ML15245A785

IR 05000263/2015007; on 06/22/2015 - 07/24/2015; Monticello Nuclear Generating Plant; Component Design Bases Inspection

ML15245A785
Person / Time
Site:	Monticello
Issue date:	09/02/2015
From:	Christine Lipa NRC/RGN-III/DRS/EB2
To:	Gardner P Northern States Power Co
References
IR 2015007
Download: ML15245A785 (33)

See also: IR 05000263/2015007

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE RD. SUITE 210

LISLE, IL 60532-4352

September 2, 2015

Mr. Peter A. Gardner

Site Vice President

Monticello Nuclear Generating Plant

Northern States Power Company, Minnesota

2807 West County Road 75

Monticello, MN 55362-9637

SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - NRC COMPONENT DESIGN

BASES INSPECTION (INSPECTION REPORT 05000263/2015007)

Dear Mr. Gardner:

On July 24, 2015, the U.S. Nuclear Regulatory Commission (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 Non-Cited Violations (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

P. Gardner -2-

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/

Christine A. Lipa, 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

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No: 50-263

License No: DPR-22

Report No: 05000263/2015007

Licensee: Northern States Power Company, Minnesota

Facility: Monticello Nuclear Generating Plant

Location: Monticello, MN

Dates: June 22, 2015, through July 24, 2015

Inspectors: A. Dunlop, Senior Engineering Inspector, Lead

B. Jose, Senior Engineering Inspector, Electrical

M. Holmberg, Senior Engineering Inspector, Mechanical

C. Phillips, Operations Inspector

S. Gardner, Electrical Contractor

G. Gardner, Mechanical Contractor

Observer: I. Khan, Engineering Inspector, Electrical

Approved by: Christine A. Lipa, Chief

Engineering Branch 2

Division of Reactor Safety

Enclosure

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, and an

associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the

failure to assure measures were established for the selection and review for suitability

of application of materials, parts, equipment and processes that were essential to the

safety-related functions of structures, systems and components. Specifically, the

licensee failed to review for suitability of application of safety-related Agastat and

General Electric relays that had exceeded their service life, a condition non-conforming

to their design basis, to justify their continued service considering in-service

deterioration. The licensee previously entered this finding into the CAP, and

completed corrective actions to replace or evaluate some relays and implemented

a program to address the remaining relays in a timely manner.

2

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, these

safety-related 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 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))

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REPORT DETAILS

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection (71111.21)

.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.

4

.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

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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.

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• 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.

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• 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.

8

• 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 125vdc 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.

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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

10

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.

11

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).

12

(2) Failure to Review for Suitability of Application of Safety-Related Relays Installed Beyond

Their Service Life

Introduction: The inspectors identified a finding of very low safety significance (Green),

and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control,

for the failure to assure measures were established for the selection and review for

suitability of application of materials, parts, equipment and processes that were essential

to the safety-related functions of SSC. Specifically, the licensee failed to review for

suitability of application of safety-related Agastat and General Electric (GE) relays that

exceeded their service life, a condition nonconforming to their design basis, to justify

their continued service considering in-service deterioration.

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 the extension of their service life. 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 the failure to review for suitability of application of

safety-related relays installed beyond their service life to justify their continued service,

considering in-service deterioration, was contrary to 10 CFR Part 50, Appendix B,

Criterion III, and a performance deficiency. The finding was determined to be more than

minor in accordance with IMC 0612, Appendix B Issue Screening, 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, these safety-related relays were installed in

protective circuits such as reactor protection system, etc., and their failure could impact

the proper operation of these protective schemes.

13

The inspectors determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1

Initial Screening and Characterization of Findings, 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 qualification deficiency of a

mitigating SSC 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 licensees current performance.

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, required,

in part, Measures shall be established to assure that the selection and review for

suitability of application of materials, parts, equipment, and processes that are essential

to the safety-related functions of SSC.

Contrary to the above, as of July 24, 2015, the licensee failed to establish measures

to ensure the selection and review for suitability of application of materials, parts,

equipment, and processes that were essential to the safety-related functions of SSC.

Specifically, the licensee failed to review for suitability of application of safety-related

Agastat and GE relays that exceeded their service life, a condition nonconforming to

their design basis, to justify their continued service considering in-service deterioration.

Because this violation was of very-low safety significance, and it was entered into the

CAP as AR 01446684, where corrective actions to replace or evaluate relays were

either already completed or scheduled for completion in a timely manner, 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 Review for Suitability of Application

Safety-Related Relays Installed Beyond Their Service Life.)

.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.

14

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

250 vdc Battery Chargers in the event of an SBO.

b. Findings

No findings were identified.

15

4. OTHER ACTIVITIES

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 05000263/2015007-01, Failure to

Review for Suitability of Application of Safety-Related Relays Installed Beyond Their

Service Life. 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

16

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

.1 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

17

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

P. Gardner, Site Vice President

S. Northavol, Vice President Nuclear Fleet Operations

T. Talyor, 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 Review for Suitability of Application of

Safety-Related Relays Installed Beyond Their Service Life

(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 Review for Suitability of Application of

Safety-Related Relays Installed Beyond Their Service Life

(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)

Attachment

LIST OF DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection. Inclusion on this list does

not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that

selected sections of portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

CALCULATIONS

Number Description or Title Revision

lnservice Testing Pump and Valve Acceptance Criteria Rounding

01-036 48

Evaluation

01-043 Verification of Torus to Drywell Vacuum Breaker Sizing Parameters 0

02-179 MNGP 125 Volt Division. I Battery Calculation 3

04-048 MNGP 250 Volt Division I Battery Calculation 2

05-128 #13 and #16 Battery Charger Sizing 0

06-104 480V MCC to Terminal Voltage Drop 3E

08-077 AOV System Calculation - RCIC 0A

09-192 Reactor Building Composite Profiles for Environmental Qualification 0

10-168 RHR and Core Spray Motor Feeder Cable Sizing 0

10-118 RCIC MOV Functional Analysis 0

11-295 MO-2075 Component Calculation 0

11-326 ND DDGV EPRI PPM Calculation 0

14-001 Monticello Stem Lubrication Study 0

14-057 Evaluation Buried Diesel Oil Overflow Line for Day Tank T-45B 1

14-073 EDG Diesel Oil Hydraulic Model 0

15-014/12 EDG Fuel Oil Piping/Cable Fire Barrier 0

92-220 Instrument Setpoint Calculation, 4.16 Kv Degraded Voltage 2

92-224 Emergency Diesel Generator Loading 006A

AC Loads Study, Degraded Voltage Setpoint, 1R Transformer,93-066 6

LOCA Load

94-094 MCC Starter Coil Pick-Up Voltages & Maximum Cable Lengths 1, 1B

95-049 Monticello Apparent Disc Coefficient of Friction Determination 3

CA 08-157 Combined AC Model Database 000-B

CA-00-003 Response Time Increase of SRV Solenoids 0

Drywell to Suppression Chamber Differential Pressure Decay Curve

CA-00-057 0

for a 1 Inch Diameter Orifice

CA-00-104 Intake Structure Minimum Water Level 0A

Determination of the Maximum Allowable Torque to Open Torus to

CA-01-037 0A

Drywell Vacuum Breakers

CA-01-053 Evaluation of the Pressure Capacity of a Door 7

CA-01-137 Evaluation of Drywell/Wetwell Vacuum Breakers 0

Maximum Allowable Leak Rate for the RCIC Minimum Flow Valve

CA-01-155 1

Air Accumulator System

CA-01-174 Minimum Required RHRSW Pressure at RHR Heat Exchanger 3

CA-01-188 RHR Motor Start Time Evaluation 0

CA-02-002 RCIC Min Flow Line Flow Rate Analysis 0

CA-02-145 HPIC and RCIC NPSH Calculations for Use in EOPs 0

CA-02-197 EQ of Dow Corning Silicone RTV Foam 1

2

CALCULATIONS

Number Description or Title Revision

CA-03-039 SRV Low-Low Setpoint 1

CA-03-041 Setpoint Calc SRV Low-Low Set Inhibit Timer 0

CA-03-06 AOV Component Calculation, CV-2104 3

CA-03-097 HPCI/RCIC Suction Head Height Difference 0

CA-03-199 Sensitivity of EOP Calculations to ECCS Pump Curve Data 0

CA-05-019 NPSH Requirements for Operating ECCS Pumps from the CST 0C

CA-05-124 Hydrogen Generation of #13 & #16 Battery Rooms 15

Core Spray and LPCI Flow Delivered to Reactor Vessel for

CA-13-055 0

Safety Analyses

CA-80-020 NPSH Requirements for RHR Pumps 0

CA-91-009 250VDC Fault Current 1

CA-92-224 Emergency Diesel Generator Loading 6

CA-94-017 Calculation of Alternate Nitrogen Operability Leakage Criteria 10

Determine the Minimum RHR Pump Flow Required During

CA-95-099 0

Testing

Stem Thrust Assessment of 3 A/D Gate Valves: MO-2075 &

CA-95-116 1

MO-2076

CA-96-079 High Energy Line RCIC HELB in the RCIC Room 1

CA-96-169 HPCI and RCIC NPSH Evaluation 3B

LLRT Test Volumes for the RCIC Air Accumulator Check Valve

CA-97-194 0

AI-612

CA-99-011 Outlier (Seismic) Evaluation of Service Water Pumps 1

EC15368 RX Bldg Envir for EPU HELB, SBA & Post LOCA 1

EC17914 Motor Control Center Thermal Lag Analysis 0

Review of TOL Performance for Auto Initiated MOVs During a

EC25687 0

Degraded Voltage Condition

Review of Protective Device Performance for Safety Related

EC25688 Continuous Duty 480V Loads During a Degraded Voltage 0

Condition

FBS-0503-1 Fuse and Breaker Coordination Study 2

PRA-CALC-

Makeup Requirements After Scram from 1775 MWth 0

II.SMR.02.001

PRA-CALC-

RCIC Min Flow Valve 0

II.SPA.02.001

EC22209 Evaluation of Agastat Service Life 0

Service Life Evaluation for Select GE HFA Century Series

EC24650 0

Relays

Engineering Evaluation Supporting 2000 Cycle Test Basis for

EC25254 0

Limitorque MOVs

EC25683 Service Life Evaluation for GE HFA Century Series Relays 0

Service Life Evaluation for Select GE HGA Century Series

EC25710 0

Relays

Service Life Evaluation for Reasonable Assurance of Agastat

EC25719 0

Function

3

CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION

Number Description or Title Date

01483209 2nd Transmittal did not have All Requested Info 06/17/15

01483311 Calcs for EDG Fuel Oil Mod Needed to be Re-sent 06/22/15

01483808 Material Storage in RCIC Cable Closet 06/23/15

01483828 Bent Rod Hanger on Conduit in RCIC Room 06/23/15

01483833 Typo Corrected on the Receipt Inspection Report from 2012 06/23/15

01484025 No Formal Testing of ECCS Corner Room Sump Capacity 06/24/15

01484043 Inspector Question Response Delay 06/25/15

01484051 Unexpected Absence of Inspector Shadow 06/25/15

01484170 CMTRs not Located- E-SRV & Valve Body & Disc of AO-2382A 06/25/15

01484177 13 Diesel Fuel Cooler Fan Power Cable Potential Damage 06/25/15

01484180 Internal Flood TCOA Scenario Insufficient 06/25/15

01484193 Incorrect Calc Given to NRC 06/25/15

01484193 Calc 02-197 not Taken to Inactive Correctly 06/25/15

01484210 CMTR for Weld Material Q12 not Provided to NRC 06/25/15

01484265 Walkdown Forms not Incorporated into Plant Records 06/26/15

01484265 DOL Walkdown Forms not Incorporated into Plant Records 06/26/15

01484364 Internal Flooding DBD not Consistent with Licensing Basis 06/26/15

01484365 03-006 Apparent Typo Error in Section 6.0 06/26/15

01484534 MOV TOL Calculations at Incorrect Status 06/29/15

01484697 Extraneous Information on NX-8685-4 06/30/15

01484859 Listed Horsepower Wrong on RHRSW Control Drawing 07/01/15

01484919 Procedure 0214 Temperature Controls Needs Enhancement 07/02/15

01485196 RCIC 4120-PM Documentation Enhancement 07/06/15

01485387 8153 Procedure Improvement Opportunities 07/07/15

01485410 Unable to Locate GE SIL 196 Evaluation Supplement 5, 11, 17 07/07/15

01485425 ECCS Corner Room Sump Pump Capacity 07/07/15

01485467 NRC Insp Question Response Delay 07/08/15

01485508 FLEX Charger Mod did not Update Procedure 8153 07/08/15

01485509 ECCS Corner Room Sump Pump Information Differences 07/08/15

01485551 Motor Program Documents not Maintained 07/08/15

01485554 Evaluate Recommendation of GE SIL 196 Supplement 1 07/08/15

01485569 RV-1524 Outlet Pipe Size not in Compliance with B31.1 Code 07/09/15

01485668 Motor Refurb Spec does not Exist for Safety-Related 07/09/15

01485693 Diesel Oil System RV Reaction Loads not In Pipe Analysis 07/09/15

01485697 AR01322841-10 Completed at Incorrect Status 07/09/15

01485786 Delay in Issuing Contract to Support Question Response 07/10/15

01485799 SRV PM Requirements not Transferred to Vendor PM 07/10/15

01486266 Drawing does not Reflect Specific Valves with Shims 07/15/15

01486343 Torus/DW Vacuum Breaker Test Report 07/16/15

01486500 VTMs not Updated to Reflect SLA Results 07/17/15

01486699 Safety Related Relays not Part of the Vendor Contact Program 07/20/15

01486828 16A-K37 Beyond Service Life 07/21/15

01486951 Some Alt N2 Bottles Stored in REC WH not Tagged 07/22/15

01486991 Question on AN2 Bottle Pressure and Gas Quality 07/22/15

01487027 Lack of Procedure Controls for AN2 Bottle Storage 07/22/15

01487139 NRC Question on Qualification of Motor Bearing 07/23/15

01487272 Component Storage in Recv Whse Questioned 07/24/15

4

CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION

Number Description or Title Date

00621181 RIS-2000-12-Resolution of Generic Safety Issue B-5 08/28/00

00630310 Lack of Ventilation to Diesel Pump Room 08/14/02

01044201 Fuel Oil House has Potential for Hazardous Environment 08/14/06

RSW Pump Flow Ref Value Discrepancy Exists Between

01169547 02/16/09

Calculation & Procedure

01182779 #13 DG Small Coolant Leak Getting Worse 05/21/09

01192708 Question Basis for Min RHRSW DP per IST Program 08/07/09

01209786 13 RHRSW Pump September Trends 12/08/09

01229823 P-109C IST Reference Value Change 04/28/10

01233820 Basis Change: Add LR Text and Attribute for OCCW PMRQ's 05/21/10

01238600 V-EF-40B Discharge Flow Out of Specification 06/2510

01242365 V-EF-40B Discharge Flow Out of Specification 07/23/10

01249264 Failed PMT on V-EF-40B 10/07/10

01278466 MO-2075 Excessive Thrust on As-Found VIPER Test 04/01/11

01289417 Possible Deficiency in EQ Testing of Limitorque Part 21 06/06/11

01289887 RV-2-71E Elevated Tailpipe Temps 06/09/11

01291640 RCE Leaking SRV RV-2-71E 08/08/11

01291959 Foreign Material Found in Spare SRV (at NWS) 06/24/11

01293850 Allowable Leakage for HPCI/RCIC Minimum Flow Valves 07/08/11

01312421 Untimely Resolution of CAP 1196513 11/09/11

01332373 2012 CDBI Motor TOLs may Trip w/ Degraded Volt 04/05/12

01332567 2012 CDBI TOL Coordination w DVR 04/06/12

01334146 ACE 2012 CDBI TS Degraded Voltage Time Value 04/17/12

01334248 Potential Margin Reduction from Degraded ECCS Pump Head 11/26/13

01338565 MOV Limiting Stroke Time Margin Issue 05/21/12

01338566 RHRSW Reduced Flow/Head Margin Issue 05/21/12

01338567 RHR Reduced Flow/Head & NSPH/Vortex Concerns 07/15/14

NRC IN 2012-14 Motor-Operated Valve Inoperable due to Stem-

01345964 07/25/12

Disc Separation

01350679 P-109C, Reference Value Change 09/06/12

01356651 Discrepancy with Sulzer Info and 4214-PM 10/26/12

01375387 SRV LLS TS Allow Values Conflict w/ SRV Mech Allow Setpoint 03/20/13

01375742 OE: NRC IN 2013-05 03/22/13

01378744 MR Evaluation, E SRV 06/28/13

01379613 SRV Actuator Testing may be Non-Conservative 04/19/13

01389246 NRC has Question on Alignment Data for HPCI and RCIC 08/05/13

01390472 OE 248697 Both Divs of RHR Inop From Leak 07/18/13

01411214 MR Evaluation, 13 DG, Non 1E 12/30/13

01417977 Failure of Drywell Vacuum Breaker to Close 02/07/14

01418471 AO-2382A Torus-to-DW Vac Brk Closed Indication Anomaly 02/11/14

01420318 Torus Vacuum Breaker Inadequate PMT 02/25/14

01420700 Small Coolant Leak on #13 DG Radiator 02/28/14

01423951 13 RHRSW Pump Exceeds MR Reliability Criteria 03/24/14

01424260 Future Preconditioning of Vacuum Brkrs Found Unacceptable 03/26/14

01426064 TS 3.6.1.7 has no Actions for Closed Vlv Brkr that Failed STP 04/09/14

01431529 Internal Flooding PAB F.P. Break Control of TCOA Inadequate 06/06/15

5

CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION

Number Description or Title Date

01434290 Coolant Leak Observed on #13 DG 06/12/14

01438672 Oil Leak Detected on 13 Diesel Generator 07/17/14

01439686 Undefined Term Used in the USAR with Regard to RCIC 11/05/14

01443013 Replacement Alternate N2 System Found Empty 08/14/14

01443073 13 Diesel Generator has a Minor Leak from Header 08/15/14

01443510 OE: 312166 Question Concerning RCIC Cooling Test 08/19/14

01448769 C&D Tech Identifies Issue with Battery Separator Plates 10/01/14

01453481 OE: NRC Part 21 C&D Technologies Batteries 10/29/14

01459466 13 Diesel (G-90) Engine Coolant Leak 12/15/14

01459539 OE: NRC PEN 50675 LaSalle RCIC Unanalyzed Condition 12/15/14

01471379 T-44 Level with 2 Pump Operation after Mod Implementation 03/24/15

01474704 Design Issue Discovered in DOL Separation Modification 04/15/15

01475109 Design Input not Considered for DOL Hydraulic Model 04/17/15

01475179 MO-2075 Exceeded App J Admin Limit 04/18/15

01475653 Leakage Found on RV-2-71E Actuator 04/22/15

01476203 Air Leaks on SRV Bellows Leak Alarm Pressure Switches 04/24/15

01476257 P-160A/B Pipe Unions Found Hand Tight 04/25/15

01477101 MOV Transient Analysis did not Consider TOL Size 04/30/15

01477714 No Formal Calc to Support MOV TOLs 05/05/15

01477916 Invalid AO-2382A Full Open Torque 05/06/15

01477935 CAs for 2012 Violation Inadequate 05/06/15

01478212 Interference on AO-2382A Vacuum Breaker Actuator 05/08/15

01479704 Circuit Protective Device Operation-Sustained Degraded Volt 05/18/15

DRAWINGS

Number Description or Title Revision

ES1506100 Fuel Transfer Pump Assembly A

M-288 Reactor BLDG.-Plan at EL. 8963 C

NE-36347-10 #142-480V MCC B42 81

NE-36394-10B RHRSW Pump P109-C ACB No. 152-507 76

NE-36399-9 Essential Bus Transfer Circuits-Division I 77

NE-36404-4B RHR Pump P-202C ACB 152-503 Control 76

NE-36438-9 11EDG Diesel Oil Pumps A and C, P-160A and C Control 82

NE-36640-5 250VDC MCC Schedule D311 76

NE-93503-3 HVAC Controls & Interlocks Scheme V201 F

NE-93504-20 EFT System HVAC Annunciator C

NE-93545 Loop Diagram Exhaust Fans V-EF-40B 4

NF-119034-2-C #12 Diesel Generator Fuel Oil System 0

NF-36298-1 Electrical Load Flow One Line Diagram 111

NF-36298-2 DC Electrical Load Distribution One Line Diagram 90

NF-36672 Standby Diesel Generators Arrangement & Piping 78

NF-95915-3 Blowdown Control System Division I Elem Diagram 76

NH-170037 Main Control Room CRV/EFT System 81

NH-36049-10 Alternate Nitrogen Supply System 78

NH-36051 Diesel Oil System 84

NH-36241-1 Reactor Pressure Relief 78

NH-36246 Residual Heat Removal System 84

6

DRAWINGS

Number Description or Title Revision

NH-36247 Residual Heat Removal System 85

NH-36251 RCIC (Steam Side) 80

NH-36252 RCIC (Water Side) 79

NH-36664 RHR Service Water & Emergency Service Water Systems 87

NH-36665 Service Water System & Make-up Intake Structure 97

NH-91177 Disc and Post for Vacuum Breaker Valve C

NQ-74976 Three Hour Fire Barrier 11 EDG Room, EDG Trench 0

MNGP Main Steam Safety/Relief Valve, Target Rock Model E

NX-15111-1

7467F, 6X10, Outline

NX-17496-3 MNGP Protective Relay Cards-4kV 11

NX-7822-22-5A RCIC Steam Supply Isolation MO-2075 Scheme A

Main Steam Safety/Relief Valves, Target Rock Model 7367F, 77

NX-7831-439

6X10, Parts List

NX-7831-539 SRV Air Actuator Model 7467F 77

NX-7905-77 600 HP RHR Pump Motor 76

NX-8685-4 Funbore Vacuum Breaker Valves E

NX-9068-37 Outline Induction Motor F

NX-9235-32 3 600# Globe Valve Motor Operated A

3 900# Gate Valve MO-2075 & MO-2076 Carbon Steel Bolted L

NX-9235-43

Bonnet

NX-9285-5 Fuel Transfer Pump Assembly 0

NX-9525-1 RHRSW Pump Assembly 76

NX-9525-8 RHRSW Pump Open Flange Column Details 76

MISCELLANEOUS

Date or

Number Description or Title

Revision

Plant Health Report - RHRSW System 06/10/15

Plant Health Report - Emergency Filtration Train 06/18/15

Plant Health Report - Reactor Core Isolation Cooling System 06/10/15

System Walkdown Observation- EDG Fuel Oil 11/26/13

System Walkdown Observation- EDG Fuel Oil 08/28/14

System Walkdown Observation- EDG Fuel Oil 12/18/14

System Walkdown Observation- EDG Fuel Oil 03/30/15

Safety Relief Valve Data Sheets-RV 1523, 1524, 1525, 1526 12/11/14

System Health Report - Auto Pressure Relief 06/10/15

System Health Report - Diesel Oil System 06/10/15

System Health Report - Primary Containment 06/10/15

Design Criteria Document - Heating, Ventilation, and Air

Conditioning System for the Main Control Room, Emergency

10040.D5.7 5

Filter Train and Technical Support Center at Monticello Nuclear

Generating Plant, Northern States Power

2015-01-030 Component Design Basis Inspection (CDBI) Readiness 0

Design Specification - Drywell to Suppression Chamber

22A1121 0

Vacuum Breakers

79M070 Design Change to Torus to Drywell Vacuum Breaker 0

98-018 EQ, General Electric Motors (50.49) 1

7

MISCELLANEOUS

Date or

Number Description or Title

Revision

98-026 Limitorque Motor Operators (50.49) 0

A.3-15-E Fire Zone 15-E Strategy 7

Contract

COC- 4(1) Gate CS 800# 01/16/15

940015040

DBD T.08 Design Basis for Internal Flooding 3

DBD-B.02.03 Reactor Core Isolation Cooling System 77

DBD-B.03.04 Residual Heat Removal System 7

DBD-B.08.01.03 Residual Heat Removal Service Water System 6

Control Room Heating, Ventilation and Emergency Filtration

DBD-B.08.13 3

System

GE Letter - Monticello Nuclear Power Station - Response to

DRF T23-00789-

NMC Question Regarding Impact of Power Rerate on 03/25/01

00

Drywell-to-Suppression Chamber Steam Bypass Leakage

EM7114T Baldor 1//.75,1760//1460RPM,3PH,60//50Hz 08/08/14

EQ 98-022 General Electric MCCs 0

FBS-0507-1 Fuse/Breaker Coordination Study, P-109C 1

FBS-4030-02-1 Fuse/Breaker Coordination Study, P-160B 0

FBS-4080-51-1 Fuse/Breaker Coordination Study, P-160-D 0

Nuclear Management Company, LLC Monticello Nuclear

GE-NE-0000-

Generating Plant Extended Power Uprate Task T0400 3

0060-9229-TR-R3

Containment System Response

Heat 001M64068 COC- 1 (217) 2 A106 Schedule 80 SMLS Pipe 03/02/15

Heat 00A132529 COC- 2 (4510) 1A106 Schedule 80 SMLS Pipe 01/02/15

LOT 59464 CMTR- 100- A105 Nuclear 90 Elbow 07/22/14

MPS-0522 Vacuum Breaker Valves 01/14/86

MPS-0567 Specification Hollow Metal Doors, Frames, Hardware 0

MPS-1010 Piping Materials, Classification and Standards for the MNGP 30

Specification for the Analysis Piping and Piping Support

MPS-1100 11

Systems

Specification for the Procurement of Emergency Diesel

MPS-2172 3

System Diesel Oil Transfer Pumps

NEDC-32514P SAFER/GESTR-LOCA Loss of Coolant Accident Analysis 1

NFPA70 National Electric Code 2011

NSP-43-103 Specification for Vacuum Breaker Replacement Parts 0

NSP-53-103 Wetwell to Drywell Vacuum Breaker Replacement Parts 1

Part 21 on C&D Technologies Battery Cells - Misaligned

OE Eval 01/15/15

Separators

P.O.49546 Schulz Certificate of Conformance, RHR Motor Overhaul 04/13/15

P.O. 205-AB841 COC- Main Steam Safety Relief Valve 11/02/74

P.O. 56112 COC-Pressure Relief Valve 02/27/15

SCR 02-0324 USAR 5.2.1.2.3, Vent and Vacuum Relief System, Rev. 19 0

SCR 14-0541 EC 23085 EDG Fuel Oil Train Separation Screening 4

Locked Valve Program Improvements and Associated USAR

SRI 96-003 0

Changes

TC-15991 RHRSW 13 Motor/Pump Curves 12/08/11

TP-ESI506100 Functional Test Procedure Fuel Transfer Pump/Motor Asm 1

8

MODIFICATIONS

Date or

Number Description or Title

Revision

DC79M070 Modify Drywell to Torus Vacuum Breakers, Add 1 and Add 2 0

EC11690 Column Gaskets not Required on RHRSW Pumps 0

EC17503 RHRSW Pump Impeller Material Change 0

EC19903 Restoration of Motor Overload Margins in MCC-134/144 0

EC22104 EDG Fuel Oil Transfer System Modification Support 7/10/14

EC23085 EDG Fuel Oil Train 0

EC23805 EDG Fuel Oil Train Separation 0

Drywell to Torus Vacuum Breakers- Remove Upper Portion of 0

EC25684

Test Actuator Piston Rod

EC25733 Alternate Nitrogen Bottle Change-out Check Valves 0

ECN25569 EDG Fuel Oil Train Separation 0

OPERABILITY EVALUATIONS

Number Description or Title Date

No Analysis Found for HELB at MO-2078 RCIC Steam

01430505-01 05/29/14

Supply and its Effect on MCC-311

Do the ESW and DGN (FSW & RSW also) Systems Remain

01431915-01 Operable While Bypassing the Basket Strainers for Periodic 06/14/14

Cleaning

01442471-01 RCIC High Steam Flow dp Switch Found Out of Tolerance 08/15/14

01478212 Past Operability AO-2382A Vacuum Breaker 06/25/15

PROCEDURES

Number Description or Title Revision

ECCS Emergency Bus Undervoltage Test and ECCS Loss of

0036-01 30

Normal Auxiliary Power Test

0137-A LLRT-LRM-Makeup Flow Method 6

0197-01 # 13 250 Vdc Battery Capacity Test 24

RCIC CV-2104 Air Accumulator Check Valve (AI-612) Leak

0255-08-ID-03 20

Rate Test

SRV Position Indication and Low Set System Instrumentation 29

0294

Checks

1136 RHR Heat Exchanger Efficiency Test 33

1374 Monthly Operability Test of No.13 Diesel Generator 19

1388 13 DG Auto Start/Loading Test 13

1401-01 Locked Valve Alignment 23

1444 Pre and Post Severe Weather Inspection Checklist 10

4050-PM Torus to Drywell Vacuum Breaker Seal Replacement 8

4280-03-PM SRV Refurbishment and As-Left Steam Testing 40

4525-PM NO. 13 & 16 Battery Charger Preventive Maintenance 12

8153 Powering Div. II 250VDC Battery Chargers from #13 Diesel 5

A.6 Acts of Nature 52

B.02.03-01 Reactor Core Isolation Cooling 5

B.03.04-01 Residual Heat Removal System 12

9

PROCEDURES

Number Description or Title Revision

B.07.01-02 Operations Manual 21

B.08.01.03-01 RHR Service Water System 10

B.08.01.03-05 RHR Service Water System - System Operation 46

B.08.07-05 Extreme Cold Weather Procedure 45

B.08.08-01 Plant Communications System 6

B.08.08-02 Plant Communications System 4

B.08.11-05 Diesel Oil System 37

B.08.13-05 Control Room H&V and EFT - System Operation 29

B.09.15-01 Non-Essential Diesel Generator 5

B.09.15-05 Non-Essential Diesel Generator 15

C.4-B.08.07.A Ventilation System Failure - Abnormal Procedures 28

C.4-B.09.02.A Abnormal Procedure, Station Blackout 46

C.4-I Plant Flooding 14

EWI-08.13.02 Motor Program 10

FP-E-MOD-02 Engineering Change Control 16

FP-E-MOD-08 Engineering Change Notices 8

FP-E-MOD-10 Modification Turnover and Closeout 13

FP-E-RTC-02 Equipment Classification 11

FP-E-SE-05 System Engineering Walkdowns 0

FP-PA-OE-01 Operating Experience Program 21

MPS-1124 Common Motor Repair and Refurbishment Specification 1

MWI-3-M-2.01 AC Electrical Load Study 14

NWS Technologies Repair of Target Rock 3 Stage Main Steam 1

NWS-R-26

Safety Relief Valves

NWS Safety Valve Test Procedure for Monticello Nuclear Plant 7

NWS-T-15

Target Rock 67F Main Steam Safety Relief Valves

OSP-AN2-0567 Monitor ADS Pneumatic Supply 7

OWI-02.03 Operator Rounds 63

OWI-03.07 Time Critical Operator Actions 10

WORK DOCUMENTS

Number Description or Title Date

00049081 0114 RCIC System Test RX Press <165 psig Cycle 05/29/15

00061207 V-EF-40B, Clean, Repair or Replace Flow Element 11/09/10

00106906 Preoperational Testing Drywell to Torus Vacuum Breakers 04/01/01

00106908 Preoperational Testing Drywell to Torus Vacuum Breakers 04/01/01

SRV Pilot and 2nd Stage Pilot Valve Assembly Inspections,

00113519 03/15/02

Refurbishment, and Steam Testing

00123486 PM 4280-1, RV-2-71E 05/21/03

00123487 PM 4280-2, RV-2-71E 05/21/03

00143657 SRV Pilot Valve Assembly (Pilot & 2nd Stage) Change Out 05/25/00

00143658 SRV Main Stage Valve Assembly 02/24/00

00387708 Test Data Evaluation for AOV CV-2104 Rising Stem Valve 04/10/13

00390096 PI-1982, Install Remaining 3 Anchor Bolts in Stanchion 09/09/09

00394602 MO-2075 Disassembly/Inspection 04/12/11

10

WORK DOCUMENTS

Number Description or Title Date

00402785 P-109C, Rebuild Spare per 4214-PM 02/16/11

00406241 V-EF-40B Discharge Flow Out of Specification 09/10/10

00414155 TD-152-503, Perform Relay PM 07/15/11

00414333 FI-9195 has Low Flow for V-EF-40B 02/02/11

00415157 TD-4KVB-21, Perform Breaker PM 12/21/12

00416774 Replace 13RHR Pump Cables 06/28/13

00422669 PMT Failure for EF-40B 07/20/11

00423219 V-EF-40B Discharge Flow is Below Spec. 06/04/11

00430036 Replace Vacuum Breaker Air Lines per EC 20501 05/31/13

00440830 MECH - Rd 1 4120-PM on S-200 (Terry Turbine) - All Steps 05/24/15

00441556 TD-Bus 15 Relays, Perform 4850-915 PM 05/21/13

00441564 TD-4KV Bus-15, Perform PM 4858-15 05/15/13

00458886 0255-08-111-1 RCIC Comprehensive PMP & VLV Tests 11/11/13

00462919 ELEC-D-52 Charger, Perform Charger 4525-PM 06/29/15

00463771 Re-Build P-202C 05/12/14

00476238 Comprehensive 13 RHR PMP & VLV Tests 12/09/14

00490342 0255-05-III-3A Comp 13 RHRSW Pump & Valve Test 09/10/14

00490607 Check Stroke Capabilities of Actuators 05/24/15

00490645 OSP-APR-0568 SRV Functional Tests 05/23/15

00490709 0255-07-IB-4 MS SRV Pilot Valve Assembly As-Found Check 05/05/15

00490762 Reactor Coolant Pressure Boundary Leak Test 05/24/15

00490806 0255-08-ID-3 RCIC CV-2104 Air Accum Check Valve LRT 05/11/15

00490905 0214 Drywell to Torus Vacuum Breaker Cycle Leakage Check 05/27/15

00490910 0127 Drywell to Torus Vacuum Breaker 05/06/15

Drywell-Torus Vacuum Breaker Inspection, Functional Tests, &

00490910 05/14/15

Calibration Maintenance of Position Indication & Alarm System

00490947 0137-07-A RX STM SUP VLV LLRT W/RX Press By Air 04/23/15

00490969 PM 4900-1 for MO-2075 02/08/04

SRV Position Indication and Low Set System Instrumentation

0491182 05/11/15

Checks

00491185 0255-07-IB-1 Main Steam AM SRV Bench Checks & Inspection 05/05/15

00491186 0131 Safety Relief Valve Bellows Monitor Check 04/27/15

00491211 # 13 250 Vdc Battery Capacity Test 04/14/15

00491333 0255-03-IA-2B Core Spray Valve Position Indication Test 05/24/15

00497230 Investigate Repair as Required AO-2382A 05/13/15

00504345 0255-08-IA-1 RCIC Quarterly Pump and Valve Tests 02/14/15

00504629 0269 Fire Protection System Valve Check 06/17/15

00505386 EC 23085 EDG Fuel Oil Train Separation and Pre-Op Testing 05/09/15

00505603 RHR Loop A Quarterly Pump and Valve Tests 03/02/15

00505605 02551-05-IA-1-1A RHRSW QRTLY PMP & Valve Test 06/06/15

00508701 Inspect and Rebuild DW to Torus Vacuum Bkr Air Actuators 05/14/15

00509926 0465-01 DIV 1 and 2 EFT Monthly Operation 05/24/15

11

WORK DOCUMENTS

Number Description or Title Date

00510950 0143 Drywell to Torus Monthly Vacuum Breaker Check 06/12/15

00510958 0255-04-IA-1-1 RHR Loop A Quarterly Pump and Valve Tests 06/08/15

00511299 1374 Monthly Oper Test of No.13 Diesel 06/15/15

00511829 0465-01 Div 1 and 2 EFT Monthly Operation 06/24/15

00513283 Operations TRB Side CK List Weekly Procedure 07/16/15

12

LIST OF ACRONYMS USED

ADAMS Agencywide Document Access Management System

AN2 Alternate Nitrogen

AOV Air-Operated Valve

ANSI American National Standards Institute

AR Action Request

CAP Corrective Action Program

CDBI Component Design Bases Inspection

CFR Code of Federal Regulations

EDG Emergency Diesel Generator

EFT Emergency Filtration Train

EQ Environmental Qualifications

GE General Electric

HELB High Energy Line Break

IEEE Institute of Electrical & Electronics Engineers

IMC Inspection Manual Chapter

IN Information Notice

IST Inservice Testing

LERF Large Early Release Frequency

LOCA Loss of Coolant Accident

MOV Motor-Operated Valve

NCV Non-Cited Violation

NPSH Net Positive Suction Head

NRC U.S. Nuclear Regulatory Commission

PARS Publicly Available Records System

PRA Probabilistic Risk Assessment

psig Pounds Per Square Inch Gauge

RCIC Reactor Core Isolation Cooling

RHR Residual Heat Removal

RIS Regulatory Issue Summary

SBO Station Blackout

SDP Significance Determination Process

SIL Service Information Letter

SRV Safety Relief Valve

SSC Systems, Structures, and Components

TIA Task Interface Agreement

TOL Thermal Overload

TS Technical Specification

USAR Updated Safety Analysis Report

URI Unresolved Item

Vac Volts Alternating Current

Vdc Volts Direct Current

13

P. Gardner -2-

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/

Christine A. Lipa, 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

DISTRIBUTION w/encl:

Janelle Jessie

RidsNrrDorlLpl3-1 Resource

RidsNrrPMMonticello

RidsNrrDirsIrib Resource

Cynthia Pederson

Darrell Roberts

Richard Skokowski

Allan Barker

Carole Ariano

Linda Linn

DRPIII

DRSIII

Jim Clay

Carmen Olteanu

ROPreports.Resource@nrc.gov

ADAMS Accession Number ML15245A785

Publicly Available Non-Publicly Available Sensitive Non-Sensitive

To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy

OFFICE RIII RIII RIII RIII

NAME BJose for ADunlop:cl ADunlop CLipa

DATE 08/27/15 09/01/15 09/02/15

OFFICIAL RECORD COPY