IR 05000331/2005013

IR 05000331-05-013 (Drs); 11/14/2005 - 11/18/2005; Duane Arnold Nuclear Power Station; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications

ML053570299
Person / Time
Site:	Duane Arnold
Issue date:	12/22/2005
From:	Hills D E Engineering Region 1 Branch 1
To:	VanMiddlesworth G D Nuclear Management Co
References
IR-05-013
Download: ML053570299 (21)

Text

December 22, 2005

SUBJECT: DUANE ARNOLD NUCLEAR POWER STATION, NRC EVALUATION OFCHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000331/2005013 (DRS)

Dear Mr. Van Middlesworth:

On November 18, 2005, the U.S. Nuclear Regulatory Commission (NRC) completed acombined baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Duane Arnold Nuclear Power Station. The enclosedreport documents the results of the inspection, which were discussed with Mr. J. Bjorseth and others of your staff at the completion of the inspection on November 18, 2005 and by telephone on December 21, 2005.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, two NRC-identified findings of very lowsafety significance were identified, both of which involved violations of NRC requirements. However, because these violations were of very low safety significance and because they were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations in accordance with Section VI.A.1 of the NRC's Enforcement Policy.In accordance with 10 CFR 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 Public G. Van Middlesworth-2-Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is 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 Nos. 50-331License Nos.

Enclosure:

Inspection Report 05000263/2005013(DRS)

G. Van Middlesworth-2-Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,David E. Hills, ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos. 50-331License Nos.

Enclosure:

Inspection Report 05000263/2005013(DRS)

G. Van Middlesworth-3-ADAMS Distribution

HKN DWS RidsNrrDirsIrib

GEG KGO GAW1 CAA1 C. Pederson, DRS DRPIII DRSIII PLB1 JRK1 ROPreports@nrc.gov (inspection reports, final SDP letters, any letter with an IR number)

EnclosureU.S. NUCLEAR REGULATORY COMMISSIONREGION IIIDocket No:50-331License No:Report No:05000331/2005013 (DRS)Licensee:Facility:Duane Arnold Nuclear Power StationLocation:Dates:November 14 through 18, 2005 Inspectors:R. Daley, Senior Reactor Inspector, Team LeaderA. Dahbur, Reactor Inspector M. Garza, Reactor Inspector M. Munir, Reactor InspectorApproved by:D. Hills, ChiefEngineering Branch 1 Division of Reactor Safety (DRS)

Enclosure 1

SUMMARY OF FINDINGS

IR 05000331/2005013 (DRS); 11/14/2005 - 11/18/2005; Duane Arnold Nuclear Power Station;Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant

Modifications.The inspection covered a one-week announced baseline inspection on evaluations of changes,tests or experiments and permanent plant modifications. The inspection was conducted by four regional based engineering inspectors. Two Green Non-Cited Violations (NCV) were 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 commercialnuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3,dated July 2000.A.Inspector-Identified and Self-Revealed FindingsGreen. The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR50.59 in that the licensee failed to perform an adequate safety evaluation review for changes made to the facility as described in the Updated Final Safety Analysis Report(UFSAR). Specifically, the licensee adversely changed the description in the UFSAR of the license basis function of the recirculation pump runback in that the recirculationrunback feature could no longer prevent a reactor scram if a feedwater pump tripped.

Within the 10 CFR 50.59 evaluation, the licensee failed to provide a basis for why this malfunction of the recirculation pumps' runback logic (equipment important to safety) did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a Structures, Systems, and Components (SSC) important to safety. Because the issue affected the NRC's ability to perform its regulatory function, thisfinding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because the inspectors could not reasonably determine that the UFSAR change, which adversely affected equipment important tosafety, would not have ultimately required NRC approval. The finding was determinedto be of very low safety significance (Green) because the recirculation runback feature was not a mitigating function. (Section 1R02.1.b.1).

Green.

A finding of very low safety significance was identified by the inspectorsassociated with a violation of 10 CFR Part 50, Appendix B, Criterion III, "DesignControl," where the licensee had not evaluated and updated the plant cable ampacitycalculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the Condenser and Heater Bays. After identification by the team, the licensee was able to demonstrate that even though the higher temperaturesdecreased the ampacity margins for the effected cabling, it did not decrease the margins to the limit where the cabling would fail if called upon to provide power to equipment important to safety.

2The finding was more than minor because it affected the mitigating system cornerstoneattribute of "Design Control." Specifically, the licensee did not account for high temperature conditions in the Condenser and Heater Bay room that adversely affected the ampacity of cabling supplying power to equipment important to safety. This finding was of very low safety significance because it screened out using the Phase 1 worksheet. Specifically, the licensee's preliminary evaluation determined that the highertemperatures would not prevent pertinent equipment from functioning.

(Section 1R17.1.b.1)

B.Licensee-Identified Violations

No findings of significance were identified.

3

REPORT DETAILS

1.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 Scope

From November 14 through 18, 2005, the inspectors reviewed three evaluationsperformed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluationswere thorough and that prior NRC approval was obtained as appropriate. The teamcould not review the minimum sample size of five evaluations, because the licensees only performed three evaluations during the biennial sample period. The inspectors also reviewed 12 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meetthe threshold to require a 10 CFR 50.59 evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. The list of documents reviewed by the inspectors is 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," Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC inRegulatory 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

b.1Updated Final Safety Analysis Report (UFSAR) Change Reducing Capability of theAutomatic Runback of the Recirculation Pumps on a Feedwater Pump TripIntroduction: The inspectors identified that the licensee did not perform an adequatesafety evaluation in accordance with 10 CFR 50.59, when the licensee made changes to the UFSAR. Specifically, the licensee failed to provide adequate bases when they determined that changes to UFSAR Section 1.3.2.8.1 "Runback of Recirculation Pumpon Feedwater Pump Trip" did not require a licensee amendment. The licensee failed to address the malfunction of the recirculation pumps' runback logic which was designed to prevent a reactor scram in the event of one feedwater pump trip. The issue was considered to be of very low safety significance, (Green) and was dispositioned as a Severity Level IV Non-Cited Violation (NCV).

4Description: During review of Duane Arnold 10 CFR 50.59 Screening Number 3409, theteam questioned changes, referenced in the screening, to the UFSAR by safety evaluation SE-98-011. The inspectors were concerned that the licensee did not provideadequate bases when they determined that changes to UFSAR Section 1.3.2.8.1"Runback of Recirculation Pump on Feedwater Pump Trip" did not require a licensee amendment. Specifically, prior to the implementation of SE-98-001, UFSAR Section 1.3.2.8.1previously stated, "automatic runback of the recirculation pumps on a feedwater pumptrip results in a reactor power reduction that is within the capabilities of the feedwatersystem with only one pump. The correction for the loss of one feedwater pump isdesigned to be fast enough to prevent a reactor scram. See section 7.9.4.3 of the initial FSAR." Based on plant response and experience, the licensee found out that this was not necessarily true. The automatic runback of the recirculation pumps on a feedwaterpump trip did not allow the feedwater system to respond fast enough to prevent areactor scram. Therefore, the licensee revised Section 1.3.2.8.1 to state, "automaticrunback of the recirculation pumps on a feedwater pump trip results in a reactor power reduction which may not be within the capabilities of the feedwater system with only onepump. The correction for the loss of one feedwater pump may not be fast enough to prevent the reactor scram." The licensee also deleted the reference to section 7.9.4.3of the UFSAR which stated that a scram will not occur with a single feedwater pump trip.In the safety evaluation for this change, the licensee answered "no" to the followingquestion, "May the proposed activity increase the probability of occurrence of amalfunction of equipment important to safety." The licensee justification for this answer stated, "The activity changed the wording in the UFSAR to describe more accurately how the plant responds on a feedwater pump trip. The Loss of Feedwater Flow transient is already described in the UFSAR Section 15.6.3 and has already been evaluated. This section analyzes the total loss of feedwater and is concluded that this transient is a non-limiting event and bounds one feedwater pump trip. No physical changes occurred in the plant as a result of the change." The inspectors questioned the correctness and the adequacy of the bases for the licensee's justification, because the malfunction of the recirculation pump runback logic, the equipment important to safety, was not addressed by the safety evaluation. The inspectors noted that this change tothe UFSAR, together with the inability of the recirculation pump runback to prevent ascram, may have resulted in a change that resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of Structures, Systems, and Components (SSC) important to safety. Following identification of this issue, the licensee entered the issue into their correctiveaction program as Action Request (AR) CAP038955.

Analysis:

The inspectors determined that this issue was a performance deficiencysince, in 1998, the licensee failed to provide adequate basis for changes made to the UFSAR in accordance with 10 CFR 50.59. Specifically, the licensee failed to provide a basis for why this malfunction of the recirculation pumps' runback logic (equipment important to safety) did not present more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the 5regulatory process, they are dispositioned using the traditional enforcement processinstead of the significance determination process (SDP). The finding was determined to be more than minor because the inspectors could not reasonably determine that thechanges to UFSAR Section 1.3.2.8.1 would not have ultimately required NRC priorapproval. The inspectors completed a significance determination of the underlying technical issueusing NRC's inspection manual chapter (IMC) 0609, Appendix A, "SignificanceDetermination of Reactor Inspection Findings for At-Power Situations." The inspectors answered "no" to the Transient Initiator screening question in the Phase 1 Screening Worksheet which states, "Does the finding contribute to both the likelihood of a reactortrip AND the likelihood that mitigation equipment or functions will not be available,"because the recirculation runback feature is not a mitigating function. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with the Enforcement Policy, the violation was therefore classified as a Severity Level IV violation.Enforcement: Title 10 CFR 50.59(d)(1) states, in part, that the licensee shall maintainrecords of changes in the facility, of changes in procedures, and of tests andexperiments. These records must include a written evaluation which provides a basis for the determination that the change, test, or experiment does not require a licenseamendment.Contrary to the above, in their safety evaluation, SE 98-011, the licensee failed toprovide an adequate basis for the determination that the revision to UFSAR Section1.3.2.8.1 was acceptable without a license amendment. Specifically, the licenseeadversely changed the description in the UFSAR of the license basis function of the recirculation pump runback in that the recirculation runback feature could no longerprevent a reactor scram if a feedwater pump tripped. Within the 10 CFR 50.59 evaluation, the licensee failed to provide a basis for why this malfunction of the recirculation pumps' runback logic (equipment important to safety) did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a Structure, System and Component (SSC) important to safety. In accordance with the Enforcement Policy, this violation of the requirements of 10 CFR 50.59 was classified as a Severity Level IV Violation because the underlying technical issue was of very low safety significance. Because this non-willful violation was non-repetitive, and was captured inthe licensee's corrective action program (CAP038955), it is considered a Non-Cited Violation consistent with VI.A.1 of the NRC Enforcement Policy (NCV ).

(NCV 05000331/2005013-01 (DRS))1R17Permanent Plant Modifications (71111.17B).1Review of Permanent Plant Modifications

a. Inspection Scope

From November 14 through 18, 2005, the inspectors reviewed eight permanent plantmodifications that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and 6complexity. As per inspection procedure 71111.17B, one modification was chosen thataffected 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 adverselyaffect any systems' safety function. Design and post-modification testing aspects wereverified to ensure the functionality of the modification, its associated system, and anysupport systems. The inspectors also verified that the modifications performed did notplace 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. Findings

b.1Failure to Consider Adverse Ampacity Effects of High Temperature Conditions in theCondenser and Heater Bay RoomIntroduction: The inspectors identified a Non-Cited Violation (NCV) having very lowsafety significance (Green) of 10 CFR 50, Appendix B Criterion III, "Design Control." Specifically, the inspectors identified that the licensee had not evaluated and updatedthe plant cable ampacity calculation to determine the potential consequences of adverseeffects to cabling due to higher temperatures in the Condenser and Heater Bays.

Description:

Engineered Maintenance Action (EMA) A69614 raised the hightemperature alarm for the Condenser and Heater Bay room from 127 degrees F to 140 degrees F. This change was performed because higher temperatures were being experienced in this area after the plant power uprate and because there was a certain amount of damaged or missing piping insulation in the area. The setpoint change was made to prevent the alarm from coming in, since the temperatures were frequently hitting or exceeding the setpoint.The modification increased the alarm setpoint, but it did not address the effects of theseheightened temperatures on the ampacity of electrical cables in the area. Since higher temperatures adversely affect the ampacity of electrical cables, the higher temperatures in the Condenser and Heaters Bay room had the potential to adversely affect the functionality and/or operability of equipment important to safety fed by cabling in theseareas. The inspectors were concerned that the possibility existed that some of theequipment that were fed by cables in the area may not function due to possible faulting of the supply cables. The licensee determined that Duane Arnold Ampacity Calculation 434-E001 assumedtemperatures of 104 degrees and 122 degrees F. This was clearly non-conservative for the Condenser and Heater Bay room. As a result of the inspectors' concern, the licensee issued corrective action document CAP038933. After performing a preliminary evaluation that assessed cabling in the area andequipment fed from that cabling, the licensee determined that there was no evidencethat safety related Structures, Systems, and Components (SSCs) would not function as required. While the higher temperatures decreased the ampacity margins for the 7effected cabling, the licensee preliminarily determined that it did not decrease themargins to the limit where the cabling would fail if called upon to provide power to equipment important to safety.Analysis: The inspectors determined that this issue was a performance deficiency sincethe licensee failed to meet the requirements of 10 CFR Part 50 Appendix B, Criterion III,"Design Control." Specifically, the licensee did not account for high temperature conditions in the Condenser and Heater Bay room that adversely affected the ampacity of cabling supplying power to equipment important to safety. The issue was more than minor because it affected the mitigating system cornerstone attribute of "DesignControl." The finding screened as having very low significance (Green) using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for the At-Power Situations," because the inspectors answered "no" to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. In particular, the licensee's preliminary evaluation determined that the higher temperatures would notprevent pertinent equipment from functioning.Enforcement: 10 CFR Part 50, Appendix B, Criterion III, "Design Control" states, in part,that measures shall be established to assure that applicable design basis are correctly translated into specifications, drawings, procedures and, instructions. Contrary to the above, the licensee did not have a design basis calculation for cable ampacity that supported the actual high temperatures that were being experienced in the Condenserand Heater Bay room. The Duane Arnold calculation that did address ampacity was significantly less conservative, since temperatures of 104 degrees and 122 degrees F were assumed while actual temperatures in the area were exceeding 127 degrees and were being allowed to go as high as 140 degrees F before alarms actuated. Because the failure to address the adverse ampacity effects of heightened temperaturesin this room was determined to be of very low safety significance and because it was entered in the licensee's corrective action program as CAP038933, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.

(NCV 05000331/2005013-02 (DRS))4.OTHER ACTIVITIES (OA)4OA2Identification and Resolution of Problems.1Routine Review of Condition Reports

a. Inspection Scope

From November 14 through 18, 2005, the inspectors ActionProcess documents (CAPs) that identified or were related to 50.59 evaluations andpermanent 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 8action system. The specific corrective action documents that were sampled andreviewed by the team are listed in the attachment to this report.

b. Findings

No findings of significance were identified.4OA6Meetings.1Exit MeetingThe inspectors presented the inspection results to Mr. J. Bjorseth and others of the licensee's staff on November 18, 2005 and by telephone on December 21, 2005.

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

KEY POINTS OF CONTACT

Licensee

S. Catron, Nuclear Safety Assurance Manager

S. Haller, Site Engineering Director

R. Murrell, Regulatory Assurance Specialist

J. Swales, Engineer

L. Swenzinski, Regulatory Assurance SpecialistNuclear Regulatory Commission

B. Burgess, Reactor Projects Branch 2

D. Hills, Chief, Engineering Branch 1

D. Spaulding, NRR Project M

anager

G. Wilson, Senior Resident Inspector

2

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None.Opened and

Closed

05000331/FIN-2005013-01NCVUFSAR Change Reducing Capability of the AutomaticRunback of the Recirculation Pumps on a Feedwater

Pump Trip

05000331/FIN-2005013-02NCVFailure to Consider Adverse Ampacity Effects of of HighTemperature Conditions in the Condenser and Heater

Bay RoomDiscussedNone.

3

LIST OF DOCUMENTS REVIEWED

The 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 NRCinspectors reviewed the documents in their entirety, but rather that selected sections or portionsof 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 (71111.02)10

CFR 50.59 Screenings3281, Revision 1; UFSAR Change Request 03-024 is Removing the Specific ComputerProgram name "STRESS" From UFSAR Table 3.8-7; dated December 1,

20033313; Establish Required Airflow After Charcoal Addition to Ensure That the SBGTSystem Meets Design Function; dated December 2,

20033551; Clarify Cable Separation Requirements for New Installations; datedFebruary 3,

20043415; Add steps to install/remove jumpers on the 45% Recirc. Runback signal; datedMay 1,

20053678; Perform Condition Evaluation to Disposition Nonconformance of "ProperClearance" Measurement of 4160 Circuit Breakers; dated March 5,

20043726; Verify Breaker Springs Charged for 1BR91/92 Crosstie; dated March 16, 2004

4101;

TRMCR-016 (TSCR-067, Amendment #254); dated June 29, 2004

4371; UFSAR Change #04-003, CAP #30681 and 31021; dated September 21,

20054453; Revision to EMA A58898 Defines the Maximum Full Load Current as Calculatedby

CAL-E02-003, Revision 1 and UFSAR, Table 8.3-2; dated October 11,

20044895;

TRMCR-015 (TSCR-064, Amendment #255); dated April 11, 2005

5132; Changing the pressure setpoint of PRV1634 based on

CAL-M01-068; dated April 25,

20055386; Add Clarifying Statement to the TRM Bases for TLCO 3.3.6; dated August 22,

20055418; Replace the Existing General Service Water Pump With a New Pump With HigherCapacity; dated August 12, 2005

410

CFR 50.59 Evaluations98-011; Revise UFSAR Section 1.3.2.8.1 "Runback of Recirculation Pump onFeedwater Pump Trip"; dated March 26, 199800-018; Safety Evaluation for EMA A50452 & A50453, Replacement of Crosby reliefvalves in HPCI and RCIC Systems; Revision 102-002; Allowance for Bypassing the RHRSW Strainer; Revision 105-001; Existing 1000 Gallon Gas Tank Relocation; dated August 30, 2005IR17Permanent Plant Modifications (71111.17B)ModificationsECP-1662; Core Spray Seal Replacement; Revision 0

ECP 1679; 1V-SF-56A Motor and Damper Control Wiring Modification; Revision 0

ECP-1696; Modify HPCI, RCIC, and RHR Piping in the Torus Airspace to EliminateFlanged Joints; dated June 15, 2004EMA A64464; Replacement of Control Building Heating Circulating Pump Motor1VHP030B-M; Revision 0EMA-A65920; Two Vent Lines Upstream of CV1621 and CV1579; dated March 9, 2004

EMA A67459; Revise Setpoint for PS2304B HPCI Booster Pump Suction Low PressureTrip in accordance with

CA-M04-11; dated February 18, 2005EMA A69614; Raising of Setpoint for the Inlet Temperature for 1VAC021 and 1VAC022from 127 degrees F to 140 degrees F; Revision 0EMA

111689; Change Setting for Ground Fault Time Delay; Revision 0Other Documents Reviewed During InspectionCorrective Action Program Documents Generated As a Result of InspectionCAP038905;

NRC 50.59 2005 Mod Inspection -

ECP 1679; dated November 15, 2005CAP038931; EMA A58898 - 2005

NRC 50.59/Mod Inspection; dated November 16, 2005CAP038932; 50.59 Screening #5418 Weakness; dated November 16, 2005

CAP038933; Condenser Bay and Heater Bay Cable Ampacity Issue from 2005 NRC50.59/Mod Inspection; dated November 16, 2005

5CAP038934; Editorial Corrections to Station Blackout Analysis Report for Power Uprate;November 16, 2005CAP038948; RCIC Vacuum Pump OOS without Compensatory Measures for LicensingCommitments; dated November 17, 2005CAP038955; Inadequate 50.59 Evaluation 98-11 (2005

NRC 50.59/Mod Inspection);dated November 17, 2005CAP038960;

CAL-M97-008 (HPCI NPSH) Contains an Apparent Discrepancy; datedNovember 17, 2005CAP038963; Ensure that the Basis for Tech Spec 3.6.2.1 Was Evaluated for PowerUprate; dated November 17, 2005Corrective Action Program Documents Reviewed During the Inspection

CAP 029587; LT4541 (RX VESSEL WIDE RANGE (FLOOD) Cable Routing ViolatesDivisional Separation; dated October 30, 2003CAP029976; STP 3.6.4.3-03 on SGTS B Stopped Due to Procedure Problems; datedDecember 1, 2003CAP030242; HPCI & RCIC Steam Line Exhaust Test Connection; dated December 31, 2003CAP030243;

RHR Flanges in Torus Air Space to be Removed; datedDecember 31, 2003CAP

030392; Potential 50.59 Screening Applicability Inconsistency; datedJanuary 14, 2005CAP031021; SSDI Unresolved Item - Station Blackout Analysis for EPU; dated March 17, 2004CAP033128; Request TRM Bases Clarification of Table 3.3.6-01 Function 1 & 2Instruments; dated September 24, 2004CAP036685; Received Condenser Area Cooler High Inlet Temperature Alarm; datedJune 4, 2005CAP

036954; NRC Violation - Inspection Report 2005-010; dated June 29, 2005CAP037401; Temporary Power Control Process Needs a Tie to 50.59 ScreeningProcess; dated August 5, 2005CAP038100; Error Found in Approved 50.59 Safety Evaluation #00-018; datedSeptember 28, 2005

6CAP

038108; 50.59/Mod Snapshot Evaluation Finding; dated September 29, 2005CA

040624; NRC Violation - TAP - Analysis Review; dated July 26, 2005CA

040625; NRC Violation - Inspection Support; dated July 26, 2005CA

040626; NRC Violation - Mark I Analysis; dated July 26, 2005CA

041038; 50.59 / Mod Snapshot Evaluation Finding; dated September 30, 2005CE

001313 for

CAP 029587; dated November 13, 2003

CE 001512; 4160 V "Prop Clearance" Measurement OOT (Spare Breaker)

CE002703; Received Condenser Area Cooler High Inlet Temperature Alarm; dated June 7, 2005COM

040508; NRC Violation - Inspection Report 2005-010; dated July 1, 2005PCR

041280; NRC Violation - Inspection Report; dated October 31, 2005Calculations434-E001; Cable Ampacity vs. Cable Tray Fill

CAL-E02-003; Single Standby Diesel Generator Static Loading for a Loss of CoolantAccident Plus a Loss Offsite Power; Revision 1CAL-E04-010; Analysis of the Diesel Loading During a LOOP LOCA for a Chiller Loadof 96KV; Revision 1CAL-M04-011; HPCI Booster Pump Low Suction Pressure Setpoint - PS2304B;Revision 1CAL-M97-008; HPCI NPSH Calculation; Revision 1DrawingsBECH-E113, Sheet 79; Heating & Ventilation Systems; Revision 8

M063-002; Control Circuit Diagram for Control Room Chiller; Revision 5Vendor Dwg. No. M073-052; RHR Pump Room Air Supply Fans

IV-SF-56A & B;Revision 12

7ProceduresACP 103.2; 10

CFR 50.59 Screening Process; Revision 25ACP 106.1; Procedure Preparation, Revision, Review, and Approval; Revision 33

ACP 106.1; Procedure Preparation, Revision, Review, and Approval; Revision 42

AOP 301.1; Station Blackout; Revision 25

AOP411; Loss of General Service Water; Revision 19

CKTBKR-G080-07; GE AM 4.16-350-2H Medium Voltage Breaker Overhaul; Revision 10DGC-E100; Design Guide for Independence of Electrical Equipment and Circuits

GMP-ELEC-04; Cable Installation; Revision 6

SPEC-E512; Cable and Wire Installation Specification; Revision 11

STP 3.6.4.3-03; Standby Gas Treatment System HEPA and Charcoal Filter EfficiencyTests; Revision 10Miscellaneous DocumentsCWO-A65920; Feedwater Discharge to Regulating Valve

CV-1621; dated May 11, 2005NEDC-32980P; Safety Analysis Report for Duane Arnold Energy Center ExtendedPower Uprate; dated November 2000NG-04-0271; Letter from Duane Arnold Energy Center to the NRC, "Additi onalInformation Regarding NRC Unresolved Item from Safety System Design andPerformance Capability Inspection"; dated J

une 4, 2004PWR No. 25713; AR OTH036960; Reconcile the ACPs for 50.59 ScreeningApplicability; dated January 19, 2004Work Order No.

1116829; Refurbish and Lubricate Breaker. Inspect/Calibrate prior toReturning to Service; dated March 12, 2002Work Order No.

1126796; Rewire TS7538C per ECP1679 and Instructions; datedJanuary 26, 2004

8

LIST OF ACRONYMS

USEDADAMSAgency-Wide Document Access and Management SystemCFRCode of Federal Regulations

DRPDivision of Reactor Projects

DRSDivision of Reactor Safety

EMAEngineered Maintenance Action

IMCInspection Manual Chapter

IRInspection Report

NCVNon-Cited Violation

NEINuclear Energy Institute

NRCNuclear Regulatory Commission

PRAProbabilistic Risk AssessmentSBLCStandby Liquid Control

SDPSignificance Determination Process

SSCStructures Systems and Components

UFSARUpdated Final Safety Analysis Report

URIU nresolved Item