ML070860170
| 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
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To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE
RIII
RIII
RIII
NAME
ADunlop: ls
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:
Report No:
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
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
Agency-Wide Document Access and Management System
Action Request
CFR
Code of Federal Regulations
Division of Reactor Projects
Division of Reactor Safety
EC
Engineering Change
Electric Power Research Institute
IMC
Inspection Manual Chapter
IR
Inspection Report
Non-Cited Violation
NEI
Nuclear Energy Institute
NRC
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
Publicly Available Records
Screening (50.59)
Setpoint Change Request
Significance Determination Process
Safety Evaluation (50.59)
TS
Technical Specifications
Updated Safety Analysis Report