IR 05000263/2015007

RE-Issued - Monticello Nuclear Generating Plant - NRC Component Design Bases Inspection (Inspection Report 05000263/2015007)

ML20086J173
Person / Time
Site:	Monticello
Issue date:	03/26/2020
From:	Karla Stoedter NRC/RGN-III/DRS/EB2
To:	Conboy T Northern States Power Company, Minnesota
References
IR 2015007
Preceding documents:	ML16095A358 (4 April 2016)
Download: ML20086J173 (33)

Text

March 26, 2020

SUBJECT:

RE-ISSUED - MONTICELLO NUCLEAR GENERATING PLANT - NRC COMPONENT DESIGN BASES INSPECTION (INSPECTION REPORT 05000263/2015007)

Dear Mr. Conboy:

In a letter dated March 23, 2020, (ML20083G903), the U.S. Nuclear Regulatory Commission (NRC) informed you that Non-Cited Violation (NCV)05000263/2015007-02 was being revised based upon the results of an independent review. The purpose of this letter is to re-issue NRC Inspection Report 05000263/2015007 in its entirety including the revised NCV.

On July 24, 2015, the NRC completed a Component Design Bases Inspection at your Monticello Nuclear Generating Plant. The enclosed report documents the inspection findings, which were discussed on July 24, 2015, with you and other members of your staff.

Based on the results of this inspection, two NRC-identified findings of very-low safety significance were identified. The findings involved violations of NRC requirements.

However, because of their very-low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as NCVs in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of these NCVs, 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 III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Monticello Nuclear Generating Plant.

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

component of the NRC's 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/

Karla K. Stoedter, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-263 License No. DPR-22 Enclosure:

Inspection Report 05000263/2015007 w/Attachment: Supplemental Information cc w/encl: Distribution via LISTSERV

SUMMARY

Inspection Report 05000263/2015007; 06/22/2015 - 07/24/2015; Monticello Nuclear Generating

Plant; Component Design Bases Inspection.

The inspection was a 3-week onsite baseline inspection that focused on the design of components. The inspection was conducted by regional engineering inspectors and two consultants. Two Green findings were identified by the inspectors. The findings were considered Non-Cited Violations (NCVs) of U.S. Nuclear Regulatory Commission (NRC)regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red), and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding having very-low safety significance, and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50,

Appendix B, Criterion III, Design Control, for the failure to assure the nitrogen supply for the alternate nitrogen (AN2) system was controlled as safety-related in system specifications, drawings, procedures, and instructions. Specifically, the licensee did not confirm effective quality assurance controls were in place to ensure the bottled nitrogen was acceptable to support the safety-related functions of this system. The licensee entered this finding into the Corrective Action Program (CAP), and subsequently contacted the commercial nitrogen gas supplier to confirm that the vendors quality controls provided a sufficient basis to conclude that the AN2 system was operable.

The finding was determined to be more than minor because if left uncorrected, the issue had the potential to lead to a more significant safety concern. Specifically, if the commercial (e.g., non-safety) gas supply vendor quality controls were not adequate to ensure contaminants such as moisture or particulates were excluded from the nitrogen gas bottles, it could potentially disable the AN2 systems capability to support manual operation of safety relief valves during post loss-of-coolant-accident mitigation. The inspectors did not identify a cross-cutting aspect associated with this finding as it did not reflect current performance. (Section 1R21.3.b.(1))

Green.

The inspectors identified a finding of very-low safety significance (Green)and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the failure to establish, implement and maintain a procedure covering preventive maintenance schedules for the inspection or replacement of non-critical safety-related relays and other components with a specified lifetime. Specifically, the licensee did not establish, implement and maintain procedures supporting a preventive maintenance schedule for various safety-related relays and motor starter contactors classified as non-critical, which had a specific lifetime. Immediate corrective actions included instituting a Relay Monitoring Program, performing generic service life evaluations on some of the safety-related Agastat and General Electric relays, and identifying and replacing relays that had exceeded vendor recommended service life. A separate action item was initiated to evaluate motor starter contactors.

The performance deficiency was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding screened as having very-low safety significance based on answering No to all the screening questions in the Mitigating Structures, Systems, and Components and Functionality Section of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power,

Exhibit 1, Mitigating Systems Screening Questions. The inspectors did not identify a cross-cutting aspect associated with this finding as it was not reflective of the licensees current performance. (Section 1R21.3.b.(2))

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection

.1 Introduction

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

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

.2 Inspection Sample Selection Process

The inspectors used information contained in the licensees PRA and the Monticello Standardized Plant Analysis Risk Model to identify internal flooding scenarios to use as the basis for component selection. Based on these scenarios, a number of risk-significant components, including those with large early release frequency (LERF)implications, were selected for the inspection.

The inspectors also used additional component information such as a margin assessment in the selection process. This design margin assessment considered original design reductions caused by design modification, power uprates, or reductions due to degraded material condition. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as performance test results, significant corrective actions, repeated maintenance activities, Maintenance Rule (a)(1) status, components requiring an operability evaluation, system health reports, and U.S. Nuclear Regulatory Commission (NRC) resident inspector input of problem areas/equipment. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. A summary of the reviews performed, and the specific inspection findings identified are included in the following sections of the report.

The inspectors also identified procedures and modifications for review associated with the selected components. In addition, the inspectors selected operating experience issues associated with the selected components.

The inspection reviewed 19 samples (5 operating experience, 13 components, and 1 component with LERF implications) as defined in Inspection Procedure 71111.21 05.

.3 Component Design

a. Inspection Scope

The inspectors reviewed the Updated Safety Analysis Report (USAR), Technical Specifications (TS), design basis documents, drawings, calculations and other available design basis information, to determine the performance requirements of the selected components. The inspectors used applicable industry standards, such as the American Society of Mechanical Engineers Code, Institute of Electrical and Electronics Engineers (IEEE) Standards, and the National Electric Code, to evaluate acceptability of the systems design. The NRC also evaluated licensee actions, if any, taken in response to NRC issued operating experience, such as Bulletins, Generic Letters, Regulatory Issue Summaries (RISs), and Information Notices (INs). The review was to verify that the selected components would function as designed when required and support proper operation of the associated systems. The attributes that were needed for a component to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes to verify that the component condition and tested capability was consistent with the design bases and was appropriate may include installed configuration, system operation, detailed design, system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation.

For each of the components selected, the inspectors reviewed the maintenance history, preventive maintenance activities, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee corrective action program documents. Field walkdowns were conducted for all accessible components to assess material condition, including age-related degradation and to verify that the as-built condition was consistent with the design. Other attributes reviewed are included as part of the scope for each individual component.

The following 14 components (samples) were reviewed:

• Non-Safeguards Diesel Generator (DG-13): The inspectors reviewed the fuel capacity of the day tank, the procedures, and equipment required for refueling the day tank to determine if the DG-13 would be able to meet its required mission time. In addition, the inspectors reviewed monthly operability testing to determine whether the DG-13 would perform as required. Maintenance records and trends were also reviewed to verify reliability. The inspectors reviewed the DG-13 ability to supply power for the safety-related inverter to Battery #13 in the event of an extended station blackout (SBO) scenario. Generator loading was reviewed for this scenario to ensure DG-13 was capable to supply the anticipated load per the operating procedures. A walk through of this scenario with licensee staff was conducted to ensure the operating procedure was adequate to perform the intended operations.

• Reactor Core Isolation Cooling Pump (P-207): The inspectors reviewed the system hydraulic calculations such as, net positive suction head (NPSH) and minimum required flow to ensure the pumps were capable of providing their function. The inspectors also reviewed the vendor manual for the pump to determine whether the pumps characteristics met the design basis requirements and these requirements were accurately incorporated in reactor core isolation cooling (RCIC) system Inservice Testing (IST) procedures. The IST results were reviewed to assess potential component degradation and impact on design margins. The operation of the pump from various suction sources was reviewed to evaluate the pumps ability to provide the required flow from each source. The inspectors reviewed the RCIC operation during SBO compared to how various RCIC subcomponents were modeled in the battery sizing calculation to verify RCIC subcomponent loading was conservative.

• Reactor Core Isolation Cooling Minimum Flow Valve (CV-2104): The inspectors reviewed the air-operated valve (AOV) calculations, including required thrust, weak link, and maximum differential pressure, to ensure the valve was capable of functioning under design and licensing bases conditions. Diagnostic and IST results, including the leak rate test of the air system up to the check valve were reviewed to verify acceptance criteria were met and performance degradation would be identified. The inspectors reviewed the capacity calculation for the safety-related air accumulator to ensure sufficient air was available for the AOV to function as required upon loss of normal air. In addition, the accumulator check valve testing was reviewed to ensure the air system capacity would remain within its design limits. The inspectors reviewed the voltage and power supply requirements and verified the minimum required voltage would be available to the valve under all postulated conditions. The inspectors also verified the operation of the valve was appropriately modelled in battery sizing calculation.

• Reactor Core Isolation Cooling Steam Supply Inboard Containment Isolation Valve (MO-2075): The inspectors reviewed the motor-operated valve (MOV)calculations, including required thrust, weak link, degraded voltage, and maximum differential pressure, to ensure the valve was capable of functioning under design and licensing bases conditions. Diagnostic, IST, and local leak rate test results were reviewed to verify acceptance criteria were met and performance degradation would be identified. The inspectors reviewed the voltage and power supply requirements and verified the minimum required voltage will be available to the valve under degraded voltage conditions.

• Residual Heat Removal Pump 13 (P-202C): The inspectors reviewed the system flow and NPSH calculations to verify the pump was capable of performing its safety-related functions. The IST results were reviewed to assess potential component degradation and impact on design margins. The IST procedures were examined to determine whether the acceptance criteria adequately evaluated pump performance. Pump operation in various modes was reviewed to evaluate the pumps ability to provide the required flow in each mode. The inspectors reviewed the periodic testing to ensure the pump interlocks would function as required. The motors fuse/breaker coordination study was examined to verify adequate coordination. The inspectors reviewed the environmental qualification (EQ) evaluation and vendor manuals to verify manufacturers requirements for cooling the motor upper bearing during a postulated event were addressed. The motor overhaul/replacement schedule and the specification for overhauling motors was reviewed to ensure the motors safety-related qualification was maintained. The inspectors compared the motor nameplate with information in the emergency diesel generator (EDG) loading calculation to ensure the correct values were incorporated into the calculation.

• Residual Heat Removal Service Water Pump 13 (P-109C): The inspectors reviewed system flow and NPSH calculations to determine whether the pump would operate at the minimum water level in the intake structure. Further, calculations and the adequacy of the differential pressure setpoint across the residual heat removal (RHR) heat exchanger were reviewed to ensure the service water side was at a higher pressure than the RHR side. The inspectors reviewed the maintenance documents for the most recent pump overhaul and the re-baselining of the pump performance curves to determine whether the rebuilt pump met design basis requirements. In addition, the inspectors reviewed completed pump surveillances for the rebuilt pump to ensure that actual performance was acceptable. The inspectors reviewed the EQ evaluation and vendor manuals to verify manufacturers requirements for cooling the motor upper bearing during a postulated event were addressed. The motors fuse/breaker coordination study was reviewed to verify adequate coordination.

The inspectors compared the motor nameplate with information in the EDG loading calculation to ensure the correct values were incorporated into the calculation. The motor overhaul/replacement schedule and the specification for overhauling motors was examined to ensure the motors safety-related qualification was maintained.

• Drywell-to-Torus Vacuum Breaker (AO-2382A): The inspectors reviewed the calculations to demonstrate the valve would function as designed following a loss-of-coolant accident (LOCA). Specifically, the inspectors reviewed calculations establishing the valve capacity (e.g., sizing) and the maximum stress on valve internal components. Additionally, the inspectors reviewed calculations establishing the acceptance criteria used in TS related surveillance tests including; the maximum allowable torque required to fully open the valve, and the differential pressure decay curve for establishing allowable seat leakage.

The inspectors also reviewed completed surveillance and maintenance records to verify acceptance criteria were met and performance degradation would be identified. The inspectors reviewed the solenoid valve voltage and power supply requirements and verified that minimum required voltage would be available under the worst-case loading conditions. The inspectors also reviewed the micro switch replacement history and the reasons for replacement.

• Safety Relief Valve (RV-2-71E): The inspectors reviewed maintenance and test procedures to determine if the procedures were adequate to ensure that the safety relief valve (SRV) would reliably function to relieve an over-pressure condition. Additionally, the inspectors reviewed the calculation demonstrating the valve had a sufficient supply of nitrogen from the safety-related alternate nitrogen (AN2) system to allow manual actuation and operation to support post-accident mitigation functions. The inspectors also reviewed completed surveillance and maintenance records to verify acceptance criteria were met and performance degradation would be identified. The inspectors reviewed the actuation of the low-low set SRV to ensure response times were within allowable values. A review of the control circuit, calculations for the setpoints, and solenoid response times was performed to ensure coordination of the low-low set SRV with the balance of mechanically operated SRVs.

• Emergency Diesel Fuel Oil System: The inspectors reviewed the modification that restored the fuel oil system to within the plants licensing basis. Specifically, the inspectors reviewed the following system components:

• Diesel Fuel Oil Transfer Pumps (P-160A-D): The inspectors reviewed the calculation to confirm these pumps developed sufficient flowrates to support the system accident mitigation function. Specifically, the inspectors reviewed the hydraulic calculation that evaluated eight operating configurations to ensure the minimum required NPSH was maintained for the limiting pump, and the pump flow capacity was sufficient to maintain the associated EDG day tank level and/or support transfer of fuel to other storage tanks.

Additionally, the inspectors reviewed the completed pre-operational pump acceptance tests and performed a visual inspection of the pumps to assess configuration and potential vulnerabilities to hazards. The inspectors reviewed the design of the EDG fuel oil system to determine whether all applicable standards and the requirements for train separation were met.

The inspectors reviewed the control and motor protection scheme for the newly installed transfer pumps and the associated calculations. Also reviewed were the cable sizing, voltage drop to motor terminals and motor control center starter coil pick-up voltages, and additional loading on the EDG by the additional transfer pump motors. The method for fire separation of Division II piping and cabling routed through the Division I EDG room was reviewed to ensure a fire in one room would not affect both EDGs.

• Diesel Fuel Oil Transfer Pump Relief Valves (RV-1523, RV-1524, RV-1525, RV-1526) and Attached Piping: The inspectors reviewed the safety relief valve design data sheet and vendor catalog information used to establish the valve lift setpoint and capacity to ensure that the relief valves provided adequate overpressure protection for the system to meet the pipe design Code (1977 Edition, Winter 1978 Addenda, ANSI B31.1 Power Piping). The inspectors reviewed the completed pre-operational acceptance testing for the relief valves and performed a visual inspection of these valves to assess configuration and potential vulnerabilities to hazards. Additionally, the inspectors reviewed the certified material test reports and certification of conformance records for the relief valves and select pipe components replaced during the relief valve installation to confirm the valve and pipe component materials met the design/fabrication Code and pipe specifications.

• 250vdc Bus (D311): The inspectors reviewed the fault current calculation and vendor documents regarding breakers contained within bus D311. The inspectors reviewed the feeder breaker calculation for sizing and protection scheme. The inspectors reviewed the environmental conditions in the RCIC room (location of D311) during a high energy line break (HELB). The inspectors reviewed the D311 cabinet and reviewed cabinet/equipment specifications for temperature and humidity to ensure equipment would function as required under worst case environmental conditions. The inspectors also considered the qualification testing and calculations regarding the HELB boundary door between the RCIC room and the torus area to verify the door would maintain an adequate boundary during a HELB event.

• 250vdc Battery (#13): The inspectors reviewed the battery sizing calculation to verify the battery has adequate capacity to cope with the most limiting accident and transient conditions, the load profile modeled was conservative compared to actual worst-case loading scenario in the plant. The inspectors also reviewed the voltage drop calculation to verify the voltages available at all components, under worst case loading conditions, were above their minimum voltage requirements.

• 250vdc Battery Charger (D-52): The inspectors reviewed the battery charger sizing calculation to verify the battery charger has sufficient capacity to supply the normal loads and fully charge the battery from a fully discharged state within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The inspectors also reviewed the scheme to supply the charger from the non-safety-related DG-13 during an extended SBO.

• 250vdc Battery Room Ventilation Fan (V-EF-40B): The inspectors reviewed calculations concerning the battery room airflow required for limiting hydrogen accumulation and the flow necessary to supply outside air across the control room emergency filtration train (EFT) system inlet radiation monitor to determine whether the current airflow met design basis requirements. The modification to the EFT system that blanked off a portion of the EFT inlet duct work was reviewed to determine whether it would interfere with the fans safety-related function. The inspectors reviewed periodic system testing and test results to verify acceptance criteria were met and performance degradation would be identified. For out of specification flow readings, the inspectors verified causes were identified and adequate corrective actions were taken. Normal and abnormal operating procedures were reviewed to ensure they were updated after the modifications. The inspectors reviewed electrical schematics to ensure adequate power was available to the fan motor and control room alarms.

• 4160vac Essential Bus 15 (A5): The inspectors reviewed the sizing and coordination of the feeder and load breakers. The degraded voltage calculation was reviewed to verify adequate voltage will be available to safety-related components during a design basis event concurrent with a degraded voltage condition. The inspectors also reviewed documents to verify that the feeder cable to the bus was adequately sized. The 125 vdc voltage drop calculation was reviewed to verify the feeder and load breaker control components will have sufficient voltage available during the worst-case loading conditions. The bus breaker/relay testing procedures were also reviewed.

b. Findings

(1) Inadequate Quality Assurance Controls for Nitrogen Supply for the Alternate Nitrogen System

Introduction:

The inspectors identified a finding of very-low safety significance (Green),and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure the nitrogen supply for the AN2 system was controlled as safety-related in system specifications, drawings, procedures, and instructions. Specifically, the licensee had not confirmed effective quality assurance controls were in place to ensure the bottled nitrogen was acceptable to support the safety-related functions of this system.

Description:

On July 23, 2015, the inspectors identified the licensee failed to control the nitrogen supply for the AN2 system as safety-related in system specifications, drawings, procedures, and instructions. In particular, the inspectors were concerned that the failure to implement adequate quality controls could result in failure of the AN2 system to function in support of accident mitigation.

The USAR Section 4.4.2.1, Safety/Relief Valves, stated, the automatic depressurization system safety/relief valves are designed to withstand a hostile environment and still perform their function for 100 days following an accident. In support of this function, a safety-related backup pneumatic supply was provided by the AN2 system, which automatically supplies pressure to 6 of the 8 SRV actuators upon loss of the non-safety related instrument nitrogen system. The USAR Section 4.4.4, stated, The bottled nitrogen supply racks used for the AN2 system are manually checked for adequate supply and pressure during plant operation at a frequency to assure minimum design capacity requirements of the system will be met, when required, assuming worst case leakage rates. To ensure an adequate supply of nitrogen to the safety-related AN2 system, the licensee determined in Calculation 94-017, Calculation of Alternate Nitrogen Operability Leakage Criteria, that in addition to the 8 installed nitrogen bottles, 59 spare nitrogen bottles charged to a minimum of 2283 psig were required. This quantity of nitrogen represented a 7-day supply, which provided time for the licensee to procure additional nitrogen from an offsite supplier.

The inspectors observed that the licensee had stored 8 spare bottles of nitrogen in the turbine building, and in excess of 51 spare bottles within the onsite shipping/receiving warehouse. These spare nitrogen bottles did not have installed pressure gauges, so the inspectors could not confirm the pressure (e.g., quantity) of nitrogen stored in the spare bottles. On August 14, 2014, during installation of spare nitrogen bottles to the AN2 system, the licensee identified two empty nitrogen bottles that prompted an apparent cause investigation documented in Action Request (AR) 01443013. As a result, the licensee determined the cause of the empty bottles was the spare nitrogen bottles were not verified fully charged prior to installation. To correct this issue, the licensee checked each bottle (with a temporary pressure gage) on a weekly basis to confirm that the spare bottles stored in the turbine building were fully charged. However, the licensee had never checked the pressure of the spare bottles in the receiving warehouse and had not determined if the empty bottles identified in 2014 were the result of an error in the gas vendors quality controls or an error in the licensees onsite inventory control process.

The inspectors observed that the nitrogen bottles stored in the receiving warehouse were not labeled as full or empty and most did not have material stock tags. Because these bottles were not procured as safety-related, the licensee did not have an inventory control procedure that required labeling nitrogen bottles as full or empty, or that prohibited storing empty nitrogen bottles with full bottles of nitrogen, or that required use of material control stock tags. The inspectors questions on inventory control prompted the licensee to measure the pressure of the spare nitrogen bottles stored in the receiving warehouse. As a result of this activity, the licensee identified one bottle with an unexpectedly low-pressure of 1800 psig. The licensee quarantined this bottle for subsequent investigation to determine the cause of the unexpected low-pressure.

In addition to the quantity of nitrogen for the AN2 system, the inspectors were concerned with the quality of the nitrogen because the licensee procured this nitrogen from a commercial gas supply vendor without performing tests to confirm the type or quality of the gas received. The inspectors were concerned that if the commercial vendor quality controls were not sufficient, the nitrogen supply may contain high moisture content, particulates, or be mixed with other gas types. In particular, if moisture levels were excessive, the water vapor would freeze during expansion of the gas at the AN2 system pressure reducers and create ice particles that could block AN2 system components (e.g., pipes or valves), and result in SRVs which could not be manually actuated.

Similarly, a high particulate concentration could block small passages in AN2 system components (e.g., pressure regulators) and restrict the flow of nitrogen resulting in SRVs, which could not be manually actuated. If the SRVs could not be operated manually, it would impair/prevent accident mitigation functions such as reactor pressure control, reactor depressurization, and alternate shutdown cooling. The inspectors concerns prompted the licensee to contact the gas supply vendor to determine what vendor controls were used to confirm the quantity and quality of the nitrogen delivered.

The commercial vendors controls included evacuation of reused bottles and sampling of the gas in one bottle from each batch (groups of 24) to confirm gas purity and lack of contaminants (e.g., moisture content). Additionally, the gas supply vendor reportedly used a closed process to fill the nitrogen bottle that did not introduce particles. The licensee concluded that the gas vendor quality controls provided a sufficient basis to conclude that the AN2 system was operable.

Title 10 CFR 50.2 states, that, safety-related structures, systems and components (SSCs) means those SSC that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary;

(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or

(3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1), or 10 CFR 100.11 of this chapter, as applicable. The licensee guidance to implement this definition existed in Attachment 2, Classification Guidance, of procedure FP-E-RTC-02, Equipment Classification, which stated, in part, Items that are either installed in safety-related systems and relied upon to provide or support the safety-related functions, or are installed in any system needed to satisfy safety-related interface requirements (e.g., isolation devices) are identified.

These items are classified as safety-related. Based upon this guidance, the nitrogen supplied by four bottles installed in each AN2 system train should have been identified as safety-related because the nitrogen was required to support the safety-related functions of the AN2 system. On drawing NH-36049-10, Alternate Nitrogen Supply System, the installed nitrogen bottles were located outside the safety-related portion of the AN2 system piping boundary and instead were identified as a special concerns item, which was defined as an item subject to augmented quality controls in FP-E-RTC-02. The licensee added the special concerns item designation for the nitrogen bottles in 1988, as a result of an NRC commitment associated with NUREG-0737, Clarification of Three Mile Island Action Plan Requirements.

However, the licensee had not procured the installed or spare nitrogen bottles under a safety-related Quality Control Program as described in 10 CFR Part 50, Appendix B.

Instead, the licensee had procured the nitrogen bottles from a commercial vendor without auditing the gas vendors quality controls and without conducting confirmatory tests to verify the type, quality or quantity of gas delivered.

The licensee initiated AR 01486991, and contacted the commercial nitrogen gas supplier to confirm that the vendors quality controls provided a sufficient basis to conclude the AN2 system was operable. In addition, the licensee identified an action to evaluate the controls in place to ensure that AN2 system nitrogen supply bottles had adequate pressure and adequate gas quality.

Analysis:

The inspectors determined the failure to demonstrate the nitrogen supply for the AN2 system was controlled as safety-related in system specifications, drawings, procedures and instructions was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and a performance deficiency. The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, because the inspectors answered Yes to the More-than-Minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, if the commercial (e.g., non-safety) gas supply vendor quality controls were not adequate to ensure contaminants such as moisture or particulates were excluded from the nitrogen gas bottles, it could potentially disable the AN2 system capability to support manual operation of SRVs during post LOCA mitigation.

The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, dated April 29, 2015, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, dated June 19, 2012, for the Mitigating Systems cornerstone. The inspectors evaluated the finding using Appendix A, The Significance Determination Process for Findings At-Power. The finding screened as very-low safety significance (Green) because the inspectors were able to answer Yes to screening Question A1 in Exhibit 2 because the finding represented a design deficiency confirmed not to result in loss of operability or functionality.

The inspectors did not identify a cross-cutting aspect associated with this finding as it did not reflect current performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2, and as specified in the license application, for those SSC to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the SSC.

Contrary to the above, as of July 23, 2015, the licensee had not established measures to assure that the design basis for the nitrogen supply to the AN2 system was correctly translated (e.g., classified/controlled as safety-related) into specifications, drawings, procedures, and instructions.

Because this violation was of very-low safety significance, and it was entered into the licensees Corrective Action Program (CAP) as AR 01486991, where the licensee contacted the supplier to confirm the vendors quality controls provided a sufficient basis to conclude the AN2 system was operable, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000263/2015007-01, Inadequate Quality Assurance Controls for Nitrogen Supply for the AN2 System).

(2) Failure to Establish a Procedure Covering Preventive Maintenance Schedules for the Inspection or Replacement of Non-Critical Safety-Related Components With a Specified Lifetime

Introduction:

The inspectors identified finding of very-low safety significance (Green)and an associated Non-Cited Violation (NCV) of TS 5.4.1, Procedures, for the failure to establish, implement and maintain a procedure covering preventive maintenance schedules for the inspection or replacement of non-critical, safety-related relays and motor starter contactors with a specified lifetime.

Description:

During the 2012 Problem Identification and Resolution inspection, Unresolved Item (URI)05000263/2012008-01 was opened related to the qualification basis for safety-related relays and motor starter contactors. The URI identified concerns with the licensee not replacing safety-related relays and motor starter contactors that were beyond the vendors recommended service life without an appropriate evaluation justifying a life extension. In addition, no procedure existed to ensure these components were expected or replaced per an approved preventive maintenance schedule. The inspectors, in consultation with Nuclear Reactor Regulation staff, issued Task Interface Agreement (TIA) 2014-01, Final Task Interface Agreement - Regulatory Position on Design Life of Safety-Related Structures, Systems, and Components Related to Unresolved Items at Donald C. Cook Nuclear Power Plant, Monticello Nuclear Generating Plant and Palisades Nuclear Plant. The TIA was issued on May 7, 2015, and concluded when a licensee becomes aware that a safety-related SSCs service life has been exceeded or information challenges the presumption that a safety-related SSC can perform its specified function, the licensee must promptly address and document this non-conforming condition in accordance with the licensees NRC approved Quality Assurance Program, the licensees operability/functionality program and the CAP. This includes completing appropriate corrective actions in a timely manner and documenting licensees evaluations justifying the service life extensions.

During this inspection, the inspectors noted the licensee previously initiated AR 01446684, which identified a number of corrective actions. Some actions were already completed and the remaining were scheduled for completion in a timely manner. Immediate corrective actions included instituting a Relay Monitoring Program, performing generic service life evaluations on some of the safety-related Agastat and GE relays, and identifying and replacing relays that had exceeded vendor recommended service life. The licensee continued to identify safety-related relays exceeding vendor recommended service life and had plans to conduct extent of condition reviews.

A separate action item was initiated to evaluate motor starter contactors.

Analysis:

The inspectors determined that the failure to establish a procedure covering preventive maintenance schedules for the inspection or replacement of non-critical, safety-related relays and motor starter contactors was contrary to TS 5.4.1, Procedures, and was a performance deficiency.

The inspectors determined that the performance deficiency was more-than-minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because the performance deficiency was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). These relays were installed in protective circuits such as reactor protection system, etc., and their failure could impact the proper operation of these protective schemes. The finding was screened in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Mitigating Systems Screening Questions, dated July 1, 2012. The finding screened as having very-low safety significance (i.e., Green) based on answering No to all the screening questions under the Mitigating SSCs and Functionality section of IMC 0609, Appendix A, Exhibit 1.

The inspectors did not identify a cross-cutting aspect associated with this finding as it did not reflect licensees current performance.

Enforcement:

Technical Specification 5.4.1, Procedures, requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance, requires, in part, that preventive maintenance schedules shall be developed for the inspection or replacement of parts that have a specific lifetime.

Contrary to the above, as of July 24, 2015, the licensee failed to establish, implement and maintain a procedure to ensure a preventive maintenance schedule was developed for the inspection or replacement of parts that have a specific lifetime. Specifically, the licensee had not established, implemented and maintained a procedure to ensure that preventive maintenance schedules were developed for specific non-critical, safety-related relays and motor starter contactors which had a specified lifetime.

Because this violation was of very-low safety significance, and it was entered into the CAP as AR 01446684, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000263/2015007-02, Failure to Establish a Procedure Covering Preventive Maintenance Schedules for the Inspection or Replacement of Parts with a Specified Lifetime.)

.4 Operating Experience

a. Inspection Scope

The inspectors reviewed five operating experience issues (samples) to ensure that NRC generic concerns had been adequately evaluated and addressed by the licensee.

The operating experience issues listed below were reviewed as part of this inspection:

• IN 2012-14, Motor-Operated Valve Inoperable Due to Stem-Disc Separation;

• IN 2013-05, Battery Expected Life and Its Potential Impact on Surveillance Requirements;

• RIS 2000-012, Resolution of Generic Safety Issue B-55, Improved Reliability of Target Rock Safety Relief Valves;

• GE Service Information Letter (SIL) 44, GE HFA Relay Coil Life; and

• GE SIL 196 - Original thru Supplement 17, Recommendations for Target Rock Main Steam Safety/Relief Valves.

b. Findings

No findings were identified.

.5 Modifications

a. Inspection Scope

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

• DC79M070, Modify Drywell to Torus Vacuum Breakers;

• EC23085, EDG Fuel Oil Train;

• EC23805, EDG Fuel Oil Train Separation; and

• EC25733, Alternate Nitrogen Bottle Change-out Check Valves.

b. Findings

No findings were identified.

.6 Operating Procedure Accident Scenarios

a. Inspection Scope

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

The following operator actions were reviewed:

• Actions to isolate flooding from plant administration building fire header;

• Actions to isolate Service Water line to 12 Main Feedwater cooler line break; and

• Actions to use the non-safety-related DG13 to provide power to the Division II 250vdc Battery Chargers in the event of an SBO.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Review of Items Entered into the Corrective Action Program

a. Inspection Scope

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

The inspectors also selected two issues identified during previous CDBIs to verify that the concern was adequately evaluated, and corrective actions were identified and implemented to resolve the concern, as necessary. The following issues were reviewed:

• NCV 05000263/2012007-03; Failure to Maintain the Degraded Voltage Function Time Delay Design: The inspectors reviewed the licensees design change that removed the 1AR transformers additional 5 second time delay and restored compliance to the TSs.

• NCV 05000263/2012007-04; Failure to Analyze Effect of Degraded Voltage on Proper Operation of Thermal Overload Relays: The inspectors reviewed three of four corrective actions completed associated with this issue. The completed issues included:

(1) EC19903 increased the margins for the subject thermal overload relay (TOL) settings;

(2) EC25687 analyzed TOL performance for MOVs during a degraded voltage with LOCA scenario; and

(3) EC25688 analyzed TOL performance for all continuous duty motors during a degraded voltage with LOCA scenario. The fourth issue to formalize the analysis was included in the Monticello Calculation Reconstitution Project with completion planned by July 2016. This was being tracked by AR 01197202 and OBN01479704-04.

b. Findings

No findings were identified.

4OA5 Other Activities

.1 (Closed) URI 05000263/2012008-01; Qualification Basis for Safety-Related Relays and

Motor Starter Contactors: This URI is closed to NCV 05000263/2015007-02, Failure to establish a procedure covering preventive maintenance schedules for the inspection or replacement of non-critical safety-related relays and other components, which had a specific lifetime. See Section 1R21.3.b.(2).

.2 (Closed) URI 05000263/2012008-02; Concern with Periodic Design Basis Testing of

Installed Relays and Motor Starter Contactors: During the 2012 Problem Identification &

Resolution inspection, the inspectors were concerned the licensee was not testing installed relays and motor starter contactors to verify their design basis capacity in accordance with IEEE Standard 336-1971 and Regulatory Guides 1.30 and 1.33. The inspectors noted that the Regulatory Guides did not contain detailed or specific testing instructions and only had general guidelines. The IEEE-336 did have detailed instructions for installation, inspection, and testing for class 1E power, instrumentation and control equipment at nuclear facilities. While reviewing the applicability section of the IEEE-336, inspectors noted the standard did not apply to periodic testing and maintenance following initial installation. The standard only applied to initial installation of new equipment or equipment modifications, or modification of power, instrumentation and control equipment and systems in a nuclear facility from the time the equipment was turned over for installation until it was declared operable for service. Therefore, the inspectors concluded the existing periodic testing and maintenance activities performed by the licensee on installed relays and motor starter contactors were adequate. No violations of NRC requirements were identified by the inspectors. Therefore, this URI is closed.

4OA6 Management Meeting

Exit Meeting Summary

On July 24, 2015, the inspectors presented the inspection results to Mr. P. Gardner, and other members of the licensee staff. The licensee acknowledged the issues presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. Several documents reviewed by the inspectors were considered proprietary information and were either returned to the licensee or handled in accordance with NRC policy on proprietary information.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

P. Gardner, Site Vice President

S. Northavol, Vice President Nuclear Fleet Operations

T. Taylor, Vice President Nuclear Oversight

H. Hanson, Jr., Plant Manager

A. Gonnering, Configuration Management Supervisor

M. Kelly, Performance Assurance Manager

M. Lingenfelter, Director of Engineering

K. Scott, Director Site Operations

A. Ward, Regulatory Affairs Manager

R. Zyduck, Design Manager

B. Halvorson, Engineering Supervisor

A. Kouba, Regulatory Affairs Manager

C. Fosaaen, Regulatory Affairs

N. Friebel, Design Engineer

D. Alstad, Design Engineer

E. Watzel, Electrical Design Engineering Supervisor

P. Young, Program Engineering Supervisor

U.S. Nuclear Regulatory Commission

K. OBrien, Director, Division of Reactor Safety

P. Zurawski, Senior Resident Inspector

P. Voss, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000263/2015007-01 NCV Inadequate Quality Assurance Controls for Nitrogen Supply for the AN2 System (Section 1R21.3.b.(1))

05000263/2015007-02 NCV Failure to Establish Procedure Covering Preventive Maintenance Schedules for Inspection or Replacement of Non-critical, Safety-related Parts with a Specified Lifetime (Section 1R21.3.b.(2))

Closed

05000263/2015007-01 NCV Inadequate Quality Assurance Controls for Nitrogen Supply for the AN2 System (Section 1R21.3.b.(1))

05000263/2015007-02 NCV Failure to Establish Procedure Covering Preventive Maintenance Schedules for Inspection or Replacement of Non-critical, Safety-related Parts with a Specified Lifetime.

(Section 1R21.3.b.(2))

05000263/2012008-01 URI Qualification Basis for Safety-Related Relays and Motor Starter Contactors (Section 4OA5)

05000263/2012008-02 URI Concern with Periodic Design Basis Testing of Installed Relays and Motor Starter Contactors (Section 4OA5)