IR 05000282/2015007

IR 05000282/2015007, 05000306/2015007; 08/03/2015 - 09/04/2015; Prairie Island Nuclear Generating Plant, Units 1 and 2; Component Design Bases Inspection

ML15288A195
Person / Time
Site:	Prairie Island
Issue date:	10/14/2015
From:	Lipa C A NRC/RGN-III/DRS/EB2
To:	Davison K Northern States Power Co
References
IR 2015007
Download: ML15288A195 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE RD. SUIT E 210 LISLE, IL 60532

-4352 October 14, 2015 Mr. Kevin Davison Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company, Minnesota 1717 Wakonade Drive East Welch, MN 55089

SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC COMPONENT DESIGN BASES INSPECTION

INSPECTION REPORT 05000 282/201500 7; 05000 306/201500 7

Dear Mr. Davison:

On September 4

, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection at your Prairie Island Nuclear Generating Plant, Units 1 and 2. The purpose of this inspection was 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. The enclosed report documents the results of this inspection, which were discussed on September 4, 2015 , with you , and other members of your staff.

This inspection examined activities conducted under your license as they relate to public health and safety to confirm compliance with the Commission's rules and regulations, and with the conditions in your license. Within these areas, the inspection consisted of a selected examination of procedures and representative records, field observations, and interviews with personnel. Based on the results of this inspection, three NRC

-identified findings of very low safety significance (Green) were identified. The issues were determined to involve 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. These NCVs are described in the subject inspection report. If you contest the subject or severity of the 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 the Prairie Island Nuclear Generating Pl ant. In addition, if you disagree with the cross

-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III; and the NRC Resident Inspector at the Prairie Island 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 NRC's 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 Nos. 50

-282, 50-306 License Nos. DPR

-42, DPR-60 Enclosure:

IR 05000 282/201500 7; 05000 306/201500 7 cc w/encl: Distribution via LISTSERV

Enclosure U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No s: 50-282; 50-306 License No s: DPR-42; DPR-60 Report No:

05000 282/201500 7; 05000 306/201500 7 Licensee: Northern States Power Company, Minnesota Facility: Prairie Island Nuclear Generating Plant, Units 1 and 2 Location: Welch, MN Dates: August 3, 2015, through September 4, 2015 Inspectors:

J. Neurauter, Senior Reactor Inspector, Lead R. Walton, Senior Operations Engineer I. Hafeez, Reactor Inspector, Electric al G. O'Dwyer, Reactor Inspector, Mechanical C. Jackel, Reactor Inspector, NSPDP Observer J. Chiloyan, Electrical Contractor J. Zudans, Mechanical Contractor Approved by:

Christine A. Lipa, Chief Engineering Branch 2 Division of Reactor Safety

SUMMARY

Inspection Report 05000 282/201500 7 , 05000 306/201500 7; 08/03/2015 - 09/04/2015; Prairie Island Nuclear Generating Plant, Units 1 and 2

Component Design Bases Inspection

. The inspection was a 3

-week on-site baseline inspection that focused on the design of components. The inspection was conducted by five regional engineering inspectors

, and two consultants. Three Green findings were identified by the team. The findings were considered N on-Cited Violation s (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" effective date December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRC's 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 System

Green.

The team identified a finding of very low safety significance

, and an associated NCV of Title 10 , Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XI, "Test Control," for the licensee's failure to have an acceptance criteria for electrical contact resistance values in preventive maintenance procedures for 4160 Vac switchgear. Specifically, the licensee's preventive maintenance Procedure PE 0 009, "4k V Switchgear Preventive Maintenance," failed to provide adequate resistance values and acceptance criteria for electrical connections at bus bar connection points and between 4kV switchgear cubicles. The licensee entered this finding into their Corrective Action Program (CAP) with a recommende d action to add acceptance criteria into Table 1 of procedure PE 0 0 09. The performance deficiency was determined to be more than minor because it was associated with the procedural quality attribute of the Mitigating Systems cornerstone

, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

The finding screened as of very low safety significance because it was a design or qualification deficiency that did not represent a loss of operability or functionality. Specifically, the licensee determined the 4160 Vac switchgear cubicles were operable using guidance from Electric Power Research Institute Technical Report 1013457. T h e finding had a cross-cutting aspect associated with resources in the area of human performance.

Specifically, the licensee management failed to ensure procedures are available to support successful work performance.

[H.1] (Section 1R21.3.b(1))

3

Green.

The team 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 licensee's failure to assure the safety-related thermal overload relay heaters were properly sized. Specifically, the licensee failed to consider the effects of the higher acceptable stroke time limits specified in motor operated valve Surveillance Test Procedure SP 1137 , "Recirculation Mode Valve Functional Test," in safety-related thermal overload sizing calculation H6.1

, "Motor Operated Valve Thermal Overload Heater Sizing for General Electric Motor Control Centers," Rev. 5. The licensee entered this finding into their CAP, and has actions in-place to stroke motor-operated valves to prevent a thermal overload relay trip.

The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance because the finding was a design deficiency confirmed not to result in a loss of safety function of a system or a train.

Specifically, the licensee performed preliminary calculations and determined the thermal overload relays were operable. The team did not identify a cross-cutting aspect associated with this finding because it was confirmed not to be reflective of current performance due to the age of the performance deficiency. (Section 1R21.3.b(2))

Green.

The team 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 licensee's failure to design all components of the replacement Containment Fan Coil Unit s in accordance with Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Specifically, the licensee failed to use Section III design rules to evaluate the Containment Fan Coil Unit header box as specified in the replacement Containment Fan Coil Unit design specification.

The licensee entered this finding into their CAP with a recommended action to perform a condition evaluation for the new Containment Fan Coil Units to be installed in the upcoming refueling outage to ensure proper design code alignment with the design specification and the design report.

The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance because it was a design or qualification deficiency that did not represent a loss of operability or functionality. Specifically, the licensee's use of design rules from American Society of Mechanical Engineers

,Section VIII

, provided reasonable assurance for the Containment Fan Coil Unit header box pressure boundary integrity.

The team did not identify a cross-cutting aspect associated with this finding because it was confirmed not to be reflective of current performance due to the age of the performance deficiency

. (Section 1R21.5.b (1))

4

REPORT DETAILS

REACTOR SAFETY

Cornerstone s: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection

(71111.21)

.1 Introduction

The objective of the Component Design Bases Inspection 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 inspect able 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 to th is report.

.2 Inspection Sample Selection Process

The team used information from the licensee's PRA and the U.S. Nuclear Regulatory Commission

's (NRC's) Standardized Plant Analysis Risk Model to select a risk-significant accident scenario and risk

-significant components. The scenario selected was a medium loss of coolant accident (LOCA).

A number of risk

-significant components that mitigate multiple accident scenarios, including those with Large Early Release Frequency (LERF) implications, were selected for the inspection.

The team also used additional component information such as a margin assessment in the selection process. This design margin assessment considered original design margin 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, NRC resident inspector input of problem areas/equipment, and system health reports. 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 team also identified procedures and modifications for review that were associated with the selected components. In addition, the team selected operating experience issues associated with the selected components

. This inspection constitute d 2 0 samples (12 regular components , 2 components with LERF implications

, and 6 operating experience) as defined in Inspection Procedure 71111.21-0 5.

.3 Component Design

a. Inspection Scope

The team reviewed the Updated Safety Analysis Report (USAR), Technical Specification (TS), design basis documents (DBDs), drawings, calculations and other available design basis information, to determine the performance requirements of the selected components. The team used applicable industry standards, such as the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, and Institute of Electrical and Electronics Engineers standards , 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 Information Notices (INs). The review verified 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 were consistent with the design bases and appropriate may have include d 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 team reviewed the maintenance history, preventive maintenance (PM) activities, system health reports, operating experience

-related information, vendor manuals, electrical and mechanical drawings, and licensee corrective action 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 1 4 components (samples) were reviewed: 4160 Volts Alternating Current (Vac) Switchgear Bus 16

The team reviewed nameplate data, design basis description and electrical calculations and drawings to confirm the bus design capability related to loading and short circuit protection and maintenance requirements were in conformance with applicable design standards.

Test procedures and associated results were reviewed to verify bus components were adequately tested and degradation would be identified.

The switchgear protective relay testing procedures and recently completed calibration test results were reviewed to verify that the acceptance criteria for tested parameters were supported by calculations or other controlled documents.

The team performed independent calculations of available fault current contributions from the emergency diesel generator and from the offsite sources for postulated phase and ground faults and compared them with th e relay settings calculations in Electrical Transient Analysis Program (ETAP) to verify the appropriateness of the applied overcurrent relay settings.

The team also reviewed the 4 kilo-volt (k V) Bus 16 loss of voltage and bus overcurrent relay settings to ensure adequate coordination was maintained between the bus overcurrent and bus under voltage relay settings to ensure the overcurrent relays function as designed during postulated electrical bus faults.

The team also reviewed the degraded voltage relay settings to verify whether they bounded the 6 TS requirements.

The team interviewed design and system engineers and operation personnel to determine whether there were any adverse operating trends or existing issues affecting bus's reliability and to assess licensee's ability to evaluate and correct problems.

Field walkdown of 4 kV switchgear bus 16 was performed to observe material condition and to verify whether breaker alignment, breaker position and status indications were consistent with plant design

drawings. 4160 Vac Breaker 16

-3 for 12 Motor-Driven Auxiliary Feedwater (MDAFW)

Pump: The team reviewed pump motor nameplate data, design basis description and electrical calculations and drawings to confirm the design basis minimum available voltage and current requirements for the 12 MDAFW pump motor were provided by the 4kV supply breaker

. The phase and ground protective relay trip setpoints were reviewed to ensure adequate margin existed for pump motor protection and coordination to ensure no undue interference when the pump motor is performing its design function. The team also reviewed the motor feeder cable ampacity for overload and short circuit withstand capability.

PM and relay calibration test records were reviewed to confirm the design basis assumptions in electrical calculations. The team performed independent calculations to determine whether the breaker overload and short circuit interrupting duty requirements were well within the breaker capacity. The team reviewed the 4k V Breaker 16

-3 maintenance procedures and test records to verify that they conf ormed to industry standards and whether recorded contact parting times were within the design assumptions stated in licensee's ETAP calculations and breaker vendor specifications. The team interviewed design and system engineers to determine whether there were any adverse operating trends or existing issues affecting 4kV Breaker 16

-3 reliability and to assess licensee's ability to evaluate and correct problems. Field walkdown of 4kV Breaker 16-3 was performed to observe material condition and to verify whether breaker alignment, breaker position and relay status indications were consistent with plant design drawings

. 4160/480 Vac Transformer 121M: The team reviewed calculations, design basi s descriptions, nameplate data and drawings to verify that the loading of Transformer 121M, the power supply breaker and the 480 Vac load was within the corresponding equipment ratings. The team reviewed design assumptions and calculations related to short circuit currents, voltage drop and protective relay settings associated with Transformer 121M to verify that they were appropriate. The team reviewed a sample of completed maintenance and functional performance test results to verify that the power supply breaker associated with Transformer 121M and the power cables were capable of supplying the power requirements of the 480 Vac loads during normal and postulated accident conditions. The team interviewed system engineers to determine whether there were any adverse operating trends or existing issues affecting Transformer 121M reliability and to assess licensee's ability to evaluate and correct problems. The team conducted field walkdown of the 4160/480 Vac Transformer 121M to verify that equipment alignment and nameplate data were consistent with design drawings and to assess the observable material condition

. 480 Vac Distribution Bus 121: The team reviewed the 480 Vac Distribution Bus 121 to determine whether it was capable of performing its design basis 7 function.

The team reviewed associated DBDs and electrical distribution calculations including load flow, voltage drop, short

-circuit, and electrical protection coordination.

This review evaluated the adequacy and appropriateness of design assumptions, evaluated if bus capacity was exceeded and determined whether bus voltages remained above minimum acceptable values under design basis conditions.

The team reviewed the trip setpoints of the overcurrent protective devices for Bus 121 supply and selected breakers at the load center to verify that the trip setpoints would not interfere with the ability of supplied equipment to perform their safety function yet ensuring the trip setpoints provided for adequate load center protection.

Additionally, the team reviewed system maintenance test results, interviewed system engineers, and conducted field walkdowns to verify that equipment alignment, nameplate data, and breaker position were consistent with design drawings and to assess the material condition of the 480 Vac Distribution Bus 121 load center.

Finally, the team reviewed corrective action documents and system health reports to determine whether there was any adverse operating trends and to assess licensee's ability to evaluate and correct problems

. 480 Vac Motor Control Center (MCC) 1A1: The team reviewed the 480 Vac MCC 1A1 to determine whether it was capable of performing its design basis function. The team reviewed the DBDs and electrical distribution calculations including load flow, voltage drop, short-circuit current, molded case circuit breaker application, thermal overload (TOL) relay heater sizing, and electrical protection coordination. This review evaluated the adequacy and appropriateness of design assumptions, evaluated whether the MCC 1A1 bus capacity was exceeded and determined whether bus voltages remained above minimum acceptable values under design basis conditions. The team reviewed the trip setpoints of the overcurrent protective devices including TOL relays to verify that the trip setpoints would not interfere with the ability of supplied equipment to perform their safety function yet ensuring the trip setpoints provided for adequate MCC protection. Finally, the team reviewed corrective action documents and system health reports to determine whether there was any adverse operating trends and to assess licensee's ability to evaluate and correct problems. Additionally, the team reviewed system maintenance test results, interviewed system engineers, and conducted field walkdowns to verify that equipment alignment, nameplate data, and breaker positions were consistent with design drawings and to assess the material condition of the 480 Vac MCC 1A1. 125 Volts Direct Current (Vdc) Distribution Panel 15

The team reviewed the 125 Vdc Distribution Panel 15 to determine whether it was capable of performing its design basis function. The team reviewed the DBDs and electrical distribution calculations including load flow, voltage drop, short-circuit current, fused-disconnect applications, and electrical protection coordination. This review evaluated the adequacy and appropriateness of design assumptions, evaluated whether the 125 Vdc Distribution Panel 15 capacity was exceeded and determined whether bus voltages remained above minimum acceptable values under design basis conditions and met voltage and current requirements of connected safety

-related 4160 Vac circuit breaker control and logic circuit loads.

The team reviewed licensee's proposed plant modification design documents to verify whether the replacement of existing obsolete fused-disconnects had 8 properly considered the original plant design basis requirements, including environmental and seismic

, to preclude any potential adverse impacts on 125 Vdc Distribution Panel 15 capacity and all of its affected loads.

Finally, the team reviewed corrective action documents and system health reports to determine whether there was any adverse operating trends and to assess licensee's ability to evaluate and correct problems. Additionally, the team reviewed system maintenance test results, interviewed system engineers, and conducted field walkdowns to verify that equipment alignment, nameplate data, and fused-disconnect positions were consistent with design drawings and to assess the material condition of the125 Vdc Distribution Panel 15

. 125 Vdc Distribution Panel 21

The team reviewed the 125 Vdc Distribution Panel 21 to determine whether it was capable of performing its design basis function. The team reviewed the DBDs and electrical distribution calculations including load flow, voltage drop, short

-circuit current, fused

-disconnect applications, and electrical protection coordination. This review evaluated the adequacy and appropriateness of design assumptions, evaluated whether the 125 Vdc Distribution Panel 21 capacity was exceeded and determined whether bus voltages remained above minimum acceptable values under design basis conditions and met voltage and current requirements of connected safety

-related 4160 Vac circuit breaker control and logic circuit loads. The team reviewed licensee's proposed plant modification design documents to verify whether the replacement of existing obsolete fused-disconnects had properly considered the original plant design basis requirements, including environmental and seismic

, to preclude any potential adverse impacts on 125 Vdc Distribution Panel 21 capacity and all of its affected loads.

Finally, the team reviewed corrective action documents and system health reports to determine whether there was any adverse operating trends and to assess licensee's ability to evaluate and correct problems.

Additionally, the team reviewed system maintenance test results, interviewed system engineers, and conducted field walkdowns to verify that equipment alignment, nameplate data, and fused

-disconnect positions were consistent with design drawings and to assess the material condition of the125 Vdc Distribution Panel 21

. 12 Component Cooling (CC) Pump: The team reviewed CC Water 12 CC pump to verify that it was capable of meeting its design basis requirements. The 12 CC pump provides intermediate cooling between heat exchangers in potentially radioactive systems and the cooling water system during normal operations and accident conditions.

The team reviewed analyses, procedures, and test results associated with operation of the 12 CC pump under postulated transient, accident, and station blackout conditions. The analyses included considerations for hydraulic performance, net positive suction head, required total developed head, and pump run-out conditions. Seismic design documentation was reviewed to verify pump design was consistent with limiting seismic conditions. The team also evaluated the chronic pump seal issues in the recent past as well as modifications to correct these problems. In

-service test in g (IST) results were reviewed to verify acceptance criteria were met and performance degradation would be identified, taking into account set

-point tolerances and instrument inaccuracies.

The team reviewed pump motor nameplate data, design basis description and electrical calculations and drawings to confirm the design basis minimum available voltage at the 12 CC pump motor terminals would be 9 adequate for starting and running under degraded voltage conditions. The phase and ground protective relay trip setpoints were reviewed to ensure adequate margin existed for pump motor protection and coordination to ensure no undue interference when the pump motor is performing its design function. The team also reviewed the motor feeder cable ampacity for overload and short circuit withstand capability. A sample of PM and relay calibration test records were reviewed to confirm the design basis assumptions in electrical calculations. The team performed independent calculations to determine if adequate time coordination margin existed between the 4kV bus undervoltage relays and the 12 CC pump motor supply feeder circuit overload current relay trip setpoints. Field walkdown of the 12 CC pump motor power supply breaker in 4k V switchgear bus16, cubicle 5 was performed to observe material condition and to verify 12 CC pump motor power supply breaker alignment and status indications were consistent with plant design drawings.

The team also conducted a detailed walkdown of the pump to assess the material and environmental conditions, and to verify that the installed configuration was consistent with system drawings, and the design and licensing bases. In addition, the team interviewed system, test and design engineers to discuss pump performance, trending and maintenance history to determine the overall condition of the pump. Finally, the team reviewed corrective action documents to evaluate whether there were any adverse trends associated with the pump and to assess the licensee's capability to evaluate and correct problems. Unit 1 Reactor Coolant Pump (RCP) Seals

The team inspected the Unit 1 RCP seal replacement to determine if the new seal designs are acceptable and will perform their safety related function, as expected.

During Prairie Island Nuclear Generating Pl ant refuel outage

1R29 , the licensee

replaced the existing Westinghouse seals with Flowserve N9000 seals.

The replacement seals are of a different design than the original Westinghouse seals and required modifications to the seal leak

-off lines, seal housing and pump coupling.

The Flowserve N

-9000 seal provides a controlled leak barrier between the pressurized reactor coolant and the primary containment.

The seal contains three hydrodynamic seal stages and a fourth "abeyance" seal

, which provides sealing when the remaining three stage s are failed. The team reviewed the USAR, TSs, TS Bases, drawings, procedures, modifications (EC 21790 and EC 25405), calculations, DBDs, the seal pressure breakdown operating trends, Reactor Coolant System makeup capability and root cause analyses associated with 12 RCP seal failures. The team also reviewed the seal maintenance history and health reports to assure that the system is being maintained at maximum levels considering open work orders (WOs) and legacy conditions and that plans for future maintenance can assure optimal system performance. The team verified that the seal design has operated as expected considering plant operating conditions as well as ongoing 12 RCP seal issues.

Finally, the team reviewed corrective action documents to evaluate whether the adverse trends associated with the 12 RCP seal were acceptably managed to verify that the licensee evaluated and corrected problems effectively.

12 Residual Heat Removal (RHR) Pump: The team reviewed pump motor nameplate data, design basis description

, electrical calculations

, and drawings to confirm the design basis minimum available voltage at the 12 RHR pump motor terminals would be adequate for starting and running under degraded voltage 10 conditions. The phase and ground protective relay trip setpoints were reviewed to ensure adequate margin existed for pump motor protection and coordination to ensure no undue interference when the pump motor is performing its design function. The team also reviewed the motor feeder cable ampacity for overload and short circuit withstand capability.

PM and relay calibration test records were reviewed to confirm the design basis assumptions in electrical calculations. The team performed independent calculations to determine if adequate time

-current coordination margin existed between the 4k V bus undervoltage relays and the 12 RHR motor supply feeder circuit overload current relay trip setpoints. Field walkdown of 4kV RHR pump motor power supply breaker in 4kV swithchgear bus 16 cubicle 6 was performed to observe material condition and to verify 12 RHR pump motor power supply breaker alignment and status indications were consistent with plant design drawings

. Containment Sump B Isolation Valves MV-32076 and MV-32078: The team reviewed the Unit 1 containment sump isolation valves

, MV-32076 (inboard containment isolation valve) and MV

-32078 (o utboard containment isolation valve), to determine if the normally closed valves in the B RHR sump are capab le of performing their design basis function to open while transferring to the recirculation mode of safety injection. The team reviewed the USAR, TSs, TS Bases, drawings, procedures, and the IST basis document to identify the performance requirements for the valve s. The team reviewed periodic motor-operated valve (MOV) diagnostic test results and stroke

-timing test data to verify acceptance criteria were met. The team evaluated whether the MOV safety functions, performance capability, torque switch configuration, and design margins were adequately monitored and maintained in accordance with the licensee's MOV program requirements. The team also reviewed MOV weak link calculations to ensure the ability of the MOV to remain structurally functional while stroking under design basis operating conditions. The team verified that the MOV valve analysis used the maximum differential pressure expected across the valve during worst case operating conditions. Additionally, the team reviewed motor nameplate data, degraded voltage and during the most limiting duty cycle operating conditions, TOL relay sizing, and voltage drop calculation results to verify that the MOV would have sufficient voltage and power available to perform its function at degraded voltage and during the most limiting duty cycle operating conditions.

The design, operation, and maintenance of the valve were discussed with the system engineer to evaluate the valve's performance history, maintenance, and overall health. The function and design of the valve enclosures were reviewed for the primary containment function and capability. The team also conducted a walkdown of th e valves and associated equipment to assess the material condition of the equipment and to evaluate whether the installed configuration was consistent with the plant drawings, procedures, and the design bases. Finally, the team reviewed corrective action documents and system health reports to evaluate whether there were any adverse trends associated with the valve s and to assess the licensee

's capability to evaluate and correct problems. 12 MDAFW Pump: The team reviewed design calculations and site procedures to verify the design bases and design assumptions were appropriately translated into these documents. Design and operational requirements were reviewed with respect to electric supply, pump flow rate, developed head, achieved system flow 11 rate, net positive suction head and minimum flow requirements. The team reviewed the adequacy of assumptions, limiting parameters, the pump's protection from the formation of air vortexes, and the adequacy of its suction sources (condensate storage tank and safety

-related cooling water discharge piping). Test procedures and recent test results were reviewed against design basis documents to verify the acceptance criteria for tested parameters were supported by calculations or other engineering documents and validated component operation under accidents and transients. This included reviewing the adequacy of pump IST. The team also reviewed operating as well as emergency operating procedures to verify selected operator actions could be accomplished

.

Unit 1 Condensate Storage Tank (CST) (LERF Implications

): The team reviewed the design basis of the tanks to verify their capability to supply the required inventory to the Auxiliary Feedwater (AFW) system during postulated transient and accident conditions. The CST level setpoint analyses were reviewed to verify the transfer of the AFW system suction from the CSTs to the safety-related Cooling Water System would occur prior to significant vortexing, which could result in air reaching the pump suction nozzle.

The team also reviewed the operator actions required to maintain the tanks above the minimum allowable level and temperature limits Safety Injection Check Valves CV 9

-5 and CV 9

-6 (LERF Implications

): The team reviewed the Unit 1 Safety Injection valves CV 9

-5 and CV 9

-6, to determine if the normally closed pressure isolation check valves in the safety injection system are capable of performing their design basis function to isolate the RHR system (Low Pressure) from the Reactor Coolant System (High Pressure), and to open to provide a flow path for low head safety injection and long term low head recirculation. The team reviewed the USAR, TSs, TS Bases, drawings, procedures, modifications, calculations, DBDs and the IST basis document to identify the performance requirements for the valves. The team reviewed periodic check valve diagnostic test results to verify acceptance criteria for leakage and full flow capability were met. The team evaluated whether the check valve safety functions, performance capability, were adequately monitored and maintained in accordance with Prairie Island's IST Program requirements. The team also reviewed the valve maintenance history and health reports to assure that the system is being maintained at maximum levels considering aggregate effects of open WOs and legacy conditions and that plans for future maintenance can assure optimal system performance. The team verified that the check valves pressure isolation function at the maximum differential pressure expected across the valves during worst case operating conditions are being maintained.

b. Findings

(1) 4160 Vac Switchgear P M Procedure Failed to Provide Adequate Resistance Values and Acceptance Criteria Introduction

A finding of very low safety significance

, and an associated Non-Cited Violation (NCV) of Title 10 , Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XI, "Test Control," was identified by the team for the licensee's failure to have 12 an acceptance criteria for electrical contact resistance values in PM procedures for 4160 Vac switchgear. Specifically, the licensee's PM Procedure PE 00 09, "4 k V Switchgear P M," failed to provide adequate resistance values and acceptance criteria for electrical connections at bus bar connection points and between 4 k V switchgear cubicles.

Description

The team reviewed the last PM procedures for the safety

-related 4160 Vac switchgear and the results for WO 00359703-1, Bus 16 Inspecti on/DOBLE last performed in July 2012. The tests were performed in accordance with Procedure PE 0009, "4 k V Switchgear P M," Revision 21. Procedure Step 7.5.5, "Measure Contact Resistance," measure d the resistance from cubicle

-to-cubicle and end

-to-end on bus stab of the cubicle for each phase when Bus Grounds are installed (i.e., the bus is grounded).

The test values and location of the tests varied depending on the live cubicle position and testing configuration.

A note prior to Procedure Step 7.5.5 states: "Typical resistance between adjacent cubicle s is less than 75 micro

-ohms." Discussions with the licensee identified that this note provide d general guidance and does not provide firm acceptance criteria for cubicle-to-cubicle, end of bus to end of bus, or situations when cubicles are bypassed due to being energized. In accordance with procedure Step 7.5.8, "if the contact resistance and the power factor are within limits," then the steps in procedure Section 7.10 may be identified as non

-applicable with the system engineer

's or maintenance supervisor's approval. However, the team noted that the WO and associated procedure PE 0009 did not have any proper acceptance criteria provided in the procedure and requested the licensee to provide the basis of the contact resistance acceptance criterion used in maintenance testing. However, the licensee could not provide a basis for the contact resistance value used in maintenance testing.

The licensee provided the team a non-controlled spread sheet with historical contact resistance readings and Electric Power Research Institute (EPRI)

Technical Report 1013457,"Nuclear Maintenance Applications Center: Switchgear and Bus Maintenance Guide." Section 6.7.1 of this EPRI report is associated with testing connection resistance and states, in

-part: "Connection resistance or the measure of resistivity associated with a connection point provides an indication of the adequacy of the connection.

Higher resistance can lead to overheating and subsequent failure of the insulation and electrical conductor.

The physical configuration of the bus, including the number of connections, can influence readings due to the cumulative resistance values at each joint.

Cumulative resistance values can be used for large bus sections with multiple connections considering individual joint resistance values.

Baseline values can be established and trended with future measurements to provide assurance of the adequacy of each connection.

This can establish a systematic approach to identify deviations or outliers.

An acceptance criteria could include a method to investigate values that deviate more than 50 percent from the lowest value." However, the licensee could not demonstrate the EPRI guidance using the 50 percent deviation from the lowest value method criteria was incorporated into licensee procedures. Therefore, the team concluded the WO and associated procedure PE 0009 did not provide proper acceptance criteria.

13 The team concluded that without a basis for the testing acceptance criteria, the licensee cannot demonstrate the 41 60 Vac switchgear bus will perform satisfactorily in service. The licensee entered this finding into their Corrective Action Program (CAP) as Action Request (AR) 01490563, and its preliminary evaluation conclude d 4kV switchgear cubicles were operable usin g the EPRI Technical Report 1013457 guidance

the licensee recommended adding acceptance criteria into Table 1 of procedure PE 0009.

Analysis:

The team determined that the failure to have an acceptance criteria for electrical contact resistance values in safety

-related preventive maintenance procedures was contrary to 10 CFR Part 50, Appendix B, Criterion XI and was a performance deficiency.

The performance deficiency was determined to be more than minor becaus e it was associated with the procedure quality 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.

Specifically, the licensee's failure to establish contact resistance acceptance criteria was a significant programmatic deficiency which would have the potential for unacceptable or degraded conditions to go undetected.

The team determined the finding could be evaluated in accordance with Inspection Manual Chapter (IMC) 0609, Appendix A, "The Significance Determination Process for Findings At

-Power," using Exhibit 2, "Mitigating Systems Screening Questions

." The finding screened as very low safety significance (Green) because it did not result in the loss of operability or functionality. Specifically, the licensee determined 4 kV switchgear cubicles were operable using Technical Report 1013457 guidance.

The team determined that this finding had a cross-cutting aspect associated with resources in the area of human performance.

Specifically, the licensee management failed to ensure procedures are available to support successful work performance.

[H.1] Enforcement

Title 10 CFR Part 50, Appendix B, Criterion XI , "Test Control," requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service and performed in accordance with written test procedures which incorporate the requirements and acceptance limits.

Contrary to the above, as of September 4, 2015, the licensee failed to provide adequate resistance values and acceptance criteria for electrical connections at bus bar connection points and between 4kV switchgear cubicles in Procedure PE 0009 used for PM testing of 4160 Vac switchgear.

Because this violation was of very low safety significance

, and was entered into the licensee's CAP as AR 01490563, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 05000282/2015007

-01; 05000306/2015007

-01 , 4160 Vac Switchgear P M Procedure Failed to Provide Adequate Resistance Values and Acceptance Criteria

)

(2) Inadequate Calculations for M OV T OL Relays

Introduction:

The team identified a finding of very low safety significance

, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." Specifically, the licensee failed to assure and verify that the TOL relays on safety

-related MOV circuits were properly sized.

14 Description

During review of 480 volt power supply circuits for MV

-32076 and MV-32078, the team noted that TOL relay evaluation Form PINGP 1111 , "MOLR Heater Sizing ," Rev. 5 , had assumed valve stroke times of 120 and 118 seconds

, respectively. However, the team noted the licensee had not considered the effects of the longer 150 second allowable stroke time identified as the Limiting Stroke Time in Surveillance Procedure SP 1137

, "Recirculation Mode Valve Functional Test," Rev.

31. The team questioned whether the non

-conservative assum ed stroke time used in TOL relay sizing calculations combined with licensee's design that leave the TOL relays in the MOV circuits continuously could result in undersized TOL relays that could trip on overcurrent and de-energize the MOV motor circuits during a design basis event and prevent the safety-related MOVs from performing their safety-related function.

" The team reviewed licensee's TOL sizing procedure H6.1

, "MOV TOL Heater Sizing for General Electric (GE) MCCs," Rev. 5, and determined the calculation failed to consider the most limiting valve stroke time and failed to demonstrate by calculation or analysis that the TOL protection w as sized properly

." The team identified that, as of September 29, 2009, the licensee had no program to ensure MOV Limiting Stroke Time and Duty Cycle parameters were identified and verified as required design inputs into the TOL sizing calculations to ensure functional reliability and accuracy of the selected TOL trip points provide adequate MOV motor circuit overcurrent protection and coordination margin to preclude spurious tripping during all design basis MOV operation.

The licensee entered this issue into their CAP as AR 01490676. The licensee is still evaluating its planned corrective actions. However, the team determined that the continued non

-compliance does not present an immediate safety concern because the licensee has actions in

-place to stroke the MOVs to prevent a TOL relay trip.

Therefore, the licensee was able to demonstrate operability in that the TOL protection would not prevent any MOVs from performing their safety function.

Analysis:

The team determined that failing to assure that TOL protection on safety-related MOV circuits w as sized properly and verified was a performance deficiency warranting a significance evaluation. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the licensee's failure to assure that TOL protection w as properly sized could affect the ability of MOVs to respond to initiating events

. The team determined the finding could be evaluated in accordance with IMC 0609, Appendix A, "The Significance Determination Process for Findings A t-Power," using Exhibit 2, "Mitigating Systems Screening Questions

." The finding screened as very low safety significance (Green) because it did not result in the loss of operability or functionality. Specifically, the licensee was able to demonstrate operability in that the TOL protection would not prevent any MOVs from performing their safety function and has corrective actions in

-place to stroke certain MOVs non-consecutively to prevent a TOL relay trip.

The team did not identify a cross

-cutting aspect associated with this finding because the finding was not representative of current performance. Specifically, the finding was related to original plant design

.

15 Enforcement

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

Contrary to the above, as of September 29, 2009, the licensee failed to ensure design control measures were in place for verifying or checking the adequacy of the design of the TOL relays for the safety

-related MOVs, MV

-320076 , "11 Containment Sump B Isolation Valve A2

," and MV-32078 , "11 Containment Sump B Isolation Valve B2

," powered from 480 volt M CCs 1KA2-B1 and 1A2

-A5 , respectively.

Because this violation was of very low safety significance

, and was entered into the licensee's CAP as AR 01490676, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 05000282/2015007

-02; 05000306/2015007

-02, Inadequate Calculations for M OV TOL Relays)

.4 Operating Experience

a. Inspection Scope

The team reviewed six 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:

Generic Letter (GL) 96-06 "Assurance of Equipment Operability and containment Integrity During Design-Basis Accident Condition

"; Institute of Nuclear Power Operations (INPO)

Event Report (IER) L2-14-46, "Multiple Electrical Faults Result in Explosion of Unit Auxiliary Transformer and Automatic Scram"

IER L3-13-04, "RCP Motor Bearing Damage

" ; IER L3-14-25 "Heavy Snow and Design Flaw Result in Dual Unit Scram

"; IN 2012-14, "M OV Inoperable Due to Stem-Disk Separation"

and IN 2014-10, "Potential Circuit Failure

-Induced Secondary Fires or Equipment Damage."

b. Findings

No findings were identified.

.5 Modifications

a. Inspection Scope

The team reviewed seven 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:

Engineering Change (EC) 23725 , "Direct Current Panel 21 Fused Switch Replacement

";

EC 18153 , "Replacement of Battery Chargers 21 and 22 and Portable Battery Charger 11P

"; EC 21790 , "RCP Seal Replacements Unit 1

"; EC 25405, "Prairie Island 12 RCP Seal Face Replacement"

EC 23833, "12 CC Pump Seal Face Material Equivalency Evaluation"

Modification 05ZC02; "Unit 2 Fan Coil Unit Cooling Faces Replacement

"; and Modification 05ZC05; "Unit 1 Containment Fan Coil Unit Cooling Faces Replacement."

b. Findings

(1) Replacement Containment Fan Coil Unit (CFCU) Component Not Designed in Accordance with ASME Section III Introduction

The team 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 licensee's failure to design all components of the replacement CFCUs in accordance with ASME Section III

. Specifically, the licensee failed to use ASME Section III design rules to evaluate the CFCU header box as specified in the replacement CFCU Design Specification.

Description

Section 5.2.3.3 of the USAR describes the Containment Vessel Air Handling System including the CFCUs. This section states, in

-part, the Containment Cooling System consists of four fan

-coil units located in the Reactor Containment Vessel. These will re

-circulate and cool the Reactor Containment Vessel atmosphere. The heat sink for the fan coils is provided by the containment and auxiliary building Chilled Water System or by the Cooling Water System. During emergency situation s, the heat sink for the fan coils is provided by the Cooling Water System.

The CFCUs are safety related and required to be operable in modes 1

- 4 by TS 3.6.5

, "Containment Spray and Cooling Systems.

" Along with the Containment Spray System

, the Containment Cooling System limits the temperature and pressure that could be experienced following LOCA or steam line break.

As a result of flow induced erosion causing pressure boundary leaks, the licensee initiated modifications to replace existing fan coil unit (FCU) cooing coils. The team reviewed modification documentation that supported CFCU replacement to verify the modification s maintained design and licensing bases with respect to the licensee's response to GL 96

-06, "Assurance of Equipment Operability and Containment Integrity During Design

-Basis Accident Conditions

." In particular, the team reviewed a sample of design documents to verify CFCU internal pressure transients due to design basis accidents postulated in GL 96

-06 were considered as a design load.

As indicated in Modification s 05ZC05, Unit 1, and 05ZC02, Unit 2, the licensee specified the replacement FCU cooling faces to be designed and fabricated in accordance with ASME Section III, Class 2, 1989 Edition (no Addenda

). The licensee noted that the use of ASME Section III was consistent with the original Specification TS-M605. The modifications utilized the original s pecification, USAR loading requir ements , GL 96-06 water hammer loads, and transient pressure loading due to two-phase flow collapse. The licensee incorporated these requirements into Design Specification M180 0001 009, Unit 1, and Design Specification M180 0001 008, Unit 2, 17 for CFCU face replacement.

The team determined that the licensee reconciled using ASME Section III, Class 2, 1989 Edition as the replacement CFCU design code in ASME Section XI repair/replacement plan documents and in modification docume nt 05ZC05, Unit 1, and in modification document 05ZC02, Unit 2.

The team reviewed a sample of design documents to verify the design requirements were incorporated into the design specification and the replacement CFCUs were designed and fabricated in accordance with the design specifications. The team verified that the design loads included USAR loading requirements, GL 96-06 water hammer loads, and transient pressure loading due to two

-phase flow collapse.

However, the team determined that the CFCU header box component was evaluated using rules from Appendix 13 of ASME Section VIII, Div. 1, "Rules for Construction of Pressure Vessels," instead of ASME Section III, Subsection NC, Class 2, as specified in the design specifications.

Specifically, Section 4 of licensee Design Specification M180 0001 008 specifies the coils shall be designed in accordance with ASME Section III, Class 2. In addition, Section 4.4 of this design specification utilized Table s NC-3321-1 and NC-33 21-2 for load combinations, service levels, and allowable stress levels for design of nozzles and nozzle connections to the header boxes. The team identified Sub-Article NC-3200 contains design rules for vessels which may be used as an alternative to the vessel design rules in NC

-3300 to evaluate the CFCU header box component. Specifically, NC

-3211-1(c) allows the designer to perform a complete stress analysis of the vessel or vessel region considering all the loadings of NC-3212 and the Design Specifications. This analysis shall be done in accordance with Section III

, Appendix XIII for all applicable stress categories. The team did not identify referral in Section III, Class 2 to ASME Section VIII for alternative vessel design rules.

Therefore, the team could not conclude the design reports for Aerofin Calculation CA-529-1158, Unit 1, and Calculation CA 1121-1, Unit 2, demonstrated replacement CFCUs were designed using the rules of ASME Section III, Class 2 in accordance with the design specifications.

The licensee captured this issue in their CAP as AR 0140769. The corrective action recommended at the time of this inspection was for the licensee to perform a condition evaluation for the new CFCUs to be installed in the upcoming refueling outage to ensure proper design code alignment with the design specification and the design report.

Analysis:

The team determined that the licensee's failure to use the design rules of ASME Section III to evaluate the replacement CFCU header box component was contrary to the replacement CFCU design specification and was a performance deficiency. Specifically, the team determined that ASME Section III, NC-3200 contains design rules for vessels which may be used as an alternative to the vessel design rules in NC-3300 to evaluate the CFCU header box component.

The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Design Control

, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee did not perform a code reconciliation to demonstrate 18 ASME Section VIII design rules are comparable to a complete stress analysis of the header box component in accordance with alternative vessel design rules specified in ASME Section III, Sub-Article NC-3200. The team determined the finding could be evaluated in accordance with IMC 0609, Appendix A, "The Significance Determination Process for Findings At

-Power," using Exhibit 2, "Mitigating Systems Screening Questions." The finding screened as very low safety significance (Green) because it did not result in the loss of operability or functionality. Specifically, the finding is a deficiency affecting the design qualification. The team determined that meeting the design rules of ASME Section VIII provided reasonable assurance for CFCU header box pressure boundary integrity.

The team did not identify a cross

-cutting aspect associated with this finding because it was confirmed not to be reflective of current performance due to the age of the performance deficiency

. Enforcement

Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and that deviations from such standards are controlled.

Contrary to the above, since 2005, the licensee failed to assure that the appropriate design standard was used to evaluate the replacement CFCU header box component

. Specifically, the licensee failed to use ASME Section III design rules as specified in the replacement CFCU design specification.

Because this violation was of very low safety significance and was entered into the licensee's CAP as AR 01490769, this violation is being treated as a n NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 05000 282/201500 7-03; 05000 306/201500 7-03 , Replacement CFCU Component Not Designed in Accordance with ASME Section III

)

.6 Operating Procedure Accident Scenarios

a. Inspection Scope

The team performed detailed review of risk

-significant, time critical operator actions (TCOA s). These actions were selected from the licensee's PRA rankings of human action importance based on risk

-achievement worth values and selected scenarios of small/medium break LOCAs. The team reviewed licensee procedures and performed plant walkdowns. The team observed the licensee administer simulator scenarios and an in-plant job performance measure (JPM) to determine whether operators were implementing the procedure steps in a timely and accurate manner and to verify the procedures were appropriate to sufficiently mitigate events.

The procedures were compared to USAR and risk assumptions. In addition, the procedures were reviewed to ensure the procedure steps would accomplish the desired result.

The following TCOAs were demonstrated, timed and reviewed against operating procedures and design documents

JPM: Secure Turbine Building Roof Exhaust Fans Following a LOCA, TCOA 8

Scenario

Complete Safety Injection Pump Recirculation Switchover, TCOA 24

and Scenario

Trip RCP's Following a Small Break LOCA, TCOA 25

.

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 team reviewed a sample of the selected component problems identified by the licensee , and entered into the CAP. The team reviewed these issues to verify an appropriate threshold for identifying issues

, and to evaluate the effectiveness of corrective actions 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 team are listed in the attachment to this report.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On September 4, 2015, the team presented the inspection results to Mr.

K. Davison, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The team asked the licensee whether any materials examined during the inspection should be considered proprietary. Several documents reviewed by the team were considered proprietary information and were either returned to the licensee or handled in accordance with NRC policy on proprietary information.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

SUPPLEMENTAL INFORMATION KEY POINTS OF CONTAC

T Licensee

K. Davison, Site Vice President

S. Sharp, Director of Site Operations

E. Blondin, Director

of Engineering

M. Molaei, Director of Nuclear Engineering

T. LaHann, Acting Design Manager

M. Pearson, Regulatory Affairs Manager

J. Connors, Fleet Design Engineering Supervisor

K. Hernandez, Engineering Supervisor

G. Carlson, Senior Licensing Engineer

P. Johnson, Regulatory Affairs

U.S. Nuclear Regulatory Commission

K. Riemer, Chief, Projects Branch 3, DRP

L. Haeg, Senior Resident Inspector

P. LaFlamme, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUS

SED Opened and Closed

05000 282/201500 7-01; 05000 306/201500 7-01 NCV 4160 Vac Switchgear Preventive Maintenance Procedure Failed to Provide Adequate Resistance Values and Acceptance Criteria

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

05000282/2015007

-02; 05000306/2015007

-02 NCV Inadequate Calculations for Motor

-Operated Valve Thermal Overload Relays

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

05000282/2015007

-03; 05000306/2015007

-03 NCV Replacement Containment Fan Cooling Unit Component Not Designed in Accordance with ASME Section III (Section 1R21.5.b(1)) Discussed None

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

Date or Rev. ENG-ME-046 MOV Target Thrust/Torque Calculation

91-02-11 Battery II Calculation

ENG-EE-168 Documentation of DCDM Software Test

ENG-EE-171 Degraded Voltage Calculation

ENG-EE-012 125 V dc System Coordination Study

ENG-EE-180 Motor Terminal Voltage for GL

96-05 0 ENG-EE-177 AC System Fault Analysis

ENG-ME-334 Section XI Design Basis Valve Limiting Stroke Time

6C ENG-EE-170 PINGP ETAP Electrical Analysis

HG.1 Motor Operated Valve Thermal Overload Heater Sizing for GE M C C s 5 91-02-21 P121 Battery Calculation

2A ENG-EE-160 UNIT 1 4 kV Safeguards Swit

chgear Protective Relay Settings and Coordination

ENG-ME-41 CC Hydraulic Model Proto

-Power Calculation 02-002 1 ENG-ME-632 Pump Nozzle Calculations, 12 CC Pump

ENG-ME-811 CC Pump Minimum Flow Evaluation

ENG-ME-482 Pump Driver Capabilities

32-9213792 GL 2008-01 Waterhammer

Pipe Stress Analysis for RHR System No. PI

-216-01 0 32-9222184 GL 2008-01 Unit 1 RHR System Void Transport Analysis.

89890-16 Prairie Island Weak

Link Calculation for Valves MV

-32075, 32076, 32178, 32179 0 89890-17 Prairie Island Weak

Link Calculation for Valves MV

-32077, 32078, 32180, 32181.

ENG-ME-083 Justification for Installation of Bonnet Vents on RHR MV

ENG-ME-334 Section XI Design Basis Valve Limiting Stroke Times

ENG-ME-460 EPRI Performance Prediction Methodology (PPM)

for Sump B MOV s 0 ENG-ME-684 Differential Pressure for the Safety Injection MOVs at PIN GP 0A 51-9216679 Prairies Island Unit 1 and 2 UT Void Size Evaluation

ENG-ME-417 Unit 1 and 2 Sump B Suction MOV Pressure Locking Analysis 1 ENG-ME-808 Design Qualification Report of the N

-9000 N-Seal Pressure Boundary 2 ENG-ME-828 FLEX Minimum RCS Makeup Flow

with RCS Unidentified Leakage 0 01Q0274-C-001 Acceptability of 5'

-wide Tube and Coupler Scaffolds

00Q0151-C-007 Minimum Seismic Ties and Bracing for Tube and Coupler Scaffolds 0

CALCULATIONS

Number Description or Title

Date or Rev. ENG-ME-644 Verification of Heat Removal Capability of AFW Pump Lube Oil Coolers

0A NSPNAD-98006 NSP Analysis of a Feedwater Line Break for Prairie Island

ENG-ME-320 AFW Pump NPSH Calculation

3A ENG-ME-454 Pressure Drop Between the SG and Safety Valve

0B ENG-ME-576 AFW Pump Acceptance Criteria

2G ENG-ME-443 Condensate Storage Tank Sizing

4F ENG-EE-170 PINGP ETAP Electrical Analysis

ENG-EE-173 D1/D2 Emergency Generator Relay Setting Calculation

8/06/09 Aerofin CA-529-1121-1 Design Report: Containment Fan Coil Unit Face Replacement, Specification M180 001 008 01 (Unit 2)

Aerofin CA

-529-1121 Design Report: Containment Fan Coil Unit Face Replacement, Specification M180 001 008 01 (Unit 2)

Aerofin CA

-529-1158 Design Report: Containment Fan Coil Unit Face Replacement, Specification M180 001 009 02 (Unit 1)

ENG-ME-242 Two Phase Flow Momentum Effects in FCU #24 (AES PI

-603866-M01, Revision 2)

ENG-CS-260 CL Waterhammer

Analysis, Penetration 37B to FCU 11 (AES PI-P-049, Revision 0)

ENG-CS-262 CL Waterhammer Analysis, Penetration 37C to FCU 12 (AES PI-P-051, Revision 0)

ENG-CS-263 CL Waterhammer Analysis, Penetration 37B to FCU 22 (AES PI-P-052, Revision 0)

PI-60 3866-P01 Fan Coil Units Nozzle Load Calculations, Unit 2

PI-603866-S01 Evaluation of Fan Coil Unit Enclosure, Unit 2 3 PI-605703-P01 Fan Coil Units Nozzle Load Calculations, Unit 1

PI-605703-M03 FCU Coil Side Pressure Drop, Unit 1

CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION

Number Description or Title

Date AR 0 1491847 CDBI 2015:

20,000 Gallon Admin Limit on CST Minimum Volume 0 9/01/15 AR 0 1491759 CDBI 2015:

ENG-ME-443 Inputs Revised without Calculation Update

08/31/15 AR 0 1491803 CDBI 2015:

TCOA 24 and 25 SEGs

9/01/15 AR 0 1491410 CDBI 2015:

CST Sizing Calculation

- "Hot Shutdown" vs "Hot Standby"

08/27/15 AR 0 1491302 CDBI 2015:

RHR Sump Recirculation Valves Duty Cycle in Question 08/26/15 AR 0 1491396 CDBI 2015:

US AR Historical Information Not Distinguished as Such 08/27/15 AR 0 149 1123 CDBI 2015:

D80 Wording on 50.59 Application Is Confusing 08/25/15 AR 0 149 1110 CDBI 2015:

D80 References Old EIR Process vs PIRE Process 08/25/15 AR 0 149 1096 CDBI 2015:

Potential Typo Errors

Found on H6.1

08/25/15 AR 0 1491104 CDBI 2015:

WO 358645 Task 01 Missing Data in Passport

08/25/15 AR 0 1490769 CDBI 2015:

Weakness in Design Specification for Installed FCUs 08/21/15

CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION

Number Description or Title

Date AR 0 1491000 CDBI 2015:

TOL Nameplate Data in Passport Parm Incorrect 08/25/15 AR 0 1490767 CDBI 2015:

High RAD Swing Gates Identified as Potential Undocumented Modification

08/21/15 AR 0 1490676 CDBI 2015:

Limiting Stroke Time Not Used in TOL Sizing Calculation

08/21/15 AR 0 1490614 CDBI 2015:

Q259 SP 11015.3 Unclear

08/20/15 AR 0 149 0560 CDBI 2015:

Inspector Provided Calculation Pages from Revision 1 Not Revision 1A

08/20/15 AR 0 149 0563 CDBI 2015:

PE 0009 Lacks Contact Resistance Acceptance Criteria

08/20/15 AR 0 149 0315 CDBI 2015:

SP 1101 Requirements for TRM 3.3.1.1 Unclear 08/19/15 AR 0 149 0267 CDBI 2015:

NRC Observation on Implementation of New Engineering Software

08/18/15 AR 0 149 0266 CDBI 2015:

New System Health Reporting Software Issue

08/18/15 AR 0 149 0194 CDBI 2015:

Enhancements Recommended for DC PNL Test Plan 08/18/15 AR 0 149 0262 CDBI 2015:

Provide NRC with Incorrect Information Q26

00/18/15 AR 0 14 89 492 CDBI 2015:

PNL 21 Temperature Limit Not Listed in VTM or Specification

08/12/15 AR 0 149 0130 CDBI 2015:

Scaffold Completion Comments Need to Be Clearer 08/18/15 AR 0 148 9459 CDBI 2015:

Insulating Blankets Installed on BKR 1H5 OCB

08/12/15 AR 0 148 9435 CDBI 2015:

IST Training Enhancement

08/12/15 AR 0 148 9439 CDBI 2015:

TS 3.7.6 Requires Further Bases/Clarification

08/12/15 AR 0 148 9360 CDBI 2015:

AFW Temperature at Low Flows May Exceed

100 F 08/11/15 AR 0 148 9273 CDBI 2015:

Out of Date Operational Aid on Cubical

08/11/15 AR 0 148 9319 CDBI 2015:

Drawing Titles on Some Drawings Are Non-Descript 08/11/15 AR 0 148 9196 CDBI 2015:

EC 21354 Didn't Account for CST

Heat Up d/t Recirc 08/10/15 AR 0 148 9070 CDBI 2015:

2M XFMR TM Power Light Out

8/0 7/15 AR 0 148 9179 CDBI 2015:

VTM Information for PNL 15 Not Found in XH

-286-14 08/10/15 AR 0 148 8919 CDBI 2015:

CT1/XFMR Housekeeping Issues

08/0 6/15 AR 01488920

CDBI 2015: 10/XFMR Housekeeping Issues

08/06/15 AR 01488918

CDBI 2015: CT11/XFMR Housekeeping Issues

08/06/15 AR 01488916

CDBI 2015: CT BATT Housekeeping Issues

08/06/15 AR 01488917

CDBI 2015: 1R/XFMR Housekeeping Issues

08/06/15 AR 01488912

CDBI 2015: Concern on Test Setup for 4kV Contact Resistance Read

08/06/15 AR 01488914

CDBI 2015: Procedure PE 0009 Improvement Suggestions

08/06/15 AR 01488850

CDBI 2015: Extra Data Coded as IST Acceptance Criteria

08/06/15 AR 0 1488700 CDBI 2015:

Error in H10.1 B for M

-32115 Basis Section

8/05/15 AR 0 1488794 CDBI 2015:

XH-1-44 Has Been Revised Twice without Incorporating Revision 77

08/06/15

CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION

Number Description or Title

Date AR 0 148 8647 CDBI 2015:

TCR Form Doesn't Match Procedure Changes for 1C20.6

08/05/15 AR 0 148 8653 CDBI 2015:

Cables in Field Do Not Show Up in Passport

8/05/15 AR 0 148 8519 CDBI 2015:

New Information in Regard to NOS CDBI FSA CAPS 0 8/0 4/15 AR 01488515

CDBI 2015: Error Found on Drawing NF

-40022-1 08/04/15 AR 01488518

CDBI 2015: Unidentified Blue Wire Run Along 11 CST Overflow 08/04/15 AR 01488513

CDBI 2015: Change Omitted for Supplemental Indication

08/04/15 AR 01488510

CDBI 2015: ET

-113N Local Control Box Has Unplugged Opening 08/04/15 AR 01488511

CDBI 2015: 16580 Cover Not Secured

08/04/15 AR 01488495

CDBI 2015: Instructional Label Left on Equipment Out of Process 08/04/15 AR 01488499

CDBI 2015: C

-27-1 Is Dripping a Small Amount of Water

08/04/15 AR 01488485

CDBI 2015: DBD SYS

-14 Reference Typo

08/04/15 CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION

Number Description or Title

Date AR 0 1362697 TOL Different for MV

-32078 Than Design

2/10/12 AR 0 1328026 OE Evaluation Regulatory Guide 1.106

Thermal Overload Protection for Electric Motors on Motor

-Operated Valves

2/04/12 AR 0 1399682 21 DC Panel Ground Detection

2/13/13 AR 0 1402767 21 DC Ground Present

2/22/13 AR 0 1448178-01 OE Item Evaluated:

NRC Information Notice 2014

-10: Potential Circuit Failure

-Induced Secondary Fires or Equipment Damage

2/08/14 AR 0 1468685 CDBI FSA Inconsistency in Seismic Qual. Tes

ting 06/30/15 AR 0 1467202 Indications of Leakage Past SI 9

-5 0 3/0 1/15 AR 0 1184681 SI 9-5 Undocumented Design Change

- Historical Issue

03/15/10 AR 0 1155124 Unauthorized Engineering Change in the Field

09/22/09 AR 0 1114840 RCP Loss of Seal Cooling Timing Concern

07/30/09 AR 0 1288291 Rags Found in Pump Bowl for 12RCP

6/0 2/11 AR 0 1303177 Single Point Vulnerability

- Reactor Coolant Pumps

2/0 9/11 AR 0 1459048 Insufficient 50.59 Support for PCR 1458690 for Unit 1 Forced Outage

1/0 2/15 AR 0 1459098 Forced Shutdown of Unit 1 Due to 12 RCP Seal Degradation

2/11/14 AR 0 1346088 OE Item Evaluated:

NRC Information Notice

2012-14: Motor-Operated Valve Inoperable Due to Stem

-Disk Separation

11/16/12 AR 0 1378410 Evaluate Component Cooling Water Pump Seal Type

04/11/13 AR 0 1459636 12 RCP Shaft Damaged by Replacement Seal Install/Replace

2/16/14 AR 0 1460093 12 RCP Sleeve Modification Performed without Proper Document 12/24/14 AR 0 1461540 Declining Performance of 12 RCP Seal

3/0 6/15 AR 0 1463641 Unit 1 Forced Outage Required due to 12 RCP Seal Leakage 05/18/15

CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION

Number Description or Title

Date AR 0 1472036 12 RCP Seal Shows Leakage

05/10/15 AR 0 1474523 Abnormal Pump Shaft "TIR" During Inspection

05/14/15 AR 0 1475138 12 RCP Flush Seal Cavity Identified 4 Portholes to

B e Flanged 04/22/15 AR 0 1476313 OPR 1475241

- 12 RCP FM

- Fleet/PORC Comments 07/10/15 AR 0 1476665 12 RCP Seal Stage 2 Not Holding DP

05/25/15 AR 0 1477308 As Found Condition of 12 RCP Seal O

-Ring 747-3 Torn 07/31/15 AR 0 1477362 As Found 12 RCP Cavity Inspect 5/2

5/0 2/15 AR 0 1477368 12 RCP Trend Data Shows Possible Vibration Issues

05/18/15 AR 0 1195148 Configuration Control Issue on MV

-032078 and XH

-1-1038 07/12/10 AR 0 1374524 GL 08-01, No Voiding Between Unit 1 Sump B MOV's

11/22/14 AR 0 1371137 OE: IER L3-13-04, Reactor Coolant Pump Motor Bearing Damage 0 5/07/13 AR 0 1431330 OE: IER L3-14-25, Heavy Snow and Design Flaw Result in Dual Unit Scram

11/03/14 AR 0 1355584 PI-11363 Reading May Be Inaccurate

10/28/12 AR 01483341

CDBI Review:

SR Motor Starts Not Evaluated at Degraded Voltage 06/18/15 AR 01483409

CDBI Review:

Survivability of Protective Devices

During DV3 06/19/15 AR 01392146

OE: Part 21 Deviation of HK Circuit Breaker Close Latch Spring 10/20/13 AR 01264091

Evaluate IER

L3-10-7 Feeder Breaker Catastrophic Failure

04/15/11 AR 01345866

TB-12-6, Rotary Style Auxiliary Switch for DHP

-VR, DH-VR 10/11/12 AR 01447385

OE: IER L2-14-46, Multiple Electrical Faults Result in Explosion of Unit Auxiliary Transformer and Automatic Scram 11/21/14 DRAWINGS Number Description or Title

Revision NF-40547-1 Wiring Diagram DC Distribution

Panels A Train

NE-40409-7 REV AF-125 VDC Panel 21

NE-4008-107 REV BV -Schematic Diagram 11 Containment Sump B ISOL. VA A

-2 MV 32076 (W 1

-8805B) 75 NE-4008-108 REV BV - Schematic Diagram Residual HT RMVL To 12 SAF INJ Pump VA MV 32207 (W 1

-8816B) 75 NF-40782-2 REV J-Interlock Logic Diagram Residual Heat Removal System Unit 1

NG-40244-3 External Connections Motor Control Center 1KA

NE-40409-8 DC Aux. & Emergency AC Schematic Diagram

NE-40006-61 Schematic Diagram

NE-40006-62 12 Residual Heat Removal Pump Bus 16 Cubicle 6

NE-40006-59 Schematic Diagram

NF-40301-1 Wiring Diagram DC Distribution Panels A Train

NE-40006-11 REV L-4160V Switchgear Safeguard Schematic Diagram

NE-40006-9 REV MP-4160V Switchgear Safeguard Schematic Diagram

NE-40006-8 REV AF-Schematic Diagram

XH-28-26 Three Line Diagram

NL-173001-1 Containment Sump B Elevation View

DRAWINGS Number Description or Title

Revision NF-39245-1 Flow Diagram Component Cooling System Unit 1

NF-39245-2 Flow Diagram Component Cooling System Unit 1 80 NF-39334-1 Residual Heat Removal Piping Unit 1

R XH-1-290 Residual Heat Removal Valve Enclosure Tank

XH-1001-315 Residual Heat Removal Valve Enclosure Tank

XH-106-13925 Valve Enclosure for MV

-32075 and MV

-32076 2 NL-173001-1 Containment Sump B Elevation View

MC-41 Containment

XH-1-44 Flow Diagram Safety Injection System Unit 1

XH-1-31 Flow Diagram Residual Heat Removal System Unit 1

XH-1-44 Flow Diagram Safety Injection System Unit 1

XH-1-45 Flow Diagram Safety Injection System Unit 1

XH-1-7 Flow Diagram Reactor Coolant System Unit 1

NF-39222 Flow Diagram Feedwater & Aux Feedwater, Unit 1

NE-40006 Sh. 8

Reserve Station Auxiliary Trans Cubicle 2 Bus 16

AF NE-40006 Sh. 59 12 Aux Feedwater

Pump Bus 16 Cub 3

NF-116752 Unit 1-Unit 2 Safeguards Consolidated Circuit Diagram

NF-239270 Prairie Island Plant and Substation Operating One Line Diagram 0 NF-40002-1 Single Line

-Metering and Relaying 1 &

Generator & Normal Incoming 4.16kV Switchgear Unit 1 and 2

77A-1 NF-40002-2 Single Line

-Metering and Relaying Safeguard and Normal

4160kV Switchgear Bus 15 &16 D1 and D2 Emergency Generator 77 NF-400 02-3 Single Line

-Metering and Relaying Safeguard and Normal

4160kV Switchgear Feeder Unit 1&

76 NF-40002-4 Single Line

-Metering and Relaying Bus 25 &

Main FDRS DSL D5 and D6

C NF-40002-5 Single Line-Metering and Relaying 4.16kV Safeguard Buses 25, 26 & 27 C NF-40021 480V Distribution Diagram Unit 1 Normal Busses

NF-40022-1 Circuit Diagram 4kV and 480V Safeguard Busses Unit 1

NF-40022-2 Circuit Diagram 4kV and 480V Safeguard Busses Unit 2

NF-40026 Motor Control Center 1A, 1AA, 1 Circuit Diagram

NF-40302-1 Wiring Diagram AC Distribution Panels 111,1111,113,1113,115,117 (A Train) 78 NF-40302-2 Wiring Diagram AC Distribution Panel 118 and 117

M NF-40302-3 Wiring Diagram AC Distribution Panels 111, 1111,113, 1113, 115, 117 (A Train)

NF-40416 480V Distribution Diagram Normal Busses Unit 2

NF-40528-1 Wiring Diagram AC Distribution Panels 211, 2111, 213, 2113, 215, 217 (A Train)

NF-40528-2 Wiring Diagram AC Distribution Panels 212, 2112, 214, 2114, 216 (B Train)

NF-40528-3 AC Distribution Panel 218 Wiring Diagram

XH-28-26 Three Line Diagram

10 CFR 50.59 DOCUMENTS (Screenings/Safety Evaluations

) Number Description or Title

Date FPE-EQV-01 Attachment 1

- Equivalency Evaluation and Change for

EC-23833 10 Screening 4454 Install Flo

wserve N-9000 Shaft Seal with Abeyance Seal on Unit 1 11 and 12 RCPs

Screening 4980 12 RCP Seal Face Replacement

Screening

3966 Scaffold Building

- G-Mod 19851 and D80 "Scaffolding, Ladders, and Cable Tray Platforms"

Screening 2364

DCP 05ZC02, Aerofin Design Report Calculation

CA-529-1121 Revision 1; Calculations PI

-603866-M01 Revision 2, PI

-6033866-M02 Revision 0, PI

-603866-M03 Revision 0, PI

-603866-P01 Revision 0, PI

-603866-S01 Revision 2, and PI

-603866-S04 Revision 0, ECN-05 1 Screening 2576

EC-410 (Modification 05ZC02), Calculations: CA

-529-1158 Revision 0 , PI-60 5703-S01 Revision

0, PI-605703-P 0 1 Revision 2, PI-605703-M0 1 Revision 0, PI

-605703-M 0 2 Revision 0, PI

-605703-M03 Revision 0 , 0 MISCELLANEOUS

Number Description or Title

Date or Rev

. DBD SYS-14 12 Component Cooling Water Pump Design

DBD SYS-15 Residual Heat Removal System

Vendor Manual

Westinghouse FDP Switches

01-69 Vendor Manual

Beckwith Electric

-Battery Transient Suppressor

2-69 Vendor Manual

Ametek 400A Battery Charger

A IN 2014-10 Potential Circuit Failure

-Induced Secondary Fires or Equipment Damage

09/16/14 RG 1.106 Thermal Overload Protection for Electric Motors on Motor

-Operated Valves

B15 Residual Heat Removal System

B18A Safety Injection System

DBD System 15

Design Basis Document for the Residual Heat Removal System 7 DBD System 18A

Design Basis Document for the Safety

Injection System

USAR Prairie Island Units 1 and 2

34P Plant System Health Report

- U1 Component Cooling

07/24/15 MV 32076 Stroke Time Trends

Various Dates

MV 32078 Stroke Time Trends

Various Dates

H 10.1 ASME Inservice Testing Program

H 5.1 Motor Operated Valve Weaklink Design Criteria

12 RCP Seal Leakoff, Bearing Temps, Seal Leakoff Temps, Seal Differential Pressure, Trends

Various Dates

B3 Reactor Coolant Pumps

DBS SYS 04

Design Basis Document for the Reactor Coolant System

Plant System Health Report

- RC Reactor Coolant

01/20/15 12 CC Pump Trends PINP

-1-145-122 Various Dates

IN 2012-14 Motor-Operated Valve Inoperable Due to Stem

-Disk Separation

07/24/12 SER Safety Evaluation Related to A

-46 Program (GL 87

-02) Implementation

8/05/98

MISCELLANEOUS

Number Description or Title

Date or Rev

. SER Safety Evaluation

- Generic Letter 81

-21, Natural Circulation Cooldown

07/18/83 2015 Period 1

System Report

- Unit 1 AFW System

04/20/15 IB-8.2.7-1 ITE Metal-Clad Switchgear

Vender M anual 1 GEI-44233B Time Overcurrent Relays

Type IAC66K

11-58 GL 96-06, Supplement 1

Assurance of Equipment Operability and containment Integrity During Design

-Basis Accident Conditions

11/13/97 GL 96-06 Assurance of Equipment Operability and containment Integrity During Design

-Basis Accident Conditions 09/30/96 TS-M605 Pioneer Service & Engineering: Technical Specification for Containment Fan Coil Units

2/15/70 M180-0001-008 Design Specification for CFCU Face Replacement, Unit

2 Item #: 2-22-066 Repair/Replacement Plan:

Fan Coil Unit, 174-011 0 5/02/05 Item #: 2-22-067 Repair/Replacement Plan: 22 Fan Coil Unit, 174

-012 05/02/05 Item #: 2-22-068 Repair/Replacement Plan: 23 Fan Coil Unit, 174

-013 05/02/05 Item #: 2-22-069 Repair/Replacement Plan: 24 Fan Coil Unit, 174

-014 05/02/05 MODIFICATIONS

Number Description or Title

Date or Rev

. EC 23725 DC Panel 21 Fused Switch Replacement

EC 18153 Replacement of Battery Chargers 21 and 22 and Portable Battery Charger 11P

EC 17202 Resolve Battery Charger Issue Per TS

Requirements

EC 21790 Reactor Coolant Pump (RCP) Seal Replacements Unit 1

EC 25405 Prairie Island 12 RCP Seal Face Replacement

EC-23833 12 CCP Seal Face Material Equivalency Evaluation

EC-21978 M395-0002-001, Rev. 0 Unit 1 RCP Replacement Seal N

-9000, Specification

EC 19851 Scaffold Building Generic Modification

05ZC05 Unit 1 Fan Coil Unit Cooling Faces Replacement

05ZC02 Unit 2 Fan Coil Unit Cooling Faces Replacement

OPERABILITY EVALUATIONS

Number Description or Title

Date AR 01491302 Prompt Operability Determination

-MV-32075 , 32076, 32077, 32078, MV

-32178, 32179, 32180, and MV

-32181 08/31/15 AR 01491302 Prompt Operability Determination

-MV-32075 , 32076, 32077, 32078, MV

-32178, 32179, 32180, and MV

-32181 0 9/03/15 AR 0 1475241 FM E Found Below 12 RCP Lower Radial Bearing

04/26/15 AR 0 1462566 12 CC Pump IB Seal Shows Signs of Leaking

01/20/15 AR 0 1462562 11 CC PMP OB Seal Leaking 1

Drop/20 Secs. 01/20/15 AR 0 1366642 Safety Concern:

CC Pump OB Seal Spraying

2/0 7/13 AR 0 1350253 21 CC Pump Inboard Seal Leak Getting Worse

9/0 6/12 AR 0 1466999 Voiding in RHR Piping

- Aggregate Review

2/22/12 SEWS GIP Rev 2, Bus 122, 48

0V Switchgear

2/14/92 PROCEDURES

Number Description or Title

Rev. PE-MCC-G7 Electrical Preventive Maintenance for GE 7700 Line MCCS

PROCEDURES

Number Description or Title

Rev. FP-PA-OE-01 Operating Experience Program

PE-0542 21 Battery Ground Detection System

1C 19.1 Containment Unit 1

1C 1.2-M5 Unit 1 Startup to Mode 5

SP-1070 Reactor Coolant System Integrity Test

SP-1092D Safety Injection Check Valve Test (Head On) Part D: Low Head SI Discharge Flow Path Verification

SP-1137 Recirculation Mode Valve Functional Test

SP-1155b CC System Quarterly Test Train B

1C1.2-M3 Unit 1 Startup to Mode 3

1C1.2-M5 Unit 1 Startup to Mode 5

1C3 AOP2 Loss of RCP Seal Cooling

1C3 AOP3 Failure of a Reactor Coolant Pump Seal

1C18 AOP2 Inadvertent

Seal Injection when Shutdown

C3 Reactor Coolant Pump

FP-PA -ARP-01 Attachment 4

- Severity and Evaluation Level Determination Matrix

FP-PA -ARP-01 CAP Evaluation Request for CAP 1463641

D-80 Scaffolding, Ladders and Cable Trays Platforms

SP 1101 12 Motor-Driven Auxiliary Feedwater Pump Quarterly Flow and Valve Test

FP-OP-OL-01 Operability/Functionality Determination

SWI O-35 Emergency Operating Procedure Verification, Validation & Maintenance

WM-0106 Panel Reports Electrical Loads from DC Nuclear Distribution Panels 12, 15, 21, and 22.

1C4-AOP1 Reactor Coolant Leak

SP 1101 12 Motor-Driven Auxiliary Feedwater Pump Quarterly Flow and Valve Test

SP 1101 12 Motor-Driven Auxiliary Feedwater Pump Quarterly Flow and Valve Test

SP 1155B CC System Quarterly Test Train B

1E-0 Reactor Trip or Safety Injection

1ES-0.3A Natural Circulation Cooldown with CRDM Fans

1ES-0.3B Natural Circulation Cooldown without CRDM Fans

1ES-0.4 Natural Circulation Cooldown with Steam Void in Vessel

H10.1 ASME In-service Testing Program

G1 Surveillance and Periodic Test Program

1C28.3 Unit 1 Condensate System

C28.6 Condensate Storage Tank Freeze Protection System

1C28.1 AOP2

Loss of Condensate Supply to AFWP Suction

H14 Procedure Writer's Manual

H14.4 Surveillance &

Periodic Test Procedure Guideline

H14.5 Maintenance Procedure Guideline

H6.1 Motor Operated Valve Thermal Overload Heater Sizing for GE MCC's 5 PE 0009 4KV Switchgear Preventive Maintenance

PEMCC-G7 MCC Electrical Preventive Maintenance for

GE 7700 Line MCCS 34

SURVEILLANCE

S (Completed)

Number Description or Title

Date or Rev

. SP 1219 Monthly 4kV Bus 16 Under

-voltage Relay Test

07/28/15 SP 1217 4k V Bus 16 Under

-voltage Relay Calibration

11/13/14 SP 1137 Recirculation Mode Valve Functional Test

WO 512240 CC System Quarterly Test Train B

- SP 1155B 0 7/0 6/15 WO 436712 01

SP 1070 Reactor Coolant System Integrity Test

11/0 6/14 WO 461954 SP 1092 D SI Check Valve Test RVSL

05/16/14 SP 1329 12 MDAFW Bearing Temperature Test 10/29/14 WO 454528 SP 1101 Comprehensive IST 11 MDAFW Pump Test 12/21/12 WO 443690 SP 2090A 21 CS Pump Quarterly Test

11/03/12 WO 500192 SP 1101 12 MDAFWP Quarterly Flow and Valve Test 11/25/14 WO 510216 SP 1101 12 MDAFWP Quarterly Flow and Valve Test 05/28/15 TRAINING DOCUMENTS

Number Description or Title

Rev. JPM P9160S

-05 Secure Turbine Roof Exhaust Fans

Following a LOCA

PI-OPS-TCOA-24 Complete SI Pump Recirculation Switchover

PI-OPS-TCOA-25 Trip RCP's Following a Small Break LOCA

WORK DOCUMENTS

Number Description or Title

Date 515803 MECH: Inspect 12 RCP Seal

01/22/15 500315-48 Bump Test 12 RCP Seal Piping EC -21790 0 8/0 7/14 500314-45 Bump Test 11 RCP Seal Piping EC -21790 0 8/0 7/14 PINGP 1198

Scaffold or Enclosure Construction Checklist Elevation 695 Containment Spray Room Unit 1 Scaffold 2568 R

11/15/14 407938 Replace SI 9

-5 Internals

10/0 9/04 D80 Att. F

Scaffolding, Ladders and Cable Trays Platforms

- WO 519170 - U1 CPR 695 AUX

1/14/15 WR 84426 PI-11363 May Be Reading Wrong 10/18/12 WO 466767 21 CSP Suction Gauge PI-11363 May Be Reading Wrong 02/15/13 WO 467389 Swap 21 CSP S

uction Gauge with 11 CSP Suction Gauge 11/02/12 WO 395828 01

BRK 111E-17, PE-MCC-G7, 11AFW to 11SG on

MV-32238 10/19/12 WO 282635 04

Bus 16 Relay Calibration:

PE 0016-01C Thru 12C and PE

0016-SYC 12/06/12

LIST OF ACRONYMS USED

ADAMS Agencywide Document Access and Management System

AFW Auxiliary Feedwater

AR Action Request

ASME American Society of Mechanical Engineers

CAP Corrective Action Program

CC Component Cooling

CFCU Containment Fan Coil Unit

CFR Code of Federal Regulations

CST Condensate Storage Tank

DBD Design Basis Document

DRP Division of Reactor Projects

DRS Division of Reactor Safety

EC Engineering Change

EPRI Electric Power Research Institute

ETAP Electrical Transient Analysis Program

FCU Fan Coil Unit

GE General Electric

GL Generic Letter

IER Institute of Nuclear Power Operations (INPO) Event Report

IMC Inspection Manual Chapter

IN Information Notice

IST In-service Testing

JPM Job Performance Measure

kV Kilo-Volt (1000 Volts)

LERF Large Early Release Frequency

LOCA Loss of Coolant Accident

MCC Motor Control Center

MDAFW Motor-Driven Auxili

ary Feedwater

MOV Motor-Operated Valve

NCV Non-Cited Violation

NRC U.S. Nuclear Regulatory Commission

PARS Publicly Available Records System

PM Preventive Maintenance

PRA Probabilistic Risk Assessment

RCP Reactor Coolant Pump

RHR Residual Heat Removal

SDP Significance Determination Process

TCOA Time Critical Operator Action TOL Thermal Overload

TS Technical Specification

U SAR Updated Safety Analysis Report

V ac Volt s Alternating Current

V dc Volts Direct Current

WO Work Order

K. Davison

-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 NRC's 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 Nos. 50

-282, 50-306 License Nos.

DPR-42, DPR-60 Enclosure:

IR 05000282/2015007; 05000306/2015007

cc w/encl: Distribution via LISTSERV

DISTRIBUTION

w/encl: Janelle Jessie

RidsNrrPMPrairieIsland Resource

RidsNrrDorlLpl3

-1 Resource

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 ML15288A195

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 JNeurauter:cl CLipa: DATE 10/14/15 10/14/15 OFFICIAL RECORD COPY