IR 05000237/2002004

IR 05000237/2002-004(DRP), IR 05000249/2002-004(DRP), Exelon Generation Company, Dresden Nuclear Power Station, Units 2 and 3, Inspection on 02/08/2002-03/31/2002 Re Identification and Resolution of Problems

ML021190480
Person / Time
Site:	Dresden
Issue date:	04/26/2002
From:	Ring M Division Reactor Projects III
To:	Skolds J Exelon Generation Co, Exelon Nuclear
References
IR-02-004
Download: ML021190480 (36)

Text

ril 26, 2002

SUBJECT:

DRESDEN NUCLEAR POWER STATION NRC INSPECTION REPORT 50-237/02-04(DRP); 50-249/02-04(DRP)

Dear Mr. Skolds:

On March 31, 2002, 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 3, 2002.

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, the inspectors identified one issue for which no risk significance or color was assigned. Additionally, the inspectors identified four issues of very low safety significance (Green). The four issues were determined to involve violations 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 deny these Non-Cited Violations, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspectors at the Dresden Nuclear Power Station. In accordance with 10 CFR 2.790 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/NRC/ADAMS/index.html (the Public Electronic Reading Room).

Sincerely,

/RA/

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

Enclosure:

Inspection Report 50-237/02-04(DRP);

50-249/02-04(DRP)

REGION III==

Docket Nos: 50-237; 50-249 License Nos: DPR-19; DPR-25 Report No: 50-237/02-04(DRP); 50-249/02-04(DRP)

Licensee: Exelon Generation Company Facility: Dresden Nuclear Power Station, Units 2 and 3 Location: 6500 North Dresden Road Morris, IL 60450 Dates: February 8, 2002, through March 31, 2002 Inspectors: D. Smith, Senior Resident Inspector B. Dickson, Resident Inspector P. Pelke, Reactor Engineer D. Pelton, Senior Operations Engineer W. Slawinski, Senior Radiation Specialist R. Zuffa, Illinois Department of Nuclear Safety Approved by: Mark Ring, Chief Branch 1 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000237-02-04(DRP), IR 05000249-02-04(DRP), on 3/31/2002, Exelon Generation Company, Dresden Nuclear Power Station, Units 2 and 3. Identification and Resolution of Problems.

The inspection was conducted by resident inspectors, one senior radiation specialist, one senior operations engineer, and one reactor engineer. The inspection identified four Green and one No Color findings, of which four were considered Non-Cited Violations. 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 are indicated by No Color or by the severity level of the applicable violation. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at http://www.nrc.gov/NRR/OVERSIGHT/index.html.

A. Inspector Identified Findings Cornerstone: Mitigating Systems

• Green. A Non-Cited Violation was identified for the licensees failure to have an adequate preventative maintenance procedure for the 480 volt motor control center (MCC) cubicles for ensuring lock washers were installed in the auxiliary control assemblies. As a result, the lock washers were not installed in 37 safety related assemblies, including one which led to the failure of the B loop recirculation pump discharge valve (NCV 50-237/02-04-03).

The finding was of very low safety significance because the reactor was in a shutdown condition, the emergency core cooling systems were not required in this condition, core spray was available, and the 36 other affected components were operable (4AO2).

• Green. A Non-Cited Violation was identified for the licensees failure to prepare supporting operability documentation for 36 safety related 480 volt MCC cubicles which had missing lock washers in their auxiliary contact assemblies (NCV 50-237/249/02-04-04).

The finding was of very low safety significance because it was determined that all 36 degraded components were operable (4OA2).

• Green. A Non-Cited Violation was identified for the licensees failure to promptly identify and correct the condition of missing reactor protection system (RPS) cable tray covers (NCV 50-249/02-04-05).

The finding was of very low safety significance because in each case two other RPS channels are routed in a different location which are sufficient to allow the RPS system to perform its intended safety function (4OA2).

• Green. A Non-Cited Violation was identified for the licensees failure to correct conditions adverse to quality when station personnel incorrectly connected test equipment to the emergency diesel generator (NCV 50-237/249/02-04-06).

The finding was of very low safety significance since the incorrect connection did not have any adverse impact on the plant (4OA2).

Cross-Cutting Issues: Corrective Actions No Color. The inspectors identified four instances where the licensee failed to promptly identify and correct conditions adverse to quality. In the first instance, during the licensee followup actions for missing lock washers in auxiliary contacts for safety related motor control center cubicles, the licensee failed to prepare supporting operability documentation for an additional 36 safety related components. In the second instance, the isolation condenser experienced a second water hammer after the licensee failed to initiate a condition report after a previous water hammer in August 2001. In the third instance, the licensee failed to promptly identify and correct the condition of missing reactor protection system cable tray covers on Unit 2 which had been identified on September 28, 2001. Finally, following the incorrect connection of a test recorder during undervoltage testing for the Unit 3 emergency diesel generator on September 24, 2000, the licensee failed to identify the full extent of condition and complete previously identified corrective actions (FIN 50-237/249/02-04-07).

The individual findings were of very low significance; however, the findings could have had a credible impact on safety or could have been a precursor to a significant event by affecting the availability, reliability, operability or functionality of mitigating equipment (4OA2).

B. Licensee Identified Findings A violation of very low significance which was identified by the licensee has been reviewed by the inspectors. This violation is listed in Section 4OA7.

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). On February 24, 2002, operators reduced load to approximately 750 MWe to perform feedwater testing and remained at 820 MWe due to oscillations on the electrohydraulic control system Number 2 control valve. The operators returned the Unit to 912 MWe on March 5, 2002.

Unit 3 began the inspection period at 822 MWe (100 percent thermal power). Unit 3 completed an 8 day maintenance outage primarily to replace 17 jet pump hold-down beams and perform preventive maintenance on the high pressure coolant injection system. The unit was taken off-line on March 16, 2002, and returned on-line on March 24, 2002.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity 1R05 Fire Protection (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, 534' Elevation, Switchgear Area (Fire Zone 8.2.6.A)

Unit 3 Turbine Building, 517' Elevation, Switchgear Area (Fire Zone 8.2.5.E)

Unit 3 Turbine Building, 538' Elevation, Reactor Feedwater Switchgear, Hydrogen Seal Area (Fire Zone 8.2.6.E)

b. Findings No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11)

.1 Written Examination and Operating Test Results a. Inspection Scope The inspectors reviewed the pass/fail results of individual written tests, operating tests, and simulator operating tests (required to be given per 10 CFR 55.59(a)(2))

administered by the licensee during calender year 2001.

b. Findings No findings of significance were identified.

.2 Observation of Licensed Operator Simulator Training a. Inspection Scope The inspectors observed Crew #1 on March 27, 2002. The scenario consisted of a reactor building vent radiation monitor failure, reactor feed pump high vibration, and loss of coolant accident in the drywell and failure to scram.

b. Findings No findings of significance were identified.

1R12 Maintenance Rule Implementation (71111.12)

a. Inspection Scope The inspectors assessed the licensees implementation of the maintenance rule by determining if systems were properly scoped within the maintenance rule. The inspectors also assessed the licensees characterization of failed structures, systems, and components, and determined whether goal setting and performance monitoring were adequate for the main steam system, station blackout diesels, and core spray system.

b. Findings During the review of the station blackout (SBO) diesel generator system, the inspectors identified that a recent SBO heating, ventilating and air conditioning temperature controller failure on February 6, 2002, had not been captured in a condition report. The licensee documented this issue as a work request. The maintenance rule process only routed condition reports contained in the maintenance rule database to the system engineers. Therefore, the inspectors were concerned that the licensees maintenance rule process would not have routed this deficiency, as documented on the work request, to the system engineer for determining whether this equipment failure constituted a maintenance rule functional failure. Additional investigation into this issue by the licensee determined that the station made a change to the corrective action program in August 2001 which allowed a work request to be generated instead of a condition report for low level equipment problems. However, the change did not take into account how the system engineer would evaluate these equipment problems for maintenance rule functional failures because the work requests did not automatically input into the maintenance rule database for routing to the system engineers. As a result of this oversight, there was a backlog of 1,700 open work requests pending reviews by the system engineers to determine if any of the identified equipment deficiencies and failures resulted in maintenance rule functional failures. The backlog review is scheduled for completion on May 10, 2002. As an interim corrective action, the maintenance rule coordinator will receive all work requests and conduct an initial review

to determine if equipment deficiencies require further review by system engineers for maintenance rule functional failure determinations. Pending the completion and review of the results of this effort, this issue will be an Unresolved Item (URI 50-237/02-04-01 and 50-249/02-04-01(DRP)).

1R13 Maintenance Risk Assessments and Emergent Work Control (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 completed this evaluation while the licensee performed surveillance testing activities on the Unit 3 isolation condenser and during maintenance activities on the 3A standby liquid control pump, the 2C containment cooling service water pump, the 2B emergency diesel generator starting air compressor, and the Unit 2 battery charger system.

b. Findings No findings of significance were identified.

1R15 Operability Evaluations (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 evaluation of the following issues:

installation of non-environmentally qualified air operated valves in the drywell equipment and floor drain system, missing lock washer in the auxiliary contacts for the 480V motor control center cubicles, and ability of the Unit 2/3 reactor building crane to function under a design load.

b. Findings No findings of significance were identified.

1R19 Post Maintenance Testing (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 the operational readiness status consistent with the design and licensing basis documents. The inspectors reviewed work activities associated with the

3B containment cooling service water pump and the Unit 3 electrohydraulic control system.

b. Findings No findings of significance were identified.

1R20 Refueling and Outage Activities (71111.20)

a. Inspection Scope The inspectors reviewed and evaluated several outage activities during the Unit 3 maintenance outage. The purpose of the outage, which was performed March 16-24, 2002, was to replace 17 jet pump hold-down beams and perform preventive maintenance activities on the high pressure coolant injection system. The evaluation was performed to ensure that the licensee appropriately considered risk factors during the development and execution of planned activities. The inspectors also ensured that technical specification requirements were verified to have been met for changing modes.

b. Findings On March 23, 2002, the licensee identified that, during startup from the maintenance outage, the high pressure coolant injection system (HPCI) was not properly aligned when reactor steam dome pressure reached 150 psig. The on-shift crew decided to leave steam isolated to the system in that steam inlet valves (3-2301-4 and 5) remained closed. This decision was based on the on-shift crews understanding that the HPCI system was inoperable due to the significant amount of maintenance that had been performed on the system during the outage. The licensee informed the resident inspectors that the on-shift crew understood the need to exceed 150 psig to obtain the appropriate plant conditions to perform the post maintenance test (low pressure test).

Therefore, when steam dome pressure reached 150 psig, the on-shift crew entered technical specification limiting condition for operation action statement 3.5.1.F for HPCI being inoperable which required immediate verification that the isolation condenser was operable and restoration of HPCI within 14 days. The decision to increase steam dome pressure above 150 psig with steam isolated to the HPCI system potentially resulted in the licensee violating technical specification limiting condition for operation 3.5.1 which required the HPCI system to be operable when steam dome pressure is equal to or greater than 150 psig. The licensee properly aligned HPCI to support post-maintenance testing approximately 43 minutes after exceeding 150 psig. The licensee planned to conduct a root cause investigation for this incident. This issue was documented in CR

1. 101056. This issue will be an Unresolved Item pending the inspectors review of the licensees completed root cause investigation (URI 50-249/02-04-02 (DRP)).

1R22 Surveillance Testing (71111.22)

a. Inspection Scope The inspectors observed surveillance testing on risk-significant equipment. The inspectors assessed whether the selected plant equipment could perform its intended safety function and satisfy the requirements contained in Technical Specifications.

Following the completion of the 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 review included surveillance testing activities for the calibration of the Unit 2 narrow range reactor pressure transmitter, the operational test of the Unit 2 station blackout diesel, the calibration of the reactor vessel high pressure scram pressure switches, and the Unit 2 condenser low vacuum pressure switch calibration and functional test.

b. Findings No findings of significance were identified.

3. RADIATION SAFETY Cornerstone: Occupational Radiation Safety 2OS1 Access Control to Radiologically Significant Areas (71121.01)

.1 Plant Walkdowns and Radiological Boundary Verifications a. Inspection Scope The inspector conducted walkdowns of the radiologically protected area to verify the adequacy of radiological area boundaries and postings. Specifically, the inspector walked down numerous radiologically significant work area boundaries (high and locked high radiation areas) in the Unit 2 and 3 Reactor Buildings (including the Unit 3 drywell),

the Turbine Buildings, and the Radwaste Building and performed confirmatory radiation measurements to determine if these areas and selected radiation areas were properly posted and controlled in accordance with 10 CFR Part 20, licensee procedures, and Technical Specifications. The inspector also evaluated the radiological condition of those areas walked down to assess the radiological housekeeping and contamination controls.

b. Findings No findings of significance were identified.

.2 High Radiation Area and Very High Radiation Area Access Control a. Inspection Scope The inspector reviewed the licensees procedures, practices and associated documentation for the control of access to radiologically significant areas (high, locked high, and very high radiation areas) and assessed compliance with Technical Specifications, procedures and the requirements of 10 CFR 20.1601 and 20.1602. In particular, the inspector reviewed the licensees practices and records for the control of keys to locked high radiation areas (LHRAs) and very high radiation areas (VHRAs), the use of access control guards to control entry into such areas, and the licensees methods for independently verifying proper closure and latching of LHRA and VHRA doors upon area egress. The inspector also observed and evaluated the adequacy of the LHRA controls implemented for access to the Unit 3 drywell and the high radiation area access controls used during reactor cavity decontamination. Additionally, radiological postings were reviewed, and access control boundaries were challenged by the inspector throughout the plant to verify that high, locked high and very high radiation areas were properly controlled.

b. Findings No findings of significance were identified.

.3 Review of Radiologically Significant Work a. Inspection Scope The inspector reviewed radiation work permit (RWP) and as-low-as-is-reasonably-achievable (ALARA) plan packages, attended the pre-job ALARA brief for cavity decontamination and observed the work activities for a job that took place in a LHRA during the Unit 3 maintenance outage (D3M09). These activities were performed to verify the adequacy of surveys, access controls, and postings; to assess the exchange of work area radiological information; and to evaluate radiation worker and radiation protection technician performance. The inspector also evaluated the licensees procedure and practices for dosimetry placement and use of multiple dosimetry in high radiation areas having significant dose gradients for compliance with the requirements of 10 CFR 20.1201 and applicable Regulatory Guides. Additionally, the licensees dose tracking and documentation practices were reviewed for recent work that involved the issuance of multiple whole body and/or extremity dosimetry to verify that worker dose was recorded consistent with 10 CFR 20.2106.

b. Findings No findings of significance were identified.

.4 Control of Non-Fuel Materials Stored in the Spent Fuel Pools a. Inspection Scope The inspector reviewed the licensees procedures for the storage of highly activated or contaminated materials (non-fuel) within the spent fuel or other storage pools and specifically evaluated the practices implemented for spent fuel pool storage of the irradiated jet pump hold-down beams that were replaced during the maintenance outage. Radiation protection (RP) and fuel handling procedures were reviewed, RP staff were interviewed, and walkdowns of the refuel floor were conducted. The inspector assessed the adequacy of the administrative and physical controls for the underwater storage of non-fuel materials to verify consistency with the licensees procedures and with Regulatory Guide 8.38, Information Notice 90-33, and applicable Health Physics Positions in NUREG/CR-5569. Procedure inconsistencies and differences in the radiological controls used for short versus longer term storage of irradiated material in the spent fuel pool were discussed with RP management, and plans to alter current practices were reviewed for adequacy relative to industry and NRC guidelines.

b. Findings No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02)

.1 Radiation Dose Goals and Trending a. Inspection Scope The inspector reviewed job specific and cumulative exposure performance for D3M09 to assess the licensees dose performance compared to pre-outage exposure goals and projections. The inspector also reviewed the licensees dose forecasting practices for radiologically significant jobs completed during the outage to determine if adequate technical bases for job dose estimates existed and to determine if prior outage experiences, resource estimates and industry operating experiences were used to establish reasonable dose estimates. Additionally, the inspector reviewed the effectiveness of the RP organizations exposure tracking for the outage to verify that the licensee could identify problems with its exposure performance and take actions to address identified deficiencies.

b. Findings No findings of significance were identified.

.2 Radiological Work Planning a. Inspection Scope The inspector reviewed the licensees procedure for ALARA Plan development and evaluated D3M09 outage ALARA plans to verify consistency with the procedure and to assess the overall adequacy of the plans relative to both licensee and industry practices.

Specifically, the inspector reviewed the ALARA plans developed for refuel floor work, for drywell activities and for outage radiography and assessed the adequacy of the radiological planning associated with each work activity.

The inspector reviewed the RWP and the ALARA plan completed for each job and assessed the radiological engineering controls and other dose mitigation techniques to verify that they included appropriate controls to reduce dose. These documents were also reviewed to determine if job history files, licensee lessons learned, and industry operating experiences were adequately integrated into each work package. Additionally, the inspector discussed ALARA planning with involved RP staff to verify that adequate interface between contractors, station work groups, and ALARA staff occurred during job planning.

b. Findings No findings of significance were identified.

.3 Implementation of ALARA Controls and Radiological Oversight of Work a. Inspection Scope The inspector evaluated the execution of the ALARA plans for jet pump hold-down beam removal, for reactor cavity decontamination and for under-vessel work in the drywell, all which were performed during D3M09. The inspector reviewed the adequacy of radiological surveys performed for these jobs, evaluated the radiological work controls, and assessed worker performance and RP staff oversight. Total effective dose equivalent (TEDE) ALARA evaluations were also assessed for technical adequacy. The inspector evaluated the licensees radiological engineering controls utilized at these work locations to determine if the controls were consistent with those specified in the ALARA plans. The inspector also observed and questioned both the RP staff that provided job coverage for these activities and the radiation workers (radworkers)

involved in selected work to verify that they had adequate knowledge of radiological work conditions and ALARA controls.

b. Findings No findings of significance were identified.

.4 Identification and Resolution of Problems a. Inspection Scope The inspector reviewed the licensees condition report (CR) database and several individual CRs related to the radiological access control and ALARA programs that were generated between December 2001 and March 19, 2002. The review was conducted to assess the effectiveness of the corrective action program to identify problems and to develop corrective actions. Selected CRs were discussed with RP staff and management to determine if problem characterization was accurate and to verify that extent of condition reviews were adequately completed or were in the process of being performed. The inspector also discussed with RP management its practice of conducting both root cause and apparent cause evaluations to determine if they were initiated at appropriate thresholds. Additionally, the inspector reviewed the preliminary results of a root cause evaluation undertaken by the licensee to assess RP performance issues to verify that the licensee was proactively evaluating problems and trends.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification (71151)

a. Inspection Scope The inspectors reviewed a sample of plant records and data against the reported performance indicators in order to determine the accuracy of the indicators.

Unit 2 and Unit 3 Safety System Unavailability, Emergency AC Power (January 2001 through December 2001)

Unit 2 and Unit 3 Unplanned Transients Per 7000 Critical Hours (October 2000 through December 2001)

b. Findings No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

a. Inspection Scope The inspectors reviewed several issues to verify that the licensee had taken effective corrective action with respect to classification and prioritization of the resolution of problems, determination of the extent of condition, evaluation and disposition of operability, and completion of corrective action in a timely manner.

b. Findings

.1 (Closed) Licensee Event Report (LER) 50-237/2001-003-00: Failure of Recirculation Pump Discharge Valve to Close Causing Low Pressure Coolant Injection (LPCI)

Inoperability.

Two Green findings involving two Non-Cited Violations were identified. The first finding involved the licensees failure to have an adequate preventative maintenance procedure for the 480 volt motor control center (MCC) cubicles ensuring lock washers were installed in the auxiliary control assemblies. The second finding involved the licensees failure to prepare any supporting operability documentation for missing lock washers in the auxiliary control assemblies for 36 safety related 480 volt MCC cubicles.

On September 2, 2001, the B loop recirculation pump discharge valve failed while the licensee was attempting to manipulate the valve with the reactor in Mode 4. The licensee performed a root cause investigation for this valve failure. The licensees investigation revealed that the failure was due to a normally closed auxiliary contact sticking open. The contact failed due to the auxiliary contact plunger arm being off its normal plastic stop which caused the contact to bind. The mispositioned plunger arm was due to a loosened plunger post that was caused by a missing lock washer. The licensee installed a new auxiliary contact assembly and satisfactorily tested the valve.

The licensee determined that an inadequate procedure led to the valve failing. The licensee performed preventive maintenance on the valve every 6 years as specified by Dresden Electrical Surveillance procedure DES 7300-05, Maintenance and Surveillance of Environmental Qualification and Safety Related 480 Volt MCC, Revision 13. The procedure did not specify checking for plunger post tightness or that a lock washer was installed. The licensee revised DES 7300-05 which included adding a picture which shows the configuration of all the parts. In determining the extent of condition during the root cause analysis for this valve failure, the licensee identified that numerous additional breakers required inspection including all the General Electric Series 7700, NEMA size 0, 1 and 2 MCC cubicles. The list of MCC cubicles that did not have a lock washer installed was provided to operations to determine any potential plant impact and to assist in determining the lock washer installation date. This issue was given action tracking item number 74173-17.

The licensee determined the extent of condition through a walkdown which was conducted October 6 - 8, 2001. Also, the licensee verified proper alignment of the contact operating post and plunger during the walkdown. A total of 74 MCC cubicles, of which 36 were safety related, were identified as having missing lock washers. The licensee determined that 25 had safety or production risk and were appropriately scheduled to have the lock washer installed. The licensee subsequently changed the schedule of several of these components. AR 000074173-19-00 dated December 18, 2001, was initiated to track completion of work orders for lock washer installation. The inspectors reviewed the list of components missing the required lock washers and the licensees set schedule to install the missing lock washers. The inspectors raised a concern about the fact that the non-safety related components were scheduled for lock washer installation in 2002 and 2003 while most of the safety related

components were scheduled in 2004. The licensee identified that some of the schedule dates had been changed without the knowledge of the Operations Department.

At the time of discovering the additional components with missing lock washers, the licensee did not prepare any supporting operability documentation. Also, the inspectors were concerned that the licensee had not documented the additional safety related and non-safety related breakers that were also missing lock washers in the original recirculation pump discharge valve failure LER.

Subsequently, the licensee initiated CR #00093478 on February 5, 2002, to re-identify the priority MCC cubicles and assign individual action tracking items to complete repair of each cubicle. The licensee characterized this issue in the CR as a root cause corrective action breakdown where there was no owner/process to ensure the corrective actions were completed satisfactorily. The licensee subsequently prepared supporting operability documentation on February 8, 2002, and performed another walkdown to compare the as-found condition with the General Electric recommended inspection criteria (auxiliary contacts were properly aligned and there were no visible gaps or space between the plunger post and the plastic plunger). The licensee concluded that all components were operable because the posts were not loose.

The inadequate procedure issue, leading to the recirculation pump discharge valve failure, was considered more than minor because it had an adverse impact on safety in that, the low pressure coolant injection system would be rendered inoperable during a loss of coolant accident with the break in the A loop of the recirculation system because the low pressure coolant injection loop selection logic would select the B loop.

The inadequate procedure issue had minimal safety significance because the reactor was in a shutdown condition, the emergency core cooling systems were not required in this condition, core spray was available, and the 36 other affected components were operable (Green).

10 CFR 50, Appendix B, Criterion V, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.

Contrary to the above, DES 7300-05 was inadequate in that it did not require the installation of a lock washer or verify the tightness of the plunger post. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation (NCV 50-249/02-04-03(DRP)) consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered into the licensees corrective action program as CR #98448 and CR #D20000-05356.

The issue of not preparing supporting operability documentation for the additional degraded components was considered more than minor because the issue could be viewed as a precursor to a significant event. Ultimately, this issue had minimal safety significance because the licensee subsequently determined that all the 36 degraded components were operable (Green).

10 CFR 50, Appendix B, Criterion XVI, requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

Station procedure LS-AA-105, Operability Determinations, Revision 0, requires that if there is a reasonable assurance that the structure, system and components are operable, but a more rigorous evaluation is warranted, then engineering prepare and review supporting operability documentation which should be completed within 3 full business days of its initiation.

Contrary to 10 CFR 50, Appendix B, Criterion XVI, the licensee failed to implement timely corrective actions in that supporting operability documentation for the additional MCC cubicles that were missing lock washers was not prepared in accordance with LS-AA-105 until February 8, 2002. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation (NCV 50-249/02-04-04(DRP))

consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered into the licensees corrective action program as CR #00074173 and CR #00093478.

.2 Ineffective Corrective Action for Missing Reactor Protection System (RPS) Cable Tray covers The inspectors identified one Green finding involving a Non-Cited Violation regarding the licensees failure to promptly identify and correct the condition of missing RPS cable tray covers.

On September 28, 2001, the inspectors identified that a number of protective covers were missing from the cable trays of the RPS instrument cable routing system. These trays were located in the Unit 2 reactor building approximately 25 off of the ground floor elevation (517). Section 7.2.5.2 of the Dresden Updated Final Safety Analysis Report (UFSAR) states in part that, All protection system wiring is run in rigid metallic conduit or solid trays with covers. Following the inspectors notification, the licensee generated CR #77244. The immediate actions taken section of CR #77244 stated that the RPS system cable trays were walked down. WO #000193336 was generated to restore the cable tray covers and CR #77244 was subsequently closed on December 5, 2001.

On March 12, 2002, the inspectors conducted a follow-up plant walkdown of other RPS cable trays passing through the Unit 2 turbine building. While touring the Unit 2 turbine building elevation 538 the inspectors noticed that the uppermost RPS cable tray (approximately 18 off the ground) was missing one of its protective covers. The inspectors were also aware that the specific area was in the near vicinity of the feedwater regulating valves; this area was considered a potential high energy line-break area. The inspectors concern was that exposed RPS cables in this area could be subject to high-energy impact from the affects of a feedwater line break. An additional concern was that the RPS cable trays located in the reactor building had their covers banded and secured from movement, and the RPS cable tray covers that were located in the high energy feedwater regulating station area were only laid over the top of the cable trays without being secured. The inspectors informed the licensee of these RPS cable tray integrity concerns. The licensee initiated CR #98992 to address and track the issue. The licensees immediate actions were to generate WO #40357 to fabricate and replace the missing cover. The inspectors noted that CR #98992 was marked No by

both the originator and supervisor for the question as to whether this was a repeat condition. By marking the CR No the management review committee (MRC) members were unaware that this issue was a repeat condition. This key information was important because when reviewing CR #98992, the MRC members debated whether a subsequent walkdown of the RPS cable trays was necessary to determine the extent of condition. In this case, the MRC conservatively recommended a sampling walkdown without being aware of the repeat condition.

On March 21, 2002, during the Unit 3 maintenance outage, the inspectors performed a Unit 3 high pressure heater bay entry and inspection. During the inspection of this area the inspectors again noticed that a portion of the RPS cable tray was missing its cable tray cover. The inspector notified the licensee who initiated CR #100368 and WO #00421697 to resolve this issue.

10 CFR 50, Appendix B, Criterion XVI, requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

Contrary to the above, following the identification of several missing RPS cable tray covers on September 28, 2001, adequate corrective action was not taken to identify all missing RPS cable tray covers. Subsequently, missing RPS cable tray covers were identified on March 12 and 22, 2002. This finding was more than minor because it could have a credible impact on safety by affecting the availability, reliability, operability or functionality of mitigating equipment. This finding was considered to be a very low safety significance because in each case two other RPS channels are routed in a different location which are sufficient to allow the RPS system to perform its intended safety function (Green). Because of the very low safety significance, this violation is being treated as a Non-Cited Violation (NCV 50-237/02-04-05 and 50-249/02-04-05(DRP)) consistent with Section VI.A.1 of the NRC Enforcement Policy.

This issue was entered into the licensees corrective action program as CR #77244, CR #98992, and CR #100368.

.3 Ineffective Correction Action Taken During Connection of Test Equipment for the Emergency Diesel Generator One Green Finding involving a Non-Cited Violation was identified regarding the licensees failure to correct conditions adverse to quality when station personnel incorrectly connected test equipment to the emergency diesel generator.

On March 6, 2002, the instrument maintenance mechanics (IMs) incorrectly connected a test recorder to the Unit 3 emergency diesel generator during the performance of Dresden Operations Surveillance DOS 6600-12, Endurance and Margin/Full Load Rejection/ECCS/Hot Restart, Revision 22. As a result, the test had to be rerun.

Procedural Step 1.a.d, on Checklist B, specified installing the load sequence recorder by positioning the conductors between the recorder and the pair termination location identified on the diesel generator chart recorder connections. The recorder connections indicated that Pen #3 should be connected to the diesel generator output voltmeter for monitoring the diesel generator frequency; however the IMs connected Pen #3 to the frequency meter. This issue was documented in CR #98448.

The licensee performed a search of CRs to determine if this error had previously occurred. The licensee determined that during undervoltage testing on Unit 3 for the refueling outage on September 24, 2000, IMs incorrectly connected test equipment for monitoring the emergency diesel generator frequency. The equipment had been connected to the frequency meter rather than the emergency diesel generator output voltmeter as specified by Dresden Operations Surveillance procedure DOS 6600-04, Bus Undervoltage and ECCS Integrated Functional Test For Unit 3 Diesel Generator, Revision 15. The load reject portion of the test was re-performed and this problem was documented in CR #D2000-05356. Procedural Step 1.d of DOS 6600-04 specified that the load sequence recorder be installed by positioning conductors between the recorder and the pair termination location identified on the Unit 3 Diesel Generator Chart Recorder. The recorder specified that Pen #3 which monitors diesel generator frequency be connected to the diesel generator 3 output voltmeter; however, the IMs connected Pen #3 to the frequency meter.

The licensee performed an apparent cause evaluation to determine the cause of the September 24, 2000 event. In this apparent cause evaluation, the licensee determined that the event occurred because the IMs doing the surveillance failed to follow procedures. To correct this issue, the licensee opened an action item to revise the procedure to include the following note: "Stackable test leads are required for Pen 1 and Pen 3 connections to common points." The "extent of condition" evaluation completed during the apparent cause evaluation identified three other procedures that needed to be revised with this note. The three other procedures were DOS 6600-03, 05, and 06. This action item was listed as complete in the licensees corrective action process.

During the investigation of the most recent event, the licensee discovered that despite this action item being listed as complete, DOS 6600-03 and 04 had not been revised.

Additionally, the licensee determined that a revision of DOS 6600-12 was not identified as part of its initial extent of condition. As a result, when DOS 6600-12 was performed on March 6, 2002, the IMs again incorrectly connected the test equipment.

10 CFR 50, Appendix B, Criterion XVI, requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

Contrary to the above, following the incorrect connection of a test recorder during undervoltage testing for the Unit 3 emergency diesel generator on September 24, 2000, adequate corrective action was not taken. Subsequently, on March 6, 2002, the IMs incorrectly connected a test recorder to the Unit 3 emergency diesel generator during the performance of Dresden Operations Surveillance DOS 6600-12, Endurance and Margin/Full Load Rejection/ECCS/Hot Restart, Revision 22. Additionally, as of March 6, 2002, the corrective actions to revise DOS 6600-03 and 04 following the September 24, 2000 event were not complete although an action item listed them as complete. This finding is more than minor because this issue can be viewed as a precursor to a significant event. Additionally, this finding was considered to be of very low safety significance since the incorrect connection did not have any adverse impact on the plant (Green). Because of the very low safety significance, this violation is being treated as a Non-Cited Violation (NCV 50-249/02-04-06(DRP)) consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered into the licensees corrective action program as CR#98448 and CR#D2000-05356.

.4 Corrective Action Cross-Cutting Issue a. Inspection Scope The inspectors reviewed the findings, as described above and Section 4OA7 below, to determine if an adverse pattern or trend was emerging in a cross-cutting area which may not be captured in individual findings.

b. Findings The inspectors identified four instances where the licensee failed to promptly identify and correct conditions adverse to quality. In the first instance, during the licensee followup actions for missing lock washers in auxiliary contacts for safety related motor control center cubicles, the licensee failed to generate supporting operability documentation for an additional 36 safety related components. In the second instance, the isolation condenser experienced a second water hammer after the licensee failed to initiate a condition report after a previous water hammer in August 2001. This instance was considered licensee identified and is described in Section 4OA7 of this report. In the third instance, the licensee failed to promptly identify and correct the condition of missing reactor protection system cable tray covers on Unit 2 which had been identified on September 28, 2001. Finally, following the incorrect connection of a test recorder during undervoltage testing for the Unit 3 emergency diesel generator on September 24, 2000, the licensee failed to identify the full extent of condition and complete previously identified corrective actions.

The individual findings were of very low significance; however, the findings could have had a credible impact on safety or could have been a precursor to a significant event by affecting the availability, reliability, operability or functionality of mitigating equipment.

This adverse corrective actions trend is not suitable for a Significance Determination Process evaluation. However, this trend has been reviewed by NRC management and is determined to be a substantive cross-cutting issue not captured in individual issues indicating an adverse performance trend, and is a Finding characterized as No Color (FIN 50-237/02-04-07 and 50-249/02-04-07(DRP)).

4OA3 Event Follow-up (71153)

.1 Review of Open Items a. Inspection Scope The inspectors reviewed licensee event reports (LERs) to ensure that issues documented in these reports were adequately addressed in the licensees corrective action program. The inspectors also interviewed plant personnel and reviewed operating and maintenance procedures to ensure that generic issues were captured appropriately.

The inspectors reviewed operator logs, the Updated Final Safety Analysis Report, and other documents to verify the statements contained in the Licensee Event Reports.

Also, the inspectors reviewed an unresolved item to determine if the licensee was in violation of any regulatory requirement.

b. Findings

.1 (CLOSED) LER 50-237/2001-002-00: Reactor Scram Due to Reactor Recirculation Pump Trip. This issue was documented in Inspection Report 50-237; 249/01-11. The inspectors verified that the licensee had implemented the corrective actions specified in the LER. The inspectors review of the implemented corrective actions did not identify any additional concerns. This LER is closed.

.2 (CLOSED) LER 50/237/2001-003-00: Failure of Recirculation Pump Discharge Valve to Close Causing Low Pressure Coolant Injection Inoperability.

See Section 4OA2 of this report. This LER is closed.

.3 (CLOSED) URI 50-249/02-03-01: Review of the licensees completed root cause report for the Unit 3 isolation condenser water hammer that resulted in piping support and heat exchanger pass plate damage.

See Section 4OA7 of this report. This URI is closed.

.4 (CLOSED) LER 50-249/2001-003-00: Reactor Scram due to Increasing Drywell Pressure.

On July 5, 2001, Dresden was manually scrammed due to increasing drywell pressure.

The rise in drywell pressure was caused by a loss of containment cooling when the Unit 3B reactor building closed cooling water (RBCCW) temperature control valve failed when the valve stem separated from the disc. This event was previously reviewed in NRC Special Inspection Report 50-249/01-16(DRP). Work orders have been completed on all of the RBCCW temperature control valves for installation of the correct retaining pin and verification that the stem is properly torqued to the disc. The RBCCW system operating procedure, DOP 3700-02, has been revised to state in Step F.4 that the preferred system lineup consists of two RBCCW pumps and two RBCCW heat exchangers. Additionally, the licensee is performing a single point failure vulnerability

study to identify other vulnerabilities to scrams from single point failures. This LER is closed.

.5 (CLOSED) LER 50-237/2001-005-00: Unit 2 Scram due to Increased First Stage Turbine Pressure On November 7, 2001, Unit 2 scrammed from 8 percent power during startup from a refueling outage. The high pressure turbine first stage pressure had risen during shell warming to defeat the stop valve closure scram bypass. This event was previously reviewed in NRC Inspection Report 50-237/2001-20. One finding was identified involving failure of the operators to maintain Unit 2 turbine first stage pressure within procedural limits and inadequate operation staffs management and oversight of the turbine shell warming evolution. The licensee implemented a number of corrective actions as documented in the LER. The inspector verified that Procedure DOP 5600-05, Main Turbine Startup, had been revised in Step G.3.n to have the operator establish an alarm of 100 psig for turbine first stage pressure during shell warming. The inspector also verified that Shift Manager panel monitoring expectations were incorporated into Operations Standing Order 01-07. This LER is closed.

4OA6 Exit Meetings The senior operations engineer presented the results of licensed operator requalification testing for calender year 2001 and applicability of NRC Inspection Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination process (SDP) to Mr. V. Castle and other members of licensee management and staff on January 9, 2002. The licensee acknowledged the findings presented. No proprietary information was identified.

The Senior Radiation Specialist presented the results of the special radiation protection inspection to Mr. P. Swafford and other members of licensee management and staff on March 22, 2002. The licensee acknowledged the findings presented. No proprietary information was identified.

The resident inspectors presented their inspection results to Mr. D. Bost and other members of licensee management at the conclusion of the inspection on April 3, 2002.

The licensee acknowledged the findings presented. No proprietary information was identified.

4OA7 Licensee Identified Violation The following finding of very low safety significance was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as a Non-Cited Violation (NCV).

(Closed) Unresolved Item URI 50-249/02-03-01: Review of the licensees completed root cause report for the Unit 3 isolation condenser water hammer that resulted in piping support and heat exchanger pass plate damage.

The inspectors reviewed the licensees completed root cause report for the Unit 3 isolation condenser water hammer that resulted in piping support and heat exchanger pass plate damage (the pass plate separates the heat exchanger inlet from the outlet).

The licensees root cause report was dated March 7, 2002. The licensee determined that on January 8, 2002, while performing DOS 0010-16, Unit 2(3) Isolation Condenser Safe Shutdown Valve Operability, a water hammer occurred while manually opening the isolation condenser inboard condensate return isolation valve 3-1301-3. Flashing of hot water trapped between the condensate return isolation valves 3-1301-3 and 3-1301-4 served as a pressure source to drive fluid into the two 12 inch isolation condenser horizontal steam headers. The water hammer occurred when the two fluid fronts collided in the steam headers. The effects of the water hammer were damage to one support, degradation and shearing of pass plate bolts internal to the isolation condenser, and bowing of the pass plates. The licensee identified two root causes for the event: (1)

inadequate design in that the current design does not provide instrument indication (pressure or temperature) for the volume between the condensate return isolation valves; and (2) procedures did not provide adequate instructions to assure proper pressure equalization across valve 3-1301-3 prior to opening the valve.

Laboratory analysis showed that although some of the sheared pass plate bolts contained older cracks from earlier stress events (i.e., earlier water hammer events), the final shear on all bolts was due to the recent water hammer event. Based on the laboratory results and thermal performance testing results of the last decade including testing after the repair, the licensee concluded that the isolation condenser was always able to perform its safety function until it was taken out of service on January 8, 2002.

In performing the root cause analysis for the January 2002 event, the licensee identified that the station had failed to take corrective actions to prevent recurrence following an identical water hammer event which had previously occurred on August 21, 2001, while performing DOS 0010-16. When valve 3-1301-3 was manually opened the non-licensed operator heard a very loud bang, exited the room and notified the control room. During this event, two engineers subsequently performed a walkdown of the piping and did not identify any external visual evidence of a water hammer. The Unit Supervisor then completed the surveillance and no condition report was written contrary to Exelon Procedure LS-AA-125, Revision 0, Corrective Action Program, which required that personnel originate a CR or inform a supervisor when an undesirable condition was recognized.

10 CFR 50, Appendix B, Criterion XVI, requires that measures shall be established to assure the conditions adverse to quality are promptly identified and corrected.

Exelon Procedure, LS-AA-125, Revision 0, Section 3.10.1 states that all Exelon Nuclear personnel are responsible for identifying conditions that have or could have an undesirable effect on performance of equipment in the power plant. Section 4.3.1.1 requires that Exelon Nuclear Personnel originate a CR or inform a supervisor when an undesirable condition is recognized. Additional guidance on when a CR should be originated can be found in Attachment 1. Attachment 1 requires a Significance Level 3 CR for inadequacy in procedures that caused or could have caused inoperability of equipment.

Contrary to the above, following a water hammer on the Unit 3 isolation condenser on August 21, 2001, neither the Unit 3 non-licensed operator nor the Unit Supervisor initiated a CR for a condition that had or could have had an undesirable effect on performance of equipment in the power plant. This violation is being treated as a Non-Cited Violation (NCV 50-249/02-04-08(DRP)). This issue was entered into the licensees corrective action program as CR #00089443.

KEY POINTS OF CONTACT Licensee R. Bauman, ISI Coordinator D. Bost, Station Manager K. Bowman, Operations Manager H. Bush, Radiation Protection Supervisor V. Castle, Training Operations Manager J. DeYoung, Corporate EP Specialist J. Ellis, Performance Monitoring Group Lead T. Fisk, Chemistry Manager M. Friedman, Emergency Preparedness Coordinator J. Ferguson, ALARA Analyst V. Gengler, Security Manager R. Geier, RV/ISI NDE Coordinator K. Hall, NDE Level III S. Hunsader, Corporate Maintenance Rule Owner T. Luke, Manager, Engineering R. May, NDE Level III C. Melgoza, ALARA Analyst J. Nalewajka, Acting Nuclear Oversight Manager D. Nestle, Radiation Protection L. Oshier, Radiation Protection Technical Support Supervisor M. Overstreet, Radiation Protection Shift Supervisor M. Phelan, Assistant Radiation Protection Manager R. Ruffin, Regulatory Assurance - NRC Coordinator R. Rybak, Acting Regulatory Assurance Manager N. Spooner, Site Maintenance Rule Coordinator W. Stoffels, Maintenance Manager P. Swafford, Site Vice President S. Taylor, Radiation Protection Manager D. VanAken, Corporate EP Specialist R. Whalen, System Engineering Manager NRC M. Ring, Chief, Division of Reactor Projects, Branch 1 D. Smith, Dresden Senior Resident Inspector B. Dickson, Dresden Resident Inspector IDNS R. Zuffa, Illinois Department of Nuclear Safety

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-237/249/02-04-01 URI Corrective Action Program Change not Factored into MR Program 50-249/02-04-02 URI Potential Violation of Technical Specification Limiting Condition for Operation 3.5.1 Due to Improper Alignment of High Pressure Coolant Injection 50-237/02-04-03 NCV Failure to Maintain Preventive Maintenance Procedure Adequate for Work on the Auxiliary Contact Assembly in the 480V Motor Control Center Cubicles 50-237/249/02-04-04 NCV Failure to Prepare Supporting Operability Documentation for Additional Safety Related Components with Missing Lock Washers in the Auxiliary Contact Assembly in the Motor Control Center Cubicle 50-237/249/02-04-05 NCV Inadequate Corrective Actions for Missing Reactor Protection System Cable Tray Covers 50-249/02-04-06 NCV Ineffective Corrective Actions for Test Equipment 50-237/249/02-04-07 FIN Four Instances Where the Licensee Failed to Identify and Implement Effective Corrective Actions 50-249/02-04-08 NCV Failure to Generate a Condition Report for a Water Hammer Event on the Isolation Condenser Closed 50-237/02-04-03 NCV Failure to Maintain Preventive Maintenance Procedure Adequate for Work on the Auxiliary Contact Assembly in the 480V Motor Control Center Cubicles 50-237/249/02-04-04 NCV Failure to Prepare Supporting Operability Documentation for Additional Safety Related Components with Missing Lock Washers in the Auxiliary Contact Assembly in the Motor Control Center Cubicle 50-237/249/02-04-05 NCV Inadequate Corrective Actions for Missing Reactor Protection System Cable Tray Covers 50-249/02-004-06 NCV Ineffective Corrective Actions for Test Equipment 50-237/249/02-04-07 FIN Four Instances Where the Licensee Failed to Identify and Implement Effective Corrective Actions

50-249/249/02-04-08 NCV Failure to Generate a Condition Report for a Water Hammer Event on the Isolation Condenser 50-249/2002-03-01 URI Review of the Licensees Completed Root Cause Report for the Unit 3 Isolation Condenser Water Hammer 50-237/2001-002-00 LER Reactor Scram Due to Reactor Recirculation Pump Trip 50/237/2001-003-00 LER Failure of Recirculation Pump Discharge Valve to Close Causing Low Pressure Coolant Injection Inoperability 50-237/2001-005-00 LER Unit 2 Scram due to Increased First Stage Turbine Pressure 50-249/2001-003-00 LER Reactor Scram due to Increasing Drywell Pressure

LIST OF ACRONYMS USED AC Alternating Current ALARA As Low As Is Reasonably Achievable AR Action Request ATI Action Tracking Item CFR Code of Federal Regulations CR Condition Report D3M09 Dresden Ninth Unit-3 Maintenance Outage DES Dresden Electrical Surveillance DIS Dresden Instrument Surveillance DOS Dresden Operating Surveillance DRP Division of Reactor Projects DRS Division of Reactor Safety HPCI High Pressure Coolant Injection IDNS Illinois Department of Nuclear Safety IM Instrument Mechanic LER Licensee Event Report LHRA Locked High Radiation Area LPCI Low Pressure Coolant Injection MCC Motor Control Center MRC Management Review Committee MWe Megawatts Electrical NCV Non-Cited Violation NRC Nuclear Regulatory Commission OA Other Activities Radworker Radiation Worker RBCCW Reactor Building Closed Cooling Water RP Radiation Protection RPS Reactor Protection System RWP Radiation Work Permit SBO Station Blackout Diesel SDP Significance Determination Process TEDE Total Effective Dose Equivalent UFSAR Updated Final Safety Analysis Report URI Unresolved Item VHRA Very High Radiation Area WO Work Order

LIST OF DOCUMENTS REVIEWED 1R05 Fire Protection CR 00099618 The Unit 2/3 Diesel Fire Pump Packing Is March 17, 2002 Smoking When Diesel Fire Pump Is Running CR 00099262 Inadvertent Start of 2/3 Diesel Fire Pump March 15, 2002 CR 00098793 Potential Time Delay in Moving Safe March 13, 2002 Shutdown Cart CR 00098540 Fire Extinguisher Missing from the Outside March 8, 2002 of the Unit 3 250vdc Battery Charger Room CR 00097716 Fire Drill Identifies Strengths and March 7, 2002 Weaknesses CR 00096359 NRC Identifies Lack of Access to Bus 31 February 26, 2002 Area for Firefighting 1R12 Maintenance Rule Implementation CR 00098406 Work Request/Work Order Not Being March 8, 2002 Reviewed for Maintenance Rule Functional Failure CR 00096806 Untimely Performance of Maintenance Rule February 26, 2002 (A)(1) Evaluation - Z7800-02 (Nonsafety-Related 480V AC Distribution)

CR 00099294 Areas for Improvement Identified During March 18, 2002 Unit 2/3 Emergency Diesel Generator Limiting Condition for Operations Action Request Untimely Performance of Maintenance Rule February 26, 2002

1. 96806 (A)(1) Evaluation - Z7800-02 (Nonsafety-Related 480V AC Distribution)

CR 00099096 Improperly Installed Switch Causes Delay in March 14, 2002 Unit 2/3 Emergency Diesel Generator Limiting Condition for Operations 1R13 Maintenance Risk Assessments and Emergent Work Control CR 00098987 Diesel Generator Surge Suppressor Test March 13, 2002 Exceeds 40 Milliamps

SWR #35900 Containment Cooling Service Water Pump Out-of-Service Due to Leak at Discharge Header WO 385627 Unit 2B Emergency Diesel Generator Starting Air Compressor Planned Maintenance and Belt Inspection WO 00405596-01 Unit 2/3 Emergency Diesel Generator Starting Air Compressors, Sample and Change Oil WO 99161667-01 Unit 2/3 Emergency Diesel Generator, Replace Auxiliary and Position Switches, Bus 40 Feed to 4KV Bus 23-1 WO 99178164-01 DEP 6600-10, Unit 2/3 Emergency Diesel Revision 1 Generator Surge Suppressor Test WO 9913746201 Unit 2 High Pressure Coolant Injection Cooler Preventative Maintenance Work WO 00365536 Replacement on Unit 2 Battery Charger WO 99176361-01 DM6600-02, Unit 2/3 Emergency Diesel Revision 18 Generator Mechanical Inspection and Preventive Maintenance 1R15 Operability Evaluations CR 00095959 ATI 90478-10 Was Not Created as Stated February 25, 2002 in Operability Evaluation 02-003 r1 CR 00095959 ATI 90478-10 Was Not Created as Stated February 25, 2002 in Operability Evaluation 02-003 r1 CR 00096237 Operability Determination Receives February 25, 2002 Quarterly Grade 3 CR 00099948 Found Broken Auxiliary Contact in Motor March 19, 2002 Control Center Bucket, 3-7838-4A3 CR 00097352 Non-environmentally Qualified Components Installed in Drywell Equipment and Floor Drain Solenoid Operated Valves Operability Evaluation General Electric Cr105X Auxiliary Contacts February 8, 2002

1. 02-004 on Size 1 Contactors

Operability Evaluation Reactor Building Crane and Superstructure Revision 0

1. 02-007 Engineering Change Reactor Building Crane and Superstructure

1. 335894 1R19 Post Maintenance Testing CR 00099617 3B Containment Cooling Service Water March 17, 2002 Pump Discharge Pipe Code Class Piping Leak CR 00100549 Missed Post Maintenance Test (VT-2) of March 22, 2002 Replaced 3-1105-B Standby Liquid Control Relief Valve WO 00419673-05 Repair of Pinhole Leak on the Discharge Elbow of the 3B Containment Cooling Service Water Pump WO 00412786-01 Replacement of Control Valve #2 Pressure Control Switch 1R20 Refueling and Outage CR 00099754 Historical Crack on JP Number 16 Riser March 18, 2002 Brace Leaf CR 00100225 Broke General Electric Tooling Foreign March 20, 2002 Material and Historical Foreign Material in Annulus Area CR 00100426 Foreign Material Found on the Reactor March 21, 2002 Cavity Bulkhead Floor Post Decon CR 00100027 Jet Pump 13 Reactor Vessel Side Set March 18, 2002 Screw Block Damage CR 00100315 3-1601-24 Initial Timing in the Alert Range March 21, 2002 CR 00099650 Foreign Material Found in Reactor Vessel March 17, 2002 CR 00100545 Load Limits in Fuel Pool and Old Procedure March 22, 2002 References CR 00100244 Refueling Interlocks Bypasses Without March 21, 2002 Configuration Control

1R22 Surveillance Test CR 00097479 Fluke 8060A Handheld Multimeters Found March 1, 2002 to Have Loose Test Jacks CR 00101358 Oil Sampled from Wrong Location March 27, 2002 CR 00098721 Missed Technical Specification March 12, 2002 Surveillance CR 00098448 Test Recorder for Unit 3 Emergency Diesel March 12, 2002 Generator Connected to Incorrect Meter CR 00098766 Pressure Switch 2-263-55B as Founds out March 12, 2002 of Tolerance CR 00097909 Safe Shutdown Unit 208 Failed Acceptance March 6, 2002 Criteria Quarterly Surveillance CR 00097820 Difficulty Connecting Emergency Response March 6, 2002 Data System During Quarterly Test CR 00097915 NRC Resident Concerns / Observations March 6, 2002 from Unit 3 Station Blackout Run CR 00097629 Isolation Condenser Inoperable after Valve March 5, 2002 Cycling Due to High Temperatures CR 00097265 Temperature Switch 3-0260-12 Found out March 1, 2002 of Technical Specification CR 00096779 Hi Production Risk Surveillance Tests February 27, 2002 Needlessly Being Performed CR 00099539 Source Range Monitor 23 Failed Dresden March 16, 2002 Operating Surveillance Procedure 700-12 WO 00393746 DIS 0500-01 Reactor Vessel High Revision-13 Pressure Scram Pressure Switch Calibration WO 99268154-01 DIS 0600-17, Unit 2 Narrow Range Reactor Pressure Transmitter Pt2-654 Calibration WO 397012 DOS 0500-06, Condenser Low Vacuum Pressure Switch and Functional Test CR 0009876 Pressure Switch 2-263-55B as Found out of Tolerance

CR 00098767 Pressure Switch 3-263-55D as Found out of Tolerance CR 00098721 DIS 0500-01 Changed from Monthly to March 11, 2002 Quarterly 1R23 Temporary Modifications CR 00095930 Temporary Modification Installed Without February 21, 2002 Completing Appropriate Paperwork 2OS1 Access Control to Radiologically Significant Areas RP-AA-460 Controls for High and Very High Radiation Revision 2 Areas DRS 5600-01 Surveillance Record for High, Locked High December 28, 2001 and Very High Radiation Area Boundary and Posting DFP 0800-39 Control of Material/ Equipment Hanging in Revision 11 Units 2 and 3 Spent Fuel Pools MA-AA-716-008 Foreign Material Exclusion Program Revision 0 RP-AA-210 Dosimetry Issue, Usage and Control Revision 3 RP-AA-210-1001 Dosimetry Logs and Forms Revision 0 CR 00099565 Doors Propped Open at the D3 Low March 16, 2002 Pressure Heater Bay CR 00099578 Reactor Cavity Not Posted Consistently March 16, 2002 with Drywell DRP 6200-08 Radiation Protection Guidelines for Work in Revision 05 the Reactor Cavity CR 00086393 Worker Enters High Radiation Area Without December 17, 2001 Brief 2OS2 As-Low-As-Is-Reasonably-Achievable Planning and Controls RP-AA-401 Operational ALARA Planning and Controls Revision 2 RP-AA-400 ALARA Program Revision 2 D3M09 Dose Performance Reports March 18 - 21, 2002

RWP #10001180 and D-3 Refuel Floor Forced Outage Revision 0 (RWP) and RP-AA-401, March 12, 2002 Attachment 2 (ALARA Plan)

(Associated ALARA Plan)

Refuel Floor Field Guide for D3M09 Undated RWP #10001011 Unit 3 Forced Outage Drywell Small Scope Revision 2 Activities RP-AA-401 ALARA Plans for IRM/SRM Replacements March 11, 2002 Attachment 2 and for Surveillance of 32 RPIS Probes (IRM/SRM Work) and March 20, 2002 (RPIS Probe Work)

RWP #10001385 and D3M09 Radiography Activities Revision 0 (RWP) and RP-AA-401, March 11, 2002 Attachment A (ALARA Plan)

(Associated ALARA Plan)

Radiation Protection Program Related CR January 2001 -

Trending Data February 15, 2002 CR 0097655 RP Improvement Opportunities March 4, 2002 CR 00099597 Spill from Condensate Prefilter Back Wash March 17, 2002 Causes Contamination CR 00099729 Individual Contaminated 90K Particle on March 19, 2002 Stomach CR 00099563 Low Level Facial Contamination March 18, 2002 CR 00086658 Spread of Contamination to Clean Areas December 18, 2001 CR 00097985 Hot Tool Return Area a Mess, Area Needs March 6, 2002 Attention CR 00098941 Numerous Low Level Contaminated March 13, 2002 Individuals CR 00099227 Radioactive Source Missing From Source March 14, 2002 Box 32