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, 2007Mr. J. ConwaySite Vice President
Monticello Nuclear Generating Plant
Nuclear Management Company, LLC
2807 West County Road 75
Monticello, MN 55362-9637SUBJECT:MONTICELLO NUCLEAR GENERATING PLANT NRC EVALUATION OFCHANGES, 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 combinedbaseline 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 andcompliance 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 the inspection, one NRC identified finding, which involved a violation ofNRC 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 licensee's
corrective action program, the NRC is treating the finding as a Non-Cited Violation (NCV) in
accordance with Section VI.A.1 of the NRC's Enforcement Policy. In accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of this letterand its enclosure will be available electronically for public inspection in the NRC PublicDocument 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, ChiefEngineering Branch 1
Division of Reactor SafetyDocket No. 50-263License No. DPR-22Enclosure:Inspection Report 05000263/2007006(DRS) w/Attachment: Supplemental Informationcc w/encl:M. Sellman, President and Chief Executive OfficerManager, 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 General's 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, ChiefEngineering Branch 1
Division of Reactor SafetyDocket No. 50-263License No. DPR-22Enclosure:Inspection Report 05000263/2007006(DRS) w/Attachment: Supplemental Informationcc w/encl:M. Sellman, President and Chief Executive OfficerManager, 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 General's OfficeDOCUMENT NAME:C:\FileNet\ML070860170.wpd
G Publicly Available
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G Non-SensitiveTo receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copyOFFICERIIIRIII RIIINAMEADunlop: lsDHillsDATE03/27/0703/27/07OFFICIAL RECORD COPY
J. Conway-3-DISTRIBUTION
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ROPreports@nrc.gov
U.S. NUCLEAR REGULATORY COMMISSIONREGION IIIDocket No:50-263License No:DPR-22Report 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 InspectorT. Bilik, Reactor InspectorObservers:V. Meghani, Reactor InspectorApproved by:D. Hills, ChiefEngineering Branch 1
Division of Reactor Safety (DRS)
Enclosure 1SUMMARY OF FINDINGSIR 05000263/2007006(DRS); 02/12/2007 through 03/01/2007; Monticello Nuclear GeneratingPlant. 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 FindingsCornerstone: Mitigating SystemsGreen. The inspectors identified a Severity Level IV NCV for an inadequate 10 CFR50.59, "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 thesignificance 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 thepotential for impacting the NRC's 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 ViolationsNo findings of significance were identified.
Enclosure 2REPORT DETAILS1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R02Evaluations of Changes, Tests, or Experiments (71111.02).1Review of 10 CFR 50.59 Evaluations and Screenings a.Inspection ScopeFrom February 12, 2007, through March 1, 2007, the inspectors reviewed twoevaluations 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.FindingsInadequate 10 CFR 50.59 Evaluation for Diesel Generator Exhaust Missile Protection Introduction: The inspectors identified an inadequate evaluation performed pursuant to10 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 3Description: 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 thequestion 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 NRC's safety evaluation for EPRI Report NP-2005, dated
October 26, 1983. The licensee's 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 NRC's safety evaluation concluded that the PRA methodology as contained in theEPRI 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 NRC's 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 plant's current licensing basis, which required a licenseamendment.Based on the above, the inspectors concluded that the licensee use of NRC's safetyevaluation on the EPRI methodology was incorrect and that the licensee's "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 EPRImethodology 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 USARchange was implemented to incorporate the use of TORMIS methodology. This finding
also affected the licensee's 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 4In response to the finding, the licensee initiated Action Request (AR) 01079705. Thelicensee determined that the NRC's 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 licensee's determination
was acceptable as the existing analysis using the TORMIS methodology did appear to
address the limitations noted in the NRC's 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 thelicensee 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 NRC's 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, "SignificanceDetermination 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
licensee's operability determination that concluded that their use of the TORMIS
methodology appeared to be consistent with the guidance provided in the NRC's 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 alicense 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 50.59evaluation (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 5because this violation was of very low safety significance and it was entered into thelicensee's 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 AR01079705.1R17Permanent Plant Modifications (71111.17B).1Review of Permanent Plant Modifications a.Inspection ScopeFrom February 12, 2007, through March 1, 2007, the inspectors reviewed tenpermanent 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 themodifications. The list of modifications and other documents reviewed by the inspectors
is included as an attachment to this report. b.FindingsNo findings of significance were identified.4.OTHER ACTIVITIES (OA)4OA2Identification and Resolution of Problems.1Routine Review of Condition Reports a.Inspection ScopeFrom February 12, 2007, through March 1, 2007, the inspectors reviewed 18 CorrectiveAction 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 6action system. The specific corrective action documents that were sampled andreviewed by the inspectors are listed in the attachment to this report. b.FindingsNo findings of significance were identified.4OA6Meetings.1Exit MeetingThe inspectors presented the inspection results to Mr. J. Grubb and others of thelicensee's 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
1SUPPLEMENTAL INFORMATIONKEY POINTS OF CONTACT
LicenseeR. Baumer, LicensingF. 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 ManagerNuclear Regulatory CommissionD. Hills, Chief, Engineering Branch 1, Division of Reactor SafetyS. Thomas, Senior Resident Inspector
L. Haeg, Resident InspectorITEMS OPENED, CLOSED, AND DISCUSSEDOpened/Closed05000263/2007006-01NCVInadequate 10 CFR 50.59 Evaluation for Diesel GeneratorExhaust Missile Protection (Section 1R21.3.b)
Attachment
2LIST OF DOCUMENTS REVIEWEDThe following is a list of licensee documents reviewed during the inspection, includingdocuments 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 statedin the inspection report.IR02Evaluation of Changes, Tests, or Experiments 71111.0210 CFR 50.59 EvaluationsSE-03-004; Diesel Exhaust Missile Protection Design Consideration; datedJuly 28, 2003SE-06-003; SBO Operator Actions Associated with the HPCI System; dated September 19, 200610 CFR 50.59 ScreeningsSCR-04-0283; SRV Air Actuator Model Change; dated November 23, 2005
SCR-04-0859; HPCI Turbine Steam Supply Low Pressure Isolation; datedSeptember 11, 2006SCR-05-0161; Set Point for RHR Minimum Flow Switches FS-10-2-121, A, B, C and D;dated August 23, 2006SCR-05-0242; Instrument Setpoint Calculation 4.16KV Degraded Voltage; datedMarch 28, 2006SCR-05-0266; ITS Setpoint Change - HPCI Steam Line Area Temperature - High;dated August 26, 2006SCR-05-0689; Calc CA-05-146, Evaluation of Wall Thinning on FW2B-10-ED; datedOctober 11, 2005SCR-05-0738; Calc CA-05-028, Evaluation of HPCI Condensate Drain Line D13-2"-HEin the HPCI Room; dated November 9, 2005SCR-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; datedNovember 15, 2005SCR-05-0822; CA-05-155, Evaluation of Offgas Stack for SSE Seismic Loads; datedDecember 22, 2005
Attachment
3SCR-06-0062; Less than Full Thread Engagement on RWCU AO Valve Actuator BonnetNuts; dated February 15, 2006SCR-06-0103; HPCI Steam Void Elimination; dated April 6, 2006
SCR-06-0106; Service Water Pump Replacement; October 30, 2006SCR-06-0165; Replace AO-1575(6) and Check Valves with Normally Closed Valve
SW-228(9); dated October 31, 2006SCR-06-0166; Replace Rotork Actuators on Five MOVs with Limitorque Actuators;dated April 26, 2006SCR-06-0310; Technical Requirements Manual - Appendix B - Secondary ContainmentIsolation Valves; dated September 12, 2006SCR-06-0557; Suppression Chamber Inspection; dated December 4, 2006
SCR-07-0043; Fuel Storage and Handling Systems, Design Basis; datedJanuary 22, 200710 CFR 50.59 Applicability DeterminationsSCR-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 HigherTemperature Rating; dated September 28, 2005SCR-05-0791; Evaluation of Fire Detector Locations in the Reactor Building; datedDecember 5, 2005SCR-05-0819; Setpoint Change Request for the Safety/Relief Valve Low-Low Set Logicto Incorporate the New Trip Settings; dated December 21, 2005SCR-05-0830; Setpoint Change Request for the 4KV Bus-15 and Bus 16 UndervoltageRelays to Incorporate the New Trip Setting; dated January 3, 2006SCR-06-0308; Update USAR for Improved Technical Specification Project; datedJuly, 29, 2006IR17Permanent Plant Modifications 71111.17BModificationsEC8819; HPCI Steam Line Area Temperature - High; dated October 27, 2006
EC7583; Degraded Voltage Relays for Safety-Related 4KV Busses ; datedAugust 7, 2006
Attachment
4Equivalency EvaluationsEC910; Replacement Blower Wheel; Revision 1EC933 (05A099); HPCI Auxiliary Lube Oil Pump; Revision 0
EC7828; Engine Driven Fuel Pump Suction Line; Revision 0Setpoint ChangesEC8818; HPCI Turbine Steam Line Pressure - Low; dated October 27, 2006EC8792; 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; datedDecember 1, 2005SCR 05-028; SRV Low Low Set Pressure Interlock; dated December 1, 2005Other Documents Reviewed During InspectionCorrective Action Program Documents Generated As a Result of InspectionAR01076896; List to NRC Screened out All 50.59 Screening using the 3283 Form;
AR01077202; SCR-05-0830 Description Contains Incorrect Value; datedFebruary 14, 2007AR01077855; Action to Correct Drawing Error was Cancelled; dated February 19, 2007
AR01078665; Error in Calculation CA-05-146, Evaluation of Wall Thinning inFW2B-10"-ED; dated February 22, 2007AR01079705; LAR Required for Use of TORMIS Code Methodology; datedFebruary 28, 2007AR01080049; 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" Feedwaterto Reactor Line; March 25, 2005AR00891838; Evidence of Water Leakage on 11 and 12 EDG Exhaust Pipe Insulation;dated September 28, 2005AR01000610; Replacement Part does not Match the Part Removed; datedOctober 10, 2005
Attachment
5AR01000746; Inconsistency Between Line Design Table and Plant; datedOctober 11, 2005AR01001520; Operation past One Cycle Not Assured for Fw Pipe; datedOctober 20, 2005AR01003632; RC-44-2 Replacement Noticed 3000 No. vs. 6000 No.; datedNovember 14, 2005AR01004032; RWC Pipe Support Discrp and Indad Thread Engage on Act Nuts; datedNovember 17, 2005AR01006064; CV-1728 Plug Replaced, No Section XI Repair/Replacement Plan; datedDecember 1, 2005AR01008347; Some SW Mods May Inadvertently Create New Problems; datedDecember 21, 2005AR01022687; 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, 2006AR01040014; Inadequate Closeout Activities for Design Change 99Q160; datedJuly 17, 2006AR01059716; 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; datedSeptember 27, 2005AR1040142; B.03.04-05 Issued Prior to Completion of Revision Process; datedJuly 18, 2006AR0780295; Revise USAR Section 10.2.4.3 to Reflect the Results of CA-95-028; datedNovember 26, 2006AR01045206; 50.59 Screening SCR-05-210 Missed USAR Impact; datedAugust 18, 2006CalculationsCA-03-038; Instrument Setpoint Calculation, 4.16 KV Loss of Voltage; Revision 1
CA-03-039; Instrument Setpoint Calculation - SRV Low-Low Set, Reactor CoolantSystem Pressure; Revision 0CA-04-110; Determination of HPCI Area High Temperature Setpoints; Revision 1
Attachment
6CA-05-108; Evaluation of Wall Thinning on FW2B-10-ED Piping; Revision 0CA-05-146; Evaluation of Wall Thinning on FW2B-10"-ED Piping; Revision 0
DrawingsEC-811-01; Monticello Nuclear Generating Plant Installation of HPCI Void Resolution;Revision 1NH-36250; Monticello Nuclear Generating Plant P&ID (Water Side) High PressureCoolant Injection System; Revision AF
Attachment
7LIST OF ACRONYMS USEDADAMSAgency-Wide Document Access and Management SystemARAction Request
CFRCode of Federal Regulations
DRPDivision of Reactor Projects
DRSDivision of Reactor Safety
EDGEmergency Diesel Generator
ECEngineering Change
EPRIElectric Power Research Institute IMCInspection Manual Chapter
IRInspection Report
NCVNon-Cited Violation
NEINuclear Energy Institute
NRCNuclear Regulatory Commission
NRROffice of Nuclear Reactor Regulation
PARSPublicly Available Records
PRAProbabilistic Risk Assessment
SCRScreening (50.59)
SCRSetpoint Change Request
SDPSignificance Determination Process
SESafety Evaluation (50.59)
TSTechnical Specifications
USARUpdated Safety Analysis Report