ML100060183

IR 05000263-09-007; on 08/31/2009 - 12/04/2009; Monticello Nuclear Generating Plant; Component Design Bases Inspection (CDBI) and Power Uprate

ML100060183
Person / Time
Site:	Monticello
Issue date:	01/06/2010
From:	Ann Marie Stone NRC/RGN-III/DRS/EB2
To:	O'Connor T Northern States Power Co
References
IR-09-007
Download: ML100060183 (48)

See also: IR 05000263/2009007

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE ROAD, SUITE 210

LISLE, IL 60532-4352

January 6, 2010

Mr. Timothy J. OConnor

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 (CDBI) INSPECTION REPORT

05000263/2009007(DRS)

Dear Mr. OConnor:

On December 4, 2009, 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 October 2, 2009,

and on December 4, 2009, with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your

license. The inspectors reviewed selected procedures and records, observed activities, and

interviewed personnel.

Based on the results of this inspection, five NRC-identified findings of very low safety-

significance were identified. The findings involved a violation 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 VI.A.1 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 a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -

Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of

Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the

Resident Inspector Office at the Monticello Nuclear Generating Plant. In addition, if you

disagree with the characterization of 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 Monticello Nuclear

Generating Plant. The information that you provide will be considered in accordance with

Inspection Manual Chapter 0305.

T. OConnor -2-

In accordance with 10 CFR 2.390 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 Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's Agencywide Documents Access and Management System (ADAMS),

accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public

Electronic Reading Room).

Sincerely,

/RA/

Ann Marie Stone, Chief

Engineering Branch 2

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure: Inspection Report 05000263/2009007

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: 05000263/2009007(DRS)

Licensee: Northern States Power Company, Minnesota

Facility: Monticello Nuclear Generating Plant

Location: Monticello, MN

Dates: August 31 through December 4, 2009

Inspectors: A. Dunlop, Senior Reactor Engineer, Lead

N. Féliz Adorno, Reactor Engineer, Mechanical

C. Brown, Reactor Engineer, Operations

J. Bozga, Reactor Engineer, Mechanical

M. Munir, Reactor Engineer, Electrical

S. Kobylarz, Electrical Contractor

H. Campbell, Mechanical Contractor

Observers: L. Jones, Reactor Engineer

S. Edmonds, Reactor Engineer

Approved by: Ann Marie Stone, Chief

Engineering Branch 2

Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000263/2009007; 08/31/2009 - 12/04/2009; Monticello Nuclear Generating Plant;

Component Design Bases Inspection (CDBI) and Power Uprate

The inspection was a 3-week onsite baseline inspection that focused on the design of

components that are risk-significant and have low design margin; and power uprate. The

inspection was conducted by regional engineering inspectors and two consultants. Five Green

findings were identified by the inspectors. The findings were considered Non-Cited Violations

(NCVs) of NRC regulations. The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination

Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a

severity level after NRC management review. The NRCs program for overseeing the safe

operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process, Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealed Findings

Cornerstone: Barrier Integrity

• Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,

Design Control, having very low safety-significance for the failure to incorporate the

actual physical configuration of the inboard main steam isolation valves (MSIVs) and the

correct pneumatic system pressure drop into the pneumatic pressure requirement

calculation for the inboard MSIVs. Specifically, the licensee failed to adjust the actuator

moving part weight to reflect that the actuator was offset by 45 degrees instead of being

mounted vertically and to correctly compute the system pressure drop. This finding was

entered into the licensees corrective action program and a preliminary calculation

performed by the licensee concluded that the valves were operable.

The finding was more than minor because it was associated with the Barrier Integrity

cornerstone attribute of structures, systems, components, and barrier performance, and

affected the cornerstone objective of providing reasonable assurance that physical

design barriers protect the public from radionuclide releases caused by accidents or

events. This finding is of very low safety-significance (Green) because there was no

actual barrier degradation. The inspectors did not identify a cross-cutting aspect

associated with this finding because this was a legacy design issue and therefore was

not reflective of current performance. (Section 1R21.3.b.(1))

Cornerstone: Mitigating Systems

• Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,

Design Control, having very low safety-significance for the failure to restore the

emergency service water (ESW) piping supports to their design specifications.

Specifically, although the licensee identified the existence of gaps between the ESW

piping supports and the baseplates, the licensee failed to recognize that this condition

did not meet seismic Category 1 design basis requirements. As a result, corrective

actions were not implemented. The licensee entered this issue into its corrective action

program and restored the supports to their design specifications.

1 Enclosure

The finding was more than minor because it was associated with the Mitigating Systems

cornerstone attribute of protection against external events and affected the cornerstone

objective of ensuring the availability of the ESW system, and ultimately the emergency

diesel generators (EDGs), to respond to initiating events to prevent undesirable

consequences. This finding is of very low safety-significance (Green) because the

design deficiency was confirmed not to result in loss of operability or functionality. This

finding has a cross-cutting aspect in the area of problem identification and resolution

because the licensee did not properly prioritize and evaluate an identified problem.

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

• Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,

Design Control, having very low safety-significance for the failure to adequately

evaluate circuit loads in determining design limits in electrical calculations. Specifically,

three examples were identified where the licensee: (1) failed to perform a calculation for

safety-related motor starters that included all control circuit loads in determining the

minimum voltage available at 120Vac starter coils, which was used to establish the coil

voltage test acceptance criteria; (2) failed to include thermal overload heater and starter

contact resistance when calculating the minimum voltage at 480Vac motor terminals;

and (3) failed to assure that the minimum voltage at the 120Vac solenoid operated

control valves was in conformance with vendor requirements. These issues were

entered into the licensees corrective action program to re-evaluate the voltage available,

and to test coils, as required, to verify the pick-up voltage.

The finding 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 safety-related equipment to respond to

initiating events to prevent undesirable consequences. This finding is of very low safety-

significance (Green) because the design deficiency was confirmed, with the exception of

ESW pump P-111B, not to result in loss of operability or functionality. Specifically, the

failure to assure adequate voltage was available at the solenoid valves coils; and to

perform periodic testing to assure the minimum voltage remained acceptable as the

components aged, did not result in an impact on current operability. With respect to the

ESW pump, it was determined that the pump would not have started under degraded

voltage condition as required such that the ESW pump was considered inoperable.

Based on a Phase III analysis, the failure of the pump to start under degraded voltage

conditions was determined to be very low safety-significance (Green). The inspectors

did not identify a cross-cutting aspect associated with this finding because this was a

legacy design issue, therefore was not reflective of current performance. (Section

1R21.3.b.(3))

• Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI,

ATest Control,@ having very low safety-significance for the failure to have adequate

testing for safety-related equipment to monitor component degradation. Specifically, the

licensee failed to verify that the motor control center contactors would continue to pick-

up under degraded voltage conditions with less than the vendors required minimum

voltage. These issues were entered into the licensees corrective action program to test

the 13 contactors as soon as practicable and to revise the maintenance procedures to

incorporate the requirements for periodic testing of contactors.

2 Enclosure

The finding 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 safety-related equipment to respond to

initiating events to prevent undesirable consequences. This finding is of very low

safety-significance (Green) because the testing deficiency was confirmed, with the

exception of ESW pump P-111B, not to result in loss of operability or functionality.

Specifically, subsequent testing confirmed for nine contactors that the safety-related

starter coils would still function at the calculated degraded voltage values. Although

three of the contactors have not been tested, they were of a different size than the failed

contactor and there appeared to be reasonable assurance based on the successful tests

that these contactors also remained operable. With respect to the ESW pump, the failed

test confirmed that the motor starter contactor would not pickup under degraded voltage

conditions due to mechanical binding of the contactor arm such that the ESW pump was

considered inoperable. Based on a Phase III analysis, the failure of the pump to start

under degraded voltage conditions was determined to be very low safety-significance

(Green). The inspectors did not identify a cross-cutting aspect associated with this

finding because this was a legacy design issue and therefore was not reflective of

current performance. (Section 1R21.3.b.(4))

Cornerstone: Mitigating Systems and Barrier Integrity

• Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,

Design Control, having very low safety-significance for the failure of two pipe supports

to meet their design requirements. Specifically, the calculation for pipe support SR-526

failed to use the minimum yield strength in determination of the allowable bending stress

of the pipe support baseplate as required in the American Institute of Steel Construction

code. In addition, the calculation for pipe support PS-16 failed to use the design basis

concrete compressive strength in determination of the anchor bolt allowable as required

in the licensees design specification. This finding was entered into the licensees

corrective action program and a preliminary analysis performed by the licensee

concluded that the pipe supports were operable but nonconforming.

The performance deficiency for pipe support SR-526 example 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 the safety-related residual heat removal and core spray pumps. This finding

is of very low safety-significance (Green) because the design deficiency was confirmed

not to result in loss of operability or functionality. The performance deficiency for pipe

support PS-16 example was more than minor because it was associated with the Barrier

Integrity cornerstone attribute of design control and affected the cornerstone objective of

providing reasonable assurance that physical design barriers protect the public from

radionuclide releases caused by accidents or events. This finding is of very low

safety-significance (Green) because there was no actual barrier degradation. The

inspectors did not identify a cross-cutting aspect associated with this finding because

this was a legacy design issue and therefore was not reflective of current performance.

(Section 4OA5.1.b.(1))

B. Licensee-Identified Violations

No violations of significance were identified.

3 Enclosure

REPORT DETAILS

1. REACTOR SAFETY

Cornerstone: 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 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 the

report.

.2 Inspection Sample Selection Process

The inspectors selected risk-significant components and operator actions for review

using information contained in the licensees PRA and the Monticello Standardized Plant

Analysis Risk (SPAR) Model, Revision 3P. In general, the selection was based upon the

components and operator actions having a risk achievement worth of greater than 1.3

and/or a risk reduction worth greater than 1.005. The operator actions selected for

review included actions taken by operators both inside and outside of the control room

during postulated accident scenarios. In addition, the inspectors selected operating

experience issues associated with the selected components.

The inspectors performed a margin assessment and detailed review of the selected risk-

significant components to verify that the design bases have been correctly implemented

and maintained. This design margin assessment considered original design reductions

caused by design modification, or 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 action, 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.

This inspection constituted 30 samples as defined in Inspection Procedure 71111.21-05.

4 Enclosure

.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 (ASME) 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 (GLs),

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,

system health reports, operating experience-related information, and licensee corrective

action program documents. Field walkdowns were conducted for all accessible

components to assess material condition 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 20 components were reviewed:

• Emergency Diesel Generator (EDG) (DG 11): The inspectors reviewed the EDG

loading calculation and vendor ratings for conformance with design basis load

requirements. The inspectors also reviewed EDG vendor de-rating requirements

for potential impact on design basis loading and operating procedures to

determine whether de-rating requirements were incorporated appropriately. In

addition, the inspectors reviewed surveillance testing to determine whether

design basis load requirements were demonstrated during periodic load testing

to satisfy TS.

• Diesel Fuel Oil Transfer and Service Pumps (P-11 and P-77): The inspectors

reviewed the system hydraulic calculations including net positive suction head

(NPSH) and vortexing to ensure that the pumps were capable of providing

sufficient flow such that the day tanks remained filled during diesel operation.

The inspection also included a review of operating procedures related to these

functions. The inspectors reviewed vendor specifications and pump curves to

make sure that these parameters had been correctly translated into calculations,

as required. In addition, design change history, corrective actions, surveillance

results, and trending data were reviewed to assess potential component

degradation and impact on design margins.

5 Enclosure

• EDG 11 Room Louvers/Fan (V-SF-10): The inspectors reviewed calculations

and operating procedures to ensure that the ventilation system was capable of

maintaining room temperatures within design limits. The review included the

control logic associated with the ventilation system to ensure it would function as

designed. Seismic Category I qualification of ventilation ductwork supports in

EDG building were also reviewed. The inspectors reviewed vendor

specifications to verify these parameters had been correctly translated into

calculations, as required. In addition, design change history, corrective actions,

surveillance results, and trending data were reviewed to assess potential

component degradation and impact on design margins.

• Emergency Service Water (ESW) Pump (P-111A): The inspectors reviewed the

system hydraulic calculations including NPSH, vortexing, and waterhammer to

ensure the pump was capable of providing sufficient flow under accident

conditions. The review also included the control logic and bay water level

setpoint associated with the pump to ensure it would function as designed.

Operating procedures related to the pump were also reviewed. The inspectors

reviewed vendor specifications and pump curves to verify these parameters had

been correctly translated into calculations, as required. In addition, design

change history, corrective actions, surveillance results, and trending data were

reviewed to assess potential component degradation and impact on design

margins. The inspectors reviewed the operability evaluation for the piping and

pipe support analysis for ESW 1B and ESW 2B to verify there was a reasonable

assurance of operability. The motor sizing and pump brake horsepower

requirements and vendor ratings were reviewed for conformance with design

basis load conditions. The inspectors also reviewed design calculations to

determine the adequacy of voltage at motor terminals and motor starter

contactors during degraded voltage conditions and the adequacy of motor feeder

cable sizing. The motor and feeder cable coordination calculation was reviewed

to determine the adequacy of protection and coordination.

• Main Steam Isolation Valve (MSIV) (AO-2-86D): The inspectors reviewed thrust

calculations associated with actuator thrust and pneumatic supply, and

environmental qualifications of the valve actuator to ensure the valve was

capable of functioning under design conditions. In addition, the inspectors

reviewed completed surveillances to ensure actual performance was acceptable

and vendor specifications to ensure parameters had been correctly translated

into calculations. Procedures related to the MSIVs were also reviewed. Design

change history, corrective actions, surveillance results, and trending data were

reviewed to assess potential component degradation, impact on design margin.

• Residual Heat Removal (RHR) Pumps (P-202A/C): The inspectors reviewed the

pumps capability to meet design basis assumptions with respect to pump flow

and pressure. Calculations, drawings, procedures, tests, and other analyses

were reviewed to verify selected calculation inputs, requirements, and

methodologies were accurate and justified, and were consistently applied. The

inspectors reviewed completed surveillance tests to confirm the acceptance

criteria and test results demonstrated the capability of the pump to provide

required flow rates. Inservice Test (IST) results were reviewed to assess

potential component degradation and impact on design margins. In addition, the

licensee responses and actions to NRC Bulletin 88-04, Potential Safety-Related

6 Enclosure

Pump Loss, were reviewed to assess implementation of operating experience.

The calculation which developed the extent of containment overpressure

required to satisfy NPSH was also reviewed, in addition to emergency operating

procedures that directly incorporated this information. The inspectors reviewed

motor sizing and pump brake horsepower requirements and vendor ratings for

conformance with design basis load conditions. The inspectors also reviewed

design calculations to determine the adequacy of voltage at motor terminals

during degraded voltage conditions and the adequacy of feeder cable sizing.

The motor and feeder cable coordination calculation was reviewed to determine

the adequacy of protection and coordination.

• RHR Pump Room Cooler (V-AC-4): The inspectors reviewed design calculations

associated with room heat loads and cooling to ensure the pump and associated

components remained within their temperature limits. Surveillance test results,

including verification of adequate air and water flows were also reviewed to

assess the capability of the room cooler to maintain temperature within

prescribed limits. In addition, two operability evaluations concerning room cooler

water flow test results and associated flow instrumentation were reviewed,

including the recently installed high accuracy flow instrumentation.

• RHR Pump Miniflow Valve (CV-1994): Setpoint and scaling calculations were

reviewed to ensure that the vendor requirements per NRC Bulletin 88-04 for

minimum flow conditions were established. Maximum expected differential

pressure calculations, coupled with valve actuator performance requirements

were also reviewed to ensure the valve was capable of functioning under design

conditions. The inspectors reviewed the electrical schematic diagrams and

investigated the adequacy of voltage at the solenoids under design basis

accident conditions. The inspectors verified the adequacy of the feeder circuits

including the circuit breakers and its settings and cables. Finally, surveillance

test results also were reviewed to assess potential component degradation and

impact on design margins.

• RHR B Loop Discharge to Torus Isolation Valve (MO-2007): The inspectors

reviewed motor-operated valve (MOV) calculations and analysis to ensure the

valve was capable of functioning under design conditions. These included

calculations for required thrust, maximum differential pressure, pressure locking,

seismic Category I qualification of pipe support SR-852, and valve weak link

analysis. The inspectors reviewed the electrical schematic diagrams and the

degraded voltage calculations for both the power and control circuits of the MOV.

The inspectors also verified the adequacy of the feeder circuits including the

circuit breakers and its settings, the power cables, and the thermal overload

relays. Diagnostic testing and IST surveillance results, including stroke time

testing, were reviewed to verify acceptance criteria were met and performance

degradation could be identified.

• RHR B Loop Low Pressure Coolant Injection Outboard Injection Valve

(MO-2013): The inspectors reviewed MOV calculations and analysis to ensure

the valve was capable of functioning under design conditions. These included

calculations for required thrust, maximum differential pressure, seismic

Category I qualification of piping and pipe supports, and valve weak link analysis.

The inspectors reviewed the electrical schematic diagrams and the degraded

7 Enclosure

voltage calculations for both the power and control circuits of the MOV. The

inspectors also verified the adequacy of the feeder circuits including the circuit

breakers and its settings, the power cables, and the thermal overload relays.

Diagnostic testing and IST surveillance results, including stroke time and leak

rate testing, were reviewed to verify acceptance criteria were met and

performance degradation could be identified.

• RHR Containment Spray Outboard Isolation Valve (MO-2020): The inspectors

reviewed MOV calculations and analysis to ensure the valve was capable of

functioning under design conditions. These included calculations for required

thrust, maximum differential pressure, seismic Category I qualification of pipe

support SR-680, and valve weak link analysis. The inspectors reviewed the

electrical schematic diagrams and the degraded voltage calculations for both the

power and control circuits of the MOV. The inspectors also verified the adequacy

of the feeder circuits including the circuit breakers and its settings, the power

cables, and the thermal overload relays. Diagnostic testing and IST surveillance

results, including stroke time and leak rate testing, were reviewed to verify

acceptance criteria were met and performance degradation could be identified.

In addition, a modification to provide operations the ability to throttle the valve

was reviewed.

• Residual Heat Removal Service Water (RHRSW) Pumps (P-109A/C): The

inspectors reviewed system hydraulic calculations including those addressing

NPSH and vortex concerns related to the intake structure and associated

required water levels. Further, calculations and the adequacy of the differential

pressure setpoint across the RHR heat exchanger were reviewed to ensure the

service water side was at a higher pressure than the RHR side. The inspectors

reviewed pump installation and manufacturing details, vendor manuals,

specifications, and pump curves to make sure that these parameters had been

correctly translated into calculations, as required. Motor sizing and pump brake

horsepower requirements and vendor ratings were reviewed for conformance

with design basis load conditions. The inspectors also reviewed design

calculations to determine the adequacy of voltage at motor terminals during

degraded voltage conditions and the adequacy of feeder cable sizing. The motor

and feeder cable coordination calculation was reviewed to determine the

adequacy of protection and coordination. In addition, the inspectors reviewed

completed pump surveillances to ensure that actual performance was acceptable

and no significant degradation was taking place.

• RHRSW Flow Control Valve (CV-1729): The inspectors reviewed maximum

expected differential pressures calculations, coupled with valve actuator

performance requirements to ensure the valve was capable of functioning under

design conditions. The calculation addressing the control feature of this valve, to

ensure a positive differential pressure of service water to RHR was also

reviewed. The inspectors reviewed the electrical schematic diagrams and

investigated the adequacy of voltage at the solenoids under design basis

accident conditions. The inspectors also verified the adequacy of the feeder

circuits including the circuit breakers and its settings and cables. Surveillance

and diagnostic test results were reviewed to ensure that actual performance was

acceptable and no significant degradation was taking place.

8 Enclosure

• Reserve Auxiliary Transformer (1AR): The inspectors reviewed one-line

diagrams and vendor test results for impedance data to confirm that correct

transformer impedances were utilized in design analyses. The inspectors

confirmed the adequacy of the overcurrent relay setting calculation for design

basis loading and protective relay setting requirements. Surveillance testing for

the overcurrent relays was reviewed for issues that affect reliability and for

conformance with relay setting cards. The inspectors reviewed the modification

that installed the transformer for potential impact on the design basis.

• 345-4.16kV Low Voltage Auxiliary Transformer (2R): The inspectors reviewed

the design basis descriptions, equipment specifications, drawings, equipment

nameplate data, voltage drop calculations, and short circuit and load flow

calculations to evaluate the capability of transformer 2R to supply the voltage and

current requirements to station safeguard loads. Protective relay trip setting

calculations were reviewed to verify whether adequate protection coordination

margins were provided. The relay settings review included the transformer

overall differential currents and ground overcurrent relays. The inspectors

reviewed the results of completed transformer preventive maintenance and relay

calibrations to verify that the test results were satisfactory.

• 4.16 kV Switchgear (Bus 15): The inspectors reviewed load flow and short circuit

current calculations to determine the design basis for maximum load, interrupting

duty and bus bracing requirements, and the switchgear equipment vendor ratings

for conformance with design basis. The coordination/protection calculation for

the incoming line and feeder breaker relay settings was reviewed for design

basis loading and protective relay setting requirements. The inspectors reviewed

surveillance testing for the overcurrent relays for issues that affect reliability and

for conformance with relay setting cards. The inspectors also reviewed selected

breaker preventive maintenance for any recurring issues that affect reliability.

• 480V Load Center 103 (LC 103): The inspectors reviewed load flow and short

circuit current calculations to determine the design basis for maximum load,

interrupting duty and bus bracing requirements and the load center equipment

vendor ratings for conformance with design basis. The coordination/protection

calculation for the incoming line and feeder breaker relay settings was reviewed

for design basis loading and protective relay setting requirements. The

inspectors reviewed surveillance and preventive maintenance testing for the

breakers for issues that affect reliability and for conformance with protective

device trip requirements.

• 480V Motor Control Center (MCC 103A): The inspectors reviewed load flow and

short circuit current calculations to determine the design basis for maximum load,

interrupting duty and bus bracing requirements and the motor control center

equipment vendor ratings for conformance with design basis. The

coordination/protection calculation for the incoming line and feeder breaker relay

settings was reviewed for design basis loading and protective relay setting

requirements. The inspectors reviewed bus and motor starter surveillance and

preventive maintenance testing for issues that affect reliability.

9 Enclosure

• 125Vdc Station Battery/Battery Charger (D11/D10): The inspectors reviewed

calculations and analyses relating to battery sizing and capacity, hydrogen

generation, station blackout (SBO) coping, and battery room transient

temperature. The review was performed to ascertain the adequacy and

appropriateness of design assumptions, and to verify that the battery was

adequately sized to support the design basis required voltage requirements of

the 125Vdc safety-related loads under both design basis accident and SBO

conditions. The inspectors reviewed calculations relating to sizing and current

limit setting to ascertain the adequacy and appropriateness of design

assumptions, and to verify that the charger was adequately sized to support the

design basis duty cycle requirements of the 125Vdc safety-related loads and the

associated battery under both normal and design basis accident conditions. The

inspectors also reviewed a sampling of completed surveillance tests, service

tests, performance discharge tests, and modified performance tests. The review

of various discharge tests was to verify that the battery capacity was adequate to

support the design basis duty cycle requirements and to verify that the battery

capacity meets TS requirements. In addition, the test procedures were reviewed

to determine whether maintenance and testing activities for the battery charger

were in accordance with vendor=s recommendations. The inspectors also

witnessed a weekly battery surveillance test. Seismic Category I qualification of

the battery charger D10 anchorage was also reviewed.

• 125Vdc Distribution Panel (D11, D111): The inspectors reviewed 125Vdc short

circuit calculations and verified that the interrupting ratings of the fuses and the

molded-case circuit breakers were well above the calculated short circuit

currents. The 125Vdc voltage calculations were reviewed to determine if

adequate voltage would be available for the breaker open and close coils and

spring charging motors. The inspectors reviewed the motor control logic

diagrams and the 125Vdc voltage drop calculation to ensure adequate voltage

would be available for the control circuit components under all design basis

conditions. The inspectors also reviewed the 125Vdc short circuit and

coordination calculations to assure coordination between the motor feed breaker

open and close control circuit fuses and 125Vdc supply breakers and to verify the

interrupting ratings of the control circuit fuses and the 125Vdc control power feed

breaker.

b. Findings

(1) Calculation Errors Associated With the Pneumatic Pressure Requirements for the

Inboard Main Steam Isolation Valves (MSIVs)

Introduction: A finding of very low safety-significance and associated Non-Cited

Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by

the inspectors for the failure to incorporate the actual physical configuration of the

inboard MSIVs and the correct pneumatic system pressure drop into the pneumatic

pressure requirement calculation for the inboard MSIVs.

Description: On September 2, 2009, the inspectors identified that the licensee failed to

adjust the weight of the actuator moving parts for the inboard MSIVs to reflect that the

actuator was offset by 45 degrees instead of being mounted vertically as assumed in

calculation CA-94-037, Calculation of Inboard MSlV Post-LOCA Closing Forces and

10 Enclosure

Pneumatic Pressure Requirements. This error affected the calculated minimum

alternate nitrogen system operating pressure required to close the inboard MSIVs.

The MSIVs were part of the primary containment with safety functions that included:

(1) providing primary containment isolation; (2) preventing core damage by limiting the

loss of coolant from the reactor vessel following a main steam line break outside the

primary containment; and (3) preventing excessive release of radioactivity to the

environs under assumed conditions of a primary steam line break outside of the primary

containment. The inboard MSIVs were wye pattern globe valves. The main disc was

attached to the lower end of the stem and moves in the valve guides at a 45 degree

angle from the inlet pipe. The valves were provided with springs that maintain the valves

in the closed position once they were shut by the valves air operator.

The normal pilot and safety grade pneumatic supplies for the inboard MSIVs was

provided by the alternate nitrogen system. The manifold and system pressures of the

alternate nitrogen system were monitored by pressure switches that give control room

annunciation on low pressure. The nominal setpoint for the pressure switches was

91 psig. However, the pressure switch has an instrument uncertainty of 3.7 psig, such

that the alarm potentially would not actuate until actual pressure was 87.3 psig.

The calculated minimum pneumatic pressure required to close the inboard MSIVs was

87 psig assuming that the valves were installed vertically. However, since the valves

were at a 45 degree angle, the closing force provided by the weight of the actuator

moving parts would be less than that assumed in the calculation. In addition, the

alternate nitrogen system pressure drop was incorrectly determined because of a

computational error. When the calculation incorporated the actual configuration of the

valves and the corrected pressure drop, the minimum required pneumatic pressure

increased to 88.84 psig, which was greater than the potential minimum pressure of 87.3

psig allowed by plant procedures.

The licensee evaluated the pneumatic pressure requirements for the inboard MSIVs and

determined that the valves were operable by removing conservatisms in the affected

calculation. The inspectors performed an independent review of the evaluation and had

no further concerns. The licensee initiated action requests (ARs) 01196394 and

01198495 to revise the affected calculation.

Analysis: The inspectors determined that the failure to incorporate the actual physical

configuration of the inboard MSIVs and the correct pneumatic system pressure drop into

the pneumatic pressure requirement calculation of the inboard MSIVs was a

performance deficiency.

The performance deficiency was determined to be more than minor because it was

associated with the Barrier Integrity cornerstone attribute of structures, systems,

components and barrier performance, and affected the cornerstone objective of

providing reasonable assurance that physical design barriers protect the public from

radionuclide releases caused by accidents or events. Specifically, the engineering

calculation error was significant enough to require the calculation to be re-performed to

assure that the minimum operating pressure of the alternate nitrogen supply was

adequate to support the MSIVs pneumatic requirements.

11 Enclosure

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, Table 3b for the Barrier

Integrity cornerstone. The finding screened as of very low safety-significance (Green)

because it was a design deficiency of the physical integrity of the reactor containment

that: (1) did not affect the barrier function of the control room against smoke or a toxic

atmosphere; (2) did not represent an actual open pathway in the physical integrity of

reactor containment; and (3) did not involve an actual reduction in function of hydrogen

igniters in the reactor containment. Specifically, the last alternate nitrogen system

surveillance demonstrated that the pressure was higher than the corrected minimum

required pressure. In addition, the licensee showed that the calculation had sufficient

conservatisms to account for the calculation errors. The inspectors performed an

independent review of this evaluation and had no further concerns.

The inspectors determined there was no cross-cutting aspect associated with this finding

because this was a legacy design issue and therefore was not reflective of current

performance.

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

in part, that measures shall be established to assure that applicable regulatory

requirements and the design basis are correctly translated into specifications, drawings,

procedures, and instructions.

Contrary to the above, as of September 4, 2009, the licensee failed to correctly translate

applicable design basis into specifications. Specifically, calculation CA-94-037 failed to

adjust the closing force provided by the weight of the actuator moving parts to reflect that

the actuator was at a 45 degree angle and to correctly compute the pneumatic system

pressure drop. Because this violation was of very low safety-significance and it was

entered into the licensees corrective action program (ARs 01196394 and 01198495),

this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC

Enforcement Policy (NCV 05000263/2009007-01).

(2) Emergency Service Water (ESW) Piping Supports Did Not Meet Seismic Category 1

Design Basis Requirements

Introduction: A finding of very low safety-significance and associated Non-Cited

Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was

identified by the inspectors for the failure to restore the ESW piping supports to their

design specifications.

Description: On September 9, 2009, the inspectors identified that the licensee failed to

recognize that the ESW discharge pipe supports did not meet their seismic Category 1

design basis requirements and that, as a result, the issue was not properly evaluated

such that corrective actions were not implemented.

During a plant walkdown, the inspectors noted that ESW piping stanchion supports

SWH-42 and SWH-47 were not in contact with their baseplates. The supports were

located on the discharge side of their respective ESW pumps near the boundary of

safety and non-safety-related piping. The gaps between the supports and the floor

baseplates were about 1/4 inch. These supports were classified seismic Category 1

because their function was to protect the safety-related piping that provides the required

12 Enclosure

cooling water to the EDGs. The EDGs were relied upon to safely shut down the reactor

upon the loss of all outside power simultaneous with a design basis accident or event.

Upon further review, it was discovered that the licensee had previously identified this

condition on at least three separate occasions.

• On February 3, 2003, AR 00633248 documented the discovery of gaps at supports

SWH-42 and SWH-47. The piping was evaluated and found to be operable

assuming that the supports were not functional. In addition, the licensee repaired the

supports by removing the gaps to restore the piping and associated supports to

within ASME Code allowables. The AR was classified Level B, which meant that the

AR documented a condition that typically results in moderate impact to the plant.

• On April 28, 2008, ARs 01135806 and 01135808 documented the discovery of the

gaps at the same supports. On this occasion, the licensee referred to the evaluation

conducted in 2003 and determined that no corrective actions were necessary

because the piping had been determined to be operable. Both ARs were classified

Level D, which meant that the ARs were considered to be associated with a

condition not adverse to quality.

• On February 26, 2009, AR 01170994 documented the rediscovery of the gaps and

noted that the condition was previously evaluated in 2003 and 2008. No corrective

actions were implemented. The AR was classified Level C, which meant that the AR

documented a condition that typically results in minor impact to the plant.

It was also noted that in 2006, two check valves were removed and an air operated

valve was replaced with a manual valve on each ESW lines under Engineering Change

(EC) 768, Service Water Check Valves Replacement, resulting in a weight reduction on

the pipe. This could have potentially resulted in the condition identified in 2008;

however, this change was not evaluated as part of the 2008 or 2009 ARs.

The licensee initiated AR 01196345 to address the condition of the supports and

AR 01196433 to document that the condition adverse to quality was not corrected when

it was identified in 2008 and 2009. The piping was evaluated by the licensee and found

to be operable assuming that the supports were not functional. The inspectors

performed an independent review of this evaluation and had no further concerns. The

licensee repaired the supports by installing shims to fill in the gaps through work

requests 49141 and 49142.

Analysis: The inspectors determined that failure to restore the ESW piping supports to

their design specifications was a performance deficiency.

The performance deficiency was determined to be more than minor because it was

associated with the Mitigating Systems cornerstone attribute of protection against

external events and affected the cornerstone objective of ensuring the availability of the

ESW system, and ultimately the EDGs, to respond to initiating events to prevent

undesirable consequences. Specifically, the gaps between the ESW piping supports

and the baseplates created a condition were the supports did not meet their seismic

Category 1 design basis requirements. Furthermore, the finding impacted the availability

of the EDGs because they cannot operate for a long period of time without cooling

provided by ESW.

13 Enclosure

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, Table 3b for the Mitigating System

cornerstone. The finding screened as very low safety-significance (Green) because the

finding was a design deficiency confirmed not to result in loss of operability or

functionality. Specifically, the licensee performed an operability evaluation that

concluded the ESW piping remained operable assuming the affected supports were not

functional. The inspectors performed an independent review of the operability

evaluation and did not have further concerns.

This finding has a cross-cutting aspect in the area of problem identification and

resolution, because the licensee did not thoroughly evaluate the issue such that the

corrective actions were inadequate. Specifically, the licensee failed to implement

corrective actions during the subsequent occasions that the condition was identified

because the problem was not properly prioritized and evaluated. The ARs initiated

subsequent to the condition being identified in 2003 were closed without corrective

actions under the premise that the condition was determined to be acceptable in 2003.

However, while the evaluation conducted in 2003 determined that the supports were

operable, it also determined that the supports needed to be restored to the original

design specifications. P.1(c)

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action,

requires, in part, that measures shall be established to assure that conditions adverse to

quality, such as failures, malfunctions, deficiencies, deviations, defective material and

equipment, and non-conformances are promptly identified and corrected.

Contrary to the above, in April 28, 2008 and February 26, 2009, the licensee failed to

promptly correct a condition adverse to quality regarding the ESW piping supports.

Specifically, although the licensee identified the gaps between the ESW piping supports

and the baseplates, it failed to recognize that this condition did not meet the seismic

Category 1 design basis requirements. Because the licensee failed to recognize the

qualification requirement, the condition was evaluated incorrectly and determined to be

acceptable such that corrective actions were not implemented. Because this violation

was of very low safety-significance and it was entered into the licensees corrective

action program (ARs 01196345 and 01196433), this violation is being treated as an

NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV

05000263/2009007-02).

(3) Failure to Adequately Evaluate Minimum Voltage Available at Safety-Related Electrical

Components

Introduction: A finding of very low safety-significance and associated Non-Cited

Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by

the inspectors for the failure to adequately evaluate circuit loads in determining design

limits in electrical calculations. Specifically, the inspectors identified three examples

associated with the following: (1) the licensees failure to perform a calculation for

safety-related motor starters that included all control circuit loads in determining the

minimum voltage available at 120Vac starter coils, which was used to establish the coil

voltage test acceptance criteria; (2) the licensees failure to include thermal overload

(TOL) heater and starter contact resistance when calculating the minimum voltage at

480Vac motor terminals; and (3) the licensees failure to assure that the minimum

14 Enclosure

voltage at the 120Vac solenoid operated control valves was in conformance with vendor

requirements.

Description: The inspectors identified three examples where the licensee failed to

adequately evaluate circuit loads in determining design limits in electrical calculations.

These examples are as follows:

• Failure to Evaluate Circuit Loads in Determining the Minimum Voltage Available At

Safety-Related Motor Starter Contactors

The inspectors reviewed the licensee evaluation on IN 2007-09, Equipment

Operability Under Degraded Voltage Conditions, and found that the licensee had not

adequately reviewed the notice for the issue on the adequacy of voltage at motor

starter coils during degraded voltage conditions. The licensee initiated AR 01196447

to perform the required review. The inspectors reviewed calculations CA-92-250,

Project 92Q600 Cable Sizing Calculation, and CA-94-094, MCC Starter Coil

Pickup Voltages and Maximum Cable Lengths, for the motor starter coil voltage for

ESW pumps P-111A and P-111B respectively. The inspectors found that calculation

CA-94-094 determined 96.32 volts were available at the P-111B motor starter

contactor coil and subsequently used 96 volts as the coil test acceptance criteria on

drawing NE-36347-1A, 480V MCC Schedules. A starter contactor coil test was

used in lieu of meeting the vendor criteria for minimum pickup voltage of 102 volts,

which was 85 percent of rated voltage. However, the inspectors found that the

calculation failed to consider the load on the control power transformer (CPT) due to

timing relay burden, indicating light load, and contact resistance in determining the

limiting voltage available at the motor starter.

The licensee performed a hand calculation during the inspection, which determined

that less than 96 volts was available when all the circuit resistance elements were

included in the voltage analysis and established 94 volts as the new acceptance

criteria for testing the coil. The contactor coil was previously tested in 1994 to pickup

at 84 volts, which provided the basis for immediate operability on AR 01200723. As

additional corrective actions, the licensee planned to test all the affected coils on

drawing NE-36347-1A. On October 15, 2009, the test for the starter contactor coil

for P-111B determined that the voltage needed for the coil to pickup was 104 volts,

not the 94 volts required by the revised calculation. The licensee subsequently

replaced the starter with a spare unit that was functionally tested at 80 volts. The

licensee sent the failed starter to a test laboratory to determine the cause of failure

and to initiate an evaluation of extent of condition.

The test laboratory determined that the cause of the failure was mechanical binding

of the starter right side contact arm and plunger binding due to abrasion and

gouging, which then required a higher voltage for the coil to function. The test lab

was not able to determine the root cause of the abrasion and binding that lead to the

failure. Since this pump was operated periodically, the normal available voltage to

the starter contactor coil was sufficient to overcome the binding effect of the contact

arm such that the pump started as required. As part of the extent of condition for this

issue, the licensee established a motor starter action plan to test 12 additional

safety-related starters. In addition, motor starter testing methodology and preventive

maintenance tasks were under review such that this issue could be identified prior to

failure.

15 Enclosure

• Failure to Fully Evaluate Power Circuit Resistance in Determining Minimum Voltage

Available at Safety-Related Motor Terminals

The inspectors reviewed calculation CA-06-104, 480V MCC to Motor Terminal

Voltage Drop, and identified that the calculation did not consider TOL heater

resistance or starter contact voltage drop in determining the voltage available at the

motor terminals. For ESW Pump P-111A, this reduced the available voltage 0.2

volts, which reduced the voltage margin available to the minimum motor voltage

(414 volts) by 12 percent. The inspectors found that the voltage available was

approximately 0.45 percent more than the minimum voltage required by the motor

based on the motor rating and the pump load requirement. The licensee initiated

AR 01197431, which determined that although the available margin had been

reduced, the operability of the pump was not affected by the calculation error.

• Failure to Assure Adequate Minimum Voltage at Safety-Related 120Vac Solenoid

Operated Control Valves

Section 8.10 of the USAR stated that the 120Vac instrument buses were maintained

between 108Vac and 132Vac (120 +/- 10 percent). Per MWI-3-M-2.01, AC Electrical

Load Study, 120Vac rated loads, including the solenoid operated control valves,

would have adequate voltage if the AC instrument buses were maintained within this

voltage range. The vendor specified a voltage range of 102V - 132V for the solenoid

operated control valves to operate properly. The published minimum operating

voltage, as well as the minimum voltage used for equipment qualification testing for

the solenoid valves (SV) was 102Vac. This value was identified in the ASCO

Nuclear Catalog-Nuclear Products Qualified to IEEE Specifications, in ASCO

Solenoid Catalog No. 31, and was used as the minimum test voltage for the ASCO

Test Report AQS-21678/TR.

The inspectors requested the voltage drop analysis for the SVs, however, the

licensee was unable to locate an analysis for the SVs. The licensee performed an

informal analysis for SV1729 and SV1994, which were in the inspectors scope, and

as part of an extent of condition, SV1728, SV1995, and SV1996. Based on this

analysis, the licensee determined that the voltages at SV1728 and SV1995 were

below the vendor specified minimum rating. As a result of the negative margin

identified for SV1728 and SV1995, the licensee tested three ASCO SVs, two of

which were the same models as the existing solenoids, to determine their minimum

pull-in and dropout voltages at various differential pressure values across the valve.

These tests concluded that the SVs were capable of energizing at voltages lower

than the specified 102Vac.

The licensee entered this issue in their corrective action program as AR 01199936 to

complete a formal voltage drop calculation for 120Vac supplies to safety-related

equipment and to restore voltage margin as a long-term solution. The licensee

documented the details of the testing for the sampled SVs in operability

recommendation OPR 199936-1. Based on the results of the tests, the licensee

determined that there was reasonable assurance that these valves would perform

their safety functions as designed and therefore the OPR classified the solenoids as

operable but nonconforming.

16 Enclosure

Analysis: The inspectors determined that the licensees failure to ensure adequate

voltage was available to energize the starter coils for 480Vac safety-related equipment

and the 120Vac solenoid valve coils was a performance deficiency because the

operability of safety-related equipment could not be assured and could have resulted in

a loss of function during a design basis accident concurrent with a degraded voltage

condition.

The performance deficiency was determined to be 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

safety-related equipment to respond to initiating events to prevent undesirable

consequences. Specifically, the failure to ensure adequate voltage was available to

energize safety-related starter coils to supply 480Vac power to motors and 120Vac to

power SV coils would have affected the availability of their respective equipment to

respond to initiating events.

The inspectors evaluated the finding, with the exception of ESW pump P-111B, using

IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for

At-Power Situations, SDP Phase 1 screening. The finding screened as very low safety-

significance (Green) because the finding was a design deficiency confirmed not to result

in loss of operability or functionality. Specifically, the failure to assure adequate voltage

was available at the starter coils and SV coils; and to perform periodic testing to assure

the minimum voltage remained acceptable as the components aged did not result in an

impact on current operability.

The inspectors evaluated the finding of ESW pump P-111B using IMC 0609,

Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power

Situations, SDP Phase 1 screening. In accordance with Table 3b, "SDP Phase 1

Screening Worksheet for Initiating Events, Mitigating Systems, and Barriers

Cornerstones," the finding affected the safety of an operating reactor, specifically, the

Mitigating Systems Cornerstone. In accordance with Table 4a, "Characterization

Worksheet for Initiating Events, Mitigating Systems, and Barriers Cornerstones," the

finding represented a potential loss of a safety function of a single train of emergency AC

power for greater than its TS allowed outage time. The inspectors contacted a Region III

Senior Reactor Analyst (SRA) to perform an SDP review of the finding.

In accordance with IMC 0609, Appendix A, the SRA determined that the Phase 2 SDP

pre-solved tables showed that having one EDG service water pump inoperable for an

entire year was a Yellow finding. The SRA determined this result to be conservative

since P-111B was not completely failed, but shown to be nonfunctional only during

certain degraded voltage conditions. Further, the voltage of interest at the P-111B

starter coil was relevant when 12 EDG was supplying its associated safety-related 4160

Vac bus (Bus 16) and also during a loss of coolant accident (LOCA) with offsite power

available (with EDG 12 operating but not supplying Bus 16) and degraded voltage

conditions prior to a loss of off-site power LOOP occurring. The SRA performed an SDP

Phase 3 analysis to further characterize the significance of the finding.

The SRA performed a Phase 3 analysis using the NRC Risk Assessment

Standardization Project Handbook and the SPAR Model for Monticello, Revision 3P,

Level 1, Change 3.45, dated August 2008. Assuming failure to start of P-111B for a one

year exposure, the delta-CDF was less than 1E-7. The dominant cut-sets involved

17 Enclosure

station blackout events with failure of emergency power sources and failure to recover

either offsite or emergency power. The SRA contacted the licensee to review their risk

analysis.

The licensee reviewed recent emergency core cooling system (ECCS) test records,

which showed that P-111B started successfully under nearly identical voltage conditions

to those that would be expected during a loss of offsite power transient with an ECCS

initiation signal present. The licensee also performed a quantitative analysis assuming

a failure to start probability for P-111B of ten times its nominal value. The resultant

delta-CDF was about 3.6 E-7/yr. Nearly all of the increased risk was associated with

internal flooding scenarios where the flood initiates at or propagates to the lower

switchgear, which in turn causes a loss of offsite power event.

Given the low numerical risk results for the core damage sequences, the SRA concluded

that the risk associated with this performance deficiency was very low safety-significance

(Green).

Based on the results of the SDP Phase I and Phase III screenings, these three

examples for this finding screened as very low safety-significance (Green). The

inspectors determined there was no cross-cutting aspect associated with this finding

because the three examples were legacy design issues and therefore was not reflective

of current performance.

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

in part, that design control measures 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 suitable testing program.

Contrary to this requirement, prior to September 30, 2009, for P-111B and

September 28, 2009, for the solenoid valve coils, the licensees design control measures

failed to verify the adequacy of minimum design voltage for the safety-related coils.

Specifically, the licensee failed to adequately establish the degraded voltage value for

safety-related starters and solenoid valve coils. In addition, periodic testing at degraded

voltage, which was lower than the vendors rated minimum voltage, was not performed

to verify that the coils remained acceptable as the components aged. However,

because this violation was of very low safety-significance and because the issue was

entered into the licensees corrective action program (AR 01200723, AR 01197431, and

AR 01199936), this violation is being treated as an NCV consistent with Section VI.A.1

of the NRC Enforcement Policy. (NCV 05000263/2009007-03)

(4) Inadequate Testing for Motor Control Center (MCC) Contactors

Introduction: A finding of very low safety-significance and associated Non-Cited

Violation of 10 CFR Part 50, Appendix B, Criterion XI, ATest Control,@ was identified by

the inspectors for the failure to have adequate testing for safety-related equipment to

monitor component degradation. Specifically, the licensee failed to verify that the MCC

contactors would continue to pick-up under degraded voltage conditions with less than

the vendors required minimum voltage.

Description: The licensees AC load study identified that the lowest voltage at the MCCs

under degraded voltage conditions was 426V. In order for MOVs and other components

18 Enclosure

to operate, the MCC contactors needed to be verified to pick-up at the MCC degraded

voltage. Calculation CA-94-094, MCC Starter Coil Pick-Up Voltages and Maximum

Cable Lengths, determined the maximum cable length for each size of control cable to

ensure reliable circuit operation of contactors under degraded voltage conditions. This

calculation was performed to ascertain the maximum cable lengths with both 75 percent

and 85 percent contactor pick-up voltage for various combinations of National Electrical

Manufacturers Association (NEMA) starter sizes and CPT sizes at degraded voltage.

The calculation determined there were 13 contactors that had less than 85 percent

pick-up voltage, although only MO-2013 was included in the inspectors scope.

MO-2013 contactor was shown to have 84.52 percent pick-up voltage. The calculation,

however, did not account for valves MO-2007 and MO-2020, which were also in the

inspectors scope.

The licensee performed an informal evaluation for MO-2007 and MO-2020 and

determined that the cable lengths for both MOVs contactors were bounded by the

maximum cable lengths determined for a NEMA size 1 starter with a 75 volt amp CPT.

The inspectors reviewed this evaluation and determined that the assessment for these

two MOVs was acceptable. This evaluation also identified that there were interposing

relays installed in the MO-2013 control circuit to facilitate the pick-up of the contactor,

which were not accounted for in the calculation CA-94-094.

The licensee provided calculation CA-92-223, Analysis for Mod 92Q520 Control Circuit

Voltage Drop, to address the interposing relay issue. This calculation determined that

the starter coil for MO-2013 would have 83 percent of the rated voltage with the

interposing relays in the circuit, which was less than the 84.52 percent identified in

calculation CA-94-094. This calculation also stated a coil pick-up voltage acceptance

criterion of 75 percent instead of 85 percent as indicated in calculation CA-94-094. The

inspectors were concerned about the discrepant acceptance criteria used for the

contactor pick-up voltage. In response, the licensee stated that during the evaluation of

IN 94-050, Failure of GE Contactors to Pull in at the Required Voltage, it was

confirmed by General Electric (GE) that the contactor coils were rated for 85 percent and

not the 75 percent pick-up voltage indicated in calculation CA-92-223. The licensee

initiated AR 01200487 to supersede calculation CA-92-223 with calculation CA-94-094.

The inspectors expressed concern with the 13 GE contactors (associated with an MOV,

fan, or pump) having less than 85 percent pick-up voltage and prompted the licensee for

justifying operability under degraded voltage conditions. The licensee stated tests were

conducted in 1976, 1977, 1994, and 2000, on a few contactor samples. The testing

indicated that the contactors were able to pick up at less than 85 percent voltage.

However, the licensee did not have a periodic testing program in place to ensure that

age related degradation did not have an adverse impact on the contactors. The licensee

provided Procedure 4847-PM, GE 7700 Line Motor Control Center Maintenance

Procedure, which indicated testing would only be performed on a starter when it was

replaced. As such, the licensee was relying on the old test data to justify the continued

operability of these contactors.

The licensee entered this issue into their corrective action program and initiated

ARs 01200539 and 01200540 to revise the maintenance procedures to incorporate the

requirements for periodic testing of contactors. In addition, the licensee initiated work

orders to test the 13 contactors as soon as practicable. As of October 15, 2009, the

licensee had tested 10 of the contactors, which verified for 9 out the 10 that the

19 Enclosure

contactors would still pick-up at the calculated degraded voltage condition. However, as

discussed in Section 1R21.3.b.(3) of this report, the starter motor contactor associated

with ESW Pump P-111B failed during this testing. Additional corrective action and

extent of condition for this issue was discussed in Section 1R21.3.b.(3) of this report.

Analysis: The inspectors determined that the failure to have an adequate testing

program for the MCC contactors was a performance deficiency, because the failure of

the contactors to pick up could have resulted in a loss of function during design basis

accident conditions. Specifically, the licensee had determined from calculations that the

voltage at the 13 contactors would be less than the vendor required 85 percent pick-up

voltage and had not implemented any periodic testing to ensure the contactors would not

degrade over time. Failure of the contactors to pick-up would prevent the associate

MOV, fan, or pump to operate as required under a degraded voltage condition.

The performance deficiency was determined to be 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

safety-related equipment to respond to initiating events to prevent undesirable

consequences. Specifically, the failure to verify that safety-related starter coils were not

degrading over time thorough periodic testing such that they would continue to function

at a lower calculated voltage. The failure of the starter coil to energize would prevent the

associate MOV, fan, or pump to operate as required under a degraded voltage condition.

The inspectors determined the finding, with the exception of ESW pump P-111B, 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, Table 3b for the Mitigating System cornerstone. The finding screened as very

low safety-significance (Green) because the finding was a design deficiency confirmed

not to result in loss of operability or functionality. Specifically, the licensee confirmed

for 9 of the 13 contactors that they would function at a lower voltage than required by the

vendor. The inspectors reviewed the results of the tests and verified that the tested

contactors were able to pick up at less than 85 percent voltage. The inspectors

determined that the schedule submitted by the licensee for testing the remaining

contactors was reasonable. Based on these test results, no apparent common cause to

the failed starter motor, and initial contactor testing performed in 1976, 1977, 1994, and

2000, there was reasonable assurance of operability for the remaining untested

contactors.

The inspectors Phase III screening of the ESW pump P-111B finding was discussed in

Section 1R21.3.b.(3) of this report, which concluded that the risk associated with this

performance deficiency was very low safety-significance (Green).

Based on the results of the SDP Phase I and Phase III screenings, this finding screened

as very low safety-significance (Green). The inspectors determined there was no

cross-cutting aspect associated with this finding because this was a legacy design issue

and therefore was not reflective of current performance.

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XI, ATest 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 is identified and performed in accordance with written test procedures, which

20 Enclosure

incorporate the requirements and acceptance limits contained in applicable design

documents.

Contrary to the above, as of October 2, 2009, the licensee failed to establish a periodic

testing program to ensure that the MCC contactors would be energized to perform its

safety-function under design basis accident conditions. Specifically, the licensee failed

to require periodic testing of contactors to ensure that age related degradation did not

have any adverse impact on contactors pick-up voltage. However, because this

violation was of very low safety-significance and because the issue was entered into the

licensees corrective action program (AR 01200539 and 01200540), this violation is

being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy.

(NCV 05000263/2009007-04)

(5) EDG Fuel Oil Supply System Design Does Not Meet Single Failure Criteria

Introduction: The inspectors identified an unresolved item (URI) regarding the EDG fuel

oil supply system. Specifically, the diesel fuel oil supply system design does not meet

the single failure criteria.

Description: The diesel fuel oil supply system consisted of two pumps, the fuel oil

transfer pump and the fuel oil service pump. However, only the fuel oil transfer pump

was considered safety-related. Nonetheless, TS Basis, Section 3.8.1, stated that both

pumps must be operable in order for both EDGs to be operable.

The original Monticello Safety Evaluation Report (SER), dated March 20, 1970, stated

that the EDGs were separate and independent with respect to fuel supplies. It also

concluded that the onsite emergency electric power system was acceptable since no

single failure should prevent power from being supplied to the engineering safety

features from onsite sources. In addition, the original Safety Analysis Report stated that

auxiliaries required to ensure continuous operation of the EDGs shall be supplied from

the essential buses or control power transformers associated with the EDGs. However,

the fuel oil service pump was powered by a non-essential motor control center that

would be load shed on an essential bus transfer load shed signal. In fact, EDG 11 does

not supply either pump without manual operator action. The fuel oil transfer pump was

normally powered by EDG 12.

In addition, the USAR, Section 10.3.1.5, stated that the safe-shutdown analysis

performed to comply with 10 CFR Part 50, Appendix R, Section G, revealed three fire

areas that required repair of equipment or use of alternate fuel oil pumping methods and

procedures for diesel oil pumping capability to the EDG day tanks to achieve cold

shutdown.

Further, the TS Basis, Section 3.8.1, described the diesel fuel oil supply piping as

redundant. However, the inspectors confirmed that the actual configuration was not

redundant in that both pumps discharge into common piping.

The inspectors also questioned if the service pump was tested to supply an adequate

flow. Specifically, TS Surveillance Requirement 3.8.1.5 verified, in part, that the fuel oil

transfer system operates to transfer fuel oil from the T-44 storage tank to the day tanks.

The TS Basis, Section 3.8.1, stated that this surveillance provides assurance that the

required fuel oil transfer pumps are operable. Operable was defined in TS 1.1, which

21 Enclosure

stated, in part, when the component is capable of performing its specified safety

function. The TS Basis, Section 3.8.1, also stated that the fuel oil transfer pump and the

fuel oil service pump are individually capable of maintaining the level in the day tank

when both EDGs are operating at full load. However, the inspectors noted that the

licensee only ensured that the fuel oil service pump transferred fuel and maintained level

in just one EDG day tank.

This issue is unresolved pending further NRC review of the licensing basis for the diesel

fuel oil transfer system and determination of NRC courses of action for resolution of the

issues. (URI 05000263/2009007-05).

(6) Inadequate Tornado Missile Protection for the EDG System Components

Introduction: The inspectors identified an unresolved item (URI) regarding the design

and licensing basis for the standby diesel generator (EDG) building ventilation system

and whether the ventilating system had to be protected from the effects of a design basis

tornado.

Description: During a walkdown of the EDG building, the inspectors noted that

temperature control dampers in the ventilation system were mounted flush with the

outside walls of the building with only a metal grating serving as a barrier for tornado

missiles. The inspectors questioned the adequacy of the design and received the

licensee position that the EDG building was a Class 1 building designed to protect the

EDG and the fuel oil day tank from tornado missiles. However, the licensee stated that

the temperature control dampers and ventilation fans (V-SF-9 and - 10) were not

designed to be protected from a tornado missile. The licensee stated that the designer

and the licensee intended on protecting Class 1 equipment required to assure safe-

shutdown of the reactor from a missile event based on the credibility of the missile. The

licensee agreed that the ventilation fans were necessary for the EDGs to be operable

and that the EDGs could only run for a short period of time without the fans running.

However, the licensee maintained that the EDGs could perform their safe-shutdown

function without the ventilation fans. The inspectors disagreed as the EDGs could not

run long enough to shut the reactor down, remove latent and decay heat, and maintain

the reactor in a cold shutdown condition following the loss of the EDG ventilation

systems. The licensee maintained that a single missile could not take out both fans as a

full height reinforced concrete wall separated the two ventilation systems.

The inspectors reviewed the original design submittal (NSP-1 dated October 17, 1969),

and found that the Class 1 equipment included the Standby Diesel Generator System,

and the Emergency Buses and other electrical gear to and including power equipment

required for safe-shutdown. The ventilation systems were not included in the list of

Class 1 equipment or in the list of Class 2 equipment. However, the licensee believed

the ventilation systems were considered Class 2 based on the last entry in the Class 2

list: All Other Piping and Equipment not listed under Class 1. After further research,

the licensee stated that there was no specific statement regarding the EDG building

ventilation in any licensing document.

The inspectors identified the following statements in various licensing documents.

• Section 2.2.4, Standby Diesel Generating Building, of the USAR stated, The

principal function of this building is to provide a safe enclosure and protection for the

22 Enclosure

standby diesel generators and portions of the power distribution systems enclosed

therein.

• Section 2.2.4.1, Structure Description, stated, in part, A north-south [sic] interior

wall of reinforced concrete extends the full height of the structure providing physical

separation of the diesel generator systems.

• The original Monticello SER, Section 3.1.2, Meteorology, stated, in part, that the

facility structures and systems, which are necessary for a safe-shutdown of the

reactor are designed to withstand the effects of wind loadings and potential missiles

resulting from a tornado.

• In the current Revision 25 of the USAR the inspectors noted the following:

a) Section 8.4.1.1, Design Basis, stated that two independent EDGs provide

redundant standby power sources.

b) Section 8.4.1.1.b stated, The EDG sets shall be complete package units with

all auxiliaries necessary to make them self-sufficient power sources capable of

automatic start at any time and capable of continued operation at rated full load

voltage and frequency until either manually or automatically stopped.

c) Section 8.4.1.1.d stated, The EDGs shall be located in Class 1 structures.

d) Section I.4.3.14, HVAC Systems, stated: The only HVAC Equipment

required for safe-shutdown are the ECCS Room Coolers, V-AC-5 (Division I)

and V-AC-4 (Division II), located in respective Reactor Building corner rooms

on the 920-foot elevation, and the EDG supply fan, V-SF-9 for EDG No. 12,

and V-SF-10 for EDG No.11.

e) USAR Table J.4.5-1, Appendix R Safe-shutdown Equipment List, identified

fans V-SF-9 and -10 as safe-shutdown equipment.

The inspectors concluded that the EDG ventilation fans were an auxiliary necessary for

the EDG system and that the term EDGs in the USAR included all of the auxiliaries for

the EDGs; therefore, the ventilation fans were necessary for safe-shutdown of the

reactor (to achieve and maintain cold shutdown).

Based on no actual licensing document specifically mentioning the EDG ventilation,

the necessity of the fans for EDG operability, the USAR references to the fans being

safe-shutdown equipment, and the NSP-1 statement that the full-height wall separated

the EDG systems, the inspectors concluded that the fans and temperature control

dampers should be considered Class 1 equipment and should have been protected from

tornado missiles, as well as the effects of tornadoes on the ventilation ducts as

described in RIS 2006-023, Post Tornado Operability of Ventilating and Air Conditioning

Systems Housed in Emergency Diesel Generator Rooms. However, because this issue

was not clearly defined in the original licensing documents, this will be an unresolved

item pending consultation with NRC headquarters for clarification of whether the EDG

building ventilation was or was not required to be a Class 1 system.

(URI 05000263/2009007-006)

23 Enclosure

.4 Operating Experience

a. Inspection Scope

The inspectors reviewed six operating experience issues 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 2006-22, New Ultra-Low-Sulfur Diesel Fuel Oil Could Adversely Impact Diesel

Engine Performance;

• IN 2007-09, Equipment Operability Under Degraded Voltage Conditions;

• IN 2008-02, Findings Identified During Component Design Bases Inspections;

• IN 2009-02, Biodiesel in Fuel Oil Could Adversely Impact Diesel Engine

Performance;

• RIS 2006-023, Post Tornado Operability of Ventilating and Air Conditioning

Systems Housed in Emergency Diesel Generator Rooms; and

• SC06-01, Worst Single Failure for Suppression Pool Temperature Analysis.

b. Findings

No findings of significance were identified.

.5 Modifications

a. Inspection Scope

The inspectors reviewed three 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:

• EC 12361, EPU-Provide Operations the Ability to Throttle MO-2020 and

MO-2021;

• EC-11734, 1AR Transformer Replacement, Monticello Plant EPU Project; and

• SRI-91-010, USAR Update Concerning Diesel Oil Day Tank.

b. Findings

No findings of significance were identified.

24 Enclosure

.6 Risk-Significant Operator Actions

a. Inspection Scope

The inspectors performed a margin assessment and detailed review of four risk-

significant, time critical operator actions. 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 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:

• Loss of Instrument Air;

• Operating Suppression Pool Cooling;

• Align Diesel Fire Pump in a Station Blackout; and

• Time Critical Manual Actions Outside the Main Control Room Before Core

Damage.

b. Findings

No findings of significance were identified.

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 that were

identified by the licensee and entered into the corrective action program. 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 corrective action program. The specific corrective action documents that were

sampled and reviewed by the inspectors are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

25 Enclosure

4OA5 Power Uprate (71004)

.1 Plant Modifications (2 samples)

a. Inspection Scope

The inspectors reviewed plant modifications for those implemented for the extended

power uprate. This includes seismic qualification of balance of plant piping and pipe

supports for extended power uprate and the control logic changes to allow operations

the ability to manually set an incremental position for the safety-related outboard

containment spray valves.

• EC 12361, EPU-Provide Operations the Ability to Throttle MO-2020 and MO-

2021; and

• EC 11126, EPU-MOD 11-Balance of Plant Piping Support Modifications.

b. Findings

(1) Pipe Support Design Deficiencies

Introduction: A finding of very low safety-significance and associated Non-Cited

Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by

the inspectors for the failure to demonstrate pipe supports met their design

requirements. Specifically, the inspectors identified that the calculation for pipe support

SR-526 failed to use the minimum yield strength in determination of the allowable

bending stress of the pipe support SR-526 baseplate as required in the American

Institute of Steel Construction (AISC) code. In addition, the inspectors identified that the

calculation for pipe support PS-16 failed to use the design basis concrete compressive

strength in determination of the anchor bolt allowable as required in the licensees

design specification.

Description: The inspectors identified two examples where pipe supports would not

meet their design requirements. These examples are as follows:

• Pipe Support SR-526

The design function of pipe support SR-526 was to hold and maintain SW6-24"-JF

discharge line in position during a seismic Category I design basis event to meet

internal flooding requirements. The failure of this pipe support may create a pipe

break in the SW6-24"-JF piping line, which could cause internal flooding and

subsequent impact on the RHR pumps P-202A/C and core spray pump P-208A.

Also, the failure of pipe support SR-526 due to a seismic Category I design basis

event could impact the RHR heat exchanger E-200A. The inspectors identified a

non-conservative technical error in pipe support SR-526 calculation CA-04-112,

Evaluation of Support SR-526 for Supplementary Load Combination. The

calculation evaluated acceptability of the baseplate connection based on an

allowable bending stress of 0.75 times the actual baseplate material yield stress.

The acceptance criterion for allowable bending stress of a pipe support baseplate

was established in Specification No. MPS-1100, Specification for the Analysis of

Piping and Pipe Support System, Revision 8. Section 8.2.2 of MPS-1100 required

26 Enclosure

the seismic Category I pipe support stress limits to be in accordance with the AISC.

The requirement in the AISC was that the allowable bending stress was 0.75 times

the specified minimum yield stress of the material. The licensee determined that

the calculation CA-04-112 specified the actual material yield stress instead of the

minimum yield stress. The use of actual material yield stress for the evaluation of

the baseplate did not meet AISC code requirements.

The licensee agreed that the use of the Certified Material Test Report or the actual

material yield stress to evaluate the pipe support baseplate was outside the

allowable bending stress acceptance criteria as described in AISC. This issue was

entered in the licensees corrective action program as ARs 01198415 and

01199540. The licensee performed an analysis and determined the pipe support

was operable but nonconforming.

• Pipe Support PS-16

The design basis function of pipe support PS-16 was to hold and maintain

PS4-18-ED main steam line in position during a seismic Category I design basis

event. The outboard MSIV AO-2-86D was located between the primary

containment penetration X-7D and pipe support PS-16. The inspectors were

concerned that, if the pipe support did not meet its seismic Category 1 design basis

requirements, the line could be damaged during a design basis event at a location

between the primary containment penetration and the MSIV. This condition could

adversely affect the capability of the outboard MSIV to perform its safety-related

function of providing primary containment isolation.

The inspectors identified a non-conservative technical error in pipe support PS-16

calculation CA-96-147, Evaluation of PS1-18ED, PS2-18ED, PS3-18ED and

PS4-18ED Outside Containment for Revised Turbine Stop Valve Closure Loads.

The calculation evaluated the acceptability of the anchor bolts based on allowables

that use age hardened concrete compressive strength of 6000 pounds per square

inch. The acceptance criteria for anchor bolt allowables were determined from

Specification No. MPS-1100. Section 8.2.7 of MPS-1100 stated the concrete

design strength for determining allowable loads for the anchor bolts shall be 3000

pounds per square inch or 4000 pounds per square inch. The licensee

determined that calculation CA-96-147 specified that use of anchor bolt allowables

based on age hardened concrete compressive strength of 6000 pounds per square

inch instead of 3000 pounds per square inch or 4000 pounds per square inch as

required in Specification No. MPS-1100.

The licensee agreed that the use of age hardened concrete compressive strength

to evaluate pipe support PS-16 anchor bolts was outside the anchor bolt

acceptance criteria established in Specification No. MPS-1100. This issue was

entered in the licensees corrective action program AR 01200215. The licensee

performed an analysis and determined the pipe support was operable but

nonconforming.

Analysis: The inspectors determined that the failure to use minimum specified yield

stress to analyze the applied bending stress on the baseplate for pipe support SR-526

as described in the AISC code and failure to use the design basis concrete compressive

27 Enclosure

strength in determination of the anchor bolt allowable for pipe support PS-16 as required

in MPS-1100 was a performance deficiency.

The performance deficiency for the pipe support SR-526 example was determined to be

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 the RHR and core spray pumps. Specifically,

compliance with seismic Category I design basis requirements (AISC code

requirements) was to ensure the pipe support SR-526 would function as required during

a seismic Category I design basis event and not cause internal flooding and subsequent

impact on the RHR pumps P-202A/C, and core spray pump P-208A.

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, Table 4a for the Mitigating Systems

cornerstone. The finding screened as of very low safety-significance (Green) because it

was a design deficiency that did not represent an actual loss of safety function. The

inspectors agreed with the licensees position that the pipe support SR-526 was

operable but nonconforming.

The performance deficiency for the pipe support PS-16 example was determined to be

more than minor because it was associated with the Barrier Integrity cornerstone

attribute of design control and affected the cornerstone objective of providing reasonable

assurance that physical design barriers protect the public from radionuclide releases

caused by accidents or events. Specifically, compliance with seismic Category I design

basis requirements was to ensure the pipe support PS-16 would function as required

during a seismic Category I design basis event and not adversely affect the outboard

MSIV AO-2-86D.

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, Table 3b for the Barrier Integrity

cornerstone. The finding screened as of very low safety-significance (Green) because it

was a design deficiency of the physical integrity of the reactor containment that: (1) did

not affect the barrier function of the control room against smoke or a toxic atmosphere;

(2) did not represent an actual open pathway in the physical integrity of reactor

containment; and (3) did not involve an actual reduction in function of hydrogen igniters

in the reactor containment. The inspectors agreed with the licensees position that the

pipe support PS-16 was operable but nonconforming.

Based on the results of the two SDP Phase I screenings, this finding screened as very

low safety-significance (Green). The inspectors determined there was no cross-cutting

aspect associated with this finding because the two examples were legacy design

issues, therefore was not reflective of current performance.

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

in part, that design control measures shall provide for verifying or checking the adequacy

of design, such as by the performance of design reviews, by the use of alternate or

simplified calculational methods, or by the performance of a suitable testing program.

28 Enclosure

Contrary to the above, on September 18, 2009 and September 29, 2009, the licensee

failed to demonstrate design adequacy of pipe support SR-526 and pipe support PS-16

was consistent with seismic Category I requirements. Specifically the performance of

design reviews for pipe support SR-526 and pipe support PS-16 were inadequate, in that

design calculation CA-04-112 did not demonstrate that the pipe support SR-526

baseplate would meet AISC code requirements and the design calculation CA-96-147

did not demonstrate the pipe support PS-16 would meet design basis anchor bolt

requirements. Because this violation was of very low safety-significance and it was

entered into the licensees corrective action program (ARs 01198415, 01199540 and

01200215), this violation is being treated as an NCV, consistent with Section VI.A.1 of

the NRC Enforcement Policy. (NCV 05000263/2009007-07)

4OA6 Meeting(s)

.1 Exit Meeting Summary

On October 2, 2009, the inspectors presented the inspection results to Mr. T. OConnor,

and other members of the licensee staff. On December 4, 2009, the inspectors

conducted a re-exit of the inspection results to Mr. T. OConnor, 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

29 Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. OConnor, Site Vice President

J. Grubb, Plant Manager

P. Albares, Operations Shift Manager

R. Anderson, System Engineering Supervisor

R. Baumer, Regulatory Affairs Compliance Engineer

V. Bhardwaj, System Engineering Manager

N. Haskell, Engineering Director

B. Halvorson, Configuration Management Supervisor

K. Jepson, Business Support Manager

S. Oswald, Regulatory Affairs Analyst

D. Pennington, Mechanical Design Engineer

G. Salamon, Corporate Regulatory Affairs Manager

R. Siepel, Electrical Design Engineer

E. Watzel, Electrical Design Engineering Supervisor

P. Young, Regulatory Programs Supervisor

Nuclear Regulatory Commission

S. Thomas, Senior Resident Inspector, DRP

L. Haeg, Resident Inspector, DRP

1 Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000263/2009007-01 NCV Calculation Errors Associated With the Pneumatic

Pressure Requirements for the Inboard MSIVs05000263/2009007-02 NCV ESW Piping Supports Did Not Meet Seismic

Category 1 Design Basis Requirements05000263/2009007-03 NCV Failure to Adequately Evaluate Minimum Voltage

Available at Safety-Related Electrical Components05000263/2009007-04 NCV Inadequate Testing for MCC Contactors05000263/2009007-05 URI EDG Fuel Oil Supply System Design does Not Meet

the Single Failure Criteria

05000263/2009007-06 URI Inadequate Tornado Missile Protection for the EDG

System Components05000263/2009007-07 NCV Pipe Support Design Deficiencies

Closed

05000263/2009007-01 NCV Calculation Errors Associated With the Pneumatic

Pressure Requirements for the Inboard MSIVs05000263/2009007-02 NCV ESW Piping Supports Did Not Meet Seismic

Category 1 Design Basis Requirements05000263/2009007-03 NCV Failure to Adequately Evaluate Minimum Voltage

Available at Safety-Related Electrical Components05000263/2009007-04 NCV Inadequate Testing for MCC Contactors05000263/2009007-07 NCV Pipe Support Design Deficiencies

Discussion

None

2 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

CA-00-104 Intake Structure Minimum Water Level 0

CA-01-036 Inservice Testing (IST) Pump and Valve Acceptance Criteria 28

Rounding Evaluation for 4Ih Ten-Year Code Interval

CA-01-062 Instrument Uncertainty Calculation 1

CA-01-113 Setpoint for RHR Minimum Flow Switches FS-10-121A, B, C, 13

and D

CA-01-174 Minimum RHRSW Pump Differential Pressure 2

CA-01-177 Determination of Containment Overpressure Required for 8

Adequate NPSH for Low Pressure ECCS Pumps Updated for

Suction Strainer Debris Loading

CA-02-179 125 Volt Div 1 Battery Calculation 1

CA-03-111 EDG Jacket Cooler Maximum Allowed Tubes Plugged 10

CA-03-145 Evaluation of SR-680 1

CA-03-146 Evaluation of New Guides on The High Point Vent 1

CA-03-200 Internal Flooding Evaluation Due To A Postulated Break In 2.5 11

Fire Line

CA-04-010 EOP Calculation Inputs 2

CA-04-103 Emergency Diesel Generator Room Heat-up Air-Cooler Fan- 12

Flow Analysis

CA-04-112 Evaluation of Support SR-526 for Supplementary Load 0

Combination

CA-04-113 Evaluation of Support SR-531 for Supplementary Load 0

Combination

CA-04-114 Evaluation of Support SWH-48A for Supplementary Load 0

Combination

CA-04-127 Determination of Maximum Allowable Outside Air Temperature 13

CA-04-133 AOV Component Calculation, AO-2-86A, B, C, D 15

CA-04-145 EDG Ventilation: Cooling Load and Air Flow Determination 12

CA-04-146 Design of Supports for HVAC Ductwork in DGB (Diesel 1

Generator Building)

CA-06-054 Evaluation of SRP Module J.3 due to Replacement of MO-2009 1

CA-06-081 Evaluation of New Support SR-852 Next to TWH-87 0

CA-06-104 480V MCC to Motor Terminal Voltage Drop 1

CA-07-047 Residual Heat Removal System MOV Performance Analysis 0

CA-08-047 D10 Battery Charger Mounting Evaluation 0

3 Attachment

CALCULATIONS

Number Description or Title Revision

CA-08-082 Evaluation of SR-537 0

CA-08-083 Evaluation of SWH-150 0

CA-08-157 Combined AC Model Database 0

CA-09-120 Evaluation of SRP Module F.2 for 120% EPU 0

CA-09-148 Cycle 25 Monticello Specific EOP Calculations, Appendix B.15 0

CA-09-177 Sizing of the Orifice in the SW lines to the RHR Pump Motor 0

Coolers

CA-88-003 RHR Pump Minimum Flow Calculation for IE Bulletin #88-04 0

Response

CA-90-018 Determination of Acceptance Criteria for RHR Pump 01/09/91

Surveillance Testing and Verification of Adequate LPCI Flow

Under Four and Two Pump Operation for SRI No.90-002 and

GE Report NEDC-31786P, Respectively

CA-90-023 Minimum Allowable Fuel Oil Storage Tank Level 2

CA-91-001 125V DC Fault Current 1

CA-91-006 125V DC Battery Charger Sizing 1

CA-91-012 125V DC SBO Load Profile Study 2

CA-91-063 Determination of Uncertainty in RHR Pump D/P Measurement 2

During Performance of Test 0255-04-111

CA-91-086 AC Load Study 1AR No.10 XFMR, LOCA Load 2 Core Spray 2

Pumps Starting

CA-91-092 Plant Fault Study, 2R XFMR, 2RS Reactor In-Line 7

CA-91-093 Plant Fault Study, 1AR Transformer Feed From #10 4

Transformer

CA-91-098 AC Load Study Short Circuit Contribution Database 7

CA-91-100 AC Load Study Transformer Database 7

CA-92-024 125 V DC Battery Sizing Calculation 5

CA-92-036 RHR Pump Room Temperature Analysis 7

CA-92-038 Determination of RHRSW Instrumentation Inaccuracies 4

CA-92-214 RHR System Motor Operated Valve Functional Analysis 11

CA-92-223 Analysis for Mod. 92Q520 Control Circuit Voltage Drop 0

CA-92-224 Emergency Diesel Generator Loading 5

CA-92-250 Project 92Q600 Cable Sizing Calculation 0

CA-92-295 Protection Settings for New LC-103 480V Switchgear Lineup 1

CA-92-299 Stem Thrust Assessment of 16 A/D Globe Valves: MO-2012 1

and MO-2013

CA-93-023 Instrument/Control Setpoint Calculation: Main Air Supply To 1

Outboard MSIV

CA-93-066 AC Loads Study, Degraded Voltage Setpoint, 1R XFMR, LOCA 6

Load

CA-93-084 Hydrogen Generation of No. 11 and No. 12 Battery Rooms 1

4 Attachment

CALCULATIONS

Number Description or Title Revision

CA-94-017 Calculation of Alternate Nitrogen System Operability Leakage 6

Criteria

CA-94-037 Calculation of Inboard MSlV Post-LOCA Closing Forces and 5

Pneumatic Pressure Requirements

CA-94-084 Determination of MSIV Stroke Time Acceptance and Setpoint 7

Bands

CA-94-094 MCC Starter Coil Pickup Voltages and Maximum Cable Lengths 0

CA-94-142 Stem Thrust Assessment of 10 A/D Gate Valves: MO-2020, 2

MO-2021, MO-2022, and MO-2023

CA-95-076 Calculation for ASCO Pressure Switch Set Points 0

CA-95-117 Stem Thrust Assessment of 12 A/D Gate Valves: MO-2006 0

and MO-2007

CA-96-012 Minimum Thrust Required to Close Inboard MSIVs and to 2

Obtain Leak Tight Seat

CA-96-015 MO-2006 and MO-2007 Pressure Locking Interim Calc. 0

CA-96-113 Temperature of RHR Rooms During DBA LOCA 0

CA-97-090 AC Voltage Study, 480V Voltage Determination During Diesel 2

LOOP/LOCA ECCS CS Start Test Conditions

CA-97-157 RHR Room Temperature Response to LOCA 2

CA-98-008 Environmental Qualification (50.49) of Automatic Valve 5

Company Air Control Assembly

CA-98-032 Environmental Qualification (50.49) of Namco EA740/EA180 11

Series Limit Switches

CA-98-033 Environmental Qualification (50.49) of Namco EC210 Quick 15

Disconnects

H20-M-001 Cooling Coil Performance at Minimum Water Flow Rate 2

CORRECTIVE ACTION PROGRAM DOCUMENTS

Number Description or Title Date

00633248 Supports SWH-42,47 Stanchion do Not Touch Baseplate 02/23/03

00839220 EDG Fuel Oil Transfer System is Vulnerable to Operator Error 06/01/05

01002620 Packing has Slight Leak on MO-2007 11/02/05

01002810 Potential NRC Violation for Sprinkler Obstruction in EDG room 11/03/05

01046881 MO-2007 12 RHR Discharge to Torus has an Oil Leak 08/28/06

01051775 New Fuel Shipment is Ultra-Low-Sulfur-Diesel 09/22/06

01051793 CST Line Above Torus has Wood Wedge Between Wall and 09/22/06

Pipe

01056182 Motor Terminal Voltages Could Drop Below 90% Rated 10/17/06

01061972 Diesel Fuel Oil Pump Disch Pressure Shows Lowering Trend 11/15/06

01103509 Station Transformer Annual Insulating Oil Dielectric Test 07/25/07

01119969 Inconsistencies Between B Manual and TS Bases 12/04/07

5 Attachment

CORRECTIVE ACTION PROGRAM DOCUMENTS

Number Description or Title Date

01130761 CV-1729 Not Controlling at 7000 gpm 03/12/08

01135806 Support SWH-42 Stanchion Does Not Touch Baseplate 04/28/08

01135808 Support SWH-47 Stanchion Does Not Touch Baseplate 04/28/08

01146078 High Voltage Spurious Trips-New D10 Charger Bench Test 07/31/08

01149737 Potential Safety Hazard from MO-2013 Insulation Lagging 09/05/08

01151194 Packing Leak on MO-2007, RHR B Outboard Torus Isolation 09/19/08

01152938 11 and 12 EDG-ESW IST Pump Trends 09/30/08

01155238 Station Evaluation Of Industry And Internal OE 10/14/08

01160133 Implementation of Time Critical Operator Actions Ctrl Process 11/21/08

01164976 Unplanned TS Action Entry due to RCIC Inoperability 01/10/09

01170994 Supports SWH-42,47, Inadequate Documentation of Issue 02/26/09

01171900 EC 12361 MO-2020 (DW Spray)-ECN Error 03/05/09

01173431 AO-2-86D Has Excessive Seat Leakage 03/18/09

01175465 OSP-DOL-0543, Rev 2 03/28/09

01186755 NRC Question Re Core Spray Flow Throttling for ASDS 06/24/09

01187408 Documentation of 1AR Transformer Operability Questions 06/29/09

01190108 Station Transformer Annual Insulating Oil Test results 07/20/09

01193840 Error/Wrong Input in Calc 06-104, 480V Motor Term Voltage 08/17/09

01194290 Error in Voltage Drop Calculation CA-06-104 08/24/09

01195281 Cracks in Concrete Pad of 1AR Transformer 08/26/09

01195865 Minimum Voltage for 2R to 1R Auto Transfer 08/31/09

CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED AS A RESULT OF THE

INSPECTION

Number Description or Title Date

01193742 Panel D21 Is Missing Two Bolts 08/15/09

01196091 Minor References List Error in EPU Calc.08-082 09/01/09

01196110 Bus Bracing Ratings Not Documented 4.16 kV Switchgear 09/02/09

and 480V Load Centers

01196224 EPU Calcs Not Listed on ADLs and Stored As Required 09/02/09

01196345 ESW Supports SWH-42 and SWH-47 are Not Functional 09/03/09

01196371 CT Ratio Discrepancy Between Relay Cards and Drawing 09/11/09

01196394 CA-94-037 Calculation Error 09/03/09

01196433 Identified Condition Adverse to Quality Not Corrected 09/03/09

01196448 Discrepancy Between Drawings of MCC 142 for B4208 09/04/09

01196451 EDG Base Tank Volume Calculation CA 90-023 09/03/09

01196495 EDG Fuel Oil XRF Pump Not Included in the EDG Loading 09/04/09

01196513 Stanchion for PI-1982 Only has 4 Bolts Installed 09/04/09

01196531 NE-93194-11 and NE-36438-9 Show Incorrect HP for P- 09/04/09

77/MTR

6 Attachment

CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED AS A RESULT OF THE

INSPECTION

Number Description or Title Date

01196811 Apparent Deficiency in Documentation of EDG Air Start 09/10/09

01196861 CA-00-104, Intake Structure Min Water Level Is Incorrect 09/08/09

01197333 No Documentation for Fuse in D11-22 (Computer Input) 09/17/09

01197339 Inconsistency in Fuse Sizing Div 1 125 VDC D111-04 Circuit 09/11/09

01197431 Undocumented Assumption in Calculation 06-104 09/12/09

01197447 Evaluation of IN 2007-09 Did Not Address Issue 09/03/09

01197792 CA-91-100 2R Impedances Are For 50 MVA, Not 25 MVA 09/15/09

01197923 Mod Did Not Assess Impact to T-44 Tank 09/15/09

01197979 Typo in Ops Man B-09.06-02 for 2R Transformer 09/16/09

01198021 Errors on MCC Schedule Drawings 09/16/09

01198033 CA-07-047 Not Updated with 2003 Test Data Correctly 10/16/09

01198046 D111 Panel Lower Right Side Wing Panel Has Gap 09/16/09

01198056 Incorrect Pressure Term Used in CA-92-214 10/16/09

01198090 Revise CA-92-224 to Remove 2750 kW and 3050 kW 09/16/09

Ratings

01198227 Update CA-03-111 to Incorporate Enhancement 09/17/09

01198266 Weak Link Labeled Incorrectly in MOV Database 09/17/09

01198415 Service Water Line Support May Not Meet Code 09/18/09

Requirements

01198495 CA-94-037 Has Error 09/18/09

01198855 50.59 Did Not Evaluate Use of Age Hardened Concrete 09/21/09

01199213 Ref. for EPU CA-08-081 Documented Incorrectly 09/23/09

01199540 Calc Accept Criteria for Suprt May Not Meet Code Reqmts 09/25/09

01199542 New Crack Not Identified in PUSAR 09/25/09

01199606 Error in 50.59 Screening for EPU EC 11126 09/25/09

01199936 Voltage Drop Evaluation for RHR/RHRSW ASCO Solenoid 09/28/09

Vlvs

01199999 Minor Error in Line Designation Table Drwg ND-57664-1 09/28/09

01200120 Procedure Question Operation of RHRSW at 6000 gpm 09/29/09

01200170 Pickup Voltage on 480V MCC Contactors Not Periodically 09/29/09

Tested

01200215 Incorrect Input was Used in Calculation 09/29/09

01200265 Incorrect Concrete Strength Chosen for 2 EPU Calcs 09/29/09

01200397 Information Lacking on 09-120 to Assess Suitability 09/30/09

01200432 Incorrect Input Used in Calculation No.09-122 09/30/09

01200487 Apparent Conflict Between Calculations 09/30/09

01200539 Revise Procedure 4847-PM (GE 7700 Line MCC) to Require 10/01/09

Contactor Pickup Voltage Test

01200540 Revise Procedure 4027-PM (Klockner-Moeller MCC B34 10/01/09

and B44) to Require Contactor Pickup Voltage Test

7 Attachment

CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED AS A RESULT OF THE

INSPECTION

Number Description or Title Date

01200546 Relays 10A-K70A and10A-K70-B are Agastat E7012AD004 10/01/09

01200682 Support SR-526 is in Contact with Insulation on RHR HX 10/01/09

01200723 Voltage at P-111B Contactor is Less Than 96 Volts 10/02/09

01200725 Instr. Uncertainty Assumed in OPR 01076631 Not Bounding 10/02/09

01202633 B4319 Contactor Failed Degraded Voltage Testing 10/15/09

DRAWINGS

Number Description or Title Revision

DNL32274 1AR Transformer Nameplate E

FSB-0503-1 Fuse Breaker Study 0

FSB-0504-1 Fuse Breaker Study 0

FSB-0507-1 Fuse Breaker Study 1

FSB-0508-1 Fuse Breaker Study 0

M-112 PandID RHR Service Water and Emergency Service Water 84

Sys

M-120 PandID Residual Heat Removal System 76

M-121 PandID Residual Heat Removal System 76

M-133 PandID Diesel Oil System 76

M-134 PandID Fire Protection System Interior Locations 70

M-143 PandID Primary Containment and Atmospheric Control 77

System

M-343 Plans and Sections, HandV Standby Diesel Generator 1/21/69

Building

M-811 PandID Service Water System and Make-Up Intake 91

Structure

NE-100346 Div I and Div II 120V Instrument AC Distribution Panel M

Schedules

NE-36347-10 No. 142-480V MCC B42 AA

NE-36347-13 No. 133 and No. 143-480V MCC B33 and B43 N

NE-36347-15 No. 134-480V MCC B34 N

NE-36347-1A 480V MCC Schedules 77

NE-36347-8 No. 133-480V MCC B33 79

NE-36394-18 Monticello Nuclear Generating Plant Emergency Service 76

Water Pumps

NE-36404-18 RHR System Auxiliary Controls L

NE-36404-18A Schematic Diagram Reactor Auxiliary Systems K

NE-36771-3 Instrument AC and Uninterruptible AC Panel Schedules 76

Y10, Y20 andY30

NE-93576 Single Line Diagram 480V MCC B34 J

8 Attachment

DRAWINGS

Number Description or Title Revision

NF-36176 Generator Aux Transformer and 4160 Volt System Buses 11 78

and 12

NF-36177 Single Line Meter and Relay Diagram 4160 Volt System 77

Buses No. 13, No. 14, No. 15, and No. 16

NF-36298-1 Electrical Load Flow One Line Diagram 81

NF-36298-2 DC Electrical Load Distribution One Line Diagram 78

NL-96763-1 Pipe Support SR-526 A

NL-96792-1 Pipe Support SR-147 A

NX-28921-3 Transformer, Class A, Core form, Outdoor, 3 Phase, 60 A

Hertz KVA-50000-Namplate Drawing

NX-7905-46-11 Elementary Diagram Residual Heat Removal System J

NX-7905-46-14E 11 RHR Containment Spray Outboard Isolation MO-2020, 76

Scheme B3339

NX-7905-46-16 Residual Heat Removal System Elementary Diagram J

NX-7905-46-17 Elementary Diagram Residual Heat Removal System 77

NX-7905-46-18 Elementary Diagram Residual Heat Removal System R

NX-7905-46-19 Elementary Diagram Residual Heat Removal System P

NX-8714-2-4 MCC B33(A) Turbine Building Elevation 971-0 H

NX-8714-2-5 No. 133-480V MCC B33 (A) Rear Essential Turbine Bldg- 78

East

NX-8714-4-2 Motor Control Center B43-B No. 143 (B) 480V H

NX-8763-23 V-AC-4 Air Cooling Unit C

NX-8763-23-1 RHR/Core Spray Pump Room Cooling Coil (V-AC-4) A

NX-9525-1 RHRSW Pumps - Vertical Pump Dimensions/Layout 76

NX-9525-18 RHRSW Pumps - Bowl Assembly A

USAR Figure Instrument AC and Uninterruptible AC Distribution System 22

8.7-1 Single Line

MISCELLANEOUS

Number Description or Title Date or

Revision

IST Stroke Time Data for MO2007, MO2013, and MO2020

14 ESW Quarterly Pump Test Data (05/22/07 - 07/22/09)

Phase 2 Risk Evaluation of CAP AR00076349, Loss of 07/14/09

Containment Overpressure Due to Spurious Operation of the

Torus and Drywell Purge and Vent Valves

4 AWI-04.05.12 Replacement of Failed Fuses 4

5828/6453-E-6 Specification for 480 Volt Motor Control Centers for the 0

Monticello Nuclear Generating Plant - Unit 1

8285 Non-Identical Fuse replacement 8

B779R3 60,000 GAL U.G. Oil TK 09/05/67

9 Attachment

MISCELLANEOUS

Number Description or Title Date or

Revision

CML Component Master List for CV-1729 Loop Components: E/P 02/21/05

1729, FT-10-97B, FI-4105, FI-10-132B, DPIC-10-130B, DPI-

10-130B, FI-7188

EC 13790 Outboard MSIV Spring Degradation Evaluation 02/23/09

EC 14932 Evaluation of Inboard MSIV Closure Margin Using Existing 09/23/09

Alternate Nitrogen System Minimum Pressure

EWI-08.15.02 Motor Operated Valve Program Engineering Standards 8

FORM 3090 IST Program Pump Data Sheet for No. 11 RHR Pump (202A) 5

Job No. 5828 Civil Structural Design Criteria for the Monticello Nuclear 1

Generating Plant-Unit 1

LER 2009-01 Containment Overpressure Not Ensured in the Appendix 09/18/09

Analysis

Letter to NRC IE Bulletin 88-04 Response 07/08/88

Letter to NRC IE Bulletin 88-04 Final Response 12/13/88

Letter To NSP Sulzer Bingham Pumps INC., (Vendor Information 11/08/88

Responding to Bulletin 88-04 for RHR Pumps)

MPS-1100 Specification for The Analysis of Piping and Pipe Support 8

Systems, NPD-M-038

MWI-3-M-2.01 AC Electrical Load Study 12

MWI-3-M-2.06 Fuse/Breaker Coordination Study and Electrical Coordination 9

NX-17496-3 MNGP Protective Relay Cards - 4KV 3

NX-17496-4 MNGP Protective Relay Cards - 480V 1

NX-7905-55 11 RHR Pump Curve Vendor Curve (No. 26653) A

NX-7905-56 11 RHR Pump Curve Vendor Test Data A

NX-7905-58 13 RHR Pump Curve Vendor Curve (No. 26680) A

NX-7905-59 Residual Heat Removal Pump 270430 A

NX-7905-70-1 11 RHRSW Pump Curve 03/31/75

NX-7905-72-1 13 RHRSW Pump Curve 03/31/75

SAR01116710 CDBI Focused Self-Assessment 12/17/08

SAR01166267 CDBI FSA Effectiveness and Inspection Readiness 07/21/09

MODIFICATIONS

Number Description or Title Date or

Revision

04Q030 EDG Ventilation System Upgrades 0

EC 11126 EPU-MOD 11-Balance of Plant Piping Support Modifications 0

EC 12361 EPU-Provide Operations the Ability to Throttle MO-2020 and 0

MO-2021

EC 14085 Security Improvements - 2009- Force on Force Project 0

EC 9419 EDG Rooms Sprinkler Modification 0

10 Attachment

MODIFICATIONS

Number Description or Title Date or

Revision

EC-11734 Modification No. 5 1AR Transformer Replacement, 4

Monticello Plant EPU Project

SRI-91-010 USAR Update Concerning Diesel Oil Day Tank 12/01/92

OPERABILITY EVALUATIONS

Number Description or Title Date

AR 01079705 LAR Required for Use of TORMIS Code Methodology 02/28/07

CA-03-124 ESW Model ESW2B-Operability Evaluation 06/24/03

CA-03-125 ESW Model ESW1B-Operability Evaluation 06/24/03

FA 010709705 Modify EDG Exhaust Silencer Lines to Restore the EDGs 10/01/09

Within a Reasonable Time Frame

OPR 1076631 Possibility of Inadequate Flow to Room Cooler 03/08/07

OPR 1101934 Operability Determination on 14 RHR Pump 07/22/07

OPR 1130761 Flow Control Problems with B RHRSW Valve CV-1729, 03/20/08

OPR 1169854-1 Outboard MSIVs Have Not Been Tested IAW Testing 02/23/09

Requirements

OPR 1199936-1 Terminal Voltage at SV-1995 and SV-1728 10/02/09

OPR1193840-1 Motors Supplied by 480V MCCs With Voltage Less than 90% 08/21/09

Rated Voltage

PROCEDURES

Number Description or Title Revision

0137-15-01 Containment Spray Loop A Isolation Valve Local Leak 10

Rate Test

0137-29 LPCI Loop B Injection Valves Local Leak Rate Test 7

0187-01 11 EDG/ESW Quarterly Pump and Valve Tests 70

0192 Diesel Fuel Quality Check 26

0255-04-IA-1-1 RHR Loop A Quarterly Pump and Valve Tests 78

0255-04-IA-1-2 RHR Loop B Quarterly Pump and Valve Tests 80

1052-03 11 Diesel Generator Auxiliary Systems Test 11

1487 Site Housekeeping Quarterly Inspection 5

4066-PM D10 Battery Charger Preventive Maintenance 0

4106-01-PM Emergency Diesel 1 Cycle Maintenance 24

4844-PM GE Thermal Overload Relay Test Procedure 20

4846-PM GE/W Molded Case Circuit Breaker Maintenance and Test 18

Procedure

4847-PM 7700 Line Motor Control Center Maintenance Procedure 19

4858-60-PM 1AR Reserve Transformer Maintenance 12

11 Attachment

PROCEDURES

Number Description or Title Revision

6-C-3 Diesel Service Oil Pump Tripped 2

6-C-6 Diesel Generator Tank T-45A Level/Flow Low 2

6-C-7 Diesel Generator Tank T-45B Level/Flow Low 3

8196 Temporary Shielding Installation 4

8900 Operation of RCIC without Electric Power 2

8-B-19 ESW Pump 11 Lo Discharge Pressure 4

8-B-20 ESW Pump 11 OL/MAN Override 4

A.6 Acts of Nature 31

B.03.04-06 Residual Heat Removal System - Figures 5

B.08.01.03-01 RHR Service Water System - Function and General 10

Description

B.08.01.03-05 RHR Service Water System-System Operation 39

B.08.04.01-05 Instrument and Service Air 10

B.08.05-05 Fire Protection - System Operation 45

B.08.11-05 Diesel Oil System 17

B.09.06-02 4.16 KV Station Auxiliary 10

B.09.13-06 Instrumentation AC and Uninterruptible AC Distribution 5

System - Figures

C.4-B.08.01.03- Loss of Instrument Air 17

A

C.4-B.09.02.B Loss of Normal Offsite Power 11

C.4-C Shutdown Outside Control Room 32

C.4-H Restoration Of Plant Loads 13

C.5.1-1000 EOP Introduction 20

C.5.1-1100 RPV Control 8

C.5-3203 Use of Alternate Injection Systems for RPV Makeup 10

C.5-3502 Containment Spray 14

FP-E-MOD-04 QF-0515a Design Input Checklist 10

FP-E-MOD-04 QF-0515b Design Input Checklist 3

FP-E-MOD-06 QF-0525 Design Description Form 2

FP-OP-COO-01 Conduct of Operations 6

FP-OP-OL-01 Operability / Functionality Determination 5

OSP-DOL-0543 Fuel Oil Receiving Quality Check 2

OSP-EDG-0540- 11 Emergency Diesel Generator 24 Month Test 2

11

OWI-03.07 Time Critical Operator Actions 0

12 Attachment

SURVEILLANCES (COMPLETED)

Number Description or Title Date or

Revision

0255-05-IA-1-1 A RHRSW Quarterly Pump And Valve Tests 05/15/09

05/27/09

06/11/09

0198-01 125 VDC Battery Capacity Test 03/31/05

4510-PM Maintenance of Onsite Batteries and Battery Chargers 01/25/07

OSP-EDG-0540-11 11 Emergency Diesel Generator 24 Month Test 04/04/09

0187-01B 11 Emergency Diesel Generator/11ESW/Monthly Pump 06/14/09

and Valve Tests 08/09/09

0187-01 11 Emergency Diesel Generator/11ESW/Quarterly 07/12/09

Pump and Valve Tests

WORK ORDERS

Number Description or Title Date

00002043 Valcor Solenoid Valve Control Assembly 12/11/01

00134749 Pre-Op Test New 125V DC Division 1 Battery Charger 08/15/08

D10

00200755 PM 4900-1, VOTES for MO2013 05/05/03

00200759 MO2020 VOTES, Inspect LS 05/16/03

00285275-01 PM 4106-01-PM, 11 EDG G-3A, Perform Electrical PM 10/09/06

00293778 P-111A, Rebuild Pump and Return to Stock 06/05/09

00306982 EC210 O Rings On Valcor Control Assembly J-box 03/14/05

00307012 11 RHR Pump Minimum Flow, (PM CV-1994 Air 03/14/05

Operator)

00307804 Make Repairs to Support SWH-47 05/28/03

00307805 Make Repairs to Support SWH-42 05/17/03

00309248 11 RHR Pump Minimum Flow, Baseline Testing on CV- 03/15/05

1994

00311664 D-10, 125VDC Charger for 11 Battery 480V Supply 07/21/04

00311666 MO-1427, RBCCW INL to DW Cooler V-CC-1 480V 03/29/05

Supply

00311667 MO-1429, RBCCW INL to DW Cooler V-CC-3 480V 03/29/05

Supply

00311670 P-88B, ECCS Area Drain Pump B (480V Supply) 05/26/04

00336541 Monitor ADS Pneumatic Supply 01/22/08

00343696-01 PM 4030-01-PM, 11 Diesel Generator Control Panel C- 03/07/09

91/93

00343716 CV-1994, Valve Position Indication Test, (0255-03-IA- 05/27/09

2B)

00343738 CV-1728 Diagnostic Test: Post Outage PM work. 04/04/09

00343827-01 PM 4851-12 (52-304) MCC-133A Feeder Breaker 03/09/09

Cubicle

13 Attachment

SURVEILLANCES (COMPLETED)

Number Description or Title Date or

Revision

00343889 Limit Switch (AO-2-86D) 04/23/09

00343994 PM 4058-6 (B RHR RM Air Cooling Unit V-AC-4) 07/24/09

00344785 Monitor ADS Pneumatic Supply 03/18/08

00349292 Monitor ADS Pneumatic Supply 05/14/08

00352042 Perform Coupling Inspection of P-77 07/15/08

00352697 Perform Coupling Inspection of P-11 07/08/08

00354013 D10 125 VDC Charger 24 Month Capacity Test 06/13/08

00354414 MO2007 Rebuild Actuator and Perform 4900-01-PM 04/04/09

00355172 CV-1728, Perform IandCPM 7070 (A)on RSW-1 09/15/08

Instruments

00356497 CV-1729, Comprehensive 11 RHRSW Pump and Valve 09/10/08

Tests

00356504 11 EDG Auxiliary Systems Test 03/29/09

00356593-01 PM 4109-01-PM, 11 EDG 12 Year Maintenance 03/07/09

00356627 Battery 11 Modified Performance Test 03/29/09

00359064-11 Perform Acceptance Testing of 1AR Transformer 07/24/09

00362759 025504-III-3A Comprehensive 13 RHR Pump and 12/09/08

Valve Test

00362760 CV-1729, Comprehensive 11 RHRSW Pump and Valve 12/26/08

Tests

00363012 Division 1 LPCI Pump Discharge Flow-Low Bypass 12/15/08

Channel Calibration, (ISP-RHR-0547-01)

00363156 0255-04-IA-11 RHR Loop A QRTRLY Pump and Valve 07/20/09

Test

00366681 CV-1729, Comprehensive 11 RHRSW Pump and Valve 03/13/09

Tests

00374280 11 and 12 125 VDC Battery Weekly Surveillance 06/30/09

00375976 11 EDG/ESW Quarterly Pump and Valve Tests 07/10/09

00377539 11 EDG/ESW Quarterly Pump and Valve Tests 08/09/09

00377780-01 Perform Major PM on Spare Breaker LCB-037 04/04/09

00379154 0255-04-IA-11 RHR Loop A QRTRLY Pump and Valve 12/09/08

Test

00385117 Diesel Fuel Oil Receiving Quality Check 05/19/09

00403241 11 RHR Pump Minimum Flow, (CV-1994 failed to 09/17/04

Operate)

09906286 Measure Air Flow on V-AC-4 and V-AC-5 07/01/09

14 Attachment

LIST OF ACRONYMS USED

ADAMS Agencywide Document Access Management System

AISC American Institute of Steel Construction

AR Action Request

ASME American Society of Mechanical Engineers

CDBI Component Design Bases Inspection

CFR Code of Federal Regulations

CPT Control Power Transformer

DRP Division of Reactor Projects

DRS Division of Reactor Safety

EC Engineering Change

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator

ESW Emergency Service Water

GE General Electric

GL Generic Letter

IEEE Institute of Electrical and Electronic Engineers

IN Information Notice

IMC Inspection Manual Chapter

IR Inspection Report

IST Inservice Test

kV Kilovolt

LC Load Center

LOCA Loss of Coolant Accident

LOOP Loss of Off-site Power

MCC Motor Control Center

MOV Motor-Operated Valve

MSIV Main Steam Isolation Valve

NCV Non-Cited Violation

NEMA National Electrical Manufactures Association

NPSH Net Positive Suction Head

NRC U.S. Nuclear Regulatory Commission

PARS Publicly Available Records

PM Preventative Maintenance

PRA Probabilistic Risk Assessment

psig Pounds Per Square Inch Gauge

RHR Residual Heat Removal

RHRSW Residual Heat Removal Service Water

RIS Regulatory Information Issue

SBO Station Blackout

SDP Significance Determination Process

SER Safety Evaluation Report

SPAR Standardized Plant Analysis Risk

SV Solenoid Valve

TOL Thermal Overload

TS Technical Specification

USAR Updated Safety Analysis Report

URI Unresolved Item

Vac Volts Alternating Current

Vdc Volts Direct Current

15 Attachment

T. OConnor -2-

In accordance with 10 CFR 2.390 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 Public Document

Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide

Documents Access and Management System (ADAMS), accessible from the NRC Web site at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Ann Marie Stone, Chief

Engineering Branch 2

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure: Inspection Report 05000263/2009007

w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ

DISTRIBUTION:

Susan Bagley

RidsNrrDorlLpl3-1 Resource

RidsNrrPMMonticello

RidsNrrDirsIrib Resource

Cynthia Pederson

Steven Orth

Jared Heck

Allan Barker

Carole Ariano

Linda Linn

DRPIII

DRSIII

Patricia Buckley

Tammy Tomczak

ROPreports Resource

DOCUMENT NAME: G:\DRS\Work in Progress\MON 2009-007 CDBI DRS IR AXD.doc

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

NAME ADunlop:ls DPassehl AMStone

DATE 12/9/09 12/10/09 01/06/10

OFFICIAL RECORD COPY