Text

IR 05000247/2015007 and 05000286/2015007; 7/20/15 to 8/20/15; Indian Point, Units 2 and 3; Component Design Bases Inspection
	ML15278A301
Person / Time
Site:	Indian Point
Issue date:	10/05/2015
From:	Paul Krohn; Engineering Region 1 Branch 2
To:	Coyle L; Entergy Nuclear Operations
References
	FOIA/PA-2016-0148 IR 2015007
	Download: ML15278A301 (52)
	v • d • e

[Type here]

UNITED STATES tober 5, 2015

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNITS 2 and 3 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000247/2015007 and 05000286/2015007

Dear Mr. Coyle:

On August 20, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Indian Point Nuclear Generating Units 2 and 3. The enclosed inspection report documents the inspection results, which were discussed on August 20, 2015, with you and other members of your staff.

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

In conducting the inspection, the team examined the adequacy of selected components to mitigate postulated transients, initiating events, and design basis accidents. The inspection involved field walkdowns, examination of selected procedures, calculations and records, and interviews with station personnel.

This report documents four NRC-identified findings that were of very low safety significance (Green). These findings were determined to involve violations of NRC requirements. However, because of the very low safety significance of the violations and because they were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Senior Resident Inspector at Indian Point. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Senior Resident Inspector at Indian Point. In accordance with 10 CFR 2.390 of the NRCs 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 NRCs 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/

Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

Enclosure:

Inspection Report 05000247/2015007 and 05000286/2015007 w/Attachment: Supplementary Information

REGION I==

Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 Report Nos. 05000247/2015007 and 05000286/2015007 Licensee: Entergy Nuclear Northeast (Entergy)

Facility: Indian Point Nuclear Generating Units 2 and 3 Location: 450 Broadway, GSB Buchanan, NY 10511-0249 Inspection Period: July 20 through August 20, 2015 Inspectors: Joseph Schoppy, Senior Reactor Inspector Division of Reactor Safety (DRS), Team Leader Juan Ayala, Reactor Inspector, DRS Javier Brand, Reactor Inspector, DRS Michael Orr, Reactor Inspector, DRS David Werkheiser, Senior Resident Inspector Division of Reactor Projects (DRP)

Gerald Nicely, NRC Electrical Contractor Matityahu Yeminy, NRC Mechanical Contractor Approved By: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety i Enclosure

SUMMARY OF FINDINGS

IR 05000247/2015007 and 05000286/2015007; 7/20/15 - 8/20/15; Indian Point Nuclear

Generating (Indian Point) Units 2 and 3; Component Design Bases Inspection.

The report covers the Component Design Bases Inspection conducted by a team of five NRC inspectors and two NRC contractors. Four findings of very low safety significance (Green) were identified, all of which were considered to be non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process. Cross-cutting aspects associated with findings are determined using IMC 0310, Components Within the Cross-Cutting Areas.

The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision

NRC-Identified Findings

Cornerstone: Mitigating Systems

Green.

The team identified a finding of very low safety significance involving a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B,

Criterion III, Design Control. Specifically, Entergy failed to verify, in design basis calculations for Unit 2, that protective device settings do not allow connected Class 1E loads to become damaged or unavailable during a design basis event: (a) under normal voltage conditions; or (b) for a sustained degraded voltage and subsequent reconnection to the emergency diesel generator concurrent with: (1) a design basis event for the degraded voltage time delay of 8.4 - 11.4 seconds, and (2) a non-accident shutdown for the degraded voltage time delay of 153 - 207 seconds. Additionally, Entergy failed to periodically test the thermal overload relays protecting safety-related motor-operated valves (MOVs) to ensure that degradation or trip setpoint drift does not affect the reliability or availability of mitigating systems when called upon to operate. After identification, Entergy entered this issue into the corrective action program, performed several additional evaluations to verify operability, declared two low pressure injection valves inoperable, and replaced fuses to restore operability to these valves.

The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power,

Exhibit 2 - Mitigating Systems Screening Questions, and concluded it required a detailed risk evaluation. The detailed risk evaluation was performed by a Region I senior reactor analyst (SRA) and concluded that the postulated inoperability of the two low pressure injections valves resulted in a change in core damage frequency of 1E-7/year, or very low safety significance (Green).

ii

The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely matter. Specifically, Entergy did not systematically and effectively evaluate NRC Regulatory Issue Summary 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages. [P.5] (Section 1R21.2.1.1.b.1)

Green.

The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of their electrical design. Specifically,

Entergy failed to verify, in design basis calculations and/or periodic testing, that adequate voltages would be available to all Class 1E motors, motor-operated valves (MOVs), static loads, and motor control center (MCC) control circuits and contactors powered from the 480 volt distribution system with the voltage at the 480 volt safety-related switchgear operating at the minimum degraded voltage dropout setting including tolerances. After identification, Entergy entered the issues into the corrective action program and performed several additional evaluations to verify adequate voltage to Class 1E motors, MOVs, static loads, and MCC control circuits.

The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power,

Exhibit 2 - Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability.

The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely matter. Specifically, Entergy did not systematically and effectively evaluate NRC Regulatory Issue Summary 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages. [P.5] (Section 1R21.2.1.1.b.2)

Green.

The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because Entergy did not verify the adequacy of the safety-related battery test program.

Specifically, Entergy did not adequately account for the effects of elevated temperature in the immediate vicinity of the No. 33 125 volts, direct current (Vdc) battery to ensure accurate and up-to-date determination of the batterys expected service life, in accordance with the vendor manual. After identification, Entergy entered this issue into the corrective action program and contacted the battery vendor for additional guidance.

iii

The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609,

Appendix A, The Significance Determination Process for Findings at Power,

Exhibit 2 - Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability.

This finding was not assigned a cross-cutting aspect because it was a historical design issue not indicative of current performance. Specifically, the associated vendor technical manual guidance was not changed within the last 3 years and there was no recent operating experience that was directly applicable to the performance deficiency.

(Section 1R21.2.1.2)

Green.

The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not promptly identify and correct a condition adverse to quality. Specifically, in April 2002, Entergy initiated a corrective action condition report (CR) to evaluate and document the seismic adequacy of a 138KV transmission tower, located in close proximity to the Unit 2 emergency diesel generator (EDG) building; however, Entergy staff closed the CR without adequately evaluating and documenting the seismic qualification concern. Entergys short-term corrective actions included initiating a corrective action CR and performing a seismic qualification evaluation.

The team determined that the inadequate resolution of the condition adverse to quality is a performance deficiency that was within Entergys ability to foresee and correct. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the protection against external factors (seismic) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems (the EDGs, in particular) that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609,

Appendix A, The Significance Determination Process for Findings at Power,

Exhibit 2 - Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a qualification deficiency confirmed not to result in a loss of operability.

The finding has a cross-cutting aspect in the area of Human Performance,

Documentation, because Entergy did not create and maintain complete, accurate, and up-to-date documentation. Specifically, Entergy did not create and maintain complete, accurate, and up-to-date design basis documentation to ensure that an adverse seismic II/I interaction would not result in the loss of the EDG safety function following a seismic event. [H.7] (Section 1R21.2.1.3)iv

Other Findings

None v

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection (IP 71111.21)

.1 Inspection Sample Selection Process

The team selected risk significant components for review using information contained in the Indian Point Nuclear Generating (Indian Point) Units 2 and 3 Probabilistic Risk Assessments (PRAs) and the U.S. Nuclear Regulatory Commissions (NRC)

Standardized Plant Analysis Risk (SPAR) models for Indian Point Units 2 and 3.

Additionally, the team referenced the Plant Risk Information e-Books (PRIBs) for Indian Point Units 2 and 3 in the selection of potential components for review. In general, the selection process focused on components 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 associated with both safety-related and non-safety related systems, and included a variety of components such as pumps, tanks, diesel engines, batteries, relays, circuit breakers, and valves.

The team initially compiled a list of components based on the risk factors previously mentioned. Additionally, the team reviewed the previous component design bases inspection (CDBI) reports and excluded the majority of those components previously inspected. The team then performed a margin assessment to narrow the focus of the inspection to 22 components and 5 operating experience (OE) items. The team selected the Unit 2 and Unit 3 main steam and feedwater (FW) piping containment penetrations and an OE sample associated with containment design pressure to review for large early release frequency (LERF) implications. The teams evaluation of possible low design margin included consideration of original design 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 (a)(1) status, operability reviews for degraded conditions, NRC resident inspector insights, system health reports, and industry OE. Finally, consideration was also given to the uniqueness and complexity of the design and the available defense-in-depth margins.

The team performed the inspection as outlined in NRC Inspection Procedure (IP)

===71111.21. This inspection 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 and OE sample are discussed in the subsequent sections of this report. Documents reviewed for this inspection are listed in the Attachment.

.2 Results of Detailed Reviews

.2.1 Results of Detailed Component Reviews (22 samples)

.2.1.1 Unit 2 Degraded Voltage Relays

a. Inspection Scope

The team reviewed related design basis support documentation, drawings, Technical Specifications (TSs), and the Updated Final Safety Analysis Report (UFSAR) to identify the function and capability of the 480 volt (V) vital bus loss of voltage (LOV) and degraded voltage protection scheme. The team reviewed the protective relay setting calculations for the LOV and degraded voltage relays (DVRs) and applicable time delay settings to ensure that Entergy used applicable industry standards in determining the channel uncertainty for the relay monitoring circuits which Entergy used to determine the LOV and DVR setpoints and time delays, including the TS nominal trip setpoints and allowable calibration values. The team verified that the DVR and LOV relays were set in accordance with calculations, and that associated periodic TS surveillance and calibration preventive maintenance (PM) activities were consistent with calculation assumptions, associated time delays, and setpoint accuracy calculations. The team reviewed plant/grid interfaces and requirements and the auxiliary power system load flow and voltage drop calculations to ensure that the LOV and DVR settings and time delays were appropriate and adequate to protect the 480V vital buses from a LOV or degraded voltage condition other than the normal bus voltage transients associated with a design basis event. The team evaluated selected portions of Entergys response to NRC Generic Letter (GL) 2006-02, Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power, to determine the stations interface and coordination with the transmission system operator for plant voltage requirements. The team also performed several visual non-intrusive inspections of the relay cabinets, indications, and 480V switchgear to assess the installation, configuration, material condition, and potential vulnerability to hazards.

b. Findings

1. Protection of Class 1E Loads during Normal Voltage and Sustained Degraded Voltage

Conditions

Introduction.

The team identified a finding of very low safety significance (Green)involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of their electrical design.

Specifically, Entergy failed to verify, in design basis calculations for Unit 2, that protective device settings do not allow connected Class 1E loads to become damaged or unavailable:

(a) under normal voltage conditions during a design basis event; or

(b) for a sustained degraded voltage and subsequent reconnection to the emergency diesel generators (EDGs) concurrent with:

(1) a design basis event for the degraded voltage time delay of 8.4 - 11.4 seconds; and

(2) a non-accident shutdown for the degraded voltage time delay of 153 - 207 seconds. Additionally, Entergy failed to periodically test the thermal overload (TOL) relays protecting safety-related MOVs to ensure that degradation or trip setpoint drift does not affect the reliability or availability of mitigating systems when called upon to operate.

Description.

In June 1977, based on industry events involving degraded offsite power systems, the NRC issued letters to all operating nuclear power plant licensees requesting that licensees install degraded voltage protection schemes (DVRs for second level of voltage protection) for the station electric power system. The NRC Letter, Statement of Staff Positions Relative to Emergency Power Systems for Operating Reactors, Position 1c(3) stated that the allowable duration of a degraded voltage at all distribution levels shall not result in failure of safety systems or components. NRC Regulatory Issue Summary (RIS) 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages, states in part that the time delay chosen should be optimized to ensure that permanently connected Class 1E loads are not damaged under sustained degraded voltage conditions (such as a sustained degraded voltage below the DVR voltage setting(s) for the duration of the time delay setting).

The team noted that, despite this guidance, Entergy did not have Unit 2 calculations to verify that connected Class 1E loads do not become damaged or unavailable:

(a) under normal voltage conditions during a design basis event; or

(b) for a sustained degraded voltage and subsequent reconnection to the EDGs concurrent with: 1) a design basis event for the degraded voltage time delay of 8.4 - 11.4 seconds, and 2) a non-accident shutdown for the degraded voltage time delay of 153 - 207 seconds. Specifically, Entergy did not verify that the protective devices and/or TOL relays, as applicable, would not trip on safety-related running or starting loads, such as motors and critical MOVs, during an accident under the voltage conditions described above. The team noted that Entergy did not verify that the trip setpoint of the protection devices was established with all uncertainties resolved in favor of completing the safety-related action. In addition, Entergy did not adequately evaluate control power circuits for the Class 1E accident-initiated motors and MOVs to ensure that their control circuit fuses would not fail (blow) if the starter did not have sufficient voltage to pick-up during the sustained degraded voltage delay period.

Additionally, Entergy failed to periodically test the TOL relays protecting safety-related MOVs to ensure that degradation or trip setpoint drift does not affect the reliability or availability of mitigating systems when called upon to operate. Regulatory Guide 1.106, Revision 1, Thermal Overload Protection for Electric Motors on Motor-Operated Valves, specified methods acceptable to the NRC staff for complying with Appendix B, Criterion XI, Test Control, with regard to the application of TOLs that are integral with the motor starter for electric motors on MOVs. These methods would ensure, in part, that the TOLs would not needlessly prevent the motor from performing its safety-related function. The guide allowed the licensee to leave the TOL in the MOV circuit continuously, provided that they were sized properly and periodically tested, in order to ensure continued functional reliability and the accuracy of the trip setpoint. Without periodic testing to determine the susceptibility to trip setpoint drift, Entergy had not verified that the applications were adequate to ensure the safety function would be met for a design basis event, which can subject the MOVs to transient voltage dips, possible stall conditions, and degraded voltage. The team identified several MOVs where the TOL setting had little or no margin for setpoint drift in order to ensure that the safety function would be met under all design basis conditions.

On August 18, 2015, Entergy initiated condition report (CR) CR IP2-2015-03688 for the sustained degraded voltage case. Entergy performed a preliminary evaluation (EC-59116) of Unit 2 safety injection (SI) actuated motors and the critical MOVs that operate during the first seconds of the accident and identified that the upstream fuses on Unit 2 MOVs 746 and 747 could potentially trip during a design basis event with a concurrent sustained degraded voltage condition. The MOVs are normally closed and have a safety function to go open during a SI to provide low pressure reactor coolant system (RCS) injection from the residual heat removal (RHR) pumps. Since Entergy no longer had reasonable assurance that the two MOVs in question would provide their associated design basis function under all conditions, operations declared both MOVs and both trains of RHR inoperable and appropriately entered TS Limiting Condition for Operation (LCO) 3.0.3 (an 8-hour shutdown action statement) on August 18, 2015. On August 18, 2015, Entergy implemented EC-59435 to replace both fuses prior to the expiration of the TS LCO. Operations were in the TS 3.0.3 LCO for approximately 48 minutes. Engineering determined that there were no protection coordination issues with the new fuses. After maintenance replaced both old fuses, operations declared both RHR trains operable.

On August 19, 2015, Entergy initiated corrective action CR IP2-2015-3725 to include a past operability review for the fuses previously installed in the MOV 746 and MOV 747 control circuits during normal voltage and sustained degraded voltage conditions. On September 9, 2015, Entergy completed their past operability evaluation, Past Operability Review of MOV-746 and 747 with Shawmut Type A4J30 Fuses during Normal and Emergency Conditions, Revision 1. For the normal voltage case, Entergy determined that the MOVs were operable and would have performed their safety function if called upon based on the margin to the fuse trip setting, review of MOV locked rotor current traces, review of fuse coordination plots, and historical MOV performance (including testing). The team performed an independent corrective action program (CAP) database search dating back to January 2000 and an Indian Point Unit 2 Licensee Event Report (LER) search dating back to January 1980 and found no evidence to support that either of the RHR MOV fuses in question had tripped under any plant conditions (including shutdown, event response, testing, or post-maintenance testing). Based on this historical review, an evaluation of the operability evaluation, and discussions with NRC MOV subject matter experts, the team determined that Entergys past operability determination for the normal voltage case was adequate.

For the sustained degraded voltage case, however, Entergys evaluation stated It has also been demonstrated that while tripping of these fuses under sustained degraded voltage conditions could occur, the possibility is very remote. Sufficient barriers, e.g. Con Edisons Real Time Contingency Analysis Program, Load Tap Changer operation, and proceduralized operator response to 480V Safeguards Bus under voltage alarms in the Central Control Room, were in place to ensure past operability of MOV-746 and 747 with the original A4J30 fuses.

NRC IMC 0326, Operability Determinations & Functionality Assessments for Conditions Adverse to Quality or Safety, Section C.06 states, the use of PRA or probabilities of occurrence of accidents or external events is not consistent with the assumption that the event occurs, and is not acceptable for making operability decisions. In addition, the team noted that Position 1 of the NRC 1977 letter states that, The operating procedures and guidelines utilized by electric utilities and their interconnected cooperative organizations minimize the probability for the above conditions to occur. However, since degradation of an offsite power system that could lead to or cause the failure of redundant safety-related electrical equipment is unacceptable, we (NRC) require the additional safety margins associated with implementation of the protective measures."

The team concluded that Entergy did not provide a sufficient technical basis for MOV-746 and 747 past operability (with the originally installed Shawmut Type A4J30 fuses) under postulated design basis accidents and concurrent degraded grid voltage conditions. The failure to ensure MOV-746 and 747 would function under all assumed design basis conditions is an NRC-identified performance deficiency.

The team noted that Entergy had missed several recent opportunities to identify and correct the above performance deficiency prior to this NRC inspection. Specifically, in April 2012, Entergy completed their internal review of NRC RIS 2011-12, Revision 1, but did not fully evaluate the issue, assess potential Entergy shortcomings relative to the guidance, and implement timely corrective actions, as appropriate. In addition, the ongoing industry discussions involving the guidance provided in NRC RIS 2011-12 and the numerous associated CDBI findings at other sites during the previous 3 years also represent missed opportunities. The team concluded that the electrical design engineering staffs awareness and application of the industry OE regarding degraded voltage calculations and assumptions during the 2015 CDBI demonstrated that the above deficiencies were indicative of current performance.

Analysis.

The performance deficiency associated with this finding was that Entergy did not verify the adequacy of their design, with respect to postulated design basis events, under normal and degraded grid voltage conditions. Specifically, some protective device settings may inadvertently permit connected Class 1E loads to become damaged or unavailable. This performance deficiency was more than minor because it was similar to NRC IMC 0612, Appendix E, Examples of Minor Issues, example 3.j. Specifically, the absence of an engineering design analysis provided reasonable doubt of operability of safety-related motor-operated valves MOV-746 and 747 under normal grid voltage conditions and under postulated DBA and concurrent degraded grid voltage conditions.

In addition, the finding was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 - Mitigating Systems Screening Questions. For the normal voltage case, Entergy determined that the MOVs were operable and would have performed their safety function. An independent review by the team determined that this operability assessment was adequate. However, for the degraded voltage case, the originally installed fast-acting Shawmut Type A4J30 fuses would likely actuate prior to the MOVs successfully stroking open. This condition results in the potential inoperability of both trains of low pressure injection/recirculation for longer than the TS LCO allowed outage time and requires a detailed risk evaluation (DRE).

The Region I senior reactor analyst (SRA) performed the DRE using Systems Analysis Programs for Hands-On Evaluation (SAPHIRE) Revision 8.1.2 and the SPAR Model for Indian Point Unit 2, Model Version 8.19, (limited use model, dated 8/28/15). Based upon recent generic thermo-hydraulic analyses the IP2 SPAR model was revised by Idaho National Laboratories (INL) to more accurately credit mitigation of medium break loss-of-coolant accidents (MLOCAs) using the safety injection pumps and associated high pressure recirculation piggyback functions achieved through operator action. The SRA made additional changes to the model to conservatively bound and evaluate the risk significance of this performance deficiency, including: basic events LPI-MOV-CC-746 and LPI-MOV-CC-747 were set to TRUE (failure to open); the exposure time was set at one year; and the truncation set at 1E-12. From the initial SPAR model results, only the large, medium and small break LOCA event tree sequences were considered for further evaluation because only these events would lead to a prompt SI system actuation signal and subsequent emergency bus loading of safety-related MOVs and pumps. The inclusion of small break LOCAs is conservative, in that, not all small break LOCAs would immediately result in an SI actuation signal. Accordingly, the cumulative conditional core damage probability for these LOCA sequences was calculated at 5.5E-6. To account for the consequential degraded grid voltage condition, this CCDP value was multiplied by 2E-2, consistent with a 2002 NRC Office of Research study (ML022120661) that approximated the probability of a LOOP given a LOCA has occurred. As a result, the SRA determined that the estimated increase in core damage frequency (CDF) associated with this performance deficiency is 1E-7/year or very low safety significance (Green). As previously stated, this risk significance approximation is overly conservative and considered a worst case bounding evaluation. The dominant core damage sequences involve large break LOCAs where early low pressure injection is needed, but compromised by the assumed failure to open of injection valves MOV-746 and 747.

In accordance with IMC 0609 guidance, the bounding 1E-7/year delta CDF value warrants a review for potential LERF and external events contributions. Based upon the Indian Point Unit 2 being a pressurized water reactor with a large dry containment, this Type A finding screens out for LERF consideration based upon Table 5.2 of IMC 0609, Appendix H, Containment Integrity Significance Determination Process. Because the conditional event sequences of interest involve loss of coolant accidents, external events coincident with or contributing to these accidents would be of extremely low probability and considered beyond the plants design basis. Accordingly, there is no external event contribution to core damage risk for this issue.

The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not systematically and effectively, collect, evaluate, and implement relevant internal and external OE in a timely matter. Specifically, Entergy did not systematically and effectively evaluate NRC RIS 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages.

[P.5]

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. The 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 calculations, or by the performance of a suitable testing program.

Contrary to the above, from June 1977 to August 18, 2015, Entergy failed to verify through calculational methods that the protective device settings do not allow connected Unit 2 Class 1E loads to become damaged or unavailable during a design basis event under normal and degraded voltage conditions. Entergys short-term corrective actions included initiating CRs, performing several additional evaluations to verify operability, declaring two valves inoperable, and replacing fuses to restore operability for the two MOVs in question. Because this violation is of very low safety significance and has been entered into Entergys CAP (CR IP2-2015-03688 and CR IP2-2015-03725), this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy.

(NCV 05000247/2015007-01, Inadequate Design Verification That Protective Device Settings Do Not Allow Connected Class 1E Loads to Become Damaged or Unavailable Under Normal and Sustained Degraded Voltage Conditions during a Design Basis Event)

.2 Adequacy of Voltage for Safety-Related Loads at the Degraded Voltage Relay Setpoint

Introduction.

The team identified a finding of very low safety significance (Green)involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy had not verified the adequacy of their electrical design.

Specifically, Entergy failed to verify, in design basis calculations and/or periodic testing, that adequate voltages would be available to all Unit 2 Class 1E motors, MOVs, static loads, and MCC control circuits and contactors powered from the 480V distribution system with the voltage at the 480V safety-related switchgear operating at the minimum degraded voltage dropout setting including tolerances.

Description.

In June 1977, based on industry events involving degraded offsite power systems, the NRC issued letters to all operating nuclear power plant licensees requesting that licensees install degraded voltage protection schemes (DVRs for second level of voltage protection) for the station electric power system. On December 29, 2011, the NRC issued RIS 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages, to clarify the NRC staffs technical position on existing regulatory requirements regarding voltage studies necessary to support the DVR setpoint.

The NRC RIS stated in part that, Licensee voltage calculations should provide the basis for their DVR settings, ensuring safety-related equipment is supplied with adequate voltage (dependent on equipment manufacturers design requirements), based on bounding conditions for the most limiting safety-related load (in terms of voltage) in the plant. The team noted that, despite the NRC RIS guidance, Entergy did not have calculations to verify that adequate voltages would be available to all Class 1E motors, MOVs, static loads, and MCC control circuits and contactors powered from the 480V distribution system with the voltage at the 480V safety-related switchgear operating at the minimum degraded voltage dropout setting including tolerances. Entergy had calculations that evaluated the Unit 2 480V switchgear motor loads and MOVs that would be actuated at the onset of a design basis event. However, Entergy had not evaluated the remaining safety-related motors, static loads, and MOVs primarily fed from MCCs 24A, 26A, 26AA, 26B, 26BB, 26C, 27A, and 29A at the degraded voltage setpoint.

Additionally, the team noted that in 2001, an Entergy safety system self-assessment initiated a corrective action CR (CR-2001-9275) related to the DVR dropout setting to assure adequate voltage to the 120V MCC control circuits for SI components under worst case loading conditions given that some MCC contactors would need to operate at lower voltages than the vendor specified. In response to this CR, Entergy developed calculation FEX-00180-00, MCC Control Circuits Voltage Evaluations, and performed a limited sample of testing on Size 1 and 2 MCC contactors to ensure adequate voltage to operate at the degraded voltage setpoint. The team noted that calculation FEX-00180-00 also recommended additionally testing of Size 3, 4 and 5 contactors; however, Entergy did not perform periodic testing of the Size 1 and 2 contactors at the lower voltages and did not perform the recommended additional testing for the Size 3, 4, and 5 contactors.

On August 18, 2015, Entergy entered these issues into their CAP as CR IP2-2015-03702 and CR IP2-2015-03706, respectively. Entergy performed preliminary evaluations under EC-59123 and EC-59368 of the motors, MOVs, static loads, and MCC control circuits and concluded that all evaluated equipment remained operable, although five MOVs were determined to have reduced margins relative to Indian Point Unit 2s design basis DVR setpoint (operable but nonconforming). Entergy also identified that most of the motors supplied from MCCs had less than the vendor recommended running voltages. Entergy determined that all the contactors that they had previously credited using the one-time testing were acceptable without crediting the testing at the lower voltages except for two of the Size 5 contactors. Entergy determined that these two Size 5 contactors, which support the primary auxiliary building (PAB) exhaust fan and the containment building purge fans, require additional testing. Based on an independent review of Entergys revised calculations, operability determinations for the nonconforming components, and the available margin, the team concluded that Entergys evaluations and short-term corrective actions were adequate and provided reasonable assurance of operability pending final resolution.

The team noted that Entergy missed several recent opportunities to identify and correct the above performance deficiencies prior to the teams identification. Specifically, in April 2012, Entergy completed their internal review of NRC RIS 2011-12, Revision 1, but did not apply sufficient engineering rigor to fully evaluate the issue, assess potential Entergy shortcomings relative to the guidance, and implement timely corrective actions, as appropriate. In addition, the ongoing industry discussions involving the guidance provided in NRC RIS 2011-12 and the numerous associated CDBI findings at other sites during the previous 3 years also represent missed opportunities. The team concluded that the electrical design engineering staffs awareness and application of the industry OE regarding degraded voltage calculations and assumptions during the 2015 CDBI demonstrated that the above deficiencies were indicative of current performance.

Analysis.

The performance deficiency associated with this finding was that Entergy did not verify the adequacy of their design with respect to ensuring in design basis calculations and/or periodic testing, that adequate voltages would be available to all Class 1E motors, MOVs, static loads, and MCC control circuits and contactors with the 480V safety-related switchgear operating at the minimum DVR dropout setting. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 - Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability.

The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not systematically and effectively, collect, evaluate, and implement relevant internal and external OE in a timely matter. Specifically, Entergy did not systematically and effectively evaluate NRC RIS 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages. [P.5]

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. The 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, from June 1977 to August 18, 2015, Entergy failed to verify the adequacy of the design to assure that adequate voltages would be available to all Unit 2 Class 1E motors, MOVs, static loads, and MCC control circuits and contactors.

Entergys short-term corrective actions included initiating corrective action CRs and performing several additional evaluations to verify adequate voltage to Class 1E motors, MOVs, static loads, and MCC control circuits. Because this violation is of very low safety significance and has been entered into Entergys CAP (CR IP2-2015-03702 and CR IP2-2015-03706), this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000247/2015007-02, Inadequate Design Verification That Adequate Voltages Would be Available to All Class 1E Motors, MOVs, Static Loads, and MCC Control Circuits and Contactors at the Minimum DVR Dropout Setting)

.2.1.2 No. 33 125 Volt Direct Current Battery

a. Inspection Scope

The team inspected the Unit 3 No. 33 station battery to assess its ability to meet the design basis function of providing a reliable source of direct current (DC) power to connected loads under operating, transient, and accident conditions. The team reviewed design calculations, drawings, and plant procedures to ensure that the battery was designed and operated in accordance with the design and licensing bases. The team also reviewed battery discharge tests to determine if the results enveloped the design discharge requirements and to verify that testing was in accordance with design calculations, TSs, and industry standards; and that the results confirmed acceptable performance of the battery. The team performed walkdowns of the battery, DC buses, battery chargers, and associated distribution panels to assess the material condition, configuration control, and operating environment. The team reviewed a sample of maintenance work orders, corrective action CRs, and system health reports to assess system performance and to ensure that Entergy identified and corrected deficiencies at an appropriate threshold. Finally, the team discussed the performance, design basis, and maintenance history of the battery with the responsible design and system engineers to assess the batterys overall reliability.

b. Findings

Introduction.

The team identified a finding of very low safety significance (Green)involving a non-cited violation of 10 CFR, Part 50, Appendix B, Criterion XI, Test Control, for Entergys failure to verify the adequacy of the safety-related battery test program. Specifically, Entergy did not adequately account for the effects of elevated temperature in the immediate vicinity of the Unit 3 No. 33 125 volts, direct current (Vdc)battery to ensure accurate and up-to-date determination of the batterys expected service life, in accordance with the vendor manual.

Description.

Battery cell design assumes that for maximum battery life the batteries are located in a temperature environment of 77ºF. Battery No. 33 is located in the No. 31 EDG room, where room temperature was consistently measured between 84ºF and 91ºF, and at times even higher. Elevated temperature above the nominal 77ºF temperature cause the battery to degrade at an accelerated rate, and must be accounted for in the expected service life of the battery. Specifically, the battery vendor technical manual, Section 4.3, states in part, Operation at higher temperature increases capacity, but reduces life approximately 50 percent for every 15F (9C) rise.

Performance testing is required every 5 years in accordance with Unit 3 TS 3.8.4.4 to verify adequate battery capacity and performance; however, Entergy performs the test every 4 years during refueling outages at Unit 3. The TSs also require more frequent testing when batteries show signs of degradation or once the battery reaches 85 percent of expected life. Entergys Unit 3 battery testing adequately measured battery capacity and health to identify potential degradation; however, Entergys battery testing and monitoring program did not derate the expected battery service life to account for operation at higher temperatures. Unit 3 is committed to Institute of the Electrical and Electronics Engineers (IEEE) Standard 450-1995.

The 1995 version does not contain the more recent specific requirements (IEEE 450-2002 and IEEE 450-2010) for age degradation due to elevated temperature, but industry technical guidance documents, including their specific battery vendor manual provides guidance on elevated temperature degradation. The team performed independent calculations of the thermal aging of the No. 33 battery using the formula in IEEE 450-2010, Annex H. The team calculated that the No. 33 batterys expected life had been reduced from 20 years to 12.1 years and that increased testing (every 2 years vice every 5 years based on the most recent capacity test) was required starting at 10.3 years (85 percent of expected life).

Based on this calculation, battery No. 33 reached 85 percent expected life on July 16, 2015; however, Entergy did not derate the batterys expected life and were unaware that the battery had entered the 2-year increased testing periodicity required by TS 3.8.4.4.

The team noted that Entergy satisfactorily completed a modified performance test on the No. 33 battery on March 5, 2015, and that the next performance test was not required until March 2017. The team determined that the No. 33 battery remained operable based on recent test results and that Entergy had not missed any increased frequency TS-required tests to date. On August 18, 2015, Entergy documented this issue in CR IP3-2015-04366 and contacted the battery vendor for additional guidance.

Analysis.

The team determined that the failure to account for elevated battery room temperature effects on the batterys expected service life was a performance deficiency.

The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 - Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability.

This finding was not assigned a cross-cutting aspect because it was a historical design issue not indicative of current performance. Specifically, the associated vendor technical manual guidance was not changed within the last 3 years and there was not recent OE that was directly applicable to the performance deficiency.

Enforcement.

10 CFR Part 50, Appendix B, Criterion XI, Design Control, requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components (SSCs) will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to the above, prior to August 18, 2015, Entergy did not establish an adequate Unit 3 test program incorporating battery service life degradation calculations for elevated temperature to ensure increased testing would be performed on station batteries when required. Entergys short-term corrective actions included initiating a corrective action CR and contacting the battery vendor for additional guidance.

Because this violation is of very low safety significance and has been entered into Entergys CAP (CR IP3-2015-04366), this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000286/2015007-03, Failure to Account for Elevated Battery Room Temperature Effects on Battery Service Life)

.2.1.3 Unit 2 Emergency Diesel Generator Fuel Oil Transfer Pumps

a. Inspection Scope

The team inspected the Unit 2 fuel oil transfer pumps (FOTPs) to evaluate whether they were capable of meeting their design basis and operational requirements to provide fuel oil to each of the three EDG fuel oil day tanks. The team reviewed the test methodology and evaluated the ability of the pumps to deliver flow rates greater than the consumption rate of the EDGs. The team evaluated the available net positive suction head (NPSH)based on the pumps location. The team performed several walkdowns of the FOTP motors, located above the fuel oil storage tank (FOST), and interviewed system and design engineers to assess the material condition, operating environment, and configuration control. Finally, the team reviewed corrective action documents and system health reports to evaluate whether there were any adverse operating trends and to assess Entergys ability to evaluate and correct problems.

b. Findings

Introduction.

The team identified a finding of very low safety significance (Green)involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not promptly identify and correct a condition adverse to quality. Specifically, in April 2002, Entergy initiated a corrective action CR to evaluate and document the seismic adequacy of a 138KV transmission tower, located in close proximity to the Unit 2 EDG building, but the Entergy staff closed the CR without adequately evaluating, documenting, and resolving the seismic qualification concern.

Description.

During a walkdown of Unit 2 FOTP motors, the team noticed that transmission tower W was located in close proximity to the building housing all three Unit 2 EDGs. The team also noticed that transmission tower V was located in very close proximity to the opposite side of the same EDG building. The team requested the seismic analysis for both transmission towers; however, Entergys staff could not find any associated analyses in their document management system. Engineerings related document search found a 2002 CR (CR IP2-2002-03770) created to evaluate and document the seismic adequacy of both towers. However, the team identified that on April 10, 2002, civil engineering closed the 2002 CR without adequately evaluating or documenting the condition. Entergy informed the team that they had recently contracted an independent engineering firm to provide a non-safety related evaluation of tower W titled Tower W Overturning Assessment for FLEX Seismic and Tornado Events. The team noted that the draft calculation had undergone several recent Entergy engineering reviews and that Entergy preliminarily concluded that the interaction ratio for the tower W anchorage had very little margin. The draft calculation for tower W preliminarily determined that seismic forces govern because the forces exerted on the anchor bolts as a result of an earthquake were greater (by 31 percent) than the forces exerted on the same bolts by the assumed wind speed. Entergy could not find any documentation to support that the transmission towers were designed for a seismic event; rather, it appeared that the transmission towers were designed for high wind, not for an earthquake. Based on the very low margin on tower W, and that it appeared as if the transmission towers were originally designed for wind loading (while seismic forces were preliminarily determined to be more limiting), the team had a reasonable doubt of operability for tower V regarding its seismic adequacy. Based on the teams concerns, Entergy initiated CR IP2-2015-03721 on August 18, 2015, to scope transmission tower V into the reconstituted seismic analysis associated with transmission tower W and document the analysis in a safety-related calculation, as appropriate.

The team noted that, although the work was ongoing at the time of the inspection, that Entergy had applied sufficient engineering rigor to their seismic evaluation of transmission tower W. However, the team noted that Entergy had missed a recent opportunity to evaluate and document in an approved calculation the seismic adequacy of transmission tower V. On September 2, 2015, engineering completed calculation IP-CALC-15-00079, Tower V Overturning Assessment for Seismic and Tornado Events, to evaluate tower V's anchorage to the concrete foundation to determine if overturning would occur during a seismic or tornado design basis event. Based on this calculation, engineering determined that tower V's anchorage to the concrete foundation would not suffer a catastrophic failure during a seismic or tornado design basis event. The team independently reviewed Entergys calculation and determined that it adequately evaluated the seismic qualification of tower V.

Analysis.

The team determined that the inadequate resolution of the condition adverse to quality was a performance deficiency that was within Entergys ability to foresee and correct. Specifically, Entergy did not adequately evaluate and document the seismic adequacy of tower V. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the protection against external factors (seismic) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems (the EDGs, in particular)that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 - Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a qualification deficiency confirmed not to result in a loss of operability.

The team determined that the issue had a cross-cutting aspect in the area of Human Performance, Documentation, because station personnel did not create and maintain complete, accurate, and up-to-date documentation. Specifically, Entergy did not create and maintain complete, accurate, and up-to-date design basis documentation demonstrating that an adverse seismic II/I interaction would not result in the loss of the EDG safety function concurrent with a loss of offsite power following a seismic event.

This performance deficiency was considered current performance because Entergy personnel missed a recent opportunity to identify and evaluate the concern during their post-Fukushima reviews, especially considering that they were evaluating the seismic adequacy of transmission tower W which, similar to tower V, was located in close proximity to the EDG building. [H.7]

Enforcement.

10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, since April 10, 2002, Entergy staff did not correct a condition adverse to quality regarding the seismic adequacy of a transmission tower that could potentially adversely impact the Unit 2 EDGs. Entergys short-term corrective actions included initiating a corrective action CR and performing a seismic qualification evaluation. Because this violation is of very low safety significance and has been entered into Entergys CAP (CR IP2-2015-03721), this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000247/2015007-04, Less Than Adequate Corrective Actions Associated with an Evaluation of the Seismic Adequacy of a 138KV Transmission Tower Located Near the Unit 2 EDG Building)

.2.1.4 Unit 2 Residual Heat Removal Supply from Reactor Coolant System Isolation Valve

(AC-MOV-731)

a. Inspection Scope

The team inspected Unit 2 MOV AC-MOV-731 to verify that it was capable of performing its design functions. The MOV has an open function to support shutdown cooling via the RHR heat exchangers (HXs) and a closed function to support containment isolation during normal power operation. 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 ensure that the MOVs operation was consistent with the design and licensing bases assumptions. The team also reviewed associated testing procedures and valve specifications to ensure consistency with design basis requirements. The team reviewed periodic verification diagnostic test results and stroke test documentation to verify that acceptance criteria were met and consistent with the design basis. Additionally, the team verified that the MOVs safety functions were maintained in accordance with NRC GL 89-10 guidance by reviewing torque switch settings, performance capability, and design margins. Finally, the team reviewed corrective action documents to verify that Entergy appropriately identified and resolved deficiencies and properly maintained the valve.

b. Findings

.2.1.1 3 Emergency Diesel Generator No. 22 (Electrical Review)

No findings were identified.

.2.2.2 NRC Information Notice 2011-12: Reactor Trips Resulting from Water Intrusion into

Electrical Equipment

a. Inspection Scope

The team assessed Entergys applicability review and disposition of NRC IN 2011-12.

The IN was issued to inform licensees about OE regarding recent events involving water intrusion into electrical equipment that resulted in reactor trips. In addition, the IN described the root causes and corrective actions taken to prevent recurrence. The team assessed Entergys evaluation of the IN as it applied to Indian Point Unit 2 and Unit 3, including their review of the electrical equipment design to ensure that it remained reliable and that there were no vulnerabilities associated with possible water intrusion events.

The inspection included a review of corrective action documents, interviews with electrical engineering and operations personnel, and a complete independent walkdown of all accessible Unit 2 and Unit 3 safety-related and non-safety related electrical panels, MCCs, and switchgear rooms.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems (IP 71152)

a. Inspection Scope

The team reviewed a sample of problems that Entergy had previously identified and entered into the 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 CRs 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 the team sampled and reviewed are listed in the

.

b. Findings

No findings were identified.

4OA6 Meetings, including Exit

On August 20, 2015, the team presented the inspection results to Mr. Larry Coyle, Site Vice President, and other members of the Entergy staff. On September 17, 2015, the team discussed the results of the teams assessment of Entergys past operability evaluation for the two low-margin MOVs (see Section 1R21.2.1.1.b.1) with Mr. Larry Coyle, Site Vice President, and other members of the Entergy staff via a telephone conference call. The team verified that no proprietary information was retained by the inspectors or documented in the report.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

K. Alfieri, Mechanical Design Engineer

V. Bacanskas, Chief Engineer, Engineering and Technical Services

F. Bloise, Electrical Design Engineer

J. Bretti, Probabilistic Risk Assessment Engineer

J. Bridges, Nuclear Plant Operator

R. Burroni, Engineering Director

T. Chan, System Engineering Supervisor

L. Coyle, Site Vice President

G. Dahl, Licensing Specialist

J. DAntonio, Maintenance Rule Coordinator

R. Drake, Civil Design Engineering Supervisor

E. Ginzburg, Design Engineer

R. Gioggia, AFW System Engineer

M. Haggstrom, City Water System Engineer

J. Hill, Instrumentation and Controls Engineering Manager

C. Hock, Control Room Supervisor

A. Kaczmarek, Design Engineering

J. Kaczor, Design Engineering

J. Kirkpatrick, Regulatory and Performance Improvement Director

B. Lawrence, RCS System Engineer

R. Machado, 138KV System Engineer

M. Maddalo, Design Engineering, Civil

S. Malinski, Design Engineering, Civil

A. Melody, EDG System Engineer

J. Miu, RHR and SI System Engineer

D. Musiyenko, Electrical Design Engineer

P. Pennacchio, Mechanical 1st Line Supervisor

E. Portanova, 125V DC System Engineer

J. Raffaele, Design Engineering Supervisor

I. Sinert, HVAC System Engineer

D. Sparozic, Instrument Air System Engineer

R. Tompkins, Mechanical Maintenance Superintendent

B. Ulrich, Control Room Supervisor

E. Varas, Sr. Nuclear Mechanic

D. Verdile, Sr. Nuclear Mechanic

R. Walpole, Regulatory Assurance Manager

W. Wittich, FIN Supervisor

C. Zannelli, ESF and Reactor Protection & Controls System Engineer

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Open and

Closed

05000247/2015007-01 NCV Inadequate design verification that protective device settings do not allow connected Class 1E loads to become damaged or unavailable under normal and sustained degraded voltage conditions during a design basis event. (Section 1R21.2.1.1.b.1)

05000247/2015007-02 NCV Inadequate design verification that adequate voltages would be available to all Class 1E motors, MOVs, static loads, and MCC control circuits and contactors at the minimum DVR dropout setting.

(Section 1R21.2.1.1.b.2)

05000286/2015007-03 NCV Failure to account for elevated battery room temperature effects on battery service life.

(Section 1R21.2.1.2)

05000247/2015007-04 NCV Less than adequate corrective actions associated with an evaluation of the seismic adequacy of a 138KV transmission tower located near the Unit 2 EDG building. (Section 1R21.2.1.3)

Discussed

050000247/2009007-03 FIN Failure to Identify Several Degraded City Water System Pipe Supports in the Utility Tunnel).

LIST OF DOCUMENTS REVIEWED