Text

IR 05000263/2011004; 07/01/2011 - 09/30/2011; Monticello Nuclear Generating Plant; Surveillance Testing
	ML11305A174
Person / Time
Site:	Monticello
Issue date:	11/01/2011
From:	Riemer K R; NRC/RGN-III/DRP/B2
To:	O'Connor T J; Northern States Power Co
References
	IR-11-004
	Download: ML11305A174 (59)
	v • d • e

November 1, 2011

Mr. Timothy Site Vice President Monticello Nuclear Generating Plant Northern States Power Company, Minnesota 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT INTEGRATED INSPECTION REPORT 05000263/2011004

Dear Mr. O'Connor:

On September 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Monticello Nuclear Generating Plant. The enclosed report documents the inspection findings, which were discussed on October 6, 2011, 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 Commission's 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, one NRC-identified finding of very low safety significance was identified. The finding involved a violation of NRC requirements. However, because of its very low safety significance, and because the issue was entered into your corrective action program, the NRC is treating the issue as a non-cited violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy. If you contest the subject or severity of this NCV, 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 cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Monticello Nuclear Generating Plant. 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 document system (ADAMS).

ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ Kenneth Riemer, Chief Branch 2 Division of Reactor Projects Docket Nos. 50-263;72-058 License No. DPR-22

Enclosure:

Inspection Report 05000263/2011004

w/Attachment:

Supplemental Information cc w/encl: Distribution via ListServ Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket Nos: 50-263;72-058 License No: DPR-22 Report No: 05000263/2011004 Licensee: Northern States Power Company, Minnesota Facility: Monticello Nuclear Generating Plant Location: Monticello, MN Dates: July 1 through September 30, 2011 Inspectors: S. Thomas, Senior Resident Inspector P. Voss, Resident Inspector K. Stoedter, Prairie Island Senior Resident Inspector P. Cardona-Morales, Reactor Engineer A. Scarbeary, Reactor Engineer J. Corujo-Sandin, Reactor Engineer N. Shah, Project Engineer M. Phalen, Senior Health Physicist, DRS S. Bell, Health Physicist, DRS M. Learn, Lead ISFSI Inspector, DNMS

Approved by: K. Riemer, Chief Branch 2 Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000263/2011004; 07/01/2011 - 09/30/2011; Monticello Nuclear Generating Plant; Surveillance Testing. This report covers a three-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. One Green finding was identified by the inspectors. The finding was considered a non-cited violation (NCV) 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reac tor Oversight Process," Revision 4, dated December 2006.

A. NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," when the licensee failed to follow the quarterly emergency diesel generator (EDG) surveillance procedure during testing of the EDG air start system. Specifically, the licensee failed to follow a procedural step that involved in-service testing of a check valve in the EDG air start system that, if degraded, could allow air to bleed out of the starting air tanks which are required for diesel generator operability. The licensee entered this issue into their corrective action program (CAP), and corrective actions for this issue included suspension of the test, performance of a Human Performance Investigation Team review, and disqualification of the individual performing the test. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having work practices components, and involving aspects associated with using human error prevention techniques during performance of work activities. H.4(a)

The inspectors determined that the licensee's failure to follow their EDG surveillance procedure was a performance deficiency, because it was the result of the failure to meet a requirement; the cause was reasonably within the licensee's ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, and determined that the issue was more than minor because the performance deficiency was associated with the Human Performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). As a result, this finding was evaluated under the Mitigating Systems Cornerstone. The inspectors applied IMC 0609,

Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," to this finding. The inspectors utilized Column 2 of the Table 4a worksheet to screen the finding. The finding was determined to have very low safety significance because the inspectors answered "No" to all five questions. (Section 1R22)

B. Licensee-Identified Violations

No violations were identified.

REPORT DETAILS

Summary of Plant Status

Monticello operated at approximately 100 percent power for a majority of the inspection period, with minor power adjustments to support routine testing and control rod manipulations, with two notable exceptions. On August 6, 2011, reactor power was reduced to approximately 45 percent to establish the appropriate plant conditions to repair a bearing on the 11 condensate pump motor. Reactor power was returned to approximately 100 percent on August 8, 2011. On August 27, 2011, additional repairs on the 11 condensate pump motor bearing required a power reduction to approximately 45 percent power. Reactor power was returned to approximately 100 percent on August 30, 2011, and remained at approximately full power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

Main article: IP 71111.04

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

high pressure coolant injection (HPCI) with reactor core isolation cooling (RCIC) out-of-service for maintenance;

'B' control room ventilation (CRV) and emergency service water (ESW) with 'A' CRV out-of-service for ESW piping modification; and

core spray train 'A' during work on core spray train 'B.' The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Safety Analysis Report (USAR), Technical Specification (TS) requirements, outstanding work orders (WOs), condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies.

The inspectors also verified that the licensee had properly identified and resolved

equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program (CAP) with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

4 Enclosure These activities constituted three partial system walkdown samples as defined in Inspection Procedure (IP) 71111.04-05.

b. Findings

No findings were identified.

.2 Semi-Annual Complete System Walkdown

a. Inspection Scope

During the week of August 29, 2011, the inspectors performed a complete system alignment inspection of the 11 emergency diesel generator (EDG) to verify the functional capability of the system. This system was selected because it was considered both safety significant and risk significant in the licensee's probabilisti c risk assessment. The inspectors walked down the system to review mechanical and electrical equipment line ups; electrical power availability; system pressure and temperature indications, as appropriate; component labeling; component lubrication; component and equipment cooling; hangers and supports; operability of support systems; and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of a sample of past and outstanding WOs was performed to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the CAP database to ensure that system equipment alignment problems were being identified and appropriately resolved. Documents reviewed are listed in the Attachment to this report. These activities constituted one complete system walkdown sample as defined in IP 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

Main article: IP 71111.05

.1 Routine Resident Inspector Tours

Main article: IP 71111.05Q

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

Fire Zone 1-A (12 residual heat removal (RHR) and core spray pump room);

Fire Zone 1-B (No. 11 RHR and core spray pump room);

Fire Zone 1-C (RCIC room);

Fire Zone 1-E (HPCI room); and

Fire Zone 23-B (intake corridor). The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability, maintained 5 Enclosure passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensee's fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plant's Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plant's ability to respond to a security event.

Using the documents listed in the Attachment to this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensee's CAP.

These activities constituted five quarterly fire protection inspection samples as defined in IP 71111.05-05.

b. Findings

No findings were identified.

.2 Annual Fire Protection Drill Observation

Main article: IP 71111.05A

a. Inspection Scope

On September 21, 2011, the inspectors observed a fire brigade activation during a scheduled fire brigade drill. Based on this observation, the inspectors evaluated the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies; openly discussed them in a self-critical manner at the drill debrief; and took appropriate corrective actions. Specific attributes evaluated were:

proper wearing of turnout gear and self-contained breathing apparatus;

proper use and layout of fire hoses;

employment of appropriate fire fighting techniques;

sufficient firefighting equipment brought to the scene;

effectiveness of fire brigade leader communications, command, and control;

search for victims and propagation of the fire into other plant areas;

smoke removal operations;

utilization of pre-planned strategies;

adherence to the pre-planned drill scenario; and

drill objectives. Documents reviewed are listed in the Attachment to this report. These activities constituted one annual fire protection inspection sample as defined in IP 71111.05-05.

b. Findings

No findings were identified.

1R06 Flooding

Main article: IP 71111.06

.1 Underground Vaults

a. Inspection Scope

The inspectors selected underground bunkers/manholes subject to flooding that contained cables whose failure could disable risk-significant equipment. The inspectors

determined that the cables were not continuously submerged, that splices were intact, and that appropriate cable support structures were in place. In those areas where

dewatering devices were used, such as a sump pump, the device was operable and level alarm circuits were set appropriately to ensure that the cables would not be submerged. In those areas without dewatering devices, the inspectors verified that drainage of the area was available, or that the cables were qualified for submergence conditions. The inspectors also reviewed the licensee's corrective action documents with respect to past submerged cable issues identified in the CAP to verify the adequacy of the corrective actions. Documents reviewed are listed in the Attachment to this report. The inspectors performed a walkdown and/or document review of the following underground bunkers/manholes subject to flooding:

2R transformer manhole NMH313 and selected independent spent fuel storage facility installation (ISFSI) manholes. This inspection constituted one underground vaults sample as defined in IP 71111.06-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

Main article: IP 71111.11

.1 Resident Inspector Quarterly Review

Main article: IP 71111.11Q

a. Inspection Scope

On August 1, 2011, the inspectors observed a crew of licensed operators in the plant's simulator during licensed operator requalification examinations to verify that operator

performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

licensed operator performance;

crew's clarity and formality of communications;

ability to take timely actions in the conservative direction;

prioritization, interpretation, and verification of annunciator alarms;

correct use and implementation of abnormal and emergency procedures;

control board manipulations;

oversight and direction from supervisors; and

ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

7 Enclosure The crew's performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report. This inspection constituted one quarterly licensed operator requalification program sample as defined in IP 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

.1 Routine Quarterly Evaluations

Main article: IP 71111.12Q

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk-significant systems:

11 condensate pump repeat motor thrust bearing failure; and

intake structure fire protection sprinkler piping blockage.

The inspectors reviewed events, such as where ineffective equipment maintenance had resulted in valid or invalid automatic actuations of engineered safeguards systems, and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

implementing appropriate work practices;

identifying and addressing common cause failures;

scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

characterizing system reliability issues for performance;

charging unavailability for performance;

trending key parameters for condition monitoring;

ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2), or appropriate and adequate goals and corrective actions for systems classified as (a)(1). The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report. This inspection constituted two quarterly maintenance effectiveness samples as defined in IP 71111.12-05.

8 Enclosure

b. Findings

The intake structure fire protection sprinkler system blockage issue was evaluated in more detail as part of a special inspection. Any findings or violations identified that are associated with this issue will be documented in Inspection Report 05000263/2011010.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed

prior to removing equipment for work:

unexpected actuation of condenser bay fire header sprinkler;

degraded instrumentation lines related to No. 11 circ water pump; and

condensate/demineralizer system wetted power supplies. These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Documents reviewed are listed in the Attachment to this report. These maintenance risk assessments and emergent work control activities constituted three samples as defined in IP 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functional Assessments

Main article: IP 71111.15

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

V-AC-5 ('A' RHR room cooling unit) cooling coil leak;

V-AC-5 cooling coil leak progressive degradation;

'G' safety relief valve filter ring installation;

11 condensate pump motor thrust bearing defect;

functionality assessment for the intake structure fire protection sprinkler header blockage;

11 SBLC oil leaks; and

11 and 12 EDGs following discovery of inadequate load reject surveillance testing. The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and USAR to the licensee's evaluations to determine

whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the Attachment to this report. These operability inspections constituted seven samples as defined in IP 71111.15-05.

b. Findings

On September 29, 2011, as a result of the licensee's discovery that they had been inadequately demonstrating that both EDGs could meet a surveillance requirement associated with load rejection, the licensee declared both the 11 and 12 EDGs inoperable and entered the applicable Limiting Condition for Operation (LCO)

Action Statements. Later that evening, the licensee requested and was granted a Notice of Enforcement Discretion (NOED) to extend the Action Completion Time for

TS 3.8.1.F to five days, in order to allow time to perform the required EDG testing (NOED 11-3-001). In accordance with the NRC's NOED process, when an NOED is granted, the inspectors open an Unresolved Item (URI) to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, including enforcement action determinations, associated with the NOED. This URI will give the inspectors an opportunity to more closely evaluate the actions taken by the site when the issue was identified. In addition, the URI will allow the inspectors to review the causal factors that resulted in the site's need for an NOED.

As part of this URI, the inspectors will evaluate the underlying technical issues associated with the EDG inoperability and potential performance deficiencies associated with the inadequate surveillance procedure. To perform this evaluation, the inspectors will review information regarding why the EDG load reject surveillance procedures were inadequate to demonstrate that the equipment could meet surveillance requirement 3.8.1.7. The inspectors will also take the opportunity to reexamine the site's actions to uncover the issue, as well as the site's actions once the issue was identified.

Further inspection activities will include gathering additional information on why the procedure was inadequate, examining the site's decision-making process for the issue, 10 Enclosure and evaluating the issue for identification credit in order to determine how to appropriately disposition the issue.

Until activities associated with the tracking and resolution of the NOED and surveillance procedure inadequacies are completed, this issue will be treated as an unresolved item

(URI 000263/2011004-01; NOED for Emergency Diesel Generator Load Rejection Surveillance Requirement 3.8.1.7). Additional information associated with the EDG NOED request is documented in Section

4OA3 of this report.

1R18 Plant Modifications

Main article: IP 71111.18

.1 Plant Modifications

a. Inspection Scope

The inspectors reviewed the following modification:

replacement of cooling coil for 'A' RHR room air cooling unit. The inspectors reviewed the licensee's commercial grade dedication and item equivalency evaluations associated with the replacement of the cooling coil for

'A' RHR room air cooling unit. The inspectors, as applicable, observed ongoing and completed work activities to ensure that the modifications were installed as directed and consistent with the design control documents; the modifications operated as expected; post-modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modifications did not impact the operability of any interfacing systems. As applicable, the inspectors verified that relevant procedure, design, and licensing documents were properly updated. Lastly, the inspectors discussed the plant modification with operations, engineering, and training personnel to ensure that the individuals were aware of how the operation with the plant modification in place could impact overall plant performance. Documents reviewed in the course of this inspection are listed in the Attachment to this report. This inspection constituted one permanent plant modification sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

.1 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance (PM) activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

11 reactor protection system motor generator (MG) Set;

MO-3502 RCIC test return isolation valve;

13 ESW - system flow and operability verification after piping modification;

diesel fire pump following its failure to start on an auto-actuation signal;

V-AC-5 room cooler subsequent to replacement; and

control rod 30-47 PM test after CRD-138 check valve repair. These activities were selected based upon the SSCs ability to impact risk. The inspectors evaluated these activities for the following (as applicable): the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test

documentation was properly evaluated. The inspectors evaluated the activities against TSs, the USAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with PM tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the Attachment to this report. This inspection constituted six PM testing samples as defined in IP 71111.19-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

Main article: IP 71111.22

.1 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural

and TS requirements:

12 EDG/12 ESW quarterly pump and valve tests (in-service test);

reactor building to torus vacuum breaker operability check (routine);

11 EDG/11 ESW quarterly pump and valve tests (in-service test); and

11/12 SBLC quarterly pump and valve tests (routine). The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

did preconditioning occur;

were the effects of the testing adequately addressed by control room personnel or engineers prior to the commencement of the testing;

were acceptance criteria clearly stated, demonstrated operational readiness, and consistent with the system design basis;

plant equipment calibration was correct, accurate, and properly documented;

as-left setpoints were within required ranges; and the calibration frequency was in accordance with TSs, the USAR, procedures, and applicable commitments;

measuring and test equipment calibration was current;

test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;

test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored

where used;

test data and results were accurate, complete, within limits, and valid;

test equipment was removed after testing;

where applicable for IST activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers Code, and reference values were consistent with the system design basis;

where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;

where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure;

where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;

prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

equipment was returned to a position or status required to support the performance of its safety functions; and

all problems identified during the testing were appropriately documented and dispositioned in the CAP. Documents reviewed are listed in the Attachment to this report. This inspection constituted two routine surveillance testing samples and two IST samples as defined in IP 71111.22, Sections -02 and -05.

b. Findings

Introduction The inspectors identified a finding of very low safety significance and non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"

when the licensee failed to follow the quarterly EDG surveillance procedure during testing of the EDG air start system. Specifically, the licensee failed to follow a

procedural step that involved IST of a check valve in the EDG air start system that, if degraded, could allow air to bleed out of the starting air tanks which are required for diesel generator operability.

Description

On July 11, 2011, the licensee performed the EDG 11 and ESW 11 pump and valve quarterly surveillance test. The first section of the surveillance procedure involved testing the EDG air start system. This section included testing of the air start compressors, air lines, and air line valves, among other components.

In this section of the procedure, the licensee was required to perform an IST of check valve GSA 32-1. Precaution 7 of Surveillance Procedure 0187-01 highlighted the importance of the IST steps, stating that "steps identified with an "IST" contain acceptance criteria that is established to prove the operational readiness of pumps and valves in the Inservice Testing Program. If the step cannot be satisfactorily completed or the measured value is outside the acceptable range, the Control Room Supervisor must be immediately notified. The Control Room Supervisor must complete Form 3107 (INSERVICE TEST DEVIATION FROM CRITERIA CONTROL ROOM SUPERVISOR'S IMMEDIATE ACTION) which initiates the corrective action process and documents the

operability determination."

Check Valve GSA-32-1, 11 EDG No. 1 air dryer 11 outlet check, was located in the air line between the 11 EDG No. 1 air compressor and the air receiver tanks for the 11 EDG No. 1 air start system. In preparation for this portion of the surveillance, the individual performing the test performed Step 6 in the procedure. Step 6 required the worker to open valve GSA 29-1, 11 EDG No. 1 air start dryer blowdown, to bleed the air pressure off the section of line between the air compressor and check valve GSA-32-1.

Following Step 6, a procedural note highlighted the importance of the next step; the note stated, "STEP 7 performs the close exercise test of GSA-32-1." Step 7 was also labeled as an "IST" step.

Step 7 of Procedure 0187-01 stated, "AFTER air pressure is bled off, THEN verify GSA-32-1, 11 EDG No. 1 AIR DRYER OUTLET CHECK, is CLOSED by no air flow from the blowdown line." However, instead of checking the blowdown line for the check valve being testing, located near GSA-32-1, the individual performing the test went and checked the blowdown line located on the opposite side of the air receiver tanks.

The individual signed off the step and left to proceed onto the next step. However, the inspectors observed that a closed valve located immediately upstream of the blowdown line checked by the individual was labeled GSA-31-2, and questioned whether the step had been performed correctly.

After being questioned by the inspector, the individual returned to the GSA-31-2 blowdown line, recognized the error, checked the blowdown line located immediately next to the GSA-31-2 line, and again proceeded onto the next step. The inspector noted that this line had a valve labeled GSA-31-1, and questioned the individual a second time. The individual returned, and again recognized his error. Finally, the inspector called his attention back to the check valve labeled "GSA-32-1," and questioned the location of the appropriate blowdown line to check. At this point the individual was able to identify the correct location of the blowdown line for this valve.

Analysis The inspectors determined that the licensee's failure to follow the EDG 11 surveillance procedure was a performance deficiency because it was the result of the failure to meet a requirement; the cause was reasonably within the licensee's ability to foresee and correct; and should have been prevented. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having work practices components, and involving aspects associated with using human error prevention techniques during performance of work activities. H.4(a)

The inspectors screened the performance deficiency per IMC 0612, "Power Reactor Inspection Reports," Appendix B, and determined that the issue was more than minor because the performance deficiency was associated with the Human Performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). As a result, this finding was evaluated under the Mitigating Systems Cornerstone. The inspectors applied IMC 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," to this finding. The inspectors utilized Column 2 of the Table 4a worksheet to screen the finding. The finding was determined to have very low safety significance because the inspectors answered 'No' to all five questions.

Enforcement Title 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affect ing quality shall be prescribed by documented procedures, of a type appropriate to the circumstances, and shall be accomplished in accordance with these procedures. Contrary to this requirement, on July 11, 2011, the licensee failed to follow Step 7 of Procedure 0187-01, "11 EDG/11 ESW Quarterly Pump and Valve Tests," a procedure affecting quality, during testing of the EDG air start system. Specifically, the licensee failed to follow a procedural step that involved the IST of a check valve in the EDG air start system that, if degraded, could allow air to bleed out of the starting air tanks that are required for diesel generator operability.

Because the violation was of very low safety significance and was entered into the licensee's corrective action program (CAP 1294118), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 05000263/2011004-02; Failure to Follow Emergency Diesel Generator Quarterly Surveillance Procedure).

1EP6 Drill Evaluation

Main article: IP 71114.06

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on July 12, 2011, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the Technical Support Center and the simulated control room to determine whether the event classification, notifications, and protective 15 Enclosure action recommendations were performed in accordance with procedures. The inspectors also attended the licensee drill critique to compare any inspector-observed weaknesses with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the CAP. As part of the inspection, the inspectors reviewed the drill package associated with the drill. This emergency preparedness drill inspection constituted one sample as defined in IP 71114.06-05.

b. Findings

No findings were identified.

.2 Training Observation

a. Inspection Scope

The inspectors observed a simulator training evolution for licensed operators on August 1, 2011, which required emergency plan implementation by a licensee operations crew. This evolution was planned to be evaluated and included in performance indicator data regarding drill and exercise performance. The inspectors observed event classification and notification activities performed by the crew. The inspectors also attended the post-evolution critique for the scenario. The focus of the inspectors' activities was to note any weaknesses and deficiencies in the crew's performance and ensure that the licensee evaluators noted the same issues and entered them into the corrective action program. As part of the inspection, the inspectors reviewed the scenario package associated with the simulator exercise. This inspection of the licensee's training evolution with emergency preparedness drill aspects constituted one sample as defined in IP 71114.06-05.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

Main article: IP 71124.01

The inspection activities supplement those documented in Inspection Report 05000263/2011003, and constitute one complete sample as defined in IP 71124.01-05.

.1 Contamination and Radioactive Material Control (02.04)

The inspectors determined whether post-job reviews were conducted and, if identified, problems were entered into the licensee's CAP.

b. Findings

No findings were identified.

.2 Verification of Dose Estimates and Exposure Tracking Systems (02.03)

a. Inspection Scope

The inspectors reviewed the assumptions and basis (including dose rate and man-hour estimates) for the current annual collective exposure estimate for reasonable accuracy for select ALARA work packages. The inspectors reviewed applicable procedures to

determine the methodology for estimating exposures from specific work activities and the intended dose outcome. The inspectors evaluated whether the licensee had established measures to track, trend and, if necessary, to reduce occupational doses for ongoing work activities. The inspectors assessed whether trigger points or criteria were established to prompt additional reviews and/or additional ALARA planning and controls. The inspectors evaluated the licensee's method of adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work were encountered. The inspectors assessed whether adjustments to exposure estimates (intended dose) were based on sound radiation protection and ALARA principles or if they were just adjusted to account for failures to control the work. The inspectors evaluated whether the frequency of these adjustments called into question the adequacy of the original ALARA planning process.

b. Findings

No findings were identified.

.3 Source Term Reduction and Control (02.04)

a. Inspection Scope

The inspectors used licensee records to determine the historical trends and current status of significant tracked plant source terms known to contribute to elevated facility aggregate exposure. The inspectors assessed whether the licensee had made allowances or developed contingency plans for expected changes in the source term as the result of changes in plant fuel performance issues or changes in plant primary chemistry.

b. Findings

No findings were identified.

18 Enclosure

2RS6 Radioactive Gaseous and Liquid Effluent Treatment

Main article: IP 71124.06

No findings were identified.

Discharges made with inoperable effluent radiation monitors, or unmonitored leakages were reviewed to ensure that an evaluation was made of the discharge to satisfy 10 CFR 20.1501 so as to account for the source term and projected doses to the public.

b. Findings

No findings were identified.

.6 Groundwater Protection Initiative Implementation (02.06)

a. Inspection Scope

The inspectors reviewed monitoring results of the Groundwater Protection Initiative to determine if the licensee has implemented its program as intended, and to identify any anomalous results. For anomalous results or missed samples, the inspectors assessed whether the licensee has identified and addressed deficiencies through its CAP. The inspectors reviewed identified leakage or spill events and entries made into 10 CFR 50.75

(g) records. The inspectors reviewed evaluations of leaks or spills, and reviewed any remediation actions taken for effectiveness. The inspectors reviewed onsite contamination events involving contamination of ground water and assessed whether the source of the leak or spill was identified and mitigated. For unmonitored spills, leaks, or unexpected liquid or gaseous discharges, the inspectors assessed whether an evaluation was performed to determine the type and amount of radioactive material that was discharged by:

Assessing whether sufficient radiological surveys were performed to evaluate the extent of the contamination and the radiological source term and assessing whether a survey/evaluation has been performed to include consideration of hard-to-detect radionuclides; and

Determining whether the licensee completed offsite notifications, as provided in its Groundwater Protection Initiative implementing procedures. The inspectors reviewed the evaluation of discharges from onsite surface water bodies that contain or potentially contain radioactivity, and the potential for ground water leakage from these onsite surface water bodies. The inspectors assessed whether the licensee is properly accounting for discharges from these surface water bodies as part of their effluent release reports.

23 Enclosure The inspectors assessed whether on-site ground water sample results and a description of any significant on-site leaks/spills into ground water for each calendar year are documented in the Annual Radiological En vironmental Operating Report for the radiological environmental monitoring program or the Annual Radiological Effluent Release Report for the radiological effluent TSs. For significant, new effluent discharge points (such as significant or continuing leakage to ground water that continues to impact the environment if not remediated), the inspectors evaluated whether the ODCM was updated to include the new release point.

b. Findings

No findings were identified.

.7 Problem Identification and Resolution (02.07)

a. Inspection Scope

Inspectors assessed whether problems associated with the effluent monitoring and control program are being identified by the licensee at an appropriate threshold and are properly addressed for resolution in the licens ee's CAP. In addition, they evaluated the appropriateness of the corrective actions for a selected sample of problems documented by the licensee involving radiation monitoring and exposure controls.

b. Findings

No findings were identified.

2RS7 Radiological Environmental Monitoring Program

Main article: IP 71124.07

This inspection constituted one complete sample as defined in IP 71124.07-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed the annual radiological environmental operating reports and the results of any licensee assessments since the last inspection to verify that the radiological environmental monitoring program was implemented in accordance with the TSs and ODCM. This review included report changes to the ODCM with respect to environmental monitoring, commitments in terms of sampling locations, monitoring and measurement frequencies, land use census, inter-laboratory comparison program, and analysis of data. The inspectors reviewed the ODCM to identify locations of environmental monitoring stations. The inspectors reviewed the FSAR for information regarding the environmental monitoring program and meteorological monitoring instrumentation.

24 Enclosure The inspectors reviewed quality assurance audit results of the program to assist in choosing inspection "smart samples" and audits and technical evaluations performed on the vendor laboratory program. The inspectors reviewed the annual effluent release report and the 10 CFR Part 61, "Licensing Requirements for Land Disposal of Radioactive Waste," report to determine if the licensee is sampling, as appropriate, for the predominant and dose-causing radionuclides likely to be released in effluents.

b. Findings

No findings were identified.

.2 Site Inspection (02.02)

a. Inspection Scope

The inspectors walked down select air sampling stations and thermoluminescent dosimeter monitoring stations to determine whether they are located as described in the ODCM and to determine the equipment material condition. Consistent with smart sampling, the air sampling stations were selected based on the locations with the highest X/Q, D/Q wind sectors, and thermoluminescent dosimeters were selected based on the most risk-significant locations (e.g., those that have the highest potential for public dose impact).

For the air samplers and thermoluminescent dosimeters selected, the inspectors reviewed the calibration and maintenance records to verify that they demonstrate adequate operability of these components. Additionally, the review included the calibration and maintenance records of select composite water samplers. The inspectors performed an assessment of whether the licensee has initiated sampling of other appropriate media upon loss of a required sampling station. The inspectors observed the collection and preparation of environmental samples from different environmental media (e.g., ground and surface water, milk, vegetation, sediment, and soil) as available to verify that environmental sampling is representative of the release pathways as specified in the ODCM and that sampling techniques are in accordance with procedures. Based on direct observation and review of records, the inspectors assessed whether the meteorological instruments were operable, calibrated, and maintained in accordance with guidance contained in the FSAR, NRC Regulatory Guide 1.23, "Meteorological Monitoring Programs for Nuclear Power Plants," and licensee procedures. The inspectors assessed whether the meteorological data readout and recording instruments in the control room and, if applicable, at the tower were operable. The inspectors evaluated whether missed and/or anomalous environmental samples were identified and reported in the annual environmental monitoring report.

The inspectors selected events that involved a missed sample; inoperable sampler; lost thermoluminescent dosimeter; or anomalous measurement to verify that the licensee has identified the cause and has implemented corrective actions. The inspectors reviewed the licensee's assessment of any positive sample results 25 Enclosure (i.e., licensed radioactive material detected above the lower limits of detection) and reviewed the associated radioactive effluent release data that was the source of the released material. Inspectors selected SSCs that involve or could reasonably involve licensed material for which there is a credible mechanism for licensed material to reach ground water, and assessed whether the licensee has implemented a sampling and monitoring program sufficient to detect leakage of these SSCs to ground water. The inspectors evaluated whether records, as required by 10 CFR 50.75(g), of leaks, spills, and remediation since the previous inspection are retained in a retrievable manner. The inspectors reviewed any significant changes made by the licensee to the ODCM as the result of changes to the land census, long-term meteorological conditions (3-year average), or modifications to the sampler stations since the last inspection.

They reviewed technical justifications for any changed sampling locations to verify that the licensee performed the reviews required to ensure that the changes did not affect its ability to monitor the impacts of radioactive effluent releases on the environment. The licensee uses a vendor laboratory to analyze the radiological environmental monitoring program samples so the inspectors reviewed the results of the vendor's quality control program, including the inter-laboratory comparison, to assess the adequacy of the vendor's program. The inspectors reviewed the results of the licensee's inter-laboratory comparison program to verify the adequacy of environmental sample analyses performed by the licensee. The inspectors assessed whether the inter-laboratory comparison test included the media/nuclide mix appropriate for the facility. If applicable, the inspectors reviewed the licensee's determination of any bias to the data and the overall effect on the radiological environmental monitoring program.

b. Findings

No findings were identified.

.3 Identification and Resolution of Problems (02.03)

b. Findings

No findings were identified.

27 Enclosure

.3 Radioactive Waste System Walkdown (02.03)

a. Inspection Scope

The inspectors walked down accessible portions of select radioactive waste processing systems to assess whether the current system configuration and operation agreed with the descriptions in the FSAR, ODCM, and process control program. The inspectors reviewed administrative and/or physical controls (i.e., drainage and isolation of the system from other systems) to assess whether the equipment which is not in service or abandoned in place would not contribute to an unmonitored release path and/or affect operating systems or be a source of unnecessary personnel exposure. The inspectors assessed whether the licensee reviewed the safety significance of systems and equipment abandoned in place in accordance with 10 CFR 50.59, "Changes, Tests, and Experiments". The inspectors reviewed the adequacy of changes made to the radioactive waste processing systems since the last inspection. The inspectors evaluated whether changes from what is described in the FSAR were reviewed and documented in accordance with 10 CFR 50.59, as appropriate, and to assess the impact on radiation doses to members of the public. The inspectors selected processes for transferring radioactive waste resin and/or sludge discharges into shipping/disposal containers and assessed whether the waste stream mixing, sampling procedures, and methodology for waste concentration averaging were consistent with the process control program, and provided representative samples of the waste product for the purposes of waste classification as described in 10 CFR 61.55, "Waste Classification."

For those systems that provide tank recirculation, the inspectors evaluated whether the tank recirculation procedures provided sufficient mixing. The inspectors assessed whether the licensee's process control program correctly described the current methods and procedures for dewatering and waste stabilization (e.g., removal of freestanding liquid).

b. Findings

No findings were identified.

.4 Waste Characterization and Classification (02.04)

a. Inspection Scope

The inspectors selected the following radioactive waste streams for review:

dry active waste (drywell, condensate, steam chase, reactor water cleanup (RWCU), and (CA) plant smears);

condensate resins; and

RWCU resins.

28 Enclosure For the waste streams listed above, the inspectors assessed whether the licensee's radiochemical sample analysis results (i.e., "10 CFR Part 61" analysis) were sufficient to support radioactive waste characterization as required by 10 CFR Part 61, "Licensing Requirements for Land Disposal of Radioactive Waste." The inspectors evaluated whether the licensee's use of scaling factors and calculations to account for difficult-to-measure radionuclides was technically sound and based on current 10 CFR Part 61 analyses for the selected radioactive waste streams. The inspectors evaluated whether changes to plant operational parameters were taken into account to:

(1) maintain the validity of the waste stream composition data between the annual or biennial sample analysis update; and

(2) assure that waste shipments continued to meet the requirements of 10 CFR Part 61 for the waste streams selected above. The inspectors evaluated whether the licensee had established and maintained an adequate quality assurance program to ensure compliance with the waste classification and characterization requirements of 10 CFR 61.55 and 10 CFR 61.56, "Waste Characteristics."

b. Findings

No findings were identified.

.5 Shipment Preparation (02.05)

a. Inspection Scope

The inspectors assessed whether the requirements of applicable transport cask certificate of compliance had been met. The inspectors evaluated whether the receiving licensee was authorized to receive the shipment packages. The inspectors evaluated whether the licensee's procedures for cask loading and closure procedures were consistent with the vendor's current approved procedures. The inspectors assessed whether the shippers were knowledgeable of the shipping regulations and whether shipping personnel demonstrated adequate skills to accomplish the package preparation requirements for public transport with respect to:

the licensee's response to NRC Bulletin 79-19, "Packaging of Low-Level Radioactive Waste for Transport and Burial," dated August 10, 1979; and

Title 49 CFR Part 172, "Hazardous Materials Table, Special Provisions, Hazardous Materials Communication, Emergency Response Information, Training Requirements, and Security Plans," Subpart H, "Training." Due to limited opportunities for direct observation, the inspectors reviewed the technical instructions presented to workers during routine training. The inspectors assessed whether the licensee's training program provided training to personnel responsible for the conduct of radioactive waste processing and radioactive material shipment preparation activities.

29 Enclosure

b. Findings

Introduction:

A URI was identified because additional information was needed to assess the significance of an issue regarding shipping and transportation of a radioactively contaminated condensate demineralizer vessel.

Description:

On July 14, 2011, it was reported to the licensee by the driver of the vehicle that there was a puncture in the side of a container package on radioactive material shipment number 11-127. The package was a Sealand box inside an enclosed conveyance. The Sealand box contained a radioactively contaminated condensate demineralizer vessel and the puncture was a nominal 4 by 6 inch hole. There was no spread of contamination as a result of the compromised package. The inspectors' initial review determined that a performance deficiency exists, in that, the shipping container contents was inappropriately braced and blocked for transport. Regulations require that licensees ensure that loads not shift under conditions normally incident to transportation. The inspectors will review the additional information provided by the licensee and determine the significance of the performance deficiency. (URI 5000263/2011004-03; Shipping and Transportation of a Radioactively Contaminated Condensate Demineralizer Vessel)

.6 Shipping Records (02.06)

a. Inspection Scope

The inspectors evaluated whether the shipping documents indicated the proper shipper name; emergency response information and a 24-hour contact telephone number; accurate curie content and volume of material; and appropriate waste classification, transport index, and UN number for the following radioactive shipments:

shipment number 09-46;

shipment number 10-11;

shipment number 10-23;

shipment number 10-35; and

shipment number 11-127.

b. Findings

No findings were identified.

.7 Identification and Resolution of Problems (02.07)

a. Inspection Scope

The inspectors assessed whether problems associated with radioactive waste processing, handling, storage, and transportation, were being identified by the licensee at an appropriate threshold, were properly characterized, and were properly addressed for resolution in the licensee CAP. Additionally, the inspectors evaluated whether the corrective actions were appropriate for a selected sample of problems documented by the licensee that involve radioactive waste processing, handling, storage, and transportation.

30 Enclosure The inspectors reviewed results of selected audits performed since the last inspection of this program and evaluated the adequacy of the licensee's corrective actions for issues identified during those audits.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

4OA1 Performance Indicator Verification

Main article: IP 71151

.1 Mitigating Systems Performance

Index - Emergency AC Power System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance Index (MSPI) - Emergency Alternating Current (AC) Power System PI for the period from the 3 rd Quarter 2010 through the 2 nd Quarter 2011. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI) Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, dated October 2009, were used. The inspectors reviewed the licensee's operator narrative logs, MSPI derivation reports, issue reports, event reports and NRC Integrated Inspection Re ports for the period of July 2010 through June 2011 to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report. This inspection constituted one MSPI emergency AC power system sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.2 Mitigating Systems Performance Index - High Pressure Injection Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the MSPI - High Pressure Injection Systems PI for the period from 3 rd Quarter 2010 through the 2 nd Quarter 2011. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, dated October 2009, were used. The inspectors reviewed the licensee's operator narrative logs, issue reports, 31 Enclosure MSPI derivation reports, event reports and NRC Integrated Inspection Reports for the period of July 2010 through July 2011 to validate the accuracy of the submittals.

The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report. This inspection constituted one MSPI high pressure injection system sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.3 Mitigating Systems Performance Index - Heat Removal System

a. Inspection Scope

The inspectors sampled licensee submittals for the MSPI - Heat Removal System PI for the period from the 3 rd Quarter 2010 through the 2 nd Quarter 2011. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, dated October 2009, were used. The inspectors reviewed the licensee's operator narrative logs, issue reports, event reports, MSPI derivation reports, and NRC Integrated Inspection Reports for the period of July 2010 through June 2011 to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report. This inspection constituted one MSPI heat removal system sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.4 Reactor Coolant System Specific Activity

a. Inspection Scope

The inspectors sampled licensee submittals for the Reactor Coolant System Specific Activity PI for Monticello Nuclear Generating Plant (MNGP) for the period from the 4 th Quarter 2010 through the 2 nd Quarter 2011

. The inspectors used PI definitions and guidance contained in the NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, dated October 2009, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensee's reactor 32 Enclosure coolant system chemistry samples, TS requirements, issue reports, event reports and NRC Integrated Inspection Reports for the period of 4 th Quarter 2010 through the 2 nd Quarter 2011 to validate the accuracy of the submittals. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample. Documents reviewed are listed in the Attachment to this report. This inspection constituted one reactor coolant system specific activity sample as defined in IP 71151-05.

4OA2 Identification and Resolution of Problems

Main article: IP 71152

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensee's CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: identification of the problem was complete and accurate; timeliness was

commensurate with the safety significance; evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.

Minor issues entered into the licensee's CAP as a result of the inspectors' observations are included in the Attachment to this report.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure, they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

No findings were identified.

35 Enclosure

.2 (Closed) Licensee Event Report (LER) 05000263/2010-002-01:

Secondary Containment Briefly Degraded; Supplement 1 The licensee originally reported this event on June 3, 2010. This event was evaluated by the inspectors and the LER was closed in Inspection Report 05000363/2010005. On June 13, 2011, the licensee issued Supplement 1 to LER 050002010-002-00.

The purpose of Supplement 1 was to document the results of a secondary containment capability test, which was performed on April 27, 2011. Based on the results of the test and subsequent engineering evaluation, the licensee concluded that secondary containment remained capable of performing its safety function in the plant configuration that existed on June 3, 2010, including considerations for penetrations that were or may have been open at that time. Based on the results of this test and engineering evaluation, the licensee concluded that since there was no loss of safety function, this event was not considered a safety system functional failure for the purposes of Reactor Oversight Process PI reporting per NEI 99-02. The inspectors reviewed the test data and subsequent engineering evaluation which supported the licensee's position that the event did not constitute a loss of the secondary containment safety function. No findings of significance were identified during this inspection. Documents reviewed as part of this inspection are listed in the Attachment to this report. This LER is closed. This event follow-up review constituted one sample as defined in IP 71153-05.

.3 (Closed) LER 05000263/2010-003-01:

Secondary Containment Briefly Degraded; Supplement 1 The licensee originally reported this event on August 5, 2010. This event was evaluated by the inspectors and the LER was closed in Inspection Report 05000363/2010005.

On June 13, 2011, the licensee issued Supplement 1 to LER 05002010-003-00. The purpose of Supplement 1 was to document the results of a secondary containment capability test, which was performed on April 27, 2011. Based on the results of the test and subsequent engineering evaluation, the licensee concluded that secondary containment remained capable of performing its safety function in the plant configuration that existed on August 5, 2010, including considerations for penetrations that were or may have been open at that time. Based on the results of this test and engineering evaluation, the licensee concluded that since there was no loss of safety function, this event was not considered a safety system functional failure for the purposes of Reactor Oversight Process PI reporting per NEI 99-02. The inspectors reviewed the test data and subsequent engineering evaluation which supported the licensee's position that the event did not constitute a loss of the secondary containment safety function. No findings of significance were identified during this inspection. Documents reviewed as part of this inspection are listed in the Attachment to this report. This LER is closed. This event follow-up review constituted one sample as defined in IP 71153-05.

36 Enclosure

.4 LER 05000263/2007-005-01:

Discovery of Appendix R - Non-Compliant Manual Actions During Review of National Fire Protection Association (NFPA) 805 The licensee issued LER 05000263/2007-005-00 regarding the use of operator manual actions in an Appendix R Section III.G.1/G.2 fire area. The LER was closed in Inspection Report 05000263/2007004 and documented as a licensee-identified violation of NRC requirements. Because the licensee was transitioning to NFPA 805 and the violation met the criteria established by the NRC's Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR Part 50.48(c)) for a licensee in NFPA 805 transition, the NRC exercised enforcement discretion to not cite the violation in accordance with the NRC's Enforcement Policy. On December 22, 2010, the licensee provided an update to LER 05000263/2007-005 to reflect their withdrawal of their letter of intent to voluntarily implement 10 CFR Part 50.48(c) at MNGP. This LER revision was evaluated and closed during the Monticello 2011 Triennial Fire Protection Inspection. The closure documentation for this LER revision is located in Section 4OA3.1 of Inspection Report 05000263/2011008 (MNGP Triennial Fire Protection Inspection Report). This event follow-up review constituted one sample as defined in IP 71153-05.

.5 (Closed) LER 05000263/2011-001-00:

Reactor Vessel Overfill in Appendix R Scenario On November 12, 2010, during a fire protection program assessment, the licensee discovered that the fire protection safe shutdown analysis did not address a postulated reactor vessel overfill event. During a potential fire, concurrent with a loss of offsite power, requiring the evacuation of the control room, both the HPCI and RCIC pumps would start if a low reactor water level setpoint was reached. Potential fire damage could result in the failure of the safety-related HPCI and RCIC high reactor water level trip circuitry leading to a reactor vessel overfill event. Specifically, level switches LS-2-3-672E (LIS) and LS 2-3-672F provide the high reactor water level trip signal. Fire damage to the switch circuitry could prevent the HPCI and RCIC high reactor water level trip, which would result in the reactor vessel overfill event. The event would continue until sufficient water filled the HPCI and RCIC steam lines to stall the HPCI and RCIC pumps. As a result of the water filled steam lines, when operations personnel manually open (i.e., lifts) the safety relief valves (SRVs) to allow for low pressure reactor water inventory makeup and decay heat removal, the SRVs may be subjected to high pressure steam/water flow. Licensee Event Report 05000263/2011-001-01 was evaluated and closed during the Monticello 2011 Triennial Fire Protection Inspection. The closure documentation for this LER revision is located in Section 4OA3.2 of Inspection Report 05000263/2011008 (MNGP Triennial Fire Protection Inspection Report). The inspectors concluded that since LER 05000263/2011-001-01 only clarified information provided in the original LER submission, the technical evaluation used to close LER 05000263/2011-001-01 was of sufficient scope to also close the original LER. This event follow-up review constituted one sample as defined in IP 71153-05.

37 Enclosure

.6 (Closed) LER 05000263/2011-003-00:

Secondary Containment Damper Icing On February 11, 2011, secondary isolation damper V-D-61 (reactor building outboard isolation damper) was discovered frozen closed due to ice buildup, with the actuator broken. The inboard damper, V-D-62, was found blocked partially open, again due to icing. The source of the icing was determined to be leakage from a heating coil collecting in the ventilation ducting which subsequently, due to outside temperatures significantly below freezing, froze when the ventilation system was placed in service. Upon discovery of the condition, the licensee entered the applicable TS action associated with a secondary containment ventilation penetration flow path with two inoperable isolation valves. Additionally, the licensee reported the event under 10 CFR 50.72(b)(3)(v)(C and D), events or conditions that at the time of discovery could have prevented fulfillment of a safety function. Corrective actions taken by the licensee included removing the ice from the dampers, repairing the actuation linkage for damper V-D-61, and performing operational testing of V-D-61 and V-D-62. The license also performed additional testing on V-D-61 to ensure that the damper's secondary containment safety function would be maintained (it would remain closed) even with the damper actuation crank arm disconnected (similar to the as found condition for this event). The licensee's evaluation of this event concluded that at least one of the two isolation dampers for this secondary containment penetration would have remained closed, in the case where the isolation of secondary containment would have been required, and therefore the secondary containment safety function was not lost during this event. The inspectors reviewed the licensee's evaluation of the event, test data, and subsequent engineering evaluation which supported the licensee's position that the event did not constitute a loss of the secondary containment safety function. No findings of significance were identified during this inspection. Documents reviewed as part of this inspection are listed in the Attachment to this report. This LER is closed. This event follow-up review constituted one sample as defined in IP 71153-05.

.7 (Closed) LER 05000263/2011-004-00:

Secondary Containment Inoperability Due to Ventilation Alignment Issue On June 8, 2011, secondary containment was declared inoperable after operators swapped the operating refuel floor supply air handling unit from V-AH-4A to V-AH-4B.

Secondary containment differential pressure reduced to

.17 inches of water column

vacuum, which did not meet TS requirements to maintain secondary containment vacuum greater than or equal to

.25 inches of water column vacuum.

Upon discovery of the condition, the licensee entered the applicable TS action associated with an inoperable secondary containment. Additionally, the licensee reported the event under 10 CFR 50.72(b)(3)(v)(C and D), events or conditions that at the time of discovery could have prevented fulfillment of a safety function. The licensee determined that the cause of the event was a malfunctioning reheat bypass damper on V-AH-4A. This malfunctioning component caused significantly different flow rates between V-AH-4A and V-AH-4B, and required a significant adjustment to the reactor building exhaust fan variable inlet vane settings to maintain the reactor building greater than

.25 inches water column vacuum the previous time (June 5, 2011) V-AH-4A was placed in service.

On June 8, 2011, when swapping from V-AH-4A to 38 Enclosure V-AH-4B, before the reactor building exhaust fan variable inlet vanes could be adjusted, secondary differential pressure changed from 0.6 inches of water column vacuum to

.17 inches of water column vacuum.

Once identified, the licensee immediately restored V-AH-4A to service, restoring secondary containment differential pressure in excess of

.25 inches water column vacuum.

Other corrective actions taken by the licensee included repair of the reheat bypass damper and implementation of a change to the operating procedure used to transfer between V-AH-4A and V-AH-4B to provide additional guidance to account for the differences in flowrates supplied by each reactor building supply fan. The inspectors reviewed the licensee's assessment of the event, subsequent corrective actions, and engineering evaluation which supported the licensee's position that the event did not constitute a loss of the secondary containment safety function. No findings of significance were identified during this inspection. Documents reviewed as part of this inspection are listed in the Attachment to this report. This LER is closed. This event follow-up review constituted one sample as defined in IP 71153-05.

.8 (Closed) LER 05000263/2011-005-00:

Power Range Monitor Channels Out of Alignment On June 30, 2011, with the reactor operating at 80 percent power, three of the four average power range monitors (APRMs) exceeded the TS requirement to maintain the absolute difference between each APRM channel and the calculated power at less than or equal to 2 percent rated thermal power while operating at equal or greater than 25 percent rated thermal power. Upon discovery of the condition, the licensee entered the applicable TS actions associated with inoperable reactor protection system instrumentation channels. Additionally, the licensee reported the event under 10 CFR 50.72(b)(3)(v)(A and D), events or conditions that at the time of discovery could have prevented fulfillment of a safety function. The licensee concluded that, even though the three APRM channels were inoperable per TSs, since two of the three affected APRMs exceeded the 2 percent power deviation in the conservative direction, the APRM safety function was preserved since a reactor scram would have initiated at a lower power level than required. Corrective actions proposed by the licensee to address the causes of the event included developing a process to enable the site to better predict APRM response due to changes in power profiles during reactivity maneuvers and revising their reactivity adjustment procedure to better identify, communicate, and prevent reactivity adjustments that could lead to multiple APRMs exceeding their maximum allowed power deviations. The inspectors reviewed the licensee's assessment of the event, subsequent corrective actions, and engineering evaluation which supported the licensee's position that the event did not constitute a loss of the safety function associated with reactor protection instrumentation inputs related to the APRMs. No findings of significance were identified during this inspection. Documents reviewed as part of this inspection are listed in the Attachment to this report. This LER is closed. This event follow-up review constituted one sample as defined in IP 71153-05.

39 Enclosure

.9 Notice of Enforcement Discretion (NOED) 11-3-001:

Monticello TS 3.8.1 Emergency Diesel Generator Load Rejection Surveillance Requirement Issue On September 27, 2011, during an engineering self-assessment, the licensee identified a potential issue associated with the testing methodology used to demonstrate each EDG's capability to withstand the rejection of an electrical load that is equivalent to the single largest post-accident electrical load. On September 29, 2011, the licensee verified that their existing surveillance test OSP-ECC-0566, "Low Pressure ECCS [emergency core cooling system ] Automatic Initiation and Loss of Auxiliary power Test," Revision 8, did not ensure that the load rejection test was performed with sufficient load to satisfy the requirements of SR 3.8.1.7 (Verify each EDG rejects a load greater than or equal to its associated single largest post-accident load and, following load rejection, the frequency is less than or equal to 67.5 Hz.). On September 29, 2011, at approximately 1700, the licensee declared both 11 and 12 EDGs inoperable and entered the Action for TS 3.8.1.E, "Two EDGs Inoperable." At approximately 2200, the licensee requested enforcement discretion to extend the Action Completion Time for TS 3.8.1.F, from twelve hours to five days, to allow time to perform the required EDG load rejection testing. At approximately 23:58, the Agency granted NOED 11-3-001. The inspector's evaluation of the issue included a review of the technical documents associated with the issue and several meetings with the licensee management and technical staff. The initial information gained by the inspectors and their assessment of the issue was communicated to senior agency managers well in advance of the licensee's NOED request, significantly contributing to the Agency's understanding and appropriate disposition of the issue. Additional information associated with the inadequate surveillance procedure and EDG operability is documented in Section 1R15 of this report. This event follow-up review constituted one sample as defined in IP 71153-05.

4OA5 Other Activities

.1 Operation of an Independent Spent Fuel Storage Facility Installation (ISFSI) at Operating Plants (60855.1)

a. Operations of an ISFSI Inspection Scope The inspectors observed and evaluated the licensee's ISFSI program to verify compliance with the applicable Certificate of Compliance (CoC) conditions, TSs, and procedures. The inspectors reviewed procedure changes related to ISFSI cask loading, movement, surveillance, and maintenance. A tour was conducted of the ISFSI pad to assess the condition of the ISFSI. No flammable or combustible materials were observed inside the ISFSI cask storage area, and inspectors reviewed evaluations of flammable materials outside the ISFSI area. Inspectors independently reviewed environmental radiation levels around the ISFSI, and reviewed the licensee's radiation monitoring program for the ISFSI.

40 Enclosure In addition, the inspectors reviewed a number of condition reports and the associated follow up actions since the last ISFSI inspection. The inspectors reviewed 10 CFR 72.48 screenings and evaluations and changes to the licensee's

10 CFR 72.212 report.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On October 6, 2011, the inspectors presented the inspection results to Mr. O'Connor and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

Radiological Environmental Monitoring Program and the Performance Indicator Verification for Barrier Integrity for Reactor Coolant System Dose Equivalent Iodine with Mr. Timothy O'Connor, the Site-Vice President, on August 19, 2011.

Radiological Hazard Assessment and Exposure Controls; Occupational ALARA Planning and Controls; Radioactive Gaseous and Liquid Effluent Treatment; and Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation; with Mr. Timothy O'Connor, the Site-Vice President, on September 16, 2011.

The operation of an ISFSI Inspection conducted an interim exit meeting on September 23, 2011. The inspectors presented the inspection results to members of the licensee management and staff. Licensee personnel acknowledged the information presented.

Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation; telephone update with Ms. S. O'Connor, Regulatory Affairs Analyst, on September 29, 2011. The inspectors confirmed that none of the potential report input discussed was considered proprietary. Proprietary material received during the inspection was returned to the licensee.

4OA7 Licensee-Identified Violations

None. ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. O'Connor, Site Vice President

J. Grubb, Plant Manager

W. Paulhardt, Operations Manager

N. Haskell, Site Engineering Director

K. Jepson, Assistant Plant Manager

S. Radebaugh, Maintenance Manager

M. Holmes, Chemistry Manager

A. Zelie, Radiation Protection Manager

P. Kissinger, Regulatory Affairs Manager

Nuclear Regulatory Commission

K. Riemer, Chief, Reactor Projects Branch 2

Attachment

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened

05000263/2011004-01 URI NOED for Emergency Diesel Generator Load Rejection Surveillance Requirement 3.8.1.7 (Section 1R15)

05000263/2011004-02 NCV Failure to Follow Emergency Diesel Generator Quarterly

Surveillance Procedure (Section 1R22)

05000263/2011004-03 URI Shipping and Transportation of a Radioactively Contaminated Condensate Demineralizer Vessel (Section 2RS8)

Closed

05000263/FIN-2011004-02 NCV Failure to Follow Emergency Diesel Generator Quarterly

Surveillance Procedure (Section 1R22)

05000263/LER-2011-001-00 LER Reactor Vessel Overfill in Appendix R Scenario (Section 4OA3.5)

05000263/LER-2011-003-00 LER Secondary Containment Damper Icing (Section 4OA3.6)

05000263/LER-2011-004-00 LER Secondary Containment Inoperability Due to Ventilation Alignment Issue (Section 4OA3.7)

05000263/LER-2011-005-00 LER Power Range Monitor Channels Out of Alignment (Section 4OA3.8)

05000263/LER-2010-002-01 LER Secondary Containment Briefly Degraded; Supplement 1 (Section 4OA3.2)

05000263/LER-2010-003-01 LER Secondary Containment Briefly Degraded; Supplement 1 (Section 4OA3.3)

Discussed

05000263/2007-005-01 LER Discovery of Appendix R - Non-Compliant Manual Actions During Review of NFPA 805 (Section 4OA3.4)

05000263/2011-001-01 LER Reactor Vessel Overfill in Appendix R Scenario; Supplement 1 (Section 4OA3.5)

Attachment

LIST OF DOCUMENTS REVIEWED

The following is a partial list of documents reviewed during the inspection.

Inclusion on this list does not imply that the NRC inspector reviewed the documents in their entirety, but rather that selected sections or 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.

Section 1R04

2154-10; HPCI System Prestart Valve Checklist; Revision 30 2118; Plant Prestart Checklist HPCI System; Revision 15 2154-35; HPCI Hydraulic Control and Lubrication System Prestart Valve Checklist; Revision 8 Operations Manual B.03.02-01; HPCI:

Function and General Description of System; Revision 9

NH-36250; P&ID (Water Side) HPCI System; Revision 79

NH-36249; P&ID (Steam Side) HPCI System; Revision 78

CAP 01299449; P-212 Packing Leak and Drip Pan Clogged

CAP 01173452; HPCI P-212 Drip Tray Appears Clogged

CAP 01299327; HPCI/RCIC Steam Trap Temps above Expected Band

201; Plant Prestart Checklist

CRV-EFT [Emergency Filtration Train] System; Revision 8

Operations Manual B.08.13-05; Control Room H&V [Heating and Ventilation] and EFT:

System Operation; Revision 19 Operations Manual B.08.13-01; Control Room H&V and EFT:

Function and General Description of System; Revision 10

NH-170037; Main Control Room CRV/EFT System; Revision 80

2154-34; ESW System Prestart Valve Checklist; Revision 26

200; Plant Prestart Checklist ESW System; Revision 2

Operations Manual B.08.01.04-05; ESW:

System Operation; Revision 22 Operations Manual B.08.01.04-01; ESW:

Function and General Description of System;

Revision 5

NH-36665; P&ID Service Water System & Make-Up Intake Structure; Revision 94

NH-36041; P&ID Service Water System; Revision 93

Station Logs; August 17, 2011 2154-14; Fuel Oil Prestart Valve Checklist; Revision 16 2124; Plant Prestart Checklist Diesel Generators and Fuel Oil System; Revision 8

2154-28; Diesel Generator Air Start System Prestart Valve Checklist; Revision 9

NH-36051; Diesel Oil System; Revision 77

NH-36664; RHR Service Water and ESW Systems; Revision 85

2154-11; Core Spray System Prestart Checklist; Revision 18 2154-26; Drywell Prestart Valve Checklist; Revision 46 Operations Manual B.03.01-01; Core Spray Cooling System- Function and General Description of System; Revision 6

Operations Manual B.03.01-02; Core Spray Cooling System - Description of Equipment;

Revision 9 Operations Manual B.03.01-05; Core Spray Cooling System - System Operation; Revision 31

NH-36248; Core Spray System; Revision 79

NH-36664; RHR Service Water and ESW Systems; Revision 85

Attachment Section 1R05

CAP 01297517; NRC Identified Transient Combustibles in A/B RHR Pump Room

CAP 01297521; NRC Identified a Pre-Fire Strategy Question

CAP 01297667; Strategy A.3-01-C FZ 1-C RCIC Room

OWI 02.03; Operator Rounds; Revision 16 Pre-Fire Strategy A.3-01-A; Fire Zone 1-A; No. 12 RHR and Core Spray Pump Room Pre-Fire Strategy A.3-01-B; Fire Zone 1-B; No. 11 RHR and Core Spray Pump Room Pre-Fire Strategy A.3-01-C; Fire Zone 1-C; RCIC Room Pre-Fire Strategy A.3-01-E; Fire Zone 1-E; HPCI Room Pre-Fire Strategy A.3-23-B; Fire Zone 23-B; Intake Corridor Station Logs; August 3, 2011 USAR Appendix J.5; Fire Protection Program; Updated Fire Hazards Analysis; Revision 26

A.3-23-B; Pre-Fire Strategies; Revision 6

Pre-Fire Strategy A.3-13-B; RFP and LO Reservoir Area; Revision 11

Pre-Fire Strategy A.3-13-C; Turbine Building (911' Elevation E. MCC [Motor Control Center]

Area); Revision 7 2176; Fire Drill Procedure; Revision 21

Fire Brigade Drill Guide 13-C.01;

MCC-133A Fire

Section 1R06

CAP 1237590; CAP Generated under

WO 404100 Cable Vault Inspection

WO 404100; Underground Vaults for Water - Inspection Manhole Inspection for Cable Condition Monitoring Program Attachment 1; Inspection results July 12, 2011

CAP 1294159; Water in Underground Vaults

WR 70045; Water in Underground Vaults; July 12, 2011

WO 408479; NMH335 Install Automatic Temporary Sump Pump

PMRQ 9344-08; Perform Visual Inspection of Manhole; Weekly inspection results;

March 22, 2009 - July 15, 2011

CAP 1209347; ISFSI Manholes Discovered with Water

EWI-08.19.01; Cable Condition Monitoring Program; Revision 3

CAP 1237590; ISFSI Manholes Found with Standing Water

CAP 1151315; Water Found in Manhole Containing Cables for 2R Feed

WO 420098; Perform Visual Inspection of Manhole Section 1R11

Simulator Guide

RQ-SS-69

Section 1R12

ODMI 11-08; CFW/P-1A/MTR No. 11 Condensate Pump Motor, Revision 0 ODMI

1299376; Bearing Defects in the Motor Thrust Bearing of P-1A 11 Condensate Pump Contractor Bearing Analysis Report:

P-1A 11 Condensate Pump; Full Spectrum Diagnostics;

August 10, 2011 WDI Analysis; 11 Condensate Pump Motor Lift Greater than the Recommended 10 to 20 mils; August 7, 2011

EWI-05.02.01; Monticello Maintenance Rule Program Document; Revision 17

Attachment B.6.5; Monticello Maintenance Rule Program System Basis Document:

Condensate and Feedwater System; Revision 2

No. 12 Condensate Pump Motor Bearing Root Cause Supporting and Refuting Evidence List

CAP 01300293; 11 Condensate Pump Vibration Monitor Spiking

CAP 01294990; Bearing Defects Identified in the 11 Condensate Pump Motor

CAP 01299376; Stage 3 Failure Condition on 11 Condensate Pump Motor Bearing

CAP 01299158; Change in 11 Condensate Motor Thrust Bearing Temperature

CAP 01298398; 11 Condensate Pump, Vibration Signature not as Expected

CAP 01298143; Vibration Levels Enter Alert Range on No. 11 Condensate Pump

CAP 01296526; Fluctuating Readings on Temporary Monitoring System

WO 00435697; P-1A Replace Motor Upper Thrust Bearing

CAP 01303209; Response to Fire Impairment Could Be More Aggressive

CAP 01203212; Opportunities Were Missed to Resolve Fire System Blockage

CAP 01302334; Sprinkler Piping In Intake Plugged With Clay-Like Debris

CAP 01180222; Intake Structure Sprinkler Blockage 0324; Fire Protection System - Sprinkler System Tests; Revision 44; August 26, 2011

Section 1R13

WO 00434880; Fire Header in Condenser Hotside Leakage

CAP 01293745; Received Condenser Hotside Fire Header Flow Alarm and Water Discharge

CAP 01296368; Bleach Corrosion Found on Stilling Wells 5 A/B Air Lines

CAP 01297614; Unidentified Tubing in Tray with Degraded CWT Air Lines

CAP 01297748; SHC Bldg (Intake Ceiling) Should Be Eval for Age Management

CAP 01294862; PLC System Rack 2 Power Supply A Failure

CAP 01294865; PLC System Rack 3 Power Supply A Failure

WO 435690-01; Replace Power Module (ES-80-3A)

WO 435677-01; Replace Power Module (ES-80-2B)

Section 1R15

Contractor Prepared Code Case N-513-3 Evaluation of Cooling Coil V-AC-5 Leaking Flaw;

July 9, 2011

CAP 01290273; V-AC-5 Cooling Coil Leak

CAP 01294288; SRV

MSB-123 has Improper Main Stage Filter Installed

CAP 01294990; P-1A, 11 Condensate Pump Motor Thrust Bearing Defect

Contractor Prepared Advanced Diagnostic Analysis Report for 11 Condensate Pump;

July 22, 2011

CAP 01298398; 11 Condensate Pump Vibration Signature not as Expected 0255-02-III; SBLC Quarterly Pump and Valve Tests; Revision 52

CAP 01285834; Elevated Vibration Levels on 11 SBLC

CAP 01286064; Abnormal Trending in SBLC Gearbox

CAP 01296385; Oil Leaking from Level Plug on 11 SBLC Pump Gearbox

CAP 01296387; 11 SBLC Failed PMT for No Oil Leaks from Crankcase

CAP 01296855; No Oil Sample Taken on 11 SBLC Gearbox DBD B.03.05; Design Basis Document for SBLC System; Revision 4

EC 18234; P-203A Oil and Vibration

Station Logs; July 26, 2011- July 27, 2011

USAR 6.6; SBLC System; Revision 25

WO 418572; P-203A, Oil Leaks from Pump Access Covers

WO 418886; 0255-02-III SBLC Inservice Test Attachment

CAP 01301421; Reached Trigger Point 2 for V-AC-5 Coil Leakage Operability Recommendation (associated with

CAP 01290273) for V-AC-9 Cooling Coil Leaks;

Revisions 1, 2, and 3 Operational Decision Making Issue Evaluation 11-11; V-AC-5, 'A' RHR Room Air Cooling Unit Operations Manual B.08.05-05; Fire Protection-System Operation; Revision 52

CAP 1302334; Sprinkler Piping in Intake Plugged with Clay-like Debris

OPS-ECC-0566; Low Pressure ECCS Automatic Initiation and Loss of Auxiliary Power Test;

Revision 8

OPS-EDG-0535-11; 11 Emergency Diesel Generator Load Rejection Test; Revision 1

CAP 01305683; CDBI FSA- Question on Definition of Post Accident Load in TS

EC 18798; 11 and 12 EDG Single Largest Load Evaluation

PRA-MEMO-11-011; Risk Assessment of Inadequate EDG Load Shed Surveillance;

September 29, 2011

Section 1R18

FP-SC-PE-01; Dedication of Commercial Grade Items and Services; Revision 4

CGD-2011-017; Cooling Coil, AAF, V-AC-5; Revision 0

IEE-2011-019; Copper Specification Change for Cooling Coil Header End Caps, V-AC-4 and

AC-5 and Weight Change; Revision 0

Quality Inspection Checklist; Cooling Coil 96 - 1/4 x 29 - 1/8 x 11", Part Number 18-80-6W5-8C

EC-18659; Evaluate Internal Flooding Impact on ECCS Corner Rooms when Hatch 6/RB is Removed under

WO 00432968

WO 432968-7; MECH - V-AC-5, Replace Cooling Coil

CAP 01301396; Documentation not Found for V-AC-5 Baffle Welds

Section 1R19

WO 403503; Replace RPS MG Set Voltage Regulator 4822-PM; Reactor Protection System MG Set Maintenance Procedure; Revision 10 B.09.12-05; Reactor Protection System Power Supplies; Revision 21 USAR 7.6; Plant Instrumentation and Control Systems, Revision 28P

CAP 01298389; NRC Question Regarding RPS MG Set Testing Power Level

WO 424843;

MO-3502 RCIC Test Return Isolation Valve PM

MEI-09.04; MOV Post-Maintenance Testing; Revision 8

4-AWI-04.05.06; Post-Maintenance Testing Revision

4900-01-PM; PM for Limotorque Motor Operated Valves; Revision 32

NH-36251; P&ID RCIC; Revision 78

CAP 01299562; Handwheel, Face Plate and Nut Found on Floor

EC 15566; EPU - ESW Discharge Piping Reroute; Revision 0

EC-15566; ESW Discharge Piping Reroute Pre-Op Test Plan; Revision 2 Operations Manual B.08.01.04-01; ESW:

Function and General Description of System;

Revision 5

NH-36665; P&ID Service Water System & Make-Up Intake Structure; Revision 94

NH-36041; P&ID Service Water System; Revision 93

255-11-III-3; 13 ESW Quarterly Pump and Valve Tests; Revision 51

CAP 01299813; Deficiencies in Freeze Seal Work Instructions and Practices

CAP 01299887; Employees not Following Action Items Identified

WO 438743; P-105 did not Auto Start when

PS-1936 Tripped

NX-74927-3; Schematic Model FS2, Positive Ground Engine Driver Fire Pump Controller;

Revision 0

Attachment CAP P-105; (Diesel Fire Pump) did not Auto Start during Testing 0266; Fire Pumps Simulated Auto-Actuation and Capability Test; Revision 53

Operations Manual B.08.05-05; Fire Protection-System Operation; Revision 52

CAP 01303133; Program Clock for DFP found ON during Investigation

FP-PE-NDE-520; Visual Examination for Leakage,

VT-2; Revision 3

NH-36664; RHRSW and ESW Systems; Revision 85

2968-02; MECH - V-AC-5, Replace Cooling Coil - PMT; September 2, 2011

2968-07;

MECH-V-AC-5, Replace Coiling Coil; September 2, 2011

0074; Control Rod Drive Exercise; Revision 54 0081; Control Rod Drive Scram Insertion Time Test; Revision 60 1054; Control Rod Drive Normal Drive Timing Test; Revision 18

24; CRD- Insert Stall Flow Test; Revision 10

2300; Reactivity Adjustment; Revision 4

CAP 01304168;

CDR-138/30-47 Clg Water Inl Chk Vlv not Functioning

NH-36244; Control Rod Hydraulic System; Revision 80

NH-36245; Control Rod Hydraulic System; Revision 77

WO 439789;

CRD-138/30-47, Repair Cooling Water Inlet Check Valve

Section 1R22

OSP-EFT-0596; Control Room Envelope In-Leakage Test; Revision 2 0187-02; 12 EDG/12 ESW Quarterly Pump and Valve Tests; Revision 75

CAP 01296286; Procedure Step Option not Performable for EDG Test 0187-01; 11 EDG / 11 ESW Quarterly Pump and Valve Tests; Revision 75

CAP 1294124; Procedure 0187-01 Lacks Detail on Step 4.b

CAP 1294118; Human Performance Error during Performance of 0187-01 Step 7

0187-01B; 11 EDG/ 11 ESW / Monthly Pump and Valve tests; Revision 14

1404-01; EDG ESW Heat Exchanger Performance Test; Revision 13

NH-36051; P&ID Diesel Oil System; Revision 77 Operations Manual B.09.08-03;

EDGs-Instrumentation and Controls; Revision 8

Operations Manual B.09.08-05; EDGs- System Operation; Revision 36

Operations Manual B.08.11-05; Diesel Oil System-System Operation; Revision 22

0141 Reactor Building to Torus Vacuum Breaker Operability Check; Revision 32

255-020-III; SBLC Quarterly Pump and Valve Tests; Revision 52

Section 2RS1

AWI-04.05.13; Control of Items in the Spent Fuel Pool; Revision 10

AWI-08.04.03; Radioactive Material Control; Revision 23 4

AWI-08.04.06; Area Control; Revision 21

AWI-08.06.02; Plant Key Control; Revision 10

AR 01259508; Review Procedure for Clearing Personnel Following PCM Alarm;

November 19, 2010

AR 01260031; Evaluation Needed as Found Gamma to Neutron Ratio in Drywell; November 23, 2010

AR 01266047; Unconditional Release Survey Requirements (Alpha); January 12, 2011 R.01.04; Control of Personnel in High Radiation and Airborne Areas; Revision 23

R.02.02; Surface Contamination Surveys; Revision 26

R.06.02; Unconditional Release of Equipment or Material; Revision 26

R.06.09; Storage and Inventory of Radioactive Material Exclusive of Approved Storage Locations; Revision 16

Attachment R.07.02; Area Posting, Special Status Signs and Hot Spot Stickers; Revision 39 R.07.03; Posting, RWP and / or Equipment Changes Due to Plant Operational Status;

Revision 15

R.08.06; Contaminated Area Control; Revision 08 R.09.52; Locked High Radiation Area Door ALARMS and Lock Operation and Functional Checks; Revision 08

R.12.02; Radiation Protection Key Control; Revision 28

R.14.13; Electronic Dosimeter Operations; Revision 10

Section 2RS2

ALARA Dose Tracking Reports; various dates 2011 Assessment 2010-004-5; Radiological Protection; January 2011

FP-RP-JPP-01; Radiation Protection Job Planning; Revision 9

EWI-08.04.04; Source Term Reduction Program; Revision 11

Monticello Nuclear Generating Plant; Long Term Dose Reduction Plan 2008 - 2012;

Revision 03 Post Job ALARA Review Work Order

383535,

385907,

394606;

MO-2373 Activities - Actuator Disassembly, Valve Disassembly, Viper Test, and Radiography; June 2011

Post Job ALARA Review Work Order

401680; Drywell Snubber Inspection / Change Out;

May 2011

Post Job ALARA Review Work Order

432925; Contingency Replace Entire E SRV; June 2011 Post Job ALARA Review Work Order

421009; T-33 Modification; dated June 2011

RWP 00001158-02; Replace Fire Detectors in Locked High Radiation Areas

RWP 00001360-00; Replace

PS-2715 and Clear Lines

RWP 00001585-00; Weigh Old Steam Dryer

Section 2RS6

0147-01; A Train Standby Gas Treatment System Filter Tests; Revisions 35 and 36 0147-02; B Train Standby Gas Treatment System Filter Tests; Revision 35 0356; Stack / Vent Iodine / Particulate Samples; Revision 20

0385-A; Drywell Particulate Monitor Functional Test; Revision 22

0466-01; A EFT Filter Efficiency and Leak Tests; Revisions 31 and 34

0466-02; B EFT Filter Efficiency and Leak Tests; Revision 30

0468-01A; EFT Charcoal Filter Laboratory Test; Revision 07

1433;

TSC-ECS Leakage Test; Revision 11 8197; Painting in Secondary Containment and Adjacent Areas; Revision 17

2008 Land Use Census and Critical Receptor Report; September 2008

2008 Radioactive Effluent Release Report, May 2009

2009 Land Use Census and Critical Receptor Report; October 2009 2009 Radioactive Effluent Release Report, May 2010 2010 Land Use Census and Critical Receptor Report; October 2010

2010 Radioactive Effluent Release Report, May 2011

AR 01200859; Abnormal Release from TBNWS; October 2, 2009

AR 01232366; Failure of

ES-17-451B Causes Reactor Bldg Vent; May 12, 2010

AR 01238399; Adverse Trend Identified for PRM Equipment Condition; June 22, 2010

AR 01263610; Received Unexpected Alarm

ANN-5-A-2; Reactor Building Vent; December 20, 2010

Attachment

MTL 720; Minneapolis Testing Laboratory - Surveillance In-Place Leak Tests of High Efficiency Particulate Air Filters and Charcoal Absorbers and Monticello Nuclear Generating Plant;

Revision 1.0

NIST Traceable Certification and Calibration Reports; Freon Monitors and Photometers; various dates Section 2RS7

AWI-08.04.12; Ground Water Protection Program; Revision 04 10

CFR 50.75.g File Index and Selected Records; various dates

CFR 61 Database Updates (2009); September 2010

2008 Annual Radiological Environmental Monitoring Program Report; May 2009

2009 Annual Radiological Environmental Monitoring Program Report; May 2010

2010 Annual Radiological Environmental Monitoring Program Report; May 2011

Assessment 2010-003-5; Radiological Protection Including Environmental Monitoring;

October 2010

AR 01265191; Missed REMP Samples for Inclusion in 2011 REMP Report; January 5, 2011 I.05.33; Weekly Radiological Environmental Monitoring Procedures; Revision 06 I.05.37; Environmental Milk Sampling; Revision 06

I.05.43; REMP Air Sampler Calibration and Maintenance; Revision 03

I.05.50; Sampling Ground Water Monitoring Wells; Revision 08

Off-Site Dose Calculation Manual; Index Revision 04 NOS Report 2010-03-032; Environmental Monitoring; September 2010 Nuclear Oversight 3rd Quarter 2010 Assessment Report for Monticello; December 2010 Nuclear Oversight 4th Quarter 2010 Assessment Report for Monticello; February 2011

NUPIC Audit; Teledyne Brown Engineering Environmental Services; March 7, 2011

Radiation Protection DRUM Report 2nd Quarter 2010; July 2010

Radiation Protection DRUM Report 3rd Quarter 2010; October 2010 Radiation Protection DRUM Report 4th Quarter 2010; January 2011 Radiation Protection DRUM Report 1st and 2nd Quarter 2011; August 2011

Teledyne Report of Analysis/Certificate of Conformance; Drywell, Condensate, Steam Chase,

RWCU, and CA Plant Smears; May 2009, October 2009 and April 2010

WO 00361586-01; Met Tower Power Transformers; September 2008

Section 2RS8

5863; Dewatering Resins in High Integrity Containers; Revision 14

AR 01294652; Radioactive Shipment Container Punctured during Transport; July 14, 2011 Assessment 2010-002-5; Radiological Protection Including Radioactive Waste Control;

June 2010

CRSD11-049; Energy Solutions Condition Report;

Excel Energy Monticello Nuclear Generating Plant 0726-07-0013 GSAP; July 14, 2011

FP-RP-RW-02; Radioactive Shipping Procedure; Revision 03

Monticello Shipment Log Index; 2009, 2010 and 2011

Northern States Power Response to NRC IE Bulletin 79-19; September 25, 1979

NOS Observation Report 2011-007; Radiation Protection Including Radioactive Waste Control;

May 2011 Ops Manual B.07.03-05; Processing of Phase Separators or Spent Resin Tank Using Centrifuge / Hopper Bypass to Radwaste

RDS-100 Dewatering System Operation; Revision 16

Radioactive Material Shipment Package 09-46; April 2009

Radioactive Material Shipment Package 10-11; March 2010

Attachment Radioactive Material Shipment Package 10-23; May 2010

Radioactive Material Shipment Package 10-35; July 2010

Radioactive Material Shipment Package 11-54; April 2011

Radioactive Material Shipment Package 11-127; July 2011 R.11.07; Sampling and Analysis of Radioactive Resins; Revision 13

R.11.08; Selection and Entry of 10 CFR Part 61 Correlation Factors; Revision 08

Section 4OA1

Monticello MSPI Derivation Report; High Pressure Injection System; Unavailability Index, Unreliability Index, and Performance Limit Exceeded; July 29, 2011 Monticello MSPI Derivation Report; Emergency AC Power System; Unavailability Index, Unreliability Index, and Performance Limit Exceeded; July 29, 2011

Monticello MSPI Derivation Report; Heat Removal System; Unavailability Index, Unreliability Index, and Performance Limit Exceeded; July 29, 2011

I.03.13; Reactor Water and Cleanup Systems Iodine; Revision 13

22; Reactor Coolant I-131 Dose Equivalent Activity; Revision 26

FG-PA-KPI-01; Performance Indicator Data Reporting; Revision 00

FP-PA-PI-02; NRC/INPO/WANO Performance Indicator Reporting; Revision 06

QF-0445; NRC/INPO/WANO Data Collection and Submittal Forms; various dates

Section 4OA2

CAP 01297053; Door-189 Improperly Classified Tornado/Missile Barrier

CAP 01300697; B.09.02-05.H.3 Guidance Contrary to Main Transformer Design C.6-520-A-04; Winding Temp Hi; Revision 0 C.6-GEN111; Main Transformer Winding Temp; Revision 1*

Operations Manual B.09.02-05; Generati on:

System Operation; Revision 20

WO 00437951; Main Transformer 3 Cooler Displaying Signs of Degradation; Revision 3

CAP 01300252; Main Transformer Cooler 1 Middle Fan Degrading

CAP 01300152; Main Transformer Cooling Fan Motors Degraded

CAP 01299684; Main Transformer 3 Cooler Displaying Signs of Degradation

CAP 01297758; 3 Main Transformer Coolers Removed from Service due to Seized Fan

ODMI

01300152-02; GSU Transformer Will Have Decreased Cooling Capacity Until Fans Can

Be Replaced

2184; Reactor Building Operator Shift Turnover Checklist; Revision 15

AWI-04.01.01; General Plant Operating Activities; Revision 55

CAP 01218663; Ops. Challenge Tracking of 4kV Flood Barrier as TMOD

CAP 01250179; 2010 Operating Burden Tracking GAR

CAP 01264309; 2011 Operator Burden Tracking GAR

CAP 01296306; Unexpected Annunciator 3-A-3, RHR HDR Pressure Received

CAP 01301423; 3-B-2 Alarm Coming Every 40 Minutes

CAP 01302582; No. 13 RHRSW Pump Differential Pressure in Alert Range

CAP 01303564; Leaking Packing on HPCI H0-7 Stop Valve

FP-OP-OB-01; Operator Burden Program; Revision 2

MNGP Plant Status Report; September 22, 2011

NH-36249; High Pressure Coolant Injection System; Revision 78 Operations Manual B.03.02-01; HPCI - Function and General Description of System; Revision 9 Operations Manual B.03.02-02; HPCI - Description of Equipment; Revision 11

Operations Manual B.03.02-05; HPCI - System Operation; Revision 39

Attachment

QF-1150;

AS-1250179-09; Operator Burden- Identification and Impact- Flood Barrier, Lower

4KV Room

QF-1150;

AS-1264309-09; Operator Burden- Identification and Impact- CDM Vessel Outlet Valves

QF-1150;

AS-1264309-11; Operator Burden- Identification and Impact- ANN 3-B-3

QF-1150;

AS-1264309-22; Operator Burden- Identification and Impact- 13 RHRSW Pump

WM-0501; Operator Burden; September 20, 2011 Revision 0

Section 4OA3

EC 18336; Secondary Containment Safety Function Evaluation with 985 Pump Room Airlock Open 2300; Reactivity Adjustment; Revision 4; August 3, 2011

2300; Reactivity Adjustment; Revision 4; August 7, 2011

EC 17785; SCT Isolation Valve V-D-61 and V-D-62 Safety System Functional Failure Evaluation; Revision 0

WO 0424023; Validate SCIV Position without Actuator Connected

CAP 01270429; V-D-61 Damper Actuator Arm Bent, Preventing Opening

CAP 01289692; Rx Bldg dP < .25" Water Vac While Swapping V-AH-A/B Fans Operations Manual B.08.07-05; Transfer between V-AH-4A and V-AH-4B; Revision 27

CAP 01292658; APRMs 1, 3, and 4, Exceed Maximum Deviation from Core Thermal Power

OPS-ECC-0566; Low Pressure ECCS Automatic Initiation and Loss of Auxiliary Power Test;

Revision 8

OPS-EDG-0535-11; 11 Emergency Diesel Generator Load Rejection Test; Revision 1

CAP 01305683; CDBI FSA- Question on Definition of Post Accident Load in TS

EC 18798; 11 and 12 EDG Single Largest Load Evaluation

PRA-MEMO-11-011; Risk Assessment of Inadequate EDG Load Shed Surveillance;

September 29, 2011

Section 4OA5

SCR-08-0348; Reactor and Core Components Handling Equipment - Tool and Equipment Operation; Revision 2

SCR-08-0465; DSC Shell Deformation for Dry Run Insert / Retrieve; Revision 0

SCR-10-0392; [FSAR Change Notices]; Revision 0

CAP 01163419; Contamination found on Transfer Trailer following Transfer of Transfer Cask to Shipping Container; December 22, 2008

CAP 01183567; QA Program Boundary AWI May be Overly-Conservative with Respect to

ISFSI; May 28, 2009

CAP 01198912; Concern with ISFSI Separation Distance Flammables; September 21, 2009

CAP 01238753; No Independent Site Supervisor for 72.48 Screening Reviews; June 24, 2010

CAP 01244081; Inadequate Guidance in B.10.01-05.H.1, Revision 0; August 3, 2010

CAP 01290843; Spent Fuel Cask Height Restriction not in USAR; June 26, 2011

CAP 01304966; Structural PM Needed of ISFSI Concrete Array; September 22, 2011

CAP 01304969; Question of Checking of Bundle Serial Number - DSC Loading;

September 22, 2011

CAP 01304997; 9506 Converts TS Gauge Pressure Using Standard Pressure; September 22, 2011

CAP 01305019; Control of Hazardous Deliveries and Proximity of ISFSI; September 22, 2011

CAP 01305022; Potential Enhancement to QA Program Boundary Procedure;

September 22, 2011

Attachment 0000-H; Operations Daily Log- Part H; Revision 94 4

AWI-06.06.01; Site Rigging, Lifting, and Material Handling Program; Revision 30

28; Radiation Protection Survey Record [ISFSI]; May, 31, 2011 9009; Procedure for Moving Fuel within the Fuel Storage Pool; Revision 20 9506; Dry Shielded Canister Sealing; Revision 5

B.10.01-05; Operations Manual Section - Dry Fuel Storage - ISFSI; Revision 0

NUC-06.02; Selecting Fuel Bundles for ISFSI Storage; Revision 0

00403551; Special Nuclear Material Physical Inventory; December 17, 2011

Calculation of Final DSC Helium Backfill Pressure Using Site Ambient Pressure from Meteorological Tower (MET100); September 22, 2011 Letter; TN Response to Questions [Regarding Helium Backfill Pressure]; September 22, 2011

Monticello Nuclear Generating Plant ISFSI 72.212 Evaluation Report; Revision 1

NUH-06-106M; Certificate of Conformance, OS197-1 On-Site Transfer Cask and Lifting Yoke;

July 29, 2008 NUHOMS Cask Number 1 Loading Report TN

NCR 2008-221; Certificate of Conformance

HUHOMS-61BT Dry Shielded Canister (Repair);

Revision 0

Attachment

LIST OF ACRONYMS

USED [[]]

AC Alternating Current

ADAMS Agencywide Document Access Management System

ALARA As-Low-As-Is-Reasonably-Achievable

APRM Average Power Range Monitor

CAP Corrective Action Program CFR Code of Federal Regulations

CoC Certificate of Compliance

CRV Control Room Ventilation

DRP Division of Reactor Projects

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator

ESW Emergency Service Water

FSAR Final Safety Analysis Report

HPCI High Pressure Coolant Injection

HRA High Radiation Area

IMC Inspection Manual Chapter

IP Inspection Procedure

ISFSI Independent Spent Fuel Storage Facility Installation

IST Inservice Test

kV Kilovolt

LER Licensee Event Report

LCO Limiting Condition for Operation

MG Motor Generator

MNGP Monticello Nuclear Generating Plant

MSPI Mitigating Systems Performance Indicator

NCV Non-Cited Violation

NEI Nuclear Energy Institute

NFPA National Fire Protection Association

NOED Notice of Enforcement Discretion

NRC [[]]

U.S. Nuclear Regulatory Commission

ODCM Offsite Dose Calculation Manual

OWA Operator Workaround

PARS Publicly Available Records System

PI Performance Indicator

PM Post Maintenance

RCA Radiological Controlled Area

RCIC Reactor Core Isolation Cooling

RHR Residual Heat Removal

RWCU Reactor Water Cleanup

SBLC Standby Liquid Control

SDP Significance Determination Process

SRV Safety Relief Valve

SSC Structure, System, and Component

TS Technical Specification

USAR Updated Safety Analysis Report

URI Unresolved Item

VHRA Very High Radiation Area WO Work Order

T. O'Connor -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 document system (ADAMS).

ADAMS is accessible from the

NRC Website

at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely, /RA/ Kenneth Riemer, Chief

Branch 2

Division of Reactor Projects

Docket Nos. 50-263;72-058

License No. DPR-22

Enclosure: Inspection Report 05000263/2011004 w/Attachment: Supplemental Information cc w/encl: Distribution via ListServ

DISTRI BUTION

See next page

DOCUMENT NAME: G:\DRPIII\1-Secy\1-Work In Progress\MON 2011 004.docx 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 [[]]

ERIII [[]]

RIII [[]]

NAME NShah:cs KRiemer

DATE 10/31/11 11/01/11

OFFICI AL

RECORD [[]]

COPY Letter to

T. O'Connor from K. Riemer dated November 1, 2011

SUBJEC T: MONTICELLO NUCLEAR GENERATING

PLANT [[]]