqs No findings were identified.

.2.3.2 Operatinq Experience Smart Sample FY 2008-01 - Neqative Trend

and Recurrinq Events Involvinq Emeroencv Diesel Generators

a. Inspection Scope

NRC OpESS FY 2008-01 is directly related to NRC Information Notice (lN) 2007-27, "Recurring Events Involving Emergency Diesel Generator Operability." The team reviewed Constellation's evaluation of lN 2007-27 and their associated corrective actions. The team reviewed Constellation's EDG system health and walkdown reports, EDG condition reports and work orders, leakage monitoring, and surveillance test results to verify that Constellation appropriately dispositioned EDG deficiencies. Additionally, the team independently walked down both EDGs on several occasions to inspect for indications of vibration-induced degradation on EDG piping and tubing and for any type of leakage (air, fuel oil, lube oil, jacket water). The team performed a post-surveillance run walkdown of the 'A' EDG on November 3, 2010, to ensure Constellation maintained appropriate configuration control and identified deficiencies at a low threshold.

Additionally, the team directly observed portions of the biennial maintenance work performed on the 'B' EDG to assess the material condition of the EDG and its support systems.

b.

Findinos No findings were identified.

.2.3.3 NRC lnformation Notice 98-02: Nuclear Power Plant Cold Weather Problems

and Protective Measures a.

lnspection Scope NRC lN 98-02 discussed the potential common-cause failure mechanisms of safety related systems and systems important to safety caused by extreme cold weather conditions. The team reviewed Constellation's evaluation of lN 98-02. The team reviewed the disposition of the information notice and conducted walkdowns of areas exposed to cold conditions. This included accompanying Constellation operations personnel on a walkdown of equipment required to protect components during winter conditions. The team also reviewed the results of periodic walkdowns by operations personnel and reviewed a sample of corrective actions generated as a result of those walkdowns to assess whether issues were appropriately identified and prioritized.

b.

Findinos No findings were identified.

.2.3.4 NRC lnformation Notice 89-44: Hvdrooen Storaoe On The Roof of The Control

Room a, Inspection Scope NRC lN 89-44 discussed potential generic problems pertaining to the storage of hydrogen in the vicinity of safety-related structures and air pathways into safety-related structures. Hydrogen is used on pressurized water reactor (PWR) plants for providing a cover gas in the volume control tank and for cooling the main turbine generator. The team reviewed the licensee's evaluation and disposition of the lN. The team reviewed the licensee's applicable procedures for hydrogen storage and makeup and performed walkdowns to assess the adequacy of the hydrogen storage methods.

b.

Findinqs No findings were identified.

OTHER ACTIVITIES

4c.42 ldentification and Resolution of Problems (lP 71152)

The team reviewed a sample of problems that Constellation had previously identified and entered into their CAP. The team reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. In addition, the team reviewed condition reports written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the CAP. The specific corrective action documents that were sampled and reviewed by the team are listed in the Attachment.

b.

Findinqs No findings were identified.

4046 Meetinqs. Includino Exit The team presented the inspection results to Mr. J. Carlin, Site Vice President, and other members of Constellation's staff at an exit meeting on November 11,2Q10. The team reviewed proprietary information, which was returned to Constellation at the end of the inspection. The team verified that none of the information in this report is proprietary.

ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Constellation Personnel

D. Crowley, EDG System Engineer

J. Jackson, Senior Licensing Engineer

D. Peters, Motor Operated Valve Engineer

R. Reissner, Senior Reactor Operator

K. Reynolds, Supervisor, Electrical Design Engineering

M. Zweigle, Supervisor, Mechanical Design Engineering

LIST OF ITEMS

Opened and Closed

0500024412010009-01 NCV Inadequate Evaluation of Breaker Coordination for Amptector Type LSG Trip Unit Discriminator Feature (1R21.2.1.1)

0500024412010009-02 NCV Inadequate Translation of NPSH Design Limits into EOPs (1R21 .2.1.2)