IR 05000382/2014007

IR 05000382/2014007; 10/06/2014 - 1/12/2015; Waterford Steam Electric Station, Unit 3; Component Design Basis Inspection

ML15022A637
Person / Time
Site:	Waterford
Issue date:	01/22/2015
From:	Anton Vegel Division of Reactor Safety IV
To:	Chisum M Entergy Operations
J. Dixon
References
EA-14-228 IR 2014-007
Download: ML15022A637 (74)

Text

January 22, 2015

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000382/2014007 AND PRELIMINARY GREATER THAN GREEN FINDING

Dear Mr. Chisum:

On November 6, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed the onsite portion of an inspection at the Waterford Steam Electric Station, Unit 3. The NRC inspectors discussed the results of this inspection with you and members of your staff at the conclusion of the onsite inspection and again on December 17, 2014, after additional in-office inspection.

A final telephonic exit meeting was conducted on January 12, 2015, with you and members of your staff. The inspectors documented the results of this inspection in the enclosed inspection report.

The enclosed report documents an NRC-identified finding that has been preliminarily determined to be of greater than very low safety significance (greater than Green), which may require additional inspections, regulatory actions, and oversight. This finding also constitutes an apparent violation of NRC requirements, which is being considered for escalated enforcement action in accordance with the NRC Enforcement Policy, which is available on the NRCs Web site at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html.

As described in Section 1R21.2.12.3 of the enclosed report, this finding is associated with through-wall corrosion on the emergency diesel generator fuel oil day tank vents. The holes in these vent pipes could allow water to enter the day tanks and contaminate the diesel fuel oil, challenging the operability and functionality of both safety-related emergency diesel generators.

While developing permanent corrective actions, your staff established compensatory measures to address the NRCs immediate safety concerns. These measures included installation of a temporary rubber wrap to cover the holes and a berm to direct water away from the vent pipes.

The preliminary risk significance of this finding was assessed using the NRCs significance determination process. This assessment was based on qualitative criteria and quantitative risk calculations. Both were required due to two areas of significant uncertainty in the risk

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E LAMAR BLVD ARLINGTON, TX 76011-4511 model: (1) the conditional probability of a loss of off-site power given a rain event of 5 inches per hour or more, and (2) the sensitivity of the stations diesel generators to water in the fuel stream.

These uncertainties in the risk model resulted in possible significance outcomes from low-to-moderate safety significance (White) to high safety significance (Red). The qualitative criteria considered are described in the preliminary detailed risk evaluation, which is attached to the enclosed report. We recognize that because of the large uncertainties associated with the preliminary risk determination for this finding, more information is needed to determine its final risk significance. If you have any further information that might clarify these uncertainties, we request that you provide it either in writing or during a scheduled regulatory conference.

We intend to complete and issue our final risk significance determination within 90 days from the date of this letter. The NRCs significance determination process is designed to encourage an open dialogue between your staff and the NRC; however, the dialogue should not affect the timeliness of our final determination.

Before the NRC makes a final decision on this matter, you may choose (1) to attend a regulatory conference, where you can present to the NRC your point of view on the facts and assumptions used to arrive at the finding and assess its significance; or (2) to submit your position on the finding to the NRC in writing. If you request a regulatory conference, it should be held within 30 days of your receipt of this letter. We encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective.

If you choose to attend a regulatory conference, it will be open for public observation. The NRC will issue a public meeting notice and press release to announce the conference. If you decide to submit only a written response, it should be sent to the NRC within 30 days of your receipt of this letter. If you choose not to request a regulatory conference or to submit a written response, you will not be allowed to appeal the NRCs final significance determination.

Please contact Mr. Eric Ruesch, Branch Chief (Acting), Engineering Branch 1, within 10 days from the issue date of this letter at 817-200-1126 or eric.ruesch@nrc.gov, and in writing, to notify the NRC of your intentions. If we have not heard from you within 10 days, we will continue with our significance determination and enforcement decision. Because the NRC has not made a final determination in this matter, no notice of violation is being issued with this report. Please be advised that the characterization of the apparent violation may change based on further NRC review.

In addition to this finding, NRC inspectors documented six findings of very low safety significance (Green) in the enclosed report. Five of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest any of these violations or their significance, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Waterford Steam Electric Station, Unit 3. If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 IV; and the NRC resident inspector at the Waterford Steam Electric Station, Unit 3.

In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management 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/

Anton Vegel, Director Division of Reactor Safety

Docket: 50-382 License: NPF-38

Enclosure:

Inspection Report 05000382/2014007 w/Attachments:

1. Supplemental Information 2. Detailed Risk Evaluation

Distribution for Waterford Steam Electric Station, Unit 3

- 1 -

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket:

05000382 License:

NPF-38 Report No.:

05000382/2014007 Licensee:

Entergy Operations, Inc.

Facility:

Waterford Steam Electric Station, Unit 3 Location:

17265 River Road Killona, LA 70057 Dates:

October 6, 2014, through January 12, 2015 Team Leader:

J. Dixon, Senior Reactor Inspector, Engineering Branch 1 Inspectors:

B. Larson, Senior Operations Engineer, Operations Branch S. Makor, Reactor Inspector, Engineering Branch 1 G. Ottenberg, Senior Reactor Inspector, Engineering Branch 1, Region II Accompanying Personnel:

G. Gardner, Contractor, Beckman and Associates P. Wagner, Contractor, Beckman and Associates Approved By:

Eric Ruesch, Chief (Acting)

Engineering Branch 1 Division of Reactor Safety

- 2 -

SUMMARY

IR 05000382/2014007; 10/06/2014 - 1/12/2015; Waterford Steam Electric Station, Unit 3;

Component Design Basis Inspection

The inspection activities described in this report were performed between October 6, 2014, and January 12, 2015, by three inspectors from the NRCs Region IV office, one inspector from the NRCs Region II office, and two contractors. Additional in-office inspection was performed through December 17, 2014. The enclosed inspection report documents one finding that has been preliminarily determined to be greater than very low safety significance (greater than Green) and may require additional inspections, regulatory actions, and oversight. The finding is also an apparent violation of NRC requirements and is being considered for escalated enforcement action in accordance with the Enforcement Policy. In addition, six findings of very low safety significance (Green) are documented in this report. Five of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310,

Components Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Cornerstone: Mitigating Systems

Green.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Action, which states, 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 nonconformance are promptly identified and corrected. Specifically, from October 27 through December 13, 2012, and on May 1, 2014, the licensee failed to identify and evaluate the impact of elevated bus voltages that exceeded the allowable voltage on the 480 VAC Class 1E Bus 3B31, a condition adverse to quality. In response to this issue, the licensee completed an operability determination with plans to evaluate any trends requiring additional actions. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2014-05458.

The team determined that the failure to identify and evaluate the impact of elevated bus voltages was a performance deficiency. This performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to identify and evaluate elevated voltages on the 480 VAC Class 1E Bus 3B31 that exceeded allowable operability limits.

In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2,

Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with trending because the licensee failed to periodically analyze information in the aggregate to identify programmatic and common cause issues (P.4). (Section 1R21.2.2)

Green.

The team identified a Green finding for inadequate station procedures for the temporary emergency diesel generators. Specifically, the licensee failed to ensure that Procedures OP-TEM-008, Emergency Diesel Generator A(B) Backup Temporary Diesel Generators, and ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4kV Buses (MODES 1-6), were maintained to ensure that the temporary diesels had enough capacity to supply auxiliary power to the required safe-shutdown loads. The team determined that the licensee failed to clearly establish appropriate instructions to ensure that operators would be running and verifying loads according to the prime rating, that three temporary diesels were capable of operating/connecting in parallel, and that required and desired loads were consistent among procedures and evaluations.

In response to this issue, the licensee evaluated and updated station procedures, specified prime loading limitations, updated vendor contracts, incorporated procedure improvements as a result of training, and updated the adverse weather procedure. This finding was entered into the licensees corrective action program as CR-WF3-2014-05662 and CR-WF3-2014-05582.

The team determined that failure to maintain procedures to ensure the temporary diesels have enough capacity to supply auxiliary power to required safe-shutdown loads was a performance deficiency. This performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences.

Specifically, the licensee failed to update Procedures OP-TEM-008 and ME-001-012, and vendor documents in accordance with engineering evaluation EC-47496 in a timely manner and prior to performance of the emergency diesel generator outage in January 2014.

In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2,

Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with teamwork because the licensee failed to ensure that individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained (H.4). (Section 1R21.2.7)

Green.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Action, which states, 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 nonconformance are promptly identified and corrected. Specifically, between October 8 and 16, 2014, the licensee failed to initiate a condition report to evaluate the lack of missile protection on the emergency diesel generator A and B storage tank vents, a nonconformance that is a condition adverse to quality. In response to this issue, the licensee performed an operability determination to address the teams concerns and initiated a separate condition report to document the failure to initiate a report for a condition adverse to quality. This finding was entered into the licensees corrective action program as CR-WF3-2014-05341 and CR-WF3-2014-05738.

The team determined that the failure to initiate a condition report to evaluate the lack of missile protection on the emergency diesel generator A and B storage tank vents for 8 days was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences.

Specifically, the licensee failed to promptly initiate and evaluate a condition adverse to quality, a design nonconformance on the emergency diesel generator A and B storage tank vents for missile protection.

In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2,

Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with work management because the licensee failed to implement a process where nuclear safety is the overriding priority and the need for coordinating with different work groups (H.5). (Section 1R21.2.12.1)

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, which states, 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. Specifically, prior to November 6, 2014, the licensee failed to verify the adequacy of design of the vents for the emergency diesel generator A and B day tanks and storage tanks to withstand impact from a wind-driven missile, or to evaluate for exemption from missile protection requirements using an approved methodology. In response to this issue, the licensee performed an evaluation using the TORMIS computer simulation code that supported a determination of operability and a future licensing basis change. TORMIS is a methodology described in Electric Power Research Institute (EPRI) Technical Report NP-2005, Tornado Missile Simulation and Design Methodology, dated August 1981, which was approved for use by Waterford in the Safety Evaluation related to License Amendment 168. This finding was entered into the licensees corrective action program as CR-WF-2014-05131, CR-WF3-2014-5341, and CR WF3-2014-5412.

The team determined that the failure to evaluate the lack of missile protection on the emergency diesel generator A and B day and storage tank vents was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to evaluate a design nonconformance on the emergency diesel generator A and B day and storage tank vents for lack of missile protection.

In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2,

Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.12.2)

TBD. The team identified an apparent violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Action, which states, 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 nonconformance are promptly identified and corrected. Specifically, prior to October 22, 2014, the licensee failed to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents, a condition adverse to quality. Prior to discovery by the team, the licensee had been unaware of the corrosion, which was significant enough that a through-wall hole had formed at the base of the each vent pipe where it penetrates the roof. Consequently, any water that collects on the roof of the building would have the potential to drain into the day tanks.

The licensee performed an immediate operability determination and concluded that the diesel and its support systems were operable based on no severe weather in the area.

While evaluating permanent corrective actions, the licensee installed a temporary repair to the vent pipes using a rubber wrap and installed a small concrete berm to minimize the potential amount of water in the immediate area. This finding was entered in to the licensees corrective action program as CR-WF3-2014-05413.

The team determined that the failure to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control and equipment performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences.

Specifically, the licensee failed to identify, evaluate, and correct through-wall corrosion on the emergency diesel generator A and B day tank vents. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened to Exhibit 4, External Events Screening Questions, because it was potentially risk-significant due to seismic, flooding, or severe weather. Per Exhibit 4 this finding screened to a Detailed Risk Evaluation because if the safety function were assumed completely failed it would degrade two trains of a multi-train system and it would degrade one or more trains of a system that supports a risk-significant system.

A Region IV senior reactor analyst performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk-important sequences included heavy-rain-induced losses of off-site power with the consequential failure of both emergency diesel generators. The ability to restore off-site power within four hours was important to avoid core damage. The finding was not significant to the large early release frequency. See Attachment 2, Detailed Risk Evaluation, for a detailed review of qualitative criteria also considered.

This finding had a cross-cutting aspect in the area of human performance associated with procedure adherence because the licensee failed to ensure that individuals follow process, procedures, and work instructions (H.8). (Section 1R21.2.12.3)

Cornerstone: Barrier Integrity

Green.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, which states 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.

Specifically, since January 18, 2006, the licensee has failed to evaluate the adequacy of design of the main feedwater isolation valve operators to provide adequate thrust in accordance with the licensees analysis methodology described in EPRI topical report TR 103237-R2, EPRI MOV Performance Prediction Program. In response to this issue, the licensee recalculated the required thrust and performed an evaluation that supported a determination that the valves remained operable. This finding was entered into the licensees corrective action program as CR-WF3-2014-05690.

The team determined that the failure to evaluate the required thrust for operation of the main feedwater isolation valves, assuming an appropriate valve-disk-to-seat coefficient of friction, was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events.

Specifically, the incorrect coefficient of friction assumption resulted in a reasonable question of operability of the main feedwater isolation valves to operate under the design basis condition of a main steam line break while auxiliary feedwater is supplying inventory to the steam generators. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012,

Exhibit 3, Barrier Integrity Screening Questions, this finding screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.15)

Green.

The team reviewed a self-revealing Green non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, which states, 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. Specifically, prior to the failure of main steam isolation valve MS-124A on January 5, 2013, the licensee failed to have an adequate weak-link evaluation for the main steam isolation valves. In response to this event, the licensee performed a seismic weak-link evaluation of the main steam isolation valves that supported a determination that the valves were operable. This finding was entered into the licensees corrective action program as CR-WF3-2014-05708.

The team determined that the failure to evaluate the main steam isolation valve maximum allowed thrust, assuming appropriate values for the structural limitations of the valve and actuator, was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events.

Specifically, the licensee used a non-conservative value for the maximum allowed thrust, and the error resulted in a failure of main steam isolation valve MS-124A, because the allowable nitrogen pressure for the valve actuator was inappropriate. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, this finding screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

(Section 1R21.2.16)

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

This inspection of component design bases verifies that plant components are maintained within their design basis. Additionally, this inspection provides monitoring of the capability of the selected components and operator actions to perform their design basis functions. As plants age, modifications may alter or disable important design features making the design bases difficult to determine or obsolete. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity cornerstones for which there are no indicators to measure performance.

1R21 Component Design Basis Inspection

.1 Overall Scope

To assess the ability of the Waterford Steam Electric Station, Unit 3, equipment, and operators to perform their required safety functions, the team inspected risk-significant components and the licensees responses to industry operating experience. The team selected risk-significant components for review using information contained in the Waterford Steam Electric Station, Unit 3, probabilistic risk assessments and the U. S. Nuclear Regulatory Commissions (NRC) standardized plant analysis risk model.

In general, the selection process focused on components that had a risk achievement worth factor greater than 1.3 or a risk reduction worth factor greater than 1.005. The items selected included components in both safety-related and nonsafety-related systems including pumps, circuit breakers, heat exchangers, transformers, and valves.

The team selected the risk-significant operating experience to be inspected based on its collective past experience.

To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed calculations to independently verify the licensees conclusions. The team also verified that the condition of the components was consistent with the design basis and that the tested capabilities met the required criteria.

The team reviewed maintenance work records, corrective action documents, and industry operating experience records to verify that licensee personnel considered degraded conditions and their impact on the components. For selected components, the team observed operators during simulator scenarios, as well as during simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design basis have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions because of modifications, and margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results; significant corrective actions; repeated maintenance; Title 10 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC resident inspector input of problem equipment; system health reports; industry operating experience; and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in-depth margins.

The inspection procedure requires a review of 15 to 25 total samples that include risk-significant and low design margin components, components that affect the large-early-release-frequency (LERF), and operating experience issues. The sample selection for this inspection was 18 components, 3 components that affect LERF, 6 operating experience items, and 3 event based activities associated with the components. The selected inspection and associated operating experience items supported risk-significant functions including the following:

a. Electrical power to mitigation systems: The team selected several components in the electrical power distribution systems to verify operability to supply alternating current (ac)and direct current

(dc) power to risk-significant and safety-related loads in support of safety system operation in response to initiating events such as loss of off-site power, station blackout, and a loss-of-coolant accident with off-site power available. As such the team selected:

• 4160 Vac Class 1E switchgear, Bus 3A3

• 480 Vac Class 1E switchgear, Bus 3B31

• Emergency Feedwater Pump Motor EMTR-3A10A

• Fast Bus Transfer Station Start-up Transformer 3B Breakers

• Static Uninterruptible Power Supply B ID-EUPSB

• Static Uninterruptible Power Supply MA ID-EUPSMA

• Temporary Emergency Diesel Generators used during Emergency Diesel Generator Maintenance

b. Components that affect LERF: The team reviewed components required to perform functions that mitigate or prevent an unmonitored release of radiation. The team selected the following components:

• Containment Atmosphere Purge Make-up Air Isolation Valves CAP-MVAAA-103, and -104; and Exhaust Isolation Valves CAP-MVAAA-203, and -204

• Maintenance Hatch CB-MEAH-0001 O-Rings

• Reactor Containment Building - Steel Containment Vessel c. Mitigating systems needed to attain safe shutdown: The team reviewed components required to perform the safe shutdown of the plant. As such the team selected:

• Containment Coolers CCS-MAHU-0001A, B, C, and D

• Emergency Diesel Generator Day Tank and Storage Tank Vents

• Emergency Feedwater Primary Isolation Valves EFW-228A and 229A

• Essential Chiller RFR-MCHL-0001A

• Main Feedwater Isolation Valve FW-184A

• Main Steam Isolation Valve MS-124A

• Shield Building Concrete Structure

• Steam Generator Power Operated Atmospheric Dump Valve MS-116A

.2.1 4160 Vac Class 1E Switchgear Bus 3A3

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings and calculations, maintenance and test procedures, and condition reports associated with 4160 Vac Class 1E Switchgear Bus 3A3. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Distribution system one-line diagrams.

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Calculations for electrical distribution, system load flow/voltage drop, short-circuit, and electrical protection to verify that bus capacity and voltages remained within minimum acceptable limits.

• The protective device settings and circuit breaker ratings to ensure adequate selective protection coordination of connected equipment during worst-case short circuit conditions.

• Procedures for preventive maintenance, inspection, and testing of the bus, transformer, and associated circuit breakers to compare maintenance practices against industry and vendor guidance.

• Cable sizing for selected loads.

• Evaluation of the most recent grid stability study.

• Waterfords response to NRC Information Notice 2010-26, Submerged Electrical Cables.

b. Findings

No findings were identified.

.2.2 480 Vac Class 1E Switchgear Bus 3B31

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with 480 Vac Class 1E Switchgear Bus 3B31. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• System load flow/voltage drop and short circuit studies, under normal and design basis accident load conditions.

• Calculations for electrical distribution, system load flow/voltage drop, short-circuit, and electrical protection to verify that bus capacity and voltages remained within minimum acceptable limits.

• The protective device settings and circuit breaker ratings to ensure adequate selective protection coordination of connected equipment during worst-case short circuit conditions.

• Vendor and station single line, schematic, wiring, and layout drawings, including available short circuit current.

• Preventative maintenance, inspection, and testing procedures, including recently completed work orders.

• Cable sizing for the load center bus.

b. Findings

Introduction.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality. Specifically, the licensee failed to identify and evaluate the impact of elevated bus voltages during the period of October 27 through December 13, 2012, and on May 1, 2014, that exceeded the allowable voltage on the 480 Vac Class 1E Bus 3B31.

Description.

The team reviewed the performance indicator data for the 480 Vac Class 1E Bus 3B31 and identified instances during refueling outages 18 and 19 where the licensee exceeded the voltage limit acceptance criteria. The team noted multiple instances where the voltage limit was exceeded, but there were two instances where a condition report had not been generated and the condition had not been evaluated. The first occurred during refueling outage 18from October 27 through December 13, 2012, the bus voltage rose above the 506 VAC limit several times with the longest period being 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The sustained voltage ranged from 508-514 VAC. The second occurred during Refueling Outage 19, on May 1, 2014, with a sustained elevated bus voltage of between 506-509 VAC for about an hour. Neither of these conditions was identified or addressed in the corrective action program until identified by the team. The lack of trending and periodically analyzing information that the licensee knows could become an issue contributed to the cross-cutting aspect.

Procedure EN-LI-102, Corrective Action Program, Revision 24, requires that a condition report be initiated promptly for conditions adverse to quality, and that operability, functionality, and immediate reportability be reviewed for the condition.

9.2, Section 4, Design and Licensing Basis Issues, specifically provides examples of adverse conditions as they concern design basis issues. The licensees corrective actions included evaluating the missed events by comparing Bus 3B31 and 3A31 voltages to ensure that Bus 3A31 was within the required band of 450-506 Vac, and performing an operability determination. The licensee generated CR-WF3-2014-05458 to address this performance deficiency and to identify any trends requiring additional actions.

Analysis.

The team determined that the failure to identify and evaluate the impact of elevated bus voltages was a performance deficiency. This performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to identify and evaluate voltages on the 480 Vac Class 1E Bus 3B31 that exceeded allowable operability limits. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with trending because the licensee failed to periodically analyze information in the aggregate to identify programmatic and common cause issues (P.4).

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states 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 nonconformance are promptly identified and corrected. Contrary to the above, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected. Specifically, during the period of October 27 through December 13, 2012, and on May 1, 2014, the licensee failed to identify and evaluate the impact of elevated bus voltages that exceeded the allowable voltage on the 480 VAC Class 1E Bus 3B31, a condition adverse to quality. In response to this issue, the licensee completed an operability determination with plans to evaluate any trends requiring additional actions. This finding was entered into the licensees corrective action program as CR WF3 2014-05458. Because this finding was of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000382/2014007-01, Failure to Identify and Evaluate Elevated Bus Voltages.

.2.3 Emergency Feedwater Pump Motor EMTR-3A10A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Emergency Feedwater Pump Motor EMTR-3A10A. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Vendor technical information and specifications, including usage in motor loading calculations, operating, maintenance, and testing procedures.

• Emergency diesel generator loading calculations associated with the motor.

• Completed surveillance and maintenance procedures to verify adequate testing.

• Electrical schematics and control wiring diagrams to verify the automatic start features.

b. Findings

No findings were identified.

.2.4 Fast Bus Transfer Station Start-up Transformer 3B Breakers

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Fast Bus Transfer Station Start-up Transformer 3B Breakers. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Distribution system one-line diagrams and relevant electrical schematics.

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Procedures for circuit breaker preventive maintenance, inspection, and testing to compare maintenance practices against industry and vendor guidance.

• Completed surveillance and maintenance procedures to verify adequate testing.

b. Findings

No findings were identified.

.2.5 Static Uninterruptible Power Supply B ID-EUPSB

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings and calculations, maintenance and test procedures, and condition reports associated with Static Uninterruptible Power Supply B ID-EUPSB. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• One-line diagrams and design basis documents for the inverters electrical distribution systems to identify requirements and interfaces.

• Preventive maintenance activities to verify the inverter system was maintained in accordance with manufacturer recommendations including replacement of age-sensitive components, and corrective action program reports to verify the monitoring of potential degradation.

• Short circuit calculations, inverter sizing calculations, coordination studies, and voltage drop calculations.

• Calculations to verify that branch circuit load and load voltage requirements had been properly translated into inverter sizing and voltage drop calculations.

• Station blackout calculations to verify that the inverters output would remain adequate to power associated instrumentation and control loads at reduced battery voltage levels.

• As part of the station blackout review, the team evaluated the adequacy of the Train B battery to supply sufficient power at the conclusion of the four hour coping cycle to energize required loads, such as the ability to flash the field of its associated emergency diesel generator.

• Procedures for circuit breaker preventive maintenance, inspection, and testing to compare maintenance practices against industry and vendor guidance.

b. Findings

No findings were identified.

.2.6 Static Uninterruptible Power Supply MA ID-EUPSMA

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings and calculations, maintenance and test procedures, and condition reports associated with Static Uninterruptible Power Supply MA ID-EUPSMA. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• One-line diagrams and design basis documents for the inverters electrical distribution systems to identify requirements and interfaces.

• Preventive maintenance activities to verify the inverter system was maintained in accordance with manufacturer recommendations including replacement of age-sensitive components, and corrective action program reports to verify the monitoring of potential degradation.

• Short circuit calculations, inverter sizing calculations, coordination studies, and voltage drop calculations.

• Calculations to verify that branch circuit load and load voltage requirements had been properly translated into inverter sizing and voltage drop calculations.

• Procedures for circuit breaker preventive maintenance, inspection, and testing to compare maintenance practices against industry and vendor guidance.

b. Findings

No findings were identified.

.2.7 Temporary Emergency Diesel Generators used During Emergency Diesel Generator

Maintenance

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the temporary emergency diesel generators used during emergency diesel generator maintenance. The team also conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Procedures for operations and maintenance

• Vendor manual and specifications

• Schematics and control wiring diagrams

• Load calculations of record and supporting documentation

• Coordination study and required/desired loads sizing

• Completed preventive maintenance and surveillance tests

• Corrective actions

b. Findings

Introduction.

The team identified a Green finding for inadequate station procedures for the temporary emergency diesel generators. Specifically, the licensee failed to ensure that Procedures OP-TEM-008, Emergency Diesel Generator A(B) Backup Temporary Diesel Generators, Revision 7, and ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4kV Buses (MODES 1-6), Revision 308, were maintained and updated to ensure that the temporary diesels had enough capacity to supply auxiliary power to the required safe-shutdown loads.

Description.

The team reviewed the loading requirements and procedures associated with the temporary emergency diesel generators. The purpose of the temporary diesels is to provide the licensee with flexibility in the performance of both corrective and preventative maintenance during power operation. The availability of the temporary diesels allows the site to extend the Technical Specification 3.8.1 allowed outage time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 10 days as approved in the safety evaluation related to License Amendment 166. The temporary diesels are commercial-grade diesels capable of supplying auxiliary power to, at a minimum, required safe-shutdown load on the emergency diesel generator train that was removed from service for maintenance.

During the review, the team identified that Procedures OP-TEM-008 and ME-001-012 had not been maintained and updated prior to the performance of the emergency diesel generator outage in January 2014, nor in a timely manner. The most current engineering evaluation EC-47496, Capability of TEDG to Start Large Motor, approved on December 19, 2013, recommended three temporary diesels connected in parallel with 70 percent average load factor and prime power ratings, and that changes be made to the vendor contract and applicable procedures. The licensee noted that two diesels connected in parallel could support the loads, but recommended that three diesels be used to increase operation margin.

The engineering evaluation also described the capability of the temporary diesel running load based on mode of operation and that they can be classified as continuous, prime, or standby. The team found that the licensee never clearly established appropriate instructions to ensure that the operators would be running and verifying the loads according to the prime rating, that three temporary diesels were capable of operating/connecting in parallel, and that the required and desired loads were consistent between procedures and evaluations. The procedure owner identified that a revision was required, but did not make any changes or inform other organizations about the necessary updates. This contributed to the cross-cutting aspect.

The licensees corrective actions included evaluating the inconsistencies in the procedures, evaluations, and contracts. As a result, the licensee generated condition report CR-WF3-2014-05662 to update ME-001-012 to specify 2500kW prime loading, update OP-TEM-008 to specify prime loading limitations, and update just-in-time training material, and condition report CR-WF3-2014-05582 to address the vendor contract and adverse weather concerns.

Analysis.

The licensees failure to include in station procedures the requirements that ensure the temporary diesels have enough capacity to supply auxiliary power to the required safe-shutdown loads was a performance deficiency. This performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to update Procedures OP-TEM-008, Emergency Diesel Generator A(B) Backup Temporary Diesel Generators, Revision 7, and ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4kV Buses (MODES 1-6), Revision 308, and vendor documents in accordance with engineering evaluation EC-47496, prior to the performance of the emergency diesel generator outage in January 2014, and in a timely manner. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with teamwork because the licensee failed to ensure that individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained (H.4).

Enforcement.

This finding does not involve enforcement action because no violation of a regulatory requirement was identified. In response to this issue, the licensee evaluated and updated station procedures, specified prime loading limitations, updated vendor contracts, incorporated procedure improvements as a result of training, and updated the adverse weather procedure. This finding was entered into the licensees corrective action program as CR-WF3-2014-05662 and CR-WF3-2014-05582. Because this finding does not involve a violation and was of very low safety significance, it is identified as FIN 05000382/2014007-02, Inadequate Station Procedures for Temporary Emergency Diesel Generator.

.2.8 Containment Atmosphere Purge Make-up Air Isolation Valves CAP-MVAAA-103 and

-104, and Exhaust Isolation Valves CAP-MVAAA-203, and -204

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, seismic weak link analysis, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Containment Atmosphere Purge Make-up Air Isolation Valves CAP-MVAAA-103 and

-104, and Exhaust Isolation Valves CAP-MVAAA-203 and -204. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Calculation EC-M97-041, Design Basis Review for Containment Purge Isolation Valves CAP-102, CAP-103, CAP-104, CAP-203, CAP-204, and CAP-205.

• Weak link analyses for the containment purge isolation valves.

• Local leak rate testing procedure and results of recent testing performed to meet the requirements of 10 CFR Part 50, Appendix J.

b. Findings

No findings were identified.

.2.9 Maintenance Hatch CB-MEAH-0001 O-Rings

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the Maintenance Hatch CB-MEAH-0001 O-Rings. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Detailed vendor drawings of the maintenance hatch and seals.

• Previous two work orders for inspection, cleaning, and buffing of the maintenance hatch seal seating surfaces.

• Containment leak-rate testing program and the past two containment integrated leak-rate tests.

• Quarterly inservice test results for the maintenance hatch seals.

b. Findings

No findings were identified.

.2.10 Reactor Containment Building - Steel Containment Vessel

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the reactor containment building, the steel containment vessel. The team conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Nuclear island and building design associated with the reactor containment building.

• Maintenance rule structural monitoring, scoping, and basis.

• Containment in-service inspection and testing program.

• Containment isolation and leakage rate testing.

• Reactor containment surface inspection layout for the inside and outside surfaces.

• Past two ASME Section XI visual inspection reports.

• Waterfords response to NRC Information Notice 2004-09, Corrosion of Steel Containment and Containment Liner.

b. Findings

No findings were identified.

.2.11 Containment Coolers CCS-MAHU-0001A, B, C, and D

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Containment Coolers CCS-MAHU-0001A, B, C, and D. The team also conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Design calculations for maximum reactor containment building temperature and pressure.

• Detailed plant drawings, purchase specifications, and operating, preventative maintenance and testing procedures.

• Control room indications for containment cooler operation and performance during normal and accident conditions.

• Waterfords response to NRC Information Notice 1996-45, Potential Common-Mode Post-Accident Failure of Containment Coolers.

b. Findings

No findings were identified.

.2.12 Emergency Diesel Generator Day Tank and Storage Tank Vents

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the emergency diesel generator day and storage tank vents. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Design basis adverse weather protection requirements.

• Normal and alternate diesel fuel oil fill procedures, including during site area flooding.

• Detailed plant drawings and operating, preventative maintenance, and testing procedures.

.1 Failure to Generate a Condition Report or to Evaluate Emergency Diesel Generator

Storage Tank Operability in a Timely Manner

Introduction.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality. Specifically, the licensee failed to initiate a condition report for a nonconformance on the emergency diesel generator A and B storage tank vents, a condition adverse to quality, for 8 days.

Description.

On October 8, 2014, during a plant walkdown the team identified that the emergency diesel generator A and B storage tank vents were outside the nuclear island plant structure and were not missile protected. The team shared this concern with the licensee and expected that a condition report and subsequent operability determination would be performed to address the apparent nonconforming condition.

On October 16, 2014, the team had still not been given a condition report that addressed the lack of missile protection. Upon questioning the licensee on the delay of the condition report, licensee personnel realized that they had still not generated a condition report. Procedure EN-LI-102, Corrective Action Program, Revision 24, requires for conditions adverse to quality that a condition report be initiated promptly/timely, and that operability, functionality, and immediate reportability determinations be reviewed for the event. Attachment 9.2, Section 4, Design and Licensing Basis Issues, specifically provides examples of adverse conditions as they concern design basis issues; a nonconforming condition is a specific example cited. Because the licensee failed to timely initiate a condition report on October 8, no determination of system operability was performed for more than a week after licensee personnel became aware of the nonconforming condition.

On October 16, 2014, the licensee initiated CR-WF3-2014-05341 and performed an operability determination to address the teams concerns. Additionally the licensee generated CR-WF3-2014-05738 to document the lack of initiating and evaluating a condition report for a nonconforming condition, a condition adverse to quality, in a timely manner.

Analysis.

The failure to initiate a condition report to evaluate the lack of missile protection on the emergency diesel generator A and B storage tank vents was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences.

Specifically, the licensee failed to initiate and evaluate a design nonconformance on the emergency diesel generator storage tank vents for missile protection for 8 days. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with work management because the licensee failed to implement a process where nuclear safety is the overriding priority and the need for coordinating with different work groups (H.5).

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states 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 nonconformance are promptly identified and corrected. The licensees measures are established by Procedure EN-LI-102 Corrective Action Program, Revision 24, which requires for conditions adverse to quality that a condition report be initiated promptly/timely, and that operability, functionality, and immediate reportability be reviewed for the event. Attachment 9.2, Section 4, Design and Licensing Basis Issues, specifically provides examples of adverse conditions as they concern design basis issues. Contrary to the above, from October 8 through October 16, 2014, the licensee failed to initiate a condition report to document and evaluate a condition adverse to quality. Specifically, the licensee failed to initiate a condition report to evaluate the lack of missile protection on the emergency diesel generator A and B storage tank vents, a nonconformance that is a condition adverse to quality, for 8 days. In response to this issue, the licensee performed an operability determination to address the teams concerns and initiated a separate condition report to document the lack of initiating and evaluating a condition report for a condition adverse to quality. This finding was entered into the licensees corrective action program as CR-WF3-2014-05341 and CR-WF3-2014-05738. Because this finding was of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000382/2014007-03, Failure to Initiate a Condition Report for a Condition Adverse to Quality.

.2 Failure to Evaluate Missile Protection Requirements for Emergency Diesel Generator

Day Tank and Storage Tank Vents

Introduction.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the missile protection requirements for the emergency diesel generator A and B day tank and storage tank vents.

Description.

On October 8, 2014, the team identified during a plant walk-down that the emergency diesel generator A and B storage tank vents were outside the nuclear island plant structure and were not missile protected. On October 22, 2014, the team questioned the missile protection of the emergency diesel generator A and B day tanks and determined that they were also not protected.

Appendix A to 10 CFR Part 50 contains the general design criteria for nuclear power plants. Specifically, Criterion 2, Design bases for protection against natural phenomena, states, in part, Structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena such astornados

[and] hurricanes, and, The design bases for these structures, systems and components shall reflect:...

(2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena. Criterion 4, Environmental and dynamic effects design bases, states, in part, Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions, and, These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles.

Upon reviewing the licensees design basis documents and general responses to General Design Criteria 2 and 4, the team determined that the licensee failed to properly protect the emergency diesel generator A and B day tank and storage tank vents from weather-related missiles. These vents do not have missile protection and the licensee does not have an exemption from the missile protection requirement. The licensee initiated CR-WF3-2014-05131, CR-WF3-2014-5341, and CR-WF3-2014-5412 to determine the protection requirements and perform an immediate operability determination. The licensee determined that missile protection was required for these locations. Subsequently, the licensee performed an evaluation using the TORMIS computer simulation code that supported a determination of operability and a future licensing basis change. TORMIS is a methodology described in Electric Power Research Institute (EPRI) Technical Report NP-2005, Tornado Missile Simulation and Design Methodology, dated August 1981, which was approved for use by Waterford in the Safety Evaluation related to License Amendment 168.

The TORMIS evaluation calculates a total probability per year of damage to all important structures, systems, and components due to a missile-generating wind event. It evaluates in the aggregate all known structures, systems, and components that are required to be missile protected and are not. This probability of damage must be less than the maximum allowed threshold provided for in the safety evaluation of 1.0 x 10-6 per year. Waterford Unit 3s TORMIS calculation resulted in a damage probability of 7.95 x 10-7 per year, which is below the allowable probability value. Therefore, now that the condition has been evaluated by an approved methodology and was determined to be acceptable, no missile protection is required to be installed on these vents.

Analysis.

The failure to evaluate the lack of missile protection on the emergency diesel generator A and B day tank and storage tank vents was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to evaluate a design nonconformance on the emergency diesel generator A and B day tank and storage tank vents for lack of missile protection. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, this finding screened as having very low safety significance (Green)because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states 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, prior to November 6, 2014, the licensee did not verify the adequacy of design of the emergency diesel generator A and B day tank and storage tank vents. Specifically, the licensee failed to have missile protection installed or an approved exemption excluding missile protection requirements. In response to this issue, the licensee performed a TORMIS evaluation that supported a determination of operability, and a licensing basis change. This finding was entered into the licensees corrective action program as CR-WF3-2014-05131, CR-WF3-2014-5341, and CR-WF3-2014-5412. Because this finding was of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000382/2014007-04, Failure to Evaluate Missile Protection Requirements for Emergency Diesel Generator Day and Storage Tank Vents.

.3 Failure to Identify and Correct Through-Wall Corrosion on Emergency Diesel Generator

A and B Day Tank Vents

Introduction.

The team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality. Specifically, the licensee failed to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents.

Description.

On October 8, 2014, the team identified during a plant walkdown that the emergency diesel generator A and B storage tank vents were outside the nuclear island plant structure and were not missile protected. On October 22, 2014, the team questioned the missile protection of the emergency diesel generator A and B day tank vents and determined that they were also not protected. The team also observed that the day tank vent pipes were significantly corroded.

Prior to discovery by the team, the licensee had not identified or evaluated the vent pipe corrosion. The team determined that the licensee failed to follow Procedure EN-LI-102, Corrective Action Program, Revision 24, which requires for conditions adverse to quality that a condition report be initiated promptly/timely, and that operability, functionality, and immediate reportability be reviewed for the condition. Attachment 9.2, Section 4, Design and Licensing Basis Issues, specifically provides examples of adverse conditions as they concern design basis issues; corrosion is a specific example cited.

The licensee documented the corrosion in CR-WF3-2014-05413 and determined that significant corrosion had occurred on both emergency diesel generator A and B day tank vent pipes. Both day tank vent pipes are located on the same roof approximately 9 feet apart. The corrosion was significant enough that a through-wall hole had formed at the base of each pipe where it penetrates the roof. Consequently, any water that collects on the roof of the building would have the potential to drain into the day tanks. The team also identified that a piece of foreign material was located approximately 2 feet from the roof drain, the roof coating was degraded, the roof drain scupper appeared to be inadequate for the size of the roof, and the roof slope appeared inconsistent with specifications in the construction drawings. The licensee documented these concerns in CR-WF3-2014-05529.

The licensees immediate operability determination concluded that the emergency diesel generators were operable since there was no severe weather in the forecast for the immediate future. Other corrective actions that the licensee performed included removing the foreign material from the roof, installing a rubber wrap around the vent pipes to cover the open holes, and creating small concrete berms immediately around the vent pipes to direct water away from the vent pipes. These corrective actions addressed the teams immediate safety concerns.

Additionally, to support the prompt operability determination, the licensee contracted with multiple engineering firms to determine the magnitude of precipitation event that would need to occur to allow water to enter the vent pipe holes. This in turn could be used to determine how much water enters the emergency diesel generator day tanks and whether the emergency diesel generators would become inoperable due to excessive water content in the fuel oil. Diesel engine vendor (Cooper Bessemer) documentation specifies that the fuel oil is limited to less than 0.1 percent water/sediment content.

The team used this specification as the limit for operability. At water/sediment content values above 0.1 percent the capacity of the emergency diesel generator would be de-rated by some amount, which would be variable based on multiple parameters. In these evaluations, the licensees contractors concluded that rainfall events less severe than those described in the updated safety analysis report would likely result in water contamination of both emergency diesel generator day tanks to approximately 5-10 percent water content, significantly greater than allowed by the vendor specification.

The team determined that had system engineering been performing walkdowns as required by EN-DC-178, System Walkdowns, Revision 7, the licensee would likely have identified the corrosion. The procedure specifically requires walking down all accessible areas of the system; it provides specific instructions for using permanently installed ladders, coordinating with other organizations, etc., to walk down accessible areas that are not normally accessed. In addition, the procedure specifically requires inspection for corrosion. The team determined that system engineering failed to follow the procedure or did not adequately implement the procedure.

Analysis.

The failure to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control and equipment performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences.

Specifically, the licensee failed to identify, evaluate, and correct through-wall corrosion on the emergency diesel generator A and B day tank vents. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the finding screened to Exhibit 4, External Events Screening Questions, because it screened as potentially risk-significant due to seismic, flooding, or severe weather. Per Exhibit 4, the finding screened to a detailed risk evaluation because if the safety functions of emergency diesel generators A and B were assumed completely lost, it would degrade two trains of a multi-train system and it would degrade one or more trains of a system that supports a risk-significant system.

A Region IV senior reactor analyst performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk-important sequences included heavy-rain-induced losses of off-site power with the consequential failure of both emergency diesel generators. The ability to restore off-site power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was important to avoid core damage. The finding was not significant to the large early release frequency. See Attachment 2, Detailed Risk Evaluation, for a detailed review of qualitative criteria also considered.

This finding had a cross-cutting aspect in the area of human performance associated with procedure adherence because the licensee failed to ensure that individuals follow process, procedures, and work instructions (H.8).

Enforcement.

The team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, 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 nonconformance are promptly identified and corrected. The licensees measures are established by Procedures EN-DC-178, System Walkdowns, which requires inspection for corrosion, and EN-LI-102 Corrective Action Program, which requires that a condition report be initiated promptly/timely for a condition adverse to quality, and that operability, functionality, and immediate reportability be reviewed. Attachment 9.2 of EN-LI-102, Section 4, Design and Licensing Basis Issues, specifically provides examples of adverse conditions as they concern design basis issues, corrosion is a specific example cited. Contrary to the above, prior to October 22, 2014, the licensee failed to identify and correct a condition adverse to quality. Specifically, the licensee failed to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents. In response to this issue, the licensee performed an immediate operability determination based on severe weather in the area, installed a temporary repair using a rubber wrap, and installed a small concrete berm to minimize the potential amount of water in the immediate area. This finding was entered into the licensees corrective action program as CR-WF3-2014-05413 and CR-WF3-2014-05529. Because this finding has been preliminarily determined to be of greater than very low safety significance (greater than Green), it will be treated as an apparent violation and tracked as: AV 05000382/2014007-05, Failure to Identify and Correct Through Wall Corrosion on Emergency Diesel Generator A and B Day Tank Vents.

.2.13 Emergency Feedwater Primary Isolation Valves EFW-228A and 229A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Emergency Feedwater Primary Isolation Valves EFW-228A and -229A. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Calculation EC-M97-054, Design Basis Review for Emergency Feedwater Valves EFW-223A, EFW-223B, EFW-224A, EFW-224B, EFW-228A, EFW-228B, EFW-229A, EFW-229B.

• Documentation regarding high-energy line break analyses and assumptions for the isolation valve areas.

• Actuator nitrogen accumulator sizing evaluations.

b. Findings

No findings were identified.

.2.14 Essential Chiller RFR-MCHL-0001A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Essential Chiller RFR-MCHL-0001A.

The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Design calculations for essential chilled water cooling loads and cooling coil performance.

• Recent condition reports concerning various issues with the essential chilled water system.

• Completed work orders for the preventative maintenance and surveillance testing.

• Operability checks.

• Modifications to the essential chillers and the chilled water system.

• Margin issues and recovery plans for the chilled water outlet temperature instrument accuracy.

b. Findings

No findings were identified.

.2.15 Main Feedwater Isolation Valve FW-184A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, seismic and design basis event loading, maintenance and test procedures, and condition reports associated with Main Feedwater Isolation Valve FW-184A. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Calculation EC-M98-003, Design Basis Review for Feedwater Isolation Valves FW-184A & B.

• Documentation regarding high-energy line break analyses and line break assumptions for the isolation valve areas.

• Weak link analyses for the main feedwater isolation valves.

• Valve closure time evaluations.

• Recent modifications to the component and supporting systems to verify the changes did not have an adverse impact on the ability of the component to perform its safety function.

• Waterfords response to NRC Information Notice 2002-26, Supplement 2, Additional Flow-Induced Vibration Failures after a Recent Power Uprate.

b. Findings

Introduction.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the thrust required to operate the main feedwater isolation valves assuming an appropriate valve-disk-to-seat coefficient of friction.

Description.

The team reviewed the licensees analysis of the main feedwater isolation valves contained in calculation EC-M98-003, Design Basis Review for Feedwater Isolation Valves FW-184A & B, Revision 2, dated January 18, 2006. In this calculation, the licensee used the evaluation methodology presented in EPRI topical report TR-103237, EPRI MOV Performance Prediction Program, Revision 2, to determine the design basis required thrust of valves FW-184A and FW-184B. The team questioned the basis of the licensees assumption in the calculation for valve disk-to-seat coefficient of friction. The coefficient of friction is an input into the calculation of the valves required thrust during design basis events. The calculated required thrust is then compared to the pneumatic/hydraulic actuator available thrust during the event to determine the valves design margin.

In response to the teams questioning, the licensee determined the assumed values in the calculation were given in EPRI TR-103237, Table 11-1, Friction Coefficients for Anchor/Darling Double Disk Gate Valve Model, which tabulates values of coefficient of friction depending on disk-to-seat contact stress and temperature. The licensee further determined that they used an incorrect value for the coefficient of friction for a main steam line break scenario in which the reactor is operating at low power, such as during reactor startup or shutdown, and the auxiliary feedwater system is supplying feedwater to the steam generators. The correct value for coefficient of friction during this scenario was higher than those assumed in the calculation due to elevated differential pressures and therefore contact stresses, as well as lower feedwater temperatures from the operation of the auxiliary feedwater system.

Following identification, the licensees corrective actions included re-performing the calculation of required thrust using the correct coefficient of friction value, and reevaluating the valves available margin during the affected scenario. The licensee determined that the valves continued to remain operable, since margin was still available, though it was reduced from approximately 21 percent to approximately 6.5 percent.

Analysis.

The failure to evaluate the main feedwater isolation valve required thrust assuming an appropriate valve disk-to-seat coefficient of friction was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the incorrect coefficient of friction assumption resulted in a reasonable question of operability of the main feedwater isolation valves to operate under design basis conditions during a main steam line break when auxiliary feedwater was supplying inventory to the steam generators. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, this finding screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, 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, since January 18, 2006, the licensee did not verify the adequacy of design of the main feedwater isolation valves. Specifically, the licensee failed to evaluate the required thrust in accordance with the licensees analysis methodology presented in EPRI TR-103237-R2, EPRI MOV Performance Prediction Program. In response to this issue, the licensee recalculated the required thrust and performed an evaluation of the remaining margin on the main feedwater isolation valves that supported an immediate operability determination. This finding was entered into the licensees corrective action program as CR-WF3-2014-05690. Because this finding was of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000382/2014007-06, Failure to Properly Evaluate Main Feedwater Isolation Valve Required Thrust.

.2.16 Main Steam Isolation Valve MS-124A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, seismic qualification reports, maintenance and test procedures, and condition reports associated with Main Steam Isolation Valve MS-124A. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Calculation EC-M98-004, Design Basis Review for Main Steam Isolation Valve MS-124A & B.

• Documentation regarding high-energy line break analyses and line break assumptions for the isolation valve areas.

• Weak link analyses for the main steam isolation valves to assess whether appropriate limits were applied to the allowable actuator setup parameters.

• Valve closure time and actuator nitrogen set-point evaluations.

• Recent modifications to the component and supporting systems to verify the changes did not have an adverse impact on the ability of the component to perform its safety function.

b. Findings

Introduction.

The team reviewed a self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to have an adequate seismic weak link evaluation for the main steam isolation valves.

Description.

During post-maintenance testing on January 5, 2013, following a preventive maintenance activity for actuator elastomer replacement, main steam isolation valve MS-124A failed. The threads of the valve actuators pneumatic/hydraulic piston sheared at the valve stem connection due to high differential pressure across the actuator piston. The licensee documented this failure in CR-WF3-2013-00107. The high differential pressure was caused by low initial nitrogen pressure when the valve was initially taken open. In response to the valves failure, the licensees corrective actions included performing a seismic weak link analysis in calculation A13068-C-001, Weak Link Analysis of 40 x 30 x 30 CL 600 Main Steam Isolation Valve.

The team reviewed the licensees analysis of the main steam isolation valves contained in calculation EC-M98-004, Design Basis Review for Main Steam Isolation Valve MS-124A & B, Revision 0, dated November 14, 2000, and noted that the seismic thrust limit in the Allowable Thrust Determination section of the calculation was listed as not available. Appendix A to 10 CFR Part 50 contains the general design criteria for nuclear power plants. Specifically, General Design Criterion 2, Design bases for protection against natural phenomena, states in part, Structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, and, The design bases for these structures, systems and components shall reflect...

(2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena.

Upon review, the team determined that prior to the failure on January 5, 2013, the limits for setting up the main steam isolation valve pneumatic/hydraulic actuator parameters, including allowable nitrogen pressure, to prevent structural damage were based on an evaluation that was performed on only the valve stem. The evaluation did not evaluate all of the valve and actuator sub-components, nor did it include consideration of seismic-event-induced stresses. After the seismic weak link analysis was completed, appropriate maximum allowable thrust values for the main steam isolation valves were determined and the corresponding allowable actuator nitrogen pressure settings were updated.

Analysis.

The failure to evaluate the main steam isolation valve maximum allowable thrust, assuming appropriate values for the structural limitations of the valve and actuator, was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee used a non-conservative value for the maximum allowed thrust, and the error resulted in the failure of main steam isolation valve MS-124A on January 5, 2013, because the allowable nitrogen pressure for the valve actuator was inappropriate. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, this finding screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, 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, prior to failure of main steam isolation valve MS-124A on January 5, 2013, the licensee did not verify or check the adequacy of its design.

Specifically, the licensee failed to have an adequate seismic weak link evaluation. In response to this event, the licensee performed a seismic weak link evaluation of the main steam isolation valves that supported a determination that the valve was operable.

This finding was entered into the licensees corrective action program as CR-WF3-2014-05708. Because this finding was of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000382/2014007-07, Failure to Properly Evaluate Main Steam Isolation Valve Weak Link.

.2.17 Shield Building Concrete Structure

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the shield building concrete structure.

The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Design calculations for the shield building.

• Nuclear Island and building design associated with the shield building.

• Maintenance rule structural monitoring, scoping, and basis.

• Procedures for preventative maintenance and inspection.

• Previous two shield building integrity inspection reports.

• Condition reports concerning the shield building concrete cracking following steam generator replacement.

• Waterfords response to NRC Information Notice 2010-20, Concrete Degradation by Alkali-Silica Reaction.

b. Findings

No findings were identified.

.2.18 Steam Generator Power Operated Atmospheric Dump Valve MS-116A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with Steam Generator Power Operated Atmospheric Dump Valve MS-116A. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Purchase specifications, purchase orders, and vendor supplied documentation verifying the valve pedigree and requirements.

• Main steam system piping isometric, detailed valve internal and pneumatic operator drawings.

• Start-up integrated test SIT-TP-707 verifying the minimum and maximum capacity of the atmospheric dump valves.

• Air-operated valve testing, maintenance, and trending program.

• Design change package for resolving valve binding during stroke testing.

b. Findings

No findings were identified.

.3.1 Inspection of NRC Information Notice 1996-45, Potential Common-Mode Post-Accident

Failure of Containment Coolers

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 1996-45, Common-Mode Post-Accident Failure of Containment Coolers, to verify the licensee performed an applicability review and took corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses a potential water-hammer issue on the cooling coils of the containment fan coolers due to steam voiding of the component cooling water inside the coils. The concerns results from post-accident conditions where high temperature containment atmosphere would be forced across the cooling coils for several seconds with no component cooling water causing the water to boil and create steam voids. The team verified that the licensees review adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.2 Inspection of NRC Information Notice 2002-26, Supplement 2, Additional Flow-Induced

Vibration Failures After a Recent Power Uprate

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2002-26, Supplement 2, Additional Flow-Induced Vibration Failures After a Recent Power Uprate, to verify the licensee performed an applicability review and took corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses concerns that could result from vibration due to changing parameters following power uprate. The team verified that the licensees review adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.3 Inspection of NRC Information Notice 2004-09, Corrosion of Steel Containment and

Containment Liner

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2004-09, Corrosion of Steel Containment and Containment Liner, to verify the licensee performed an applicability review and took corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses concerns that corrosion in the vicinity of the moisture barrier at the floor-to-containment junction could result in corrosion and thinning of the steel containment structure. The team verified that the licensees review adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.4 Inspection of NRC Information Notice 2010-26, Submerged Electrical Cables

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2010-26, Submerged Electrical Cables, to verify the licensee performed an applicability review and took corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses electrical cable degradation and potential failures resulting from cables being submerged for extended periods of time. The team verified that the licensees cable monitoring program adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.5 Inspection of NRC Information Notice 2011-20, Concrete Degradation by Alkali-Silica

Reaction

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2011-20, Concrete Degradation by Alkali-Silica Reaction, to verify the licensee performed an applicability review and took corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses potential degradation of mechanical properties of concrete due to Alkali-Silica reaction. The team verified that the licensees review, as documented in CR-WF3-2012-00569, adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.6 Inspection of NRC Information Notice 2013-04, Shield Building Concrete Subsurface

Laminar Cracking Caused by Moisture Intrusion and Freezing

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2013-04, Shield Building Concrete Subsurface Laminar Cracking Caused by Moisture Intrusion and Freezing, to verify that the licensee performed an applicability review and took appropriate corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses the potential for subsurface cracking in concrete shield building structures during some winter weather conditions.

The team verified that the licensees review, as documented in CR-WF3-2012-06038 and CR-WF3-2012-07645; and calculation EC 41691, adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.4 Results of Reviews for Operator Actions

a. Inspection Scope

The team selected risk-significant components and operator actions for review using information contained in the licensees probabilistic risk assessment. This included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1 x 10-6.

For the review of operator actions, the team observed operators during a simulator scenario associated with the selected components as well as observing simulated actions in the plant.

The selected operator actions were:

• Scenario: The crew was provided turnover with the plant at 100 percent rated thermal power and emergency diesel generator B out of service. One temporary emergency diesel generator was onsite and available to supply loads through Bus 22B.

After the crew took the shift, a reactor coolant pump was tripped which resulted in a reactor trip. Within seconds of the reactor trip, a loss of off-site power occurred. Emergency Diesel Generator A started but did not load onto Bus 3A due to two failures: output voltage was out of range low requiring manual action to restore, and Tie Breaker 3-2 failed to open automatically, requiring operator action to open.

Following completion of both operator actions, Emergency Diesel Generator A output breaker automatically closed. A non-licensed auxiliary operator was expected to be dispatched to acknowledge the emergency diesel generator A alarms resulting from degraded jacket water flow causing a Jacket Water Temperature off-normal alarm on the local annunciator panel and the A EDG Trouble alarm in the control room. After several minutes, the High Jacket Water Temperature alarmed on the local annunciator panel and the crew was expected to stop the running emergency diesel generator, placing the unit in a station blackout.

The crew was expected to enter Procedure OP-902-005, Station Blackout Recovery, and take appropriate actions. The crew was then expected to enter Procedure OP-TEM-008, Emergency Diesel Generator A(B) Backup Temporary Diesel Generators (TED), and align the temporary emergency diesel generator to supply Bus 3B from Bus 2B.

After the crew restored power to Bus 3B from the temporary emergency diesel generator, off-site power was restored and the crew was expected to restore power to Bus 3A from off-site using Appendix 12 of Procedure OP-902-009, Standard Appendices.

• In-plant job performance measures included:

o Reduce unnecessary loads during station blackout

o Transfer emergency feedwater suction to auxiliary component cooling wet towers

o Local manual operation of emergency feedwater flow control valve

o Local operation of steam generator atmospheric dump valve

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

4OA2 Problem Identification and Resolution

The team reviewed action requests associated with the selected components, operator actions and operating experience notifications. Any related findings are documented in prior sections of this report.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On November 6, 2014, the team leader presented the inspection results to Mr. M. Chisum, Site Vice President, and other members of the licensee staff. On December 17, 2014, following additional in-office inspection, the team leader presented the updated inspection results to Mr. M. Chisum, and other members of the licensee staff. On January 12, 2015, the team leader presented the final inspection results to Mr. M. Chisum and other members of the licensee staff.

The licensee acknowledged the findings during each meeting. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed; no proprietary information has been included in this report.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Barreto, Design Engineering, Engineer

B. Briner, Systems Engineering, Engineer

M. Chisum, Site Vice President

W. Crowley, Operations, Procedures

K. Dolese, Systems Engineering, Engineer

D. Gallodoro, Design Engineering, Engineer

R. Gilmore, Acting Engineering Director

A. Griffin, Systems Engineering, Engineer

M. Haydel, Design Engineering Manager

J. Hoss, Design Engineering, Engineer

J. Jarrell, Regulatory Assurance Manager

B. Lindsey, Operations Manager

D. Litolff, Operations, Control Room Supervisor

C. Lunk, Systems Engineering, Engineer

L. Milster, Regulatory Assurance, Licensing Engineer

N. Petit, Design Engineering Mechanical/Civil, Supervisor

S. Picard, Design Engineering, Engineer

C. Pickering, Design Engineering Programs, Supervisor

C. Pratt, Operations, Auxiliary Operator

J. Russo, Design Engineering Electrical/I&C, Supervisor

J. Signorelli, Operations Training, Superintendent

L. Smith, Systems Engineering, Engineer

M. Thigpen, Design Engineering, Engineer

R. Tran, Design Engineering, Engineer

E. Wilbur, Systems Engineering, Engineer

J. Williams, Regulatory Assurance, Specialist

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000382/2014007-05 AV Failure to Identify and Correct Through Wall Corrosion on Emergency Diesel Generator A and B Day Tank Vents (Section 1R21.2.12.3)

Opened and Closed

05000382/2014007-01 NCV Failure to Identify and Evaluate Elevated Bus Voltages (Section 1R21.2.2)

05000382/2014007-02 FIN Inadequate Station Procedures for Temporary Emergency Diesel Generator (Section 1R21.2.7)

05000382/2014007-03 NCV Failure to Initiate a Condition Report for a Condition Adverse to Quality (Section 1R21.2.12.1)

Opened and Closed

05000382/2014007-04 NCV Failure to Evaluate Missile Protection Requirements for Emergency Diesel Generator Day and Storage Tank Vents (Section 1R21.2.12.2)

05000382/2014007-06 NCV Failure to Properly Evaluate Main Feedwater Isolation Valve Required Thrust (Section 1R21.2.15)

05000382/2014007-07 NCV Failure to Properly Evaluate Main Steam Isolation Valve Weak Link (Section 1R21.2.16)