ML070860170

IR 05000263-07-006( Drs); 02/12/2007 Through 03/01/2007; Monticello Nuclear Generating Plant. Evaluations of Changes, Tests, Experiments and Permanent Plant Modifications

ML070860170
Person / Time
Site:	Monticello
Issue date:	03/27/2007
From:	Dave Hills NRC/RGN-III/DRS/EB1
To:	Conway J Nuclear Management Co
References
IR-07-006
Download: ML070860170 (19)

See also: IR 05000263/2007006

Text

March 27, 2007

Mr. J. Conway

Site Vice President

Monticello Nuclear Generating Plant

Nuclear Management Company, LLC

2807 West County Road 75

Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT NRC EVALUATION OF

CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT

MODIFICATIONS BASELINE INSPECTION REPORT 05000263/2007006(DRS)

Dear Mr. Conway:

On March 1, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed a combined

baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant

Modifications at the Monticello Nuclear Generating Plant. The enclosed report documents the

results of the inspection, which were discussed with Mr. J. Grubb, and others of your staff at the

completion of the inspection on March 1, 2007.

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

compliance with the Commissions Rules and Regulations, and with the conditions of your

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

interviewed personnel.

Based on the results of the inspection, one NRC identified finding, which involved a violation of

NRC requirements of very low safety significance, was identified. Because of the very low

safety significance of the violation and the fact that the issue was entered into the licensees

corrective action program, the NRC is treating the finding as a Non-Cited Violation (NCV) in

accordance with Section VI.A.1 of the NRCs Enforcement Policy.

In accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of this letter

and its enclosure will be available electronically for public inspection in the NRC Public

Document Room, or from the Publicly Available Records (PARS) component of NRC's

J. Conway

-2-

document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-

rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

David E. Hills, Chief

Engineering Branch 1

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure:

Inspection Report 05000263/2007006(DRS)

w/Attachment: Supplemental Information

cc w/encl:

M. Sellman, President and Chief Executive Officer

Manager, Nuclear Safety Assessment

J. Rogoff, Vice President, Counsel, and Secretary

Nuclear Asset Manager, Xcel Energy, Inc.

State Liaison Officer, Minnesota Department of Health

R. Nelson, President

Minnesota Environmental Control Citizens

Association (MECCA)

Commissioner, Minnesota Pollution Control Agency

D. Gruber, Auditor/Treasurer,

Wright County Government Center

Commissioner, Minnesota Department of Commerce

Manager - Environmental Protection Division

Minnesota Attorney Generals Office

J. Conway

-2-

document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-

rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

David E. Hills, Chief

Engineering Branch 1

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure:

Inspection Report 05000263/2007006(DRS)

w/Attachment: Supplemental Information

cc w/encl:

M. Sellman, President and Chief Executive Officer

Manager, Nuclear Safety Assessment

J. Rogoff, Vice President, Counsel, and Secretary

Nuclear Asset Manager, Xcel Energy, Inc.

State Liaison Officer, Minnesota Department of Health

R. Nelson, President

Minnesota Environmental Control Citizens

Association (MECCA)

Commissioner, Minnesota Pollution Control Agency

D. Gruber, Auditor/Treasurer,

Wright County Government Center

Commissioner, Minnesota Department of Commerce

Manager - Environmental Protection Division

Minnesota Attorney Generals Office

DOCUMENT NAME:C:\\FileNet\\ML070860170.wpd

G Publicly Available

G Non-Publicly Available

G Sensitive

G Non-Sensitive

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

OFFICE

RIII

RIII

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NAME

ADunlop: ls

DHills

DATE

03/27/07

03/27/07

OFFICIAL RECORD COPY

J. Conway

-3-

DISTRIBUTION:

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ROPreports@nrc.gov

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No:

50-263

License No:

DPR-22

Report No:

05000263/2007006(DRS)

Licensee:

Nuclear Management Company, LLC

Facility:

Monticello Nuclear Generating Plant

Location:

Monticello, Minnesota

Dates:

February 12, 2007 through March 1, 2007

Inspectors:

A. Dunlop, Senior Reactor Inspector

T. Bilik, Reactor Inspector

Observers:

V. Meghani, Reactor Inspector

Approved by:

D. Hills, Chief

Engineering Branch 1

Division of Reactor Safety (DRS)

Enclosure

1

SUMMARY OF FINDINGS

IR 05000263/2007006(DRS); 02/12/2007 through 03/01/2007; Monticello Nuclear Generating

Plant. Evaluations of Changes, Tests, Experiments and Permanent plant modifications.

The inspection covered a 2-week announced baseline inspection on evaluations of changes,

tests, or experiments and permanent plant modifications. The inspection was conducted by

two regional based engineering inspectors. One Green finding associated with a Non-Cited

Violation (NCV) was identified. The significance of most findings is indicated by their color

(Green, White, Yellow, Red) using Inspection Manual Chapter 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, Reactor Oversight Process, Revision 3; dated July 2000.

A.

Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green. The inspectors identified a Severity Level IV NCV for an inadequate 10 CFR 50.59, Changes, Tests, and Experiments, evaluation resulting in failure to receive

prior NRC approval for changes in licensed activities associated with protection of

the emergency diesel generator exhaust stacks against tornado generated missiles.

Specifically, the licensee did not provide an adequate response to the question posed

in 10 CFR 50.59(c)(2)(viii), and did not demonstrate that the proposed change did not

result in a departure from a method of evaluation described in the Final Safety Analysis

Report (as updated) used in establishing the design bases or in the safety analyses. As

part of the corrective actions, the licensee verified that the emergency diesel generators

remained operable and initiated actions to submit a licensee amendment request for use

of the new methodology.

Because the Significance Determination Process is not designed to assess the

significance of violations that potentially impact or impede the regulatory process, this

issue was dispositioned using the traditional enforcement process in accordance with

Section IV of the NRC Enforcement Policy. However, the results of the violation, that is,

the failure to demonstrate that the proposed change did not result in a departure from a

method of evaluation, were assessed using the Significance Determination Process.

The finding was determined to be greater than minor because the change had the

potential for impacting the NRCs ability to perform its regulatory function as the

inspectors determined the change would have required prior NRC approval. The

finding was of very low safety significance based on the completed analysis for the

emergency diesel generator exhausts. This was determined to be a Severity Level IV

NCV of 10 CFR 50.59. (Section 1R02)

B.

Licensee-Identified Violations

No findings of significance were identified.

Enclosure

2

REPORT DETAILS

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R02

Evaluations of Changes, Tests, or Experiments (71111.02)

.1

Review of 10 CFR 50.59 Evaluations and Screenings

a.

Inspection Scope

From February 12, 2007, through March 1, 2007, the inspectors reviewed two

evaluations performed pursuant to 10 CFR 50.59. The inspectors reviewed the

evaluations to confirm that they were thorough and that prior NRC approval was

obtained as appropriate. The inspector could not review the minimum sample size of

five evaluations because the licensee only performed one evaluation during the biennial

sample period. One additional safety evaluation was reviewed that was performed in

the previous sample period for a total of two samples. The inspectors also reviewed

18 screenings where licensee personnel had determined that a 10 CFR 50.59

evaluation was not necessary. In addition, seven applicability determinations were

reviewed to verify they did not meet the applicability requirements for a screening. The

evaluations and screenings were chosen based on risk significance, safety significance,

and complexity. The list of documents reviewed by the inspectors are included as an

attachment to this report.

The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for

10 CFR 50.59 Implementation, and Revision 1, to determine acceptability of the

completed evaluations, and screenings. The NEI document was endorsed by the

NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59,

Changes, Tests, and Experiments, dated November 2000. The inspectors also

consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.

b.

Findings

Inadequate 10 CFR 50.59 Evaluation for Diesel Generator Exhaust Missile Protection

Introduction: The inspectors identified an inadequate evaluation performed pursuant to

10 CFR 50.59 associated with the vulnerability of the emergency diesel generator (EDG)

exhaust stacks to tornado generated missiles. Specifically, the licensee did not provide

an adequate response to the question posed in 10 CFR 50.59(c)(2)(viii) and did not

demonstrate that the proposed change did not result in a departure from a method of

evaluation described in the USAR used in establishing the design bases or in the safety

analyses. This issue was considered to be of very low safety significance (Green) and

was dispositioned as a Severity Level IV Non-Cited Violation (NCV).

Enclosure

3

Description: The inspectors reviewed 10 CFR 50.59 safety evaluation (SE)03-004,

concerning the utilization of the TORMIS probabilistic risk assessment (PRA)

methodology (per Electric Power Research Institute (EPRI) Report NP-2005,

Volumes 1 and 2). This methodology was to verify that the risk from tornado

generated missiles was sufficiently small to justify leaving the EDG exhaust

unprotected. On page 7 of SE 03-004 in Section III.8, the licensee responded to the

question posed in 10 CFR 50.59(c)(2)(viii). This question asked, "Does the proposed

change result in a departure from a method of evaluation described in the Final Safety

Analysis Report (as updated) used in establishing the design bases or in the safety

analyses? The licensee justified the No answer to this question by citing the NRC

acceptance of the EPRI methodology for specific plant features and subject to resolution

of specific concerns in the NRCs safety evaluation for EPRI Report NP-2005, dated

October 26, 1983. The licensees evaluation included addressing the specific

concerns and stated that the resolutions of these concerns for the Monticello plant

were consistent with those accepted by the NRC for the D. C. Cook Nuclear Plant

(Amendment No. 247 to DPR-58 and Amendment No. 228 to DPR-74).

The NRCs safety evaluation concluded that the PRA methodology as contained in the

EPRI report was an acceptable probabilistic approach for demonstrating compliance

with the requirements of General Design Criteria 2 and 3 regarding protection of

safety-related plant features from the effects of tornado and high wind generated

missiles, but subject to the additional concerns identified. It further stated that use of

the EPRI or any tornado missile probabilistic study should be limited to the evaluation of

specific plant feature where additional costly tornado missile protective barriers or

alternative systems were under consideration. The inspectors contacted the staff in the

Office of Nuclear Reactor Regulation (NRR) to determine the basis for the NRCs safety

evaluation and the acceptability of the licensee using this methodology that was not in

accordance with the current licensing basis. Based on this discussion, although the

methodology had been reviewed and could be used as a basis for not having to

physically protect specific plant features from tornado generated missiles, it was

considered a change to the plants current licensing basis, which required a license

amendment.

Based on the above, the inspectors concluded that the licensee use of NRCs safety

evaluation on the EPRI methodology was incorrect and that the licensees No answer

to 10 CFR 50.59(c)(2)(viii), and the conclusion that no activity requiring prior NRC

approval per 10 CFR 50.59 was identified were not justified.

The inspectors also determined that the results of the calculations based on the EPRI

methodology discussed above were utilized for responses to the questions for

10 CFR 50.59(c)(2) (i) through (vi) in Section III of the SE 03-004 and that a USAR

change was implemented to incorporate the use of TORMIS methodology. This finding

also affected the licensees 10 CFR 50.59 screening SCR-04-0069, Revision 0, which

was used to screen out activities involving subsequent application of the EPRI

methodology for evaluation of other plant specific features from tornado generated

missiles.

Enclosure

4

In response to the finding, the licensee initiated Action Request (AR) 01079705. The

licensee determined that the NRCs 1983 safety evaluation endorsing the EPRI TORMIS

methodology was misinterpreted by the licensee as a generic NRC approval for use and

was inappropriately used in the 50.59 evaluation to conclude that prior NRC approval

was not required. The licensee determined the EDGs remained operable based on the

existing completed analysis and acceptance of similar technical approach by the NRC

for other operating plants. The inspectors concluded that the licensees determination

was acceptable as the existing analysis using the TORMIS methodology did appear to

address the limitations noted in the NRCs safety evaluation. The AR also

recommended an action to submit an license amendment request to the NRC to

incorporate the TORMIS methodology into the license basis for all the affected plant

specific features.

Analysis: This issue was determined to involve a performance deficiency because the

licensee incorrectly concluded that the TORMIS methodology had been approved for

generic application and consequently concluded that prior NRC approval was not

required when such a conclusion could not be supported by the documented 50.59

evaluation. Because violations of 10 CFR 50.59 are considered to be violations that

potentially impede or impact the regulatory process, they are dispositioned using the

traditional enforcement process instead of the significance determination process (SDP)

described in Inspection Manual Chapter (IMC) 0609, "Significance Determination

Process. The finding was determined to be greater than minor because the change

had the potential for impacting the NRCs ability to perform its regulatory function as the

inspectors determined the change would have required prior NRC approval.

The inspectors evaluated the finding using IMC 0609, Appendix A, Significance

Determination of Reactor Inspection Findings for At-Power Situations, Phase 1

screening, and determined that the finding screened as Green because it was not a

design issue resulting in loss of function per Part 9900, Technical Guidance,

Operability Determinations, and Functionality Assessments for Resolution of Degraded,

or Nonconforming Conditions Adverse to Quality or Safety, did not represent an actual

loss of a system safety function, did not result in exceeding a technical specification

allowed outage time, and did not affect external event mitigation. This was based on the

licensees operability determination that concluded that their use of the TORMIS

methodology appeared to be consistent with the guidance provided in the NRCs safety

evaluation of the methodology and that NRC had accepted its use at other plants when

used for the intended purpose. The inspectors did not identify a cross-cutting aspect

with this finding.

Enforcement: Title 10 CFR 50.59(c)(2)(viii) states, in part, that a licensee shall obtain a

license amendment pursuant to Section 50.90 prior to implementing a proposed change,

test, or experiment if the change, test, or experiment would result in a departure from a

method of evaluation described in the Final Safety Analysis Report (as updated) used in

establishing the design bases or in the safety analyses.

Contrary to the above, on July 28, 2003, the licensee approved a 10 CFR 50.59

evaluation (SE-03-004) incorporating a change to the tornado missile protection

methodology for the EDG exhaust system, which resulted in a departure from a method

of evaluation described in the USAR, without obtaining a license amendment. However,

Enclosure

5

because this violation was of very low safety significance and it was entered into the

licensees corrective action program, this Severity Level IV violation is being treated as

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

(NCV 05000263/2007006-01(DRS)). The licensee entered the finding into their

corrective action program as AR01079705.

1R17

Permanent Plant Modifications (71111.17B)

.1

Review of Permanent Plant Modifications

a.

Inspection Scope

From February 12, 2007, through March 1, 2007, the inspectors reviewed ten

permanent plant modifications that had been installed in the plant during the last two

years. This included two engineering changes, three equivalency evaluations, and five

setpoint changes. The modifications were chosen based upon risk significance, safety

significance, and complexity. As per inspection procedure 71111.17B, two modifications

were chosen that affected the barrier integrity cornerstone. The inspectors reviewed the

modifications to verify that the completed design changes were in accordance with the

specified design requirements, and the licensing bases, and to confirm that the changes

did not adversely affect any systems' safety function. Design and post-modification

testing aspects were verified to ensure the functionality of the modification, its

associated system, and any support systems. The inspectors also verified that the

modifications performed did not place the plant in an increased risk configuration.

The inspectors also used applicable industry standards to evaluate acceptability of the

modifications. The list of modifications and other documents reviewed by the inspectors

is included as an attachment to this report.

b.

Findings

No findings of significance were identified.

4.

OTHER ACTIVITIES (OA)

4OA2 Identification and Resolution of Problems

.1

Routine Review of Condition Reports

a.

Inspection Scope

From February 12, 2007, through March 1, 2007, the inspectors reviewed 18 Corrective

Action Process documents that identified or were related to 10 CFR 50.59 evaluations

and permanent plant modifications. The inspectors reviewed these documents to

evaluate the effectiveness of corrective actions related to permanent plant modifications

and evaluations for changes, tests, or experiments issues. In addition, corrective action

documents written on issues identified during the inspection were reviewed to verify

adequate problem identification and incorporation of the problems into the corrective

Enclosure

6

action system. The specific corrective action documents that were sampled and

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

b.

Findings

No findings of significance were identified.

4OA6 Meetings

.1

Exit Meeting

The inspectors presented the inspection results to Mr. J. Grubb and others of the

licensees staff, on March 1, 2007. Licensee personnel acknowledged the inspection

results presented. Licensee personnel were asked to identify any documents, materials,

or information provided during the inspection that were considered proprietary. No

proprietary information was identified.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Attachment

1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. Baumer, Licensing

F. Domke, Electrical Design Supervisor

J. Grubb, Engineering Director

B. Guldemond, Nuclear Safety Assurance Manager

N. Haskell, Engineering Design Manager

T. Hurrle, Configuration Management Supervisor

D. Nordell, Configuration Management Engineer

J. Ohotto, Design Engineering Supervisor

D. Pennington, Design Engineer

B. Sawatzke, Plant Manager

Nuclear Regulatory Commission

D. Hills, Chief, Engineering Branch 1, Division of Reactor Safety

S. Thomas, Senior Resident Inspector

L. Haeg, Resident Inspector

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened/Closed

05000263/2007006-01

NCV

Inadequate 10 CFR 50.59 Evaluation for Diesel Generator

Exhaust Missile Protection (Section 1R21.3.b)

Attachment

2

LIST OF DOCUMENTS REVIEWED

The following is a list of licensee documents reviewed during the inspection, including

documents prepared by others for the licensee. Inclusion on this list does not imply that NRC

inspectors 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 in this list does not imply NRC acceptance of the document, unless specifically stated

in the inspection report.

IR02

Evaluation of Changes, Tests, or Experiments 71111.02

10 CFR 50.59 Evaluations

SE-03-004; Diesel Exhaust Missile Protection Design Consideration; dated

July 28, 2003

SE-06-003; SBO Operator Actions Associated with the HPCI System; dated

September 19, 2006

10 CFR 50.59 Screenings

SCR-04-0283; SRV Air Actuator Model Change; dated November 23, 2005

SCR-04-0859; HPCI Turbine Steam Supply Low Pressure Isolation; dated

September 11, 2006

SCR-05-0161; Set Point for RHR Minimum Flow Switches FS-10-2-121, A, B, C and D;

dated August 23, 2006

SCR-05-0242; Instrument Setpoint Calculation 4.16KV Degraded Voltage; dated

March 28, 2006

SCR-05-0266; ITS Setpoint Change - HPCI Steam Line Area Temperature - High;

dated August 26, 2006

SCR-05-0689; Calc CA-05-146, Evaluation of Wall Thinning on FW2B-10-ED; dated

October 11, 2005

SCR-05-0738; Calc CA-05-028, Evaluation of HPCI Condensate Drain Line D13-2"-HE

in the HPCI Room; dated November 9, 2005

SCR-05-0739; Calc 05-147, Evaluation of HPCI Module E.2; dated November 9, 2005

SCR-05-0757; Chilled Water Expansion Tank V-CT-1 Replacement; dated

November 15, 2005

SCR-05-0822; CA-05-155, Evaluation of Offgas Stack for SSE Seismic Loads; dated

December 22, 2005

Attachment

3

SCR-06-0062; Less than Full Thread Engagement on RWCU AO Valve Actuator Bonnet

Nuts; dated February 15, 2006

SCR-06-0103; HPCI Steam Void Elimination; dated April 6, 2006

SCR-06-0106; Service Water Pump Replacement; October 30, 2006

SCR-06-0165; Replace AO-1575(6) and Check Valves with Normally Closed Valve

SW-228(9); dated October 31, 2006

SCR-06-0166; Replace Rotork Actuators on Five MOVs with Limitorque Actuators;

dated April 26, 2006

SCR-06-0310; Technical Requirements Manual - Appendix B - Secondary Containment

Isolation Valves; dated September 12, 2006

SCR-06-0557; Suppression Chamber Inspection; dated December 4, 2006

SCR-07-0043; Fuel Storage and Handling Systems, Design Basis; dated

January 22, 2007

10 CFR 50.59 Applicability Determinations

SCR-05-0645; Drawing Classification Level Change to 3'; dated September 19, 2005

SCR-05-0657; Combustible Loading Calculation; dated September 22, 2005

SCR-05-0663; Replace Fusible Link on V-DF-SBGT-2 with One of a Higher

Temperature Rating; dated September 28, 2005

SCR-05-0791; Evaluation of Fire Detector Locations in the Reactor Building; dated

December 5, 2005

SCR-05-0819; Setpoint Change Request for the Safety/Relief Valve Low-Low Set Logic

to Incorporate the New Trip Settings; dated December 21, 2005

SCR-05-0830; Setpoint Change Request for the 4KV Bus-15 and Bus 16 Undervoltage

Relays to Incorporate the New Trip Setting; dated January 3, 2006

SCR-06-0308; Update USAR for Improved Technical Specification Project; dated

July, 29, 2006

IR17

Permanent Plant Modifications 71111.17B

Modifications

EC8819; HPCI Steam Line Area Temperature - High; dated October 27, 2006

EC7583; Degraded Voltage Relays for Safety-Related 4KV Busses ; dated

August 7, 2006

Attachment

4

Equivalency Evaluations

EC910; Replacement Blower Wheel; Revision 1

EC933 (05A099); HPCI Auxiliary Lube Oil Pump; Revision 0

EC7828; Engine Driven Fuel Pump Suction Line; Revision 0

Setpoint Changes

EC8818; HPCI Turbine Steam Line Pressure - Low; dated October 27, 2006

EC8792; LPCI Pump Discharge Flow - Low; dated October 27, 2006

SCR 05-022; 4KV Bus-15 and Bus-16 Undervoltage Relays; dated December 1, 2005

SCR 05-023; Main Steam Line Steam Chase High Temp Group 1 Isolation; dated

December 1, 2005

SCR 05-028; SRV Low Low Set Pressure Interlock; dated December 1, 2005

Other Documents Reviewed During Inspection

Corrective Action Program Documents Generated As a Result of Inspection

AR01076896; List to NRC Screened out All 50.59 Screening using the 3283 Form;

AR01077202; SCR-05-0830 Description Contains Incorrect Value; dated

February 14, 2007

AR01077855; Action to Correct Drawing Error was Cancelled; dated February 19, 2007

AR01078665; Error in Calculation CA-05-146, Evaluation of Wall Thinning in

FW2B-10"-ED; dated February 22, 2007

AR01079705; LAR Required for Use of TORMIS Code Methodology; dated

February 28, 2007

AR01080049; SCR-05-0161 Activity Incorrectly Categorized; dated March 1, 2007

Corrective Action Program Documents Reviewed During the Inspection

AR00824446; NDE Thickness < 87.5 percent TNOM on FW2B-10"-ED, B Feedwater

to Reactor Line; March 25, 2005

AR00891838; Evidence of Water Leakage on 11 and 12 EDG Exhaust Pipe Insulation;

dated September 28, 2005

AR01000610; Replacement Part does not Match the Part Removed; dated

October 10, 2005

Attachment

5

AR01000746; Inconsistency Between Line Design Table and Plant; dated

October 11, 2005

AR01001520; Operation past One Cycle Not Assured for Fw Pipe; dated

October 20, 2005

AR01003632; RC-44-2 Replacement Noticed 3000 No. vs. 6000 No.; dated

November 14, 2005

AR01004032; RWC Pipe Support Discrp and Indad Thread Engage on Act Nuts; dated

November 17, 2005

AR01006064; CV-1728 Plug Replaced, No Section XI Repair/Replacement Plan; dated

December 1, 2005

AR01008347; Some SW Mods May Inadvertently Create New Problems; dated

December 21, 2005

AR01022687; SW 1-18"-JF Does Not Meet Class 1 Design Criteria ; dated April 6, 2006

AR01026395; Potential Exists for Failure to Manually Start ECCS Room Coolers; dated

April 26, 2006

AR01040014; Inadequate Closeout Activities for Design Change 99Q160; dated

July 17, 2006

AR01059716; Change to PM Frequency not Considered; dated November 3, 2006

AR01059908; Adverse Trend in Modification Implementation; dated November 6, 2006

AR00891237; No Column Gaskets Found on RHRSW Pump Columns; dated

September 27, 2005

AR1040142; B.03.04-05 Issued Prior to Completion of Revision Process; dated

July 18, 2006

AR0780295; Revise USAR Section 10.2.4.3 to Reflect the Results of CA-95-028; dated

November 26, 2006

AR01045206; 50.59 Screening SCR-05-210 Missed USAR Impact; dated

August 18, 2006

Calculations

CA-03-038; Instrument Setpoint Calculation, 4.16 KV Loss of Voltage; Revision 1

CA-03-039; Instrument Setpoint Calculation - SRV Low-Low Set, Reactor Coolant

System Pressure; Revision 0

CA-04-110; Determination of HPCI Area High Temperature Setpoints; Revision 1

Attachment

6

CA-05-108; Evaluation of Wall Thinning on FW2B-10-ED Piping; Revision 0

CA-05-146; Evaluation of Wall Thinning on FW2B-10"-ED Piping; Revision 0

Drawings

EC-811-01; Monticello Nuclear Generating Plant Installation of HPCI Void Resolution;

Revision 1

NH-36250; Monticello Nuclear Generating Plant P&ID (Water Side) High Pressure

Coolant Injection System; Revision AF

Attachment

7

LIST OF ACRONYMS USED

ADAMS

Agency-Wide Document Access and Management System

AR

Action Request

CFR

Code of Federal Regulations

DRP

Division of Reactor Projects

DRS

Division of Reactor Safety

EDG

Emergency Diesel Generator

EC

Engineering Change

EPRI

Electric Power Research Institute

IMC

Inspection Manual Chapter

IR

Inspection Report

NCV

Non-Cited Violation

NEI

Nuclear Energy Institute

NRC

Nuclear Regulatory Commission

NRR

Office of Nuclear Reactor Regulation

PARS

Publicly Available Records

PRA

Probabilistic Risk Assessment

SCR

Screening (50.59)

SCR

Setpoint Change Request

SDP

Significance Determination Process

SE

Safety Evaluation (50.59)

TS

Technical Specifications

USAR

Updated Safety Analysis Report