IR 05000237/2005003

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IR 05000237-05-003; IR 05000249-05-003, IR 07200037-05-003, on 01/01/2005 - 03/31/2005 for Dresden, Units 1, 2 and 3; Post Maintenance Testing, Surveillance Testing, Routine Integrated Report
ML051240413
Person / Time
Site: Dresden  
Issue date: 04/29/2005
From: Ring M
NRC/RGN-III/DRP/RPB1
To: Crane C
Exelon Generation Co, Exelon Nuclear
References
FOIA/PA-2010-0209 IR-05-003
Download: ML051240413 (34)
v • d • e


Text

April 29, 2005

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 NRC INTEGRATED INSPECTION REPORT 05000237/2005003; 05000249/2005003; 0720037/2005003

Dear Mr. Crane:

On March 31, 2005, the NRC completed an inspection at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed report presents the inspection findings which were discussed with Mr. D. Bost and other members of your staff on April 15, 2005.

The inspection examined activities conducted under your license as they relate to safety and to 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 this inspection, one NRC identified finding and one self-revealed finding of very low safety significance were identified. Each of these findings involved a violation of NRC requirements. However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations, in accordance with Section VI.A.1 of the NRCs Enforcement Policy.

If you contest any Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555-0001; and the NRC Resident Inspector at the Dresden Nuclear Power Station. In accordance with 10 CFR 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 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/

Mark A. Ring, Chief Branch 1 Division of Reactor Projects Docket Nos. 50-237; 50-249;72-037 License Nos. DPR-19; DPR-25;DPR-02

Enclosure:

Inspection Report 05000237/2005003; 05000249/2005003; 0720037/2005003 w/Attachment: Supplemental Information

REGION III==

Docket Nos:

50-237; 50-249;72-037 License Nos:

DPR-19; DPR-25; DPR-02 Report No:

05000237/2005003; 05000249/2005003 07200037/2005003 Licensee:

Exelon Generation Company Facility:

Dresden Nuclear Power Station, Units 1, 2 and 3 Location:

6500 North Dresden Road Morris, IL 60450 Dates:

January 1 through March 31, 2005 Inspectors:

C. Phillips, Senior Resident Inspector D. Smith, Senior Resident Inspector M. Sheikh, Resident Inspector D. Tharp, Resident Inspector, Clinton W. Slawinski, Senior Radiation Specialist R. Landsman, Project Inspector, Decommissioning Branch M. Gryglak, Decommissioning Inspector, Decommissioning Branch R. Schulz, Illinois Emergency Management Agency Approved by:

Mark Ring, Chief Branch 1 Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000237/2005003; IR 05000249/2005003, IR 07200037/2005003 01/01/2005 - 03/31/2005,

Exelon Generation Company, Dresden Nuclear Power Station, Units 1, 2 and 3; Post Maintenance Testing, Surveillance Testing, routine integrated report.

This report covers a 3-month period of baseline resident inspection; announced baseline inspections on radiation material processing and transportation; and announced inspection of independent spent fuel storage installation activities. The inspection was conducted by Region III inspectors and the resident inspectors. Two findings or violations 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 severity level after NRC management review. The NRCs 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 Findings

Cornerstone: Mitigation Systems

Green.

On March 3, 2005, a performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified by the inspectors. The licensee had implemented inadequate corrective actions for a significant condition adverse to quality that occurred on January 19, 2001; and no corrective actions were assigned to prevent recurrence of a significant condition adverse to quality that occurred on November 29, 2004. Both events involved the failure of the Unit 2 emergency diesel generator air start regulating valve due to corrosion build up on the valve stem. The primary cause of this finding was related to the cross-cutting issue of Problem Identification and Resolution.

The finding was greater than minor because it impacted the Mitigating System Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events and because it affected the reliability of a safety related component. After the inspectors questioned the lack of corrective actions for the November 29, 2004 event, the licensee created an action item to review the cause of the event and create corrective actions. In addition, the licensee wrote IR [Issue Report] 308526, IR 277466 Significance Not Properly Identified. The purpose of this IR was to identify why this event was not entered into the corrective action system. This review had not been completed by the end of the inspection period. The finding was of very low safety significance because the emergency diesel generator started upon demand. (Section 1R19)

Green.

On January 5, 2005, a performance deficiency involving a Non-Cited Violation of Technical Specification 5.4.1 was self revealed when instrument maintenance technicians were performing Dresden Instrument Surveillance 700-02, APRM/RBM

[average power range monitor/rod block monitor] Flow Instrumentation Total Drive Flow Adjustment, Revision 16. The technicians misadjusted the recirculation flow signal to the reactor protection system which required entry into Technical Specification 3.3.1.1

Limiting Condition for Operation A.1 and C.1 for Average Power Range Monitor Channels 1, 2, and 3 Flow Bias Trips. The instrument maintenance technicians were using the averaging function of a Fluke 189 digital multi-meter. The technicians had not been trained on how to use the function and the procedure did not provide instructions on how to use the multi-meter. The mis-use of the averaging function resulted in the adjusting the recirculation flow converter signal too high.

The finding was greater than minor because it impacted the Mitigating System Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events and because it affected the procedure quality of a surveillance test procedure. The finding was of very low safety significance because it impacted the reactor protection system for a time period of less than 1 minute. The surveillance test procedure was changed to include instructions on how to use the averaging function of a digital multi-meter and the instrument maintenance technicians were briefed on this event and trained on how to use the averaging function of the digital multi-meter. (Section 1R22)

B.

Licensee Identified Findings No findings of significance were identified.

REPORT DETAILS

Summary of Plant Status

Unit 2 began the inspection period at 912 MWe (95 percent thermal power and 100 percent of rated electrical capacity).

  • Beginning February 2, 2005, load was reduced at various times to investigate higher than expected reactor pressure vessel steam moisture carryover. The unit returned to full power on February 19, 2005.
  • On March 12, 2005, load was reduced to 72 percent power for turbine valve testing, control scram time testing, and control rod pattern adjustment. The unit returned to full power on March 13, 2005.
  • On March 24, 2005, the unit scrammed due to the malfunction of the A electro-hydraulic control pressure regulator, and returned to full power March 27, 2005.
  • On March 29, 2005, load was reduced to 79 percent power for control rod pattern adjustment, and returned to full power on March 30, 2005.

Unit 3 began the inspection period at 822 MWe (100 percent thermal power).

  • On February 27, 2005, load was reduced to 55 percent to perform turbine valve testing, control rod scram time testing, and a control rod pattern adjustment; and returned to full power on February 28,

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R04 Equipment Alignment

Main article: IP 71111.04Q

.1 Partial System Walkdowns

a. Inspection Scope

The inspectors selected a redundant or backup system to an out-of-service or degraded train, reviewed documents to determine correct system lineup, and verified critical portions of the system configuration. Instrumentation valve configurations and appropriate meter indications were also observed. The inspectors observed various support system parameters to determine the operational status. Control room switch positions for the systems were observed.

Other conditions, such as adequacy of housekeeping, the absence of ignition sources, and proper labeling were also evaluated.

The inspectors performed partial equipment alignment walkdowns of the:

This represented three inspection samples.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

Main article: IP 71111.05

a. Inspection Scope

The inspectors toured plant areas important to safety to assess the material condition, operating lineup, and operational effectiveness of the fire protection system and features. The review included control of transient combustibles and ignition sources, fire suppression systems, manual fire fighting equipment and capability, passive fire protection features, including fire doors, and compensatory measures. The following areas were walked down:

  • Unit 2 turbine building, elevation 517' computer room and auxiliary electrical room, (Fire Zone 6.2);
  • Unit 2 turbine building, elevation 495' containment cooling service water pumps, (Fire Zone 8.2.2.A);
  • Unit 2 reactor building, elevation 589' isolation condenser area, (Fire Zone 1.1.2.5.A);
  • Unit 3 turbine building, elevation 495' containment cooling service water pumps, (Fire Zone 8.2.2.B);
  • Unit 2/3 turbine building, elevation 534' Lube oil reservoir area, (Fire Zone 8.2.6.C);
  • Unit 3 reactor building, elevation 589' isolation condenser area, (Fire Zone 1.1.1.5A)
  • Unit 2/3 crib house, elevation 517', ground floor, (Fire Zone 11.3); and
  • Unit 2/3 crib house, elevation 509', service water pumps room, (Fire Zone 11.3).

This represented ten inspection samples.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

Main article: IP 71111.11Q

a. Inspection Scope

The inspectors observed an evaluation of an operating crew on February 7, 2005. The scenario consisted of a loss of instrument air (recoverable), feedwater level control system setpoint drift, feedwater system vibration, unisolable feed line break in drywell, and high pressure coolant injection system start failure. The inspectors verified that the operators were able to complete the tasks in accordance with applicable plant procedures and that the success criteria as established in the job performance measures were satisfied. The inspectors observed the licensees evaluators to ensure that no inappropriate cues were provided by the evaluators while assessing the operators' performance. In addition, the inspectors verified that issue reports written regarding licensed operator requalification training were entered into the licensees corrective action program with the appropriate significance characterization.

This represented one inspection sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

a. Inspection Scope

The inspectors reviewed the licensee's handling of performance issues and the associated implementation of the Maintenance Rule (10 CFR 50.65) to evaluate maintenance effectiveness for the selected systems. The following system were selected based on being designated as risk significant under the Maintenance Rule, being in the increased monitoring (Maintenance Rule category a(1)) group, or due to an inspector identified issue or problem that potentially impacted system work practices, reliability, or common cause failures:

  • Unit 3 isolation condenser system.

The inspectors verified the licensee's categorization of specific issues, including evaluation of the performance criteria, appropriate work practices, identification of common cause errors, extent of condition, and trending of key parameters.

Additionally, the inspectors reviewed the licensee's implementation of the maintenance rule requirements, including a review of scoping, goal-setting, performance monitoring, short-term and long-term corrective actions, functional failure determinations associated with the condition reports reviewed, and current equipment performance status.

This represented two inspection samples.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

a. Inspection Scope

The inspectors evaluated the effectiveness of the risk assessments performed before maintenance activities were conducted on structures, systems, and components and verified how the licensee managed the risk. The inspectors evaluated whether the licensee had taken the necessary steps to plan and control emergent work activities. The inspectors also verified that equipment necessary to complete planned contingency actions was staged and available.

The inspectors completed evaluations of maintenance activities on the:

This represented six inspection samples.

b. Findings

No findings of significance were identified.

1R14 Personnel Performance Related to Non-routine Evolutions and Events

Main article: IP 71111.14

Unit 2 Forced Outage On March 24, 2005, Unit 2 scrammed due to the malfunction of the A electro-hydraulic control system pressure regulator, and returned to full power March 27, 2005.

The inspectors performed the following activities daily:

  • attended control room operator and outage management turnover meetings to verify that the current shutdown risk status was well understood and communicated;
  • performed walkdowns of the main control room to observe the alignment of systems important to safe/shutdown risk condition;
  • observed portions of power ascension; and
  • ensured that Technical Specification requirements were verified to have been met for changing modes.

b.

Finding No findings of significance were identified.

1R15 Operability Evaluations

Main article: IP 71111.15

a. Inspection Scope

The inspectors reviewed operability evaluations to ensure that operability was properly justified and the component or system remained available, such that no unrecognized increase in risk occurred. The review included issues involving the operability of:

  • Unit 2 steam dryer high moisture carry over (ECs 346368 and 353315);
  • Unit 2 core spray line break detection (IR 294408);
  • Common mode electrical bus failure of 23, 24, 33, and 34 (IR 297540);
  • 2B containment cooling service water heat exchanger outlet valve controller (IR 301450);
  • Non-conforming condition discovered on containment cooling service water room coolers (IR 278234);
  • Secondary containment differential pressure slow to recover and causing high standby gas treatment flow (IR 298901); and

This represented eight inspection samples.

b. Findings

No findings of significance were identified.

1R16 Operator Workarounds

Main article: IP 71111.16

a. Inspection Scope

The inspectors assessed one operator workaround issue to determine the potential effects on the functionality of the corresponding mitigating system:

  • Dresden Operating Abnormal Procedure 5700-0, Loss of Heating Steam Boilers During this inspection, the inspectors reviewed numerous existing problems with the heating steam boilers and heating steam piping in regard to how the loss of heating steam would impact safety related equipment.

This represented one inspection sample.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modification

Main article: IP 71111.17A

a. Inspection Scope

The inspectors reviewed Modification Engineering Change (EC) 6602, Core Spray Lower Sectional Replacement, and the associated installation work orders. The inspectors interviewed several members of the licensees engineering staff and the General Electric on-site liaison and project manager for the modification installation.

This represented one inspection sample.

b. Findings

Introduction:

The inspectors identified an unresolved item regarding the installation and testing of a modification to core spray piping inside the vessel and outside the core shroud of Unit 3.

Description:

During the Fall 2004 refueling outage, the licensee performed a core spray lower sectional piping replacement modification (EC 6602). The sectional replacement was intended to replace piping inside the reactor vessel annulus that was susceptible to intergranular stress corrosion cracking (IGSCC) due to the environment and original welded materials (Type 304 stainless steel). The IGSCC could result in leakage from the piping into the annulus rendering core spray inoperable. The modification was designed by General Electric using IGSCC resistant materials with mechanical connections in lieu of eight previously welded connections to further mitigate susceptibility to IGSCC.

Title 10 of the Code of Federal Regulation Part 50, Appendix B, Criterion III, Design Control, states in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.

The Exelon Quality Assurance Manual, Topical Report, Revision 75, states in part in Chapter 11, Test Control, Section 2.1.1 that the test program covers all required tests including the demonstration of satisfactory performance following plant maintenance or modifications. Section 2.7 states in part, The Company performs test following plant modification or significant changes in operating procedures to confirm that the modification or changes produces the expected results.

The licensee did not perform any post modification testing to verify that leakage from core spray piping into the annulus was within the design limits. The inspectors considered that the failure to perform a performance test might be considered minor provided that the licensee could provide adequate assurance that the modification had been installed properly to preclude leakage into the annulus. The inspectors wanted to see documented evidence within the installation work instructions that the modification was properly installed.

As part of Modification EC 6602, the inspectors reviewed GENE-0000-0021-4342-04, Dresden Nuclear Power Station, Unit 3 Core Spray Line Lower Sectional Replacement Leakage Analysis, Revision 1. This analysis listed maximum gaps where mechanical components were joined together to prevent excess leakage outside the core shroud. These gaps, if exceeded, could render one or more trains of the core spray system inoperable. The work instructions used to perform the modification installation did not include verification that the maximum gaps between mechanical joints described in the leakage analysis were not exceeded. The inspectors asked the licensee to demonstrate how they ensured that the maximum gaps between mechanical joints described in the leakage analysis were not exceeded. The licensee was unable to show documentation that verified that the modification was installed with less than or equal to the maximum gaps between mechanical joints described in the leakage analysis.

The licensees ability to satisfactorily demonstrate the proper installation of the core spray piping to ensure that leakage from the piping into the annulus region did not render either or both trains of the core spray system inoperable was an Unresolved Item. (URI 05000249/2005003-01)

1R19 Post Maintenance Testing

Main article: IP 71111.19

a. Inspection Scope

The inspectors reviewed post-maintenance test results to confirm that the tests were adequate for the scope of the maintenance completed and that the test data met the acceptance criteria. The inspectors also reviewed the tests to determine if the systems were restored to operational readiness status consistent with the design and licensing basis documents. The inspectors reviewed post-maintenance testing activities associated with the following:

  • Unit 2 containment cooling service water system following flow controller replacement;

This represented seven inspection samples.

b. Findings

Introduction:

A Green finding involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified by the inspectors. The inspectors identified that there were inadequate corrective actions to prevent recurrence for a significant condition adverse to quality that occurred on January 19, 2001; and there were no corrective actions assigned to prevent recurrence of a significant condition adverse to quality that occurred on November 29, 2004. Both events involved the failure of the Unit 2 emergency diesel generator air start regulating valve due to corrosion build up on the valve stem.

Description:

On November 29, 2004, the Unit 2 Emergency Diesel Generator (EDG) was started for the scheduled monthly operability surveillance. The starting air regulating valve failed such that it was relieving pressure via the main valve seat, through a pressure relief tube. The pressure relief path reduced starting air system pressure to about 150 psig before operators isolated the system. Although the regulating valve blew down, the engine started normally and was subsequently secured due to the air regulating valve failure. The licensee wrote issue report (IR) 277466 which stated that the air start regulating valve was leaking. The licensee replaced the air start regulating valve on the Unit 2 EDG and also the regulating valves on the Units 3 and 2/3 EDGs. The regulating valve was sent out to undergo testing at the Exelon PowerLabs Division under project number DRE-34545. The PowerLabs report was sent to the licensee on December 3, 2004. The report stated that several problems were found internal to the regulating valve, but that the primary cause of the failure was due to the build up of aluminum oxide deposits on the valve stem from the corrosion of the internals of the regulator materials.

The regulating valve maintained a pressure of at least 175 psig downstream of the valve to allow starting of the EDG. When the EDG started and the downstream pressure began to drop, the air regulating valve opened attempting to maintain the downstream pressure. The valve had a reverse seating disc. Air pressure acts on a diaphragm and opens the valve against pressure from a spring underneath the valve disc. The valve stem was hollow. A relief mechanism attached to the diaphragm fit inside the valve stem. If downstream pressure became too high the diaphragm would lift the relief mechanism out of the valve stem providing a relief port out of the hollow valve steam until down stream pressure decreased. The failure occurred because corrosion products on the main valve stem prevented the closing of the main valve seat under normal spring pressure. When pressure equalized across the valve and the diaphragm returned to its normal position the relief device came out of the stuck valve stem causing the upstream air to blow down through the stuck open valve and out the relief device. The build up of the corrosion products was considered a time dependent mechanism. The PowerLabs report made several recommendations as to how to address the problem.

The regulating valve that failed was last replaced in February 2003, as a two year preventive maintenance item. The valve had also been replaced in January 2001, after failing in service under similar circumstances as the November 2004 regulating valve failure. The inspectors attempted to review the corrective action documents from the January 2001 failure. The inspectors identified that this failure had not been entered into the corrective action program in 2001. The system engineer took the initiative to send the valve to Engine Systems, Inc. for a failure analysis. The manufacturer had issued a 10 CFR Part 21 report for a different failure mechanism in September 2002. The valves that were replaced in February 2003 were purchased in November 2002 after the Part 21 was issued.

Therefore, this failure was not related to the 2002 Part 21.

In February 2005, the inspectors were observing scheduled maintenance on the Unit 2 EDG and observed that a hoist mounted on a rail in the overhead of the Unit 2 EDG room was making hard contact with the air start line piping. The inspectors were concerned about the potential for foreign material, caused by corrosion of the air start line, to enter and impact the operation of the air start regulating valve. The inspectors attempted to review the corrective actions associated with IR 277466 from the regulating valve failure in November 2004.

The inspectors identified that there were no assignments to perform a root or apparent cause of the failure nor were there any corrective actions associated with the regulating valve failure in the corrective action program. Sending the valve to PowerLabs was done on the initiative of the system engineer. However, there was no assigned action item in the corrective action program to evaluate what had been found during the PowerLabs review of the failure and formulate any necessary corrective actions.

Analysis:

The inspectors determined that the failure to have adequate corrective actions associated with repetitive failures of safety-related equipment was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on June 20, 2003, because it impacted the Mitigating System Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events and because it affected the reliability of a safety related component. The EDG started before the regulating valve failed. Upon the regulating valve failing, air pressure in the receiver blew down to about 150 psig, from a normal pressure range of 225-260 psig, before the operator could shut the air receiver isolation valve and stop the reduction in the air receiver pressure.

The design of the air start system is to have sufficient air available to attempt to start the EDG three times. An air pressure of about 175 psig was necessary to start the U2 EDG per the basis of Technical Specification 3.8.3, Action C.1.

Therefore, if the EDG had failed to start, the air receiver capacity would not have been sufficient to make another start attempt. The primary cause of this finding was related to the cross-cutting issue of Problem Identification and Resolution.

The inspectors completed a Phase 1 significance determination of this issue using IMC 0609, Significance Determination Process, Appendix A, Attachment 1, dated December 1, 2004. The inspectors concluded that the finding impacted the mitigating system cornerstone. The inspectors answered NO to all five questions under the Mitigating System Cornerstone column, and the issue screened as having very low safety significance (Green).

Enforcement:

10 CFR Part 50, Appendix B, Criterion XVI required, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management. The licensees Quality Assurance Topical Report (NO-AA-10), Revision 75, Appendix D, Paragraph 2.117, defined a significant condition adverse to quality, in part, as large deviations from expected plant performance of safety-related structures, systems, or components; and recurring deficiencies or errors that cannot be dispositioned or brought into conformance by established corrective action systems. Contrary to the above, the inspectors identified that on January 19, 2001, the Unit 2 EDG air start regulating valve failed due to corrosion of its valve stem which was a large deviation from expected plant performance. The licensee took action to assure the cause of the condition was determined but took inadequate action to prevent recurrence. The Unit 2 EDG air start regulating valve again failed on November 29, 2004, for the same cause as the January 19, 2001 event. The licensee took action to determine the cause of the failure, but as of March 3, 2005, took no corrective actions to prevent recurrence nor assigned any actions to evaluate the cause of the failure to assess corrective actions to prevent recurrence. After the inspectors questioned the lack of corrective actions, the licensee created an action item to review the cause of the event and create corrective actions. In addition, the licensee wrote IR 308526, IR 277466 Significance Not Properly Identified. The purpose of this IR was to identify why this event was not properly entered into the corrective action system. This review had not been completed by the end of the inspection period. Because this violation was of very low safety significance and it was entered into the licensees corrective action program (IR 277466 and 308526), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000237/2005003-02)

1R22 Surveillance Testing

Main article: IP 71111.22

a. Inspection Scope

The inspectors observed surveillance testing on risk-significant equipment and reviewed test results. The inspectors assessed whether the selected plant equipment could perform its intended safety function and satisfy the requirements contained in TS. Following the completion of each test, the inspectors determined that the test equipment was removed and the equipment returned to a condition in which it could perform its intended safety function.

The inspectors observed surveillance testing activities and/or reviewed completed packages for the tests, listed below, related to systems in the Initiating Event, Mitigating Systems, and Barrier Integrity Cornerstones:

  • Unit 2 DIS 700-02, Average Power Range Monitor/Rod Block Monitor Flow Instrumentation Drive Flow Adjustment;
  • Unit 2 DIS 1500-7, Reactor Vessel Pressure Switch Calibration;
  • Units 2/3 DOS 6600-08, Diesel Generator Cooling Water Pump Quarterly and In-Service Testing;
  • Unit 3 DIS 0250-02, Main Steam Line Low Pressure Isolation Switch Calibration.

This represented seven inspection samples.

b. Findings

Introduction:

A Green finding involving a Non-Cited Violation of Technical Specification 5.4.1 was self revealed when instrument maintenance technicians were performing surveillance procedure DIS 700-02, APRM/RBM [average power range monitor/rod block monitor] Flow Instrumentation Total Drive Flow Adjustment, Revision 16. The technicians misadjusted the recirculation flow signal to the reactor protection system which required entry into technical specification 3.3.1.1 Limiting Condition of Operation, Actions A.1 and C.1 for APRM Channels 1, 2, and 3 Flow Bias Trips.

Description:

Instrument maintenance technicians were adjusting a recirculation flow converter using a digital multi-meter (DMM) with an averaging function. As the technicians adjusted the flow converter signal from a low set point to a high set point they did not reset the averaging function of the DMM. The control room operators received the Unit 2 neutron monitor flow off normal alarm during the adjustment. The technicians were not trained on how to use the DMM and the associated averaging function and thought that the DMM would self-adjust when they increased the flow signal. The surveillance procedure did not have any instructions on how to use the averaging function of the DMM.

Analysis:

The inspectors determined that the failure to have adequate instructions for performing a surveillance test that impacted safety-related equipment was a performance deficiency warranting a significance evaluation.

The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on June 20, 2003, because it impacted the Mitigating System Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events and because it affected the procedure quality of a surveillance test procedure.

The inspectors completed a Phase 1 significance determination of this issue using IMC 0609, Significance Determination Process, Appendix A, Attachment 1, dated December 1, 2004. The inspectors concluded that the finding impacted the mitigating systems cornerstone because the instrument maintenance technicians adjusted the core flow signal for the APRM flow biased trip setpoint too high. This action impacted, but did not prevent, the ability of the reactor protection system to respond to a transient temporarily. The inspectors answered NO to all five questions under the Mitigating System Cornerstone column, and the issue screened as having very low safety significance (Green).

Enforcement:

Technical Specification 5.4.1 required, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978, Paragraph 8.b.2.l listed surveillance tests for the reactor protection system. Contrary to the above, on January 5, 2005, Surveillance test procedure DIS 0700-02, APRM/RBM Flow Instrumentation Total Drive Flow Adjustment, Revision 16, allowed instrument maintenance technicians to use a DMM, Fluke Model 189, to perform the test procedure without adequate instructions as to how to use the averaging function of the DMM. The instrument maintenance technicians were not trained prior to the performance of the surveillance test procedure on how to use the averaging function. As immediate corrective action, the flow converter setpoints were returned to within specification about 1 minute after the Rod Out Block and Neutron Monitor Flow Unit Off Normal, alarm annunciated and identified the problem. The surveillance procedure was changed to include instructions on how to use the averaging function of the DMM Fluke Model 189, and all instrument maintenance technicians were briefed of this event and trained on how to use the DMM averaging function. Because this violation was of very low safety significance and it was entered into the licensees corrective action program (IR 287839), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000249/2005003-03)

1R23 Temporary Modification

Main article: IP 71111.23

a. Inspection Scope

The inspectors screened two active temporary modification and assessed the effect of the temporary modification on safety-related systems. The inspectors also determined if the installation was consistent with system design:

  • Overcurrent trip for buses 23, 24, 33, and 34 (temporary change configuration package [TCCP] 353670 and TCCP 353669); and
  • Addition of a Temporary Sample Point for U2 Feedwater Heater Drain Chemistry to troubleshoot the increased moisture carryover on U2 (TCCP EC 353770).

This represented two inspection samples.

b. Findings

No findings of significance were identified.

1EP6 Drill and Training Evaluations

Main article: IP 71114.06

February 7, 2005, Emergency Preparedness Performance Indicator

a. Inspection Scope

The inspectors observed station personnel during a licensee-only-participation emergency preparedness training exercise on February 7, 2005, to determine the effectiveness of drill participants and the adequacy of the licensees critique in identifying weaknesses and failures. The drill scenario involved the loss of instrument air (recoverable), feedwater level control system setpoint drift, feedwater system vibration, unisolable feed line break in drywell, and high pressure coolant injection system start failure.

This represented one inspection sample.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Public Radiation Safety

2PS2 Radioactive Material Processing and Transportation (71122.02)

.1 Radioactive Waste System Description and Waste Generation

a. Inspection Scope

The inspectors reviewed the liquid and solid radioactive waste system descriptions in the Updated Final Safety Analysis Report (UFSAR) and the 2002 and 2003 Annual Radioactive Effluent Release Reports for information on the types and amounts of radioactive waste (radwaste) generated and disposed.

The inspectors reviewed the scope of the licensees audit/self-assessment activities with regard to radioactive material processing and transportation programs to determine if those activities satisfied the requirements of 10 CFR 20.1101(c).

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.2 Radioactive Waste System Walkdowns

a. Inspection Scope

The inspectors walked down portions of the solid radwaste processing systems to verify that these systems were consistent with the descriptions in the UFSAR and the Process Control Program, and to assess the material condition and operability of the systems. The status of radioactive waste process equipment that was not operational or was abandoned in-place was reviewed along with the licensees administrative and physical controls in order to ensure that the equipment would not contribute to an unmonitored release, adversely affect operating systems, or be a source of unnecessary personnel exposure. The inspectors also discussed with radiation protection management its ongoing efforts to consolidate the large number of satellite radioactive material/radwaste storage areas and plans to address the lack of a complete inventory for some of the waste storage bays in the Radwaste Building.

The inspectors reviewed the licensees processes for transferring waste resin into shipping containers (and for dewatering) to determine if appropriate waste stream mixing and sampling was performed so as to obtain representative waste stream samples for analysis. The inspectors evaluated a problem which occurred in 2004, that involved a small quantity of spent resin found loose inside a shipping cask (outside the primary container) upon its receipt at a low level waste disposal facility to ensure that the licensees resin sluicing, waste processing and packaging methods were adequate. The inspectors also reviewed the licensees practices for the collection of area smear surveys to represent the dry-active waste (DAW) stream and the method used for determining the radionuclide mix of various filter media to ensure they were representative of the intended radwaste stream. Additionally, the inspectors reviewed the methodologies for quantifying gamma emitting radionuclide waste stream content, for determining waste stream tritium concentrations and for waste concentration averaging to ensure that representative samples of the waste products were provided for the purposes of waste classification pursuant to 10 CFR 61.55.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Waste Characterization and Classification

a. Inspection Scope

The inspectors reviewed the licensees methods and procedures for determining the classification of radioactive waste shipments including the use of scaling factors to quantify difficult-to-measure radionuclides (e.g., pure alpha or beta emitting radionuclides and those that decay by electron capture). The inspectors reviewed the licensees radiochemical sample analysis results for each of the licensees waste streams which consisted of ion exchange (bead-type) resins, various filter media, concentrator waste, DAW, and activated metals. The reviews were conducted to verify that the licensees program assured compliance with 10 CFR 61.55 and 10 CFR 61.56, as required by Appendix G of 10 CFR Part 20. The inspectors also reviewed the licensees waste characterization and classification program to ensure that reactor coolant chemistry data was periodically evaluated to account for changing operational parameters that could potentially affect waste stream classification and thus validate the continued use of existing scaling factors between annual sample analysis updates.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.4 Shipment Preparation and Shipment Manifests

a. Inspection Scope

The inspectors reviewed the documentation of shipment packaging, surveying, package labeling and marking, vehicle checks and placarding, emergency instructions, and licensee verification of shipment readiness for nine non-excepted radioactive material and radwaste shipments made in 2003 and 2004.

The inspectors verified that the requirements specified in the Certificate of Compliance were met for selected shipments made in Type B casks. These shipments included:

  • Contaminated Equipment in a Strong Tight Container;
  • Dewatered Spent Resin in a Type B Cask;
  • Spent Filters in a Type B Cask;
  • Irradiated Metals in a Type B Cask;
  • Irradiated Metals in a Type A Container;
  • Spent Resin in a Strong Tight Container;
  • Spent Resin in a Type B Cask; and
  • Dry Active Waste in a Sea-Land (Strong Tight) Container.

The inspectors selectively verified that the requirements of 10 CFR Parts 20 and 61, and those of the Department of Transportation (DOT) in 49 CFR Parts 170-189 were met for each shipment. Specifically, records were reviewed and some of the staff involved in shipment activities were interviewed to verify that packages were labeled and marked properly, that package and transport vehicle surveys were performed with appropriate instrumentation and survey results satisfied DOT requirements, and that the quantity and type of radionuclides in each shipment were determined accurately. The inspectors also verified that shipment manifests were completed in accordance with DOT and NRC requirements, included the required emergency response information, that the recipient was authorized to receive the shipment, and that shipments were tracked as required by 10 CFR Part 20. The inspectors also reviewed the licensees transportation security plan required by 49 CFR 172.800/802 and discussed its implementation with the licensees shipping specialist.

The inspectors observed a radiation protection technician perform surveys of an outgoing Type B shipment of spent resin which was being prepared for shipment to a waste processor. The technician and the licensees primary shipper were questioned by the inspectors to verify that they had adequate skills to accomplish shipment related tasks, to determine if the shippers were knowledgeable of the applicable regulations and whether shipping personnel demonstrated adequate skills to satisfy package preparation requirements for public transport with respect to NRC Bulletin 79-19, Packaging of Low-Level Radioactive Waste for Transport and Burial, and 49 CFR Part 172 Subpart H. Additionally, the lesson plan used for training station laborers involved in radioactive material shipping was reviewed for compliance with the hazardous material training requirements of 49 CFR 172.704.

These reviews represented two inspection samples.

b. Findings

No findings of significance were identified.

.5 Identification and Resolution of Problems for Radwaste Processing and

Transportation

a. Inspection Scope

The inspectors reviewed selected condition reports, an audit and a self-assessment report that addressed the radioactive waste and radioactive materials shipping program since the last inspection to verify that the licensee had effectively implemented the corrective action program and that problems were identified, characterized, prioritized, and corrected. The inspectors also verified that the licensee's oversight mechanisms collectively were capable of identifying repetitive deficiencies or significant individual deficiencies in problem identification and resolution.

The inspectors also selectively reviewed other corrective action reports generated since the previous inspection that dealt with the radioactive material or radwaste shipping program, and interviewed staff and reviewed documents to determine if the following activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk:

  • Initial problem identification, characterization, and tracking;
  • Disposition of operability/reportability issues;
  • Evaluation of safety significance/risk and priority for resolution;
  • Identification of repetitive problems;
  • Identification of contributing causes;
  • Identification and implementation of effective corrective actions; and
  • Implementation/consideration of risk significant operational experience feedback.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES (OA)

4OA2 Identification and Resolution of Problems

Main article: IP 71152

.1 Routine Quarterly Review

a. Inspection Scope

As discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees corrective action system at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Minor issues entered into the licensees corrective action system as a result of inspectors observations are generally denoted in the report. In addition, in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. This review was accomplished by reviewing daily issue reports and attending daily issue report review meetings.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

Main article: IP 71153

.1 (Closed) LER 50-237;249/2004-006-00: Units 2 and 3 Main Turbine Generator

Rotor Cracks On October 31, 2004, during a Unit 3 refueling outage, the licensee inspected the Unit 3 main turbine generator rotor and identified a crack in the shaft near the rotor coupling. Both units had been experiencing increasing trends in vibration levels on main turbine generator bearings 9 and 10 since May 2004. The licensee removed Unit 2 from service to conduct an inspection of its rotor and identified a crack in the same general location and of a similar configuration. The licensee determined the cause of the rotor cracks to be intermittent oscillating torsional loading on the generator rotor which produced a torsional fatigue failure mode. However, the licensee was unable to determine the cause of the intermittent oscillating torsional loading. Both main turbine generator rotors were sent off-site for inspection and replacement of the cracked end of the shaft with new stub shafts. The LER was reviewed by the inspectors and no findings were identified. This LER is closed.

.2 (Closed) Unresolved Item 05000237/2004010-01 (DRP); 05000249/2004010-01

(DRP): Units 2 and 3 UFSAR indicate that there are multiple pumps available to supply make up water to the Unit 2 and Unit 3 Isolation Condensers The Updated Final Safety Analysis Report (UFSAR) Section 3.4.1.1, External Flood Protection Measures, stated that if forecasted flood levels exceed 517 feet, 150 gallon per minute emergency makeup pumps are connected to the fire system. The licensee has since went to the use of one portable diesel driven pump in response to a flooding event. However, the 10 CFR 50.59 screening performed did not address the change in the number and type of pumps.

Upon further investigation into this issue, the inspectors concluded that the use of only one pump to supply makeup water to both the Unit 2 and Unit 3 Isolation condensers during a design basis flood event was acceptable as determined in calculation DRE99-0035. Subsequently, the licensee submitted an UFSAR change to reflect the use of one pump. The inspectors reviewed the licensees 10 CFR 50.59 review of the change to the UFSAR and had no concerns. This item is considered closed.

4OA4 Cross-Cutting Findings

.1 A finding described in 1R19 of this report had, as its primary cause, Problem

Identification and Resolution, in that there were inadequate corrective actions to prevent recurrence for a significant condition adverse to quality that occurred on January 19, 2001; and there were no corrective actions assigned to prevent recurrence of a significant condition adverse to quality that occurred on November 29, 2004. Both events involved the failure of the Unit 2 emergency diesel generator air start regulating valve due to corrosion build up on the valve stem.

4OA5 Other Activities

.1 Preoperational Testing of an Independent Spent Fuel Storage Installation (ISFSI)

(60854.1)

a. Inspection Scope

The inspectors evaluated whether the licensee had effectively implemented procedures to weld the multi-purpose canister (MPC) lid and secondly, to remove the lid. The inspection focused on the activities of a new welding contractor, PCI Energy Services, for the Dresden and Quad Cities Station new ISFSI campaigns.

The inspectors observed welding, non-destructive examination (NDE), and lid removal dry-runs on MPC mockups in the contractors shop. The inspectors verified that the work was performed as specified in the general project instructions, PI-900343-04, Closure Welding of Multi-Purpose Canister, and PI-900343-05, Canister Cutter-Operation. The inspectors reviewed the procedures and compared them to the requirements specified in the Certificate of Compliance, the Technical Specifications, and the Safety Analysis Report.

b. Findings

No findings of significance were identified.

4OA6 Meetings

.1 Interim Exit Meetings

Interim exit was conducted for:

  • Public radiation safety radioactive waste processing and transportation program inspection with Mr. D. Wozniak on February 4, 2005, and a followup telephone conversation with Mr. S. Taylor on February 9, 2005.
  • Independent Spent Fuel Storage Installation with the Dry Cask Project Managers, Mr. M. Mikota of Dresden, and Mr. D. Moore of Quad Cities, on March 16, 2005.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. Bost, Site Vice President
D. Wozniak, Plant Manager
S. Bell, Shipping Specialist
H. Bush, Radiological Engineering Manager
R. Conklin, Radiation Protection Supervisor
J. Fox, Design Engineer
R. Gadbois, Operations Director
D. Galanis, Design Engineering Manager
V. Gengler, Dresden Site Security Director
J. Griffin, Regulatory Assurance - NRC Coordinator
P. Salas, Regulatory Assurance Manager
R. Kalb, Chemistry ODCM Coordinator
T. Loch, Supervisor, Design Engineering
M. McGivern, System Engineer
M. Mikota, Dry Cask Project Manager, Dresden
D. Moore, Dry Cask Project Manager, Quad Cities
D. Nestle, Radiation Protection Technical Manager
M. Overstreet, Radiation Protection Supervisor
R. Quick, Security Manager
N. Spooner, Site Maintenance Rule Coordinator
B. Surges, Operations Requalification Training Supervisor
G. Bockholdt, Maintenance Director
S. Taylor, Radiation Protection Director

NRC

M. Ring, Chief, Division of Reactor Projects, Branch 1

IEMA

R. Schulz, Illinois Emergency Management Agency
R. Zuffa, Resident Inspector Section Head, Illinois Emergency Management Agency

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000249/2005003-01 URI Install U3 Core Spray Lower Sectional Replacement
05000237/2005003-02 NCV Inadequate Corrective Actions for a Significant Condition Adverse to Quality Involving the Failure of the Unit 2 Emergency Diesel Generator Air Start Regulator
05000249/2005003-03 NCV Performance Deficiency While Performing Surveillance Procedure DIS 700-02, APRM/RBM [average power range monitor/rod block monitor] Flow Instrumentation Total Drive Flow Adjustment, Revision 16

Closed

05000237/2005003-02 NCV Inadequate Corrective Actions for a Significant Condition Adverse to Quality Involving the Failure of the Unit 2 Emergency Diesel Generator Air Start Regulator
05000249/2005003-03 NCV Performance Deficiency While Performing Surveillance Procedure DIS 700-02, APRM/RBM [average power range monitor/rod block monitor] Flow Instrumentation Total Drive Flow Adjustment, Revision 16 50-237;50-249/2004-006-00 LER Units 2 and 3 Main Turbine Generator Rotor Cracks
05000237/2004010-01 URI Units 2 and 3 UFSAR indicate that there are multiple
05000249/2004010-01 pumps available to supply make up water to the Unit 2 and Unit 3 Isolation Condensers

Discussed

None

LIST OF ACRONYMS USED CFR Code of Federal Regulations CR Condition Report DAW Dry-Active Waste DIS Dresden Instrument Surveillance DOA Dresden Operating Abnormal Procedure DOS Dresden Operating Surveillance DOT Department of Transportation DRP Division of Reactor Projects DRS Division of Reactor Safety EC Engineering Change EDG Emergency Diesel Generator IEMA Illinois Emergency Management Agency IMC Inspection Manual Chapter LER Licensee Event Report MWe megawatts electrical NCV Non-Cited Violation NRC Nuclear Regulatory Commission OA Other Activities PI Performance Indicator radwaste Radioactive Waste SDP Significance Determination Process TCCP Temporary Change Configuration Package TS&P Test, Specification, and Procedure UFSAR Updated Final Safety Analysis Report URI Unresolved Item WO Work Order

LIST OF DOCUMENTS REVIEWED

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