IR 05000456/2001013

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IR 05000456/2001-013(DRP), IR 05000457/2001-013(DRP), Exelon Generation Company, LLC, Braidwood Station, Units 1 & 2, Inspection on 11/20-12/29/2001 Re Post Maintenance Testing. One Noncited Violation Noted
ML020280507
Person / Time
Site: Braidwood  Constellation icon.png
Issue date: 01/28/2002
From: Ann Marie Stone
Division Reactor Projects III
To: Kingsley O
Exelon Generation Co, Exelon Nuclear
References
noed-2001-6-005 IR-01-013
Download: ML020280507 (27)
v • d • e


Text

ary 28, 2002

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 NRC INSPECTION REPORT 50-456/01-13(DRP); 50-457/01-13(DRP)

Dear Mr. Kingsley:

On December 29, 2001, the NRC completed an inspection at your Braidwood Station, Units 1 and 2. The enclosed report documents the inspection findings which were discussed on January 8, 2002, with Mr. J. von Suskil and other members of your staff.

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. Specifically, this inspection focused on resident and regional specialist inspection activities.

Based on the results of this inspection, one finding of very low safety significance (Green) was identified. This issue was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it was entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation, in accordance with Section V1.A.1 of the NRCs Enforcement Policy.

If you contest a 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 Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, Region III, Resident Inspector and the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001. In accordance with 10 CFR 2.790 of the NRCs 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 NRCs 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/

Ann Marie Stone, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77

Enclosure:

Inspection Report 50-456/01-13(DRP);

50-457/01-13(DRP)

REGION III==

Docket Nos: 50-456; 50-457 License Nos: NPF-72; NPF-77 Report Nos: 50-456/01-13(DRP); 50-457/01-13(DRP)

Licensee: Exelon Generation Company, LLC Facility: Braidwood Station, Units 1 and 2 Location: 35100 S. Route 53 Suite 84 Braceville, IL 60407-9617 Dates: November 20 through December 29, 2001 Inspectors: C. Phillips, Senior Resident Inspector N. Shah, Resident Inspector R. Jickling, Plant Support Inspector J. Roman, Illinois Department of Nuclear Safety Approved by: Ann Marie Stone, Chief Branch 3 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000456-01-13(DRP), 05000457-01-13(DRP); on 11/20-12/29/01, Exelon Generation Company, LLC; Braidwood Station; Units 1 & 2. Post Maintenance Testing.

This report covers a 6-week routine resident inspectors inspection and a baseline emergency preparedness inspection. The inspection was conducted by resident and specialist inspectors.

One Green finding was identified. This finding involved a Non-Cited Violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). 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. Findings for which the SDP does not apply are indicated by No Color or by the severity level of the applicable violations.

A. Inspector Identified Findings Cornerstone: Initiating Events and Mitigating Systems

  • Green. When placing shutdown cooling in service at the start of a forced outage on November 7, 2001, the 2B residual heat removal pump seized and was inoperable. The licensee determined that the failure of the 2B residual heat removal pump was due to a combination of a maintenance error which left the clearance between the pump impeller and the stuffing box extension wear ring less than that required and temperature transients when placing the RH pump in the shutdown cooling mode at a high temperature.

This finding was determined to be of very low safety significance because the B train of residual heat removal was inoperable for less than the Technical Specification allow outage time. A Non-Cited Violation of Technical Specification 5.4.1 was identified.

(Section 1R19).

B. Licensee Identified Violations Violations of very low significance which were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 40A7 of this report.

Report Details Summary of Plant Status Both units operated at full power throughout the inspection period.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems and Emergency Preparedness 1R04 Equipment Alignment (71111-04)

a. Inspection Scope The inspectors verified the alignment of the following systems while the alternate trains were out-of-service for planned maintenance:

The inspectors also verified the alignment of the Units 1 and 2 outside and auxiliary building fire protection ring headers, during scheduled maintenance on the fire protection pumps.

The inspectors performed a partial walkdown of the accessible portions of these systems and observed the system (electrical and mechanical) lineup and selected, system operating parameters (i.e., pump and bearing lube oil levels, room temperature, electrical breaker position, etc). The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), Technical Specifications, system drawings, condition reports and station procedures, as applicable. As necessary, the inspectors also interviewed licensee engineering, maintenance and operations staff.

In addition, the inspectors reviewed selected issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance.

b. Findings No findings of significance were identified.

1R05 Fire Protection (71111-05)

a. Inspection Scope The inspectors observed a fire drill on November 26, 2001. The inspectors evaluated the performance of the fire brigade in accordance with the inspection module against the criteria established in Fire Drill Scenario No. 20.11.12.01, Electrical Fire 0wx04J,

OP-AA-201-003, Fire Drill Performance, Revision 3; and Braidwood Fire Protection Report, Section 3.2, Amendment 13.

b. Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments And Emergent Work Control (71111-13)

a. Inspection Scope The inspectors reviewed the licensees assessment and management of plant risk for planned maintenance and/or surveillance activities:

The inspectors attended shift briefings and daily status meetings to verify that the licensee took actions to maintain a heightened level of awareness of the plant risk status among plant personnel. The inspectors also evaluated the availability of redundant train equipment. In particular, the inspectors observed whether licensee operating and engineering staff were aware of the licensees revised probabilistic risk assessment model which was issued on June 28, 2000. The inspectors also reviewed Nuclear Station Procedure WC-AA-103, On-Line Maintenance, Revision 3, and evaluated licensee compliance with that procedure.

In addition, the inspectors reviewed selected issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance.

b. Findings No findings of significance were identified.

1R14 Personnel Performance During Non-Routine Plant Evolutions and Events (71111-14)

a. Inspection Scope The inspectors evaluated personnel performance in response to non-routine plant evolutions and events. Specifically, the inspectors verified that licensee personnel responded in accordance with station procedures and training, that problems with the licensees response were entered into the corrective actions program, that the corrective actions were appropriate, and as applicable, that the licensee notified offsite agencies

and documented the events in Licensee Event Reports (LERs) as discussed in Section 4OA3.

b. Findings No findings of significance were identified.

1R15 Operability Evaluations (71111-15)

a. Inspection Scope The inspectors reviewed and evaluated the following operability evaluations:

The inspectors also reviewed the technical adequacy of the evaluations against the Technical Specification, UFSAR, and other design information; determined whether compensatory measures, if needed, were taken; and determined whether the evaluations were consistent with the requirements of LS-AA-105, Operability Determination Process, Revision 0.

In addition, the inspectors reviewed selected issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance.

b. Findings The licensee identified that the 10 CR 50.59 screening performed for a change to a procedure described in the UFSAR was inadequate, in that, a full evaluation was required.

As discussed in Section 1R19, on November 7, 2001, the 2B RH pump seized and became inoperable. The licensee determined that the failure resulted from (1) a maintenance error which caused the as-left clearance between the pump impeller and stuffing box extension wear ring to be less than required and (2) temperature transients when placing the RH pump in the shutdown cooling mode at a high temperature. The licensee performed an extent of condition and determined that the 1B RH pump would be susceptible to the same failure because the as-left clearance between the impeller and wear ring was also less than required.

The licensee performed an operability evaluation and determined that change to Braidwood Operating Procedure BWOP RH-6, Placing The RH System In Shutdown Cooling, was necessary to ensure operability. On November 15, 2001, the licensee revised the procedure to delay placing RH in shutdown cooling until the reactor coolant system was below 260 degrees Fahrenheit. This would minimize the temperature transient and prevent the failure. On November 26, 2001, the licensee documented in CR 00083865 that the 10 CFR 50.59 screening performed for the procedure change was inadequate. Specifically, the original screening assumed that there was no impact on the UFSAR and did not identify that UFSAR Section 5.4.7.2.7 assumed that RH was

placed into service for shutdown cooling at 350 degrees Fahrenheit for the natural circulation without letdown analysis.

The operability evaluation stated that reducing the temperature at which RH was placed on shutdown cooling would increase steaming to the atmosphere during some accident scenarios. This would result in the release of greater amounts of activity which could impact calculated offsite dose rates. The operability evaluation stated that a 50 to 80 percent margin between the calculated offsite dose rates for the impacted accident scenarios and the legal limit existed; therefore, the additional steaming caused by placing shutdown cooling on at a lower temperature was not a problem. However, 10 CFR 50.59 requires that the licensee evaluate whether or not a change to a procedure described in the UFSAR will have more than a minimal amount of impact on the consequences of an accident before the procedure change is implemented. This issue is an unresolved item (URI 50-456/457/01-13-01(DRP)) pending determination on whether the changes to BwOPRH-6 resulted in more than a minimal amount of impact on the consequences of any accident scenarios described in Chapter 15 of the UFSAR.

1R16 Operator Workarounds (71111-16)

a. Inspection Scope The inspectors accompanied a station auxiliary operator during routine rounds of the Unit 1 auxiliary building to observe whether there were any challenges or workarounds that may affect the operators ability to control the plant and respond to transients.

b. Findings No findings of significance were identified.

1R19 Post Maintenance Testing (71111-19)

a. Inspection Scope The inspectors reviewed the post-maintenance testing associated with the following components:

  • Unit 1B RH pump For each activity, the inspectors reviewed the applicable sections of the Technical Specification and UFSAR, and observed portions of the maintenance work. The inspectors also evaluated the adequacy of work controls (including Foreign Material Exclusion controls), reviewed post-maintenance test data, and conducted walkdowns to verify system restoration after the testing was completed.

In addition, the inspectors reviewed selected issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance.

b. Findings A finding of very low safety significance (Green) was self revealed when the 2B RH pump seized and became inoperable on November 7, 2001. The licensee was in the process of placing shutdown cooling in service at the start of a forced outage when the pump failed. The licensee determined that the failure of the 2B RH pump was due to (1) a failure to follow maintenance procedures for the as-left clearance between the 2B RH pump impeller and the stuffing box extension wear ring and (2) placing the RH pump in service for shutdown cooling at a high temperature. The inspectors determined that this failure to follow procedures was a Non-Cite Violation of Technical Specification 5.4.1.

As documented in Condition Report 00082906, the licensee identified that in 1996 the as-left clearance between the pump impeller and the stuffing box extension wear ring was less than required by the work package reassembly instructions due to a math error. In addition, the licensees procedure allowed operators to place RH in shutdown cooling at 360 degrees Fahrenheit which caused thermal heat up transients resulting in permanent deformation of the stuffing box extension. The reduced as-left clearance and placing the RH pump in service for shutdown cooling at a high temperature caused the seizure. The licensee also identified that Westinghouse Technical Bulletin ESBU-TB-9603-R0, RH Pump Operating Recommendations, stated that a rapid temperature increase from normal system temperature to 350 degrees Fahrenheit will cause a temporary reduction in the impeller to wear ring clearance. The bulletin also stated that this condition makes the pump vulnerable to seizure if other adversities existed. The licensee took no action after the receipt of this bulletin to address the rapid heat-up of the impeller and the stuffing box extension when initiating shutdown cooling.

The licensees corrective actions included changing the Braidwood maintenance procedure for the reassembly of the RH pump to require documentation of the math performed to calculate the clearance and to change the shutdown cooling procedure to include a pump warm up when placing shutdown cooling in service.

This finding was considered more than minor, as the failure to follow a procedure resulted in the failure of the 2B RH pump which had an actual impact on safety. The inspectors entered the significance determination process using Manual Chapter 0609, Appendix A, Significance Determination For Reactor Inspection Findings For At-Power Situations. The inspectors determined that the failure of the 2B RH pump impacted the mitigation system cornerstone only. The inspectors answered no to all five questions in the Phase I analysis under the mitigating systems cornerstone which resulted in the finding screening out as very low safety significance (Green). The inspectors also entered the significance determination process using Manual Chapter 0609, Appendix G, Shutdown Operations - Pressurized Water Reactor Hot Shutdown Operation. Since more than two heat removal paths consisting of any combination of reactor coolant system loops and RH systems existed at all times, the issued screened out as very low safety significance (Green).

Technical Specification 5.4.1, states, in part, that applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 shall be establihsed and maintained. Paragraph 9.a. of this Regulatory Guide, states, in part, that

maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures. Work package 950017181-01 included BwMP 3100-046, Residual Heat Removal Pump Disassembly, Inspection, and Reassembly, Revision 7. Procedure BwMP 3100-046, Step F.4.d.1. required the clearance between the stuffing box extension and the impeller to be greater than

.025 inches. Contrary to this, on February 1, 1996, licensee personnel failed to perform maintenance on the safety-related 2B RH pump in accordance with written procedures described in maintenance work package 950017181-01. Specifically, the clearance between the stuffing box extension and the pump impeller was left at 0.19 inches when the minimum procedurally required clearance was 0.25 inches.

However, because this violation was of very low risk significance, was non-repetitive, and was captured in the licensees corrective action program (CR 82906), it is considered a Non-Cited Violation consistent with Section VI.A of the NRC enforcement policy (NCV 50-456/457-01-13-02(DRP)).

1R23 Temporary Plant Modifications (71111-23)

a. Inspection Scope The inspectors evaluated the licensees installation of the following temporary modifications:

  • Increase the setpoint of the Unit 2 reactor head vent high temperature switches;
  • Lift the leads of Units 1 and 2 core exit thermocouples in panels 1PA51J, 1PA52J, 2PA51J and 2PA51J; and
  • Temporary leak sealant repair of Unit 1 pressurizer steam space sample line.

Specifically, the inspectors reviewed the UFSAR to determine whether the licensee adequately addressed system operability, design requirements, configuration control, risk significance, and post-installation testing.

The inspectors also reviewed selected issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance.

b. Findings No findings of significance were identified.

1EP2 Alert and Notification System (ANS) Testing (71114-02)

a. Inspection Scope The Regional Plant Support inspector discussed with Emergency Preparedness (EP)

staff the design, equipment, and periodic testing of the public ANS for the Braidwood reactor facility emergency planning zone to verify that the system was properly tested and maintained. The inspector also reviewed procedures and records for a 9 month period ending September 2001 related to ANS testing, annual preventive maintenance, and non-scheduled maintenance. The inspector reviewed the licensees criteria for

determining whether each model of siren installed in the emergency planning zone would perform as expected if fully activated. Records used to document and trend component failures for each model of installed siren were also reviewed to ensure that corrective actions were taken for test failures or system anomalies.

b. Findings No findings of significance were identified.

1EP3 Emergency Response Organization (ERO) Augmentation Testing (71114-03)

a. Inspection Scope The Regional Plant Support inspector reviewed the licensees ERO augmentation testing to verify that the licensee maintained and tested its ability to staff the ERO during an emergency in a timely manner. Specifically, the inspector reviewed semi-annual, off-hours staff augmentation drill procedures, related January 11, December 9 and December 18, 2001, drill records, primary and backup provisions for off-hours notification of the Braidwood reactor facility emergency responders, and the current ERO rosters for Braidwood. The inspector reviewed and discussed the facility EP staffs provisions for maintaining ERO call out lists.

b. Findings No findings of significance were identified.

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies (71114-05)

a. Inspection Scope The Regional Plant Support inspector reviewed the Nuclear Oversight staffs 2001 audits and field observations to ensure that these audits complied with the requirements of 10 CFR 50.54(t) and that the licensee adequately identified and corrected deficiencies. The inspector also reviewed the EP staffs self-assessments and critiques to evaluate the EP staffs efforts to identify and correct weaknesses and deficiencies.

Additionally, the inspector reviewed a sample of EP items, CRs, and ARs related to the facilitys EP program to determine whether corrective actions were acceptably completed.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES 40A1 Performance Indicator Verification (71151)

a. Inspection Scope The inspectors reviewed whether the licensee was accurately reporting data for the following performance indicators:

  • High pressure safety injection system unavailability;
  • Safety system functional failures; and

The inspectors reviewed system operating logs and licensee monthly operating reports submitted to the NRC, and interviewed licensee engineering and operations staff to determine whether the performance indicator data was being collected and reported consistent with the guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 1.

The Regional Plant Support inspector verified that the licensee had accurately reported these indicators: ANS, ERO Drill Participation, and Drill and Exercise Performance, for the EP cornerstone. Specifically, the inspector reviewed the licensees PI records, data reported to the NRC, and condition reports for the period January 2001 through September 2001 to identify any occurrences that were not identified by the licensee.

Records of relevant control room simulator training sessions, periodic ANS tests, and excerpts of drill and exercise scenario and evaluations were also reviewed.

b. Findings No findings of significance were identified.

4OA3 Event Follow-up (71153)

The inspectors reviewed licensee event reports and other items using Inspection Procedure 71153.

(Closed) LER 50-456/2001-001-00: Three main steam safety valves exceeded the Technical Specification limit by greater than 3%. During testing from September 19-20, 2001, the licensee identified that 3 of the 20 steam generator safety relief valves had lifted 3-4% above their required setpoint. Two of the valves had apparently failed due to oxide bonding between the nozzle and disk seating surfaces, while the third apparently failed due to setpoint drift. Additionally, the licensee determined that the test data was within the acceptance criteria for the applicable accident scenarios discussed in the UFSAR. Subsequently, the licensee adjusted and successfully retested the affected relief valves. The inspectors determined that the licensees evaluation and testing were acceptable and that the remaining safety relief valves were capable of performing their required safety function. Consequently, this item was considered of very low safety significance and no findings were identified.

(Closed) LER 50-457/2001-002-00: Main steam isolation valves (MSIVs) not stroke tested as required. This issue is discussed in Section 4OA5 of this report, under URI 50-456/457/01-10-03.

4OA5 Other (Closed) URI 50-456/457/01-10-03(DRP): Failure to perform required testing of the Units 1 and 2 MSIVs. On September 26, 2001, the licensee identified that both units MSIVs were not tested in Mode 3 as required by Technical Specifications. Because Unit 1 was already shutdown for a planned refueling outage, the licensee requested a Notice of Enforcement Discretion (NOED No. 01-6-005) for Unit 2. The NRC approved this NOED on September 27, 2001. The licensee determined the Mode 3 testing requirement was specifically stated in the Improved Technical Specifications, which was implemented in February 19, 1999. Prior to this date, the licensees Technical Specification did not explicitly require that the testing be performed in Mode 3; with testing typically occurring in Modes 4 or 5. The inspectors determined that the root cause was an administrative oversight during the change process to the Improved Technical Specifications. Subsequently the inspectors observed that the Units 1 and 2 MSIVs were successfully tested in Mode 3 on October 6 and November 16, 2001, respectively. The failure to perform the testing in Mode 3 as required in TS 3.7.2.1 constituted a violation of minor significance that is not subject to enforcement actions in accordance with Section IV of the NRCs Enforcement Policy. This violation was captured in the licensees corrective action program (CR 76608).

4OA6 Meetings

.1 Exit Meeting The inspectors presented the inspection results to Mr. J. von Suskil and other members of licensee management at the conclusion of the inspection on January 8, 2002. The licensee acknowledged the findings presented. No proprietary information was identified.

.2 Interim Exit Meeting The inspectors presented the results of the Emergency Preparedness program and performance indicators inspection, to Mr. K. Schwartz and other members of licensee management at the conclusion of the inspection on December 21, 2001. The licensee acknowledged the findings presented. No proprietary information was identified.

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

NCV Tracking Number Requirement Licensee Failed to Meet NCV 50-456/457-01-13-03 Technical Specification 5.4.1 requires, in part, that written procedures be established, implemented, and maintained covering those activities listed in Regulatory Guide 1.33, Appendix A, Revision 2, February 1978. Section 8 of Appendix A to this Regulatory Guide requires, in part, that procedures of a type appropriate to the circumstances be provided to ensure that measuring and testing devices are properly controlled, calibrated and adjusted. On November 28, 2001, the licensee identified that measurement and test equipment were not being properly recorded as required by Step 4.7.3 of station procedure MA-AA-716, Control of Portable Measurement and Test Equipment Program, Revision 1. Reference condition report No. 84419. This is being treated as a Non-Cited Violation.

KEY POINTS OF CONTACT Licensee K. Aleshire, Emergency Preparedness Coordinator J. Bailey, Regulatory Assurance - NRC Coordinator G. Baker, Security Manager S. Butler, Corrective Action Program Coordinator G. Dudek, Operations Manager C. Dunn, Engineering Director A. Ferko, Regulatory Assurance Manager M. Finney, Radiation Protection Engineering Supervisor R. Graham, Work Management Director L. Guthrie, Maintenance Director F. Lentine, Design Engineering Manager K. Schwartz, Plant Manager J. von Suskil, Site Vice President Nuclear Regulatory Commission M. Chawla, Project Manager, NRR A. Stone, Chief, Reactor Projects Branch 3 LIST OF ITEMS OPENED AND CLOSED Opened 50-456/457/01-13-01 URI Potential inadequate 10 CFR 50.59 evaluation for 1B RH pump 50-456/457/01-13-02 NCV Failure to perform RH pump maintenance in accordance with procedure 50-456/457/01-13-03 NCV Failure to properly record M&TE Closed 50-456/457/01-13-02 NCV failure to perform RH pump maintenance 50-456/457/01-13-03 NCV failure to properly record M&TE 50-456/2001-001-00 LER MSSVs exceeded the Technical Specification limit 50-457/2001-002-00 LER MSIVs not stroke tested in Mode 3 as required 50-456/457/01-10-03 URI Failure to perform required testing of the MSIVs

LIST OF ACRONYMS AND INITIALISMS USED ADAMS Agencywide Documents Access and Management System ANS Alert and Notification System AR Action Request BwAR Braidwood Annunciator Response Procedure BwMP Braidwood Maintenance Procedure BwOP Braidwood Operating Procedure BwOSR Braidwood Operability Surveillance Requirement BwVS Braidwood Engineering Surveillance CC Component Cooling Water CFR Code of Federal Regulations CR Condition Report DG Diesel Generator EP Emergency Preparedness ERO Emergency Response Organization LER Licensee Event Report M&TE Measurement and Test Equipment MSIV Main Steam Isolation Valve NOED Notice of Enforcement Discretion NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulations OOS Out-of-Service PARS Publicly Available Records PI Performance Indicator RH Residual Heat Removal ROP Reactor Oversight Process SDP Significant Determination Process SI Safety Injection UFSAR Updated Final Safety Analysis Report VIO Violation

LIST OF DOCUMENTS REVIEWED 1R04 Equipment Alignment BwOP DG-1 Diesel Generator Alignment to Standby Revision 13 Condition Reference Use BwOP DG-E2 Electrical Lineup -1B Diesel Generator Revision 2E4 Reference Use BwOP DG-M2 Operating Mechanical Lineup 1B D/G Revision 9 Reference Use BwOP FP-M6 Operating Mechanical Lineup Unit 0 Aux Revision 2 Bldg Ring Hdr Operating BwOP FP-M1 Operating Mechanical Lineup Unit 0 Lsh and Revision 6 Outside Ring Hdr Operating M-97 Diagram of Diesel Generator Room 1A & 1B April 12, 1996 Ventilation System Unit 1 AR 00083325 Maintenance Configuration Control During November 18, 2001 FP Work (PI&R)

CR A2001-00815 Unintentional Rotation of U1 C/D Traveling March 19, 2001 Screens During Maintenance (PI&R)

CR 00075347 2CV8519 Found Open (PI&R) October 19, 2001 CR 00075650 Lowering Level Trend in Unit 1 RWST (PI&R) September 17, 2001 CR 00077043 OOS Air Isolation Valve Found Open on RTS October 12, 2001 (PI&R)

CR 00077672 Paperwork for TMOD EC 332832 Was Not December 5, 2001 Completed Properly (PI&R)

1R13 Maintenance Risk Assessments And Emergency Work Control BwVSR SI-1 ECCS Injection Line Depressurization with Revision 0 Optional Leakage Test of SI9849A/B/C/D and SI8956A/B/C/D WO 00384177 ECCS Inj Line Depress w/Optional Lk November 28, 2001 1R14 Personnel Performance During Nonroutine Plant Evolutions And Events LER 457/01-002-00 MSIVs Not Stroke Timed in Mode 3 as November 26, 2001 Required

CR 00076608 MSIVs Not Stroke Timed in Mode 3 as September 26, 2001 Required WO 00366963 MSIV Full Stroke Test 1BwOSR 3.7.2.1 October 6, 2001 WO 99162844 01 MSIV Full Stroke Test 1BwOSR 3.7.2.1 September 22, 2001 WO 99228191 01 MSIV Full Stroke Test Unit 2 November 16, 2001 1R15 Operability Evaluations AR 00083865 Inadequate 50.59 Screening for Placing RH November 13, 2001 In Service BRW-S-2001-614 50.59 Screening Form BwOP RH-6 Revision 0 ATI 00081944 2B Residual Heat Removal Pump Failure November 15, 2001 Due to Contact Between Pump Impeller and Stuffing Box Extension Upper Wear Ring ESBU-TB-96-3-RO Westinghouse Technical Bulletin, RH Pump June 20, 1996 Operating Recommendations BwOP RH-6 Placing the RH System in Shutdown Cooling Revision 24 1R19 Post Maintenance Testing SPP-01-016 Residual Heat Removal Pump 2RH01PB Revision 0 Post Maintenance Testing WO 00373399 ASME Srv Rqmts for Residual Heat Removal November 15, 2001 Pump 4561018800400 IEB 88-04: Potential Safety-Related Pump May 5, 1988 Loss ITR 01-081 Revise Technical Requirements Manual November 13, 2001 (TRM) Miscellaneous Test Requirement TSR 2.5.c.4 to Provide Additional Flexibility in Satisfying the Surveillance Requirement While Meeting the Original Intent CR 00081944 2B RH Pump Tripped on Phase C November 7, 2001 Overcurrent CR 00082906 Math Error in Work Package Resulting in November 10, 2001 Below Spec Clearance CF A2001-00037 M&TE Not Properly Logged in EWCS (PI&R) January 5, 2001

AR 00084419 M&TE Used in WOs not Properly November 28, 2001 Documented in Passport (PI&R)

MA-AA-716-040 Control of Portable Measurement and Test Revision 1 Equipment Program 1R23 Temporary Plant Modifications CC-AA-404 Maintenance Specification: Application Revision 2 Selection, Evaluation and Control of Leak Sealant Injection and Temporary Leak Repair NES-MS-03.3 Injected Leak Sealant Application Revision 0 Part 9900 Technical On-Line Leak Sealing Guidelines for ASME No Date Guidance Code Class 1 and 2 Components WO 00369393-01 Install Furmanite Clamp on 1PS01BA - 3/8 October 26, 2001 WO 00369393-03 Fittings Leaking on 1PS01BA - 3/8, Install October 26, 2001 TMod (ECC333613)

WO 00382512-01 Incorporate Setpoint Scaling Temporary November 21, 2001 Modification No. 333958 Revision 1 WO 00382513 Incorporate Setpoint Scaling Temporary November 21, 2001 Modification No. 333958 Revision 1 TMod DCP 9900659 Lift Leads of Core Exit Thermocouples in Revision 0 Panels 2PA51J, 2PA52J TMod DCP EC42446 Lift Leads of Core Exit Thermocouples in Revision 1 Panels 2PA51J, 2PA52J TMod DCP Revise Setpoints for Reactor Head Vent Revision 0 EC333958 Temperature Switches TSH-RC017 and 2TSH-RC018 CC-AA-112 Temporary Configuration Changes Revision 4 LS-AA-106 Plant Operations Review Committee Revision 0 AR 00083400 Reactor Head Vent High Temperature Alarm November 19, 2001 EC 0000333958-001 Revise Setpoint of Reactor Head Vent Temp November 20, 2001 Switches 2TSH-RC017 and 2TSH-RC018 BwAR 2-14-E4 RX Head Vent Temp High Revision 7 71151 Performance Indicator Verification

RS-AA-122-14 Performance Indicator - Unplanned Power Revision 2 Changes per 7000 Critical Hours RS-AA-122-103 Performance Indicator - Safety System Revision 2 Functional Failures RS-AA-122-104 Performance Indicator - Safety system Unavailability (HPSI/HPCI, RHR, AFW/RCIC, Revision 3 EDG)

CS-AA-2080 Monthly Performance Indicator Data June 25, 2001 Elements for Safety System Functional Failures LS-AA-2030 Monthly Performance Indicator Data June 25, 2001 Elements for Unplanned Power Changes per 7000 Critical Hours LS-AA-2050 Monthly Performance Indicator Data June 25, 2001 Elements for Safety System Unavailability-High Pressure Injection (BWR) or High Pressure Safety Injection (PWR)

Exelon Nuclear Performance Summary: Revision 1 Braidwood P.2: Power History Curves AR 00088344 Unresolved Issues Resulted After NRC December 28, 2001 Resident Rev. of HPSI Sys RS-AA-122-109 Performance Indicator - ERO Drill January-July 2001 Participation RS-AA-122-110 Performance Indicator -ANS Reliability January-July 2001 Monthly LS-AA-2110 Monthly Performance Indicator Data June 2001 Elements for ERO Drill Participation LS-AA-2130 Monthly Performance Indicator Data August - September, Elements for ANS Reliability 2001 EP-AA-120-1001 NRC DEP PI Data Summary January - September 2001 Drill/ Exercise Nuclear Accident Reporting January -

System Forms September, 2001 Braidwood 2001 EP Drill/Exercise Schedule AR 74592 Improper Implementation of EP Drill/Exercise July 1, 2001 PI

1EP2 Alert and Notification System (ANS) Testing Braidwood Off-Site Siren Test Plan Revision 2 Siren Daily Operability Data Sheets Siren Monthly Operability Reports 2001 Exelon Semi-Annual Siren Report January 1 - June 30, 2001 CR 00084351 Siren Monthly Reporting Data November 29, 2001 1EP3 Emergency Response Organization (ERO) Augmentation Testing Section B Exelon Nuc. Standardized Rad. Emergency Revision 11 Plan Section E Exelon Nuc. Standardized Rad. Emergency Revision 11 Plan Dialogic Comm. Corp. Hot Site Agreement January 19, 2001 EP-AA-112-100 Att. 2 - ERO Augmentation Revision 2 EP-AA-112-1001 Att. 2 - Conduct Of Augmentation Drills Revision 0 Braidwood ERO Quarterly Roster December 13, 2001 Questions and Answers Regarding ERO May 10, 2001 Call-Outs and the Computerized DCC Communicator Sys.

Report On ERO Off-Hours Augmentation January 10, 2001 Drill -

Report On ERO Off-Hours Augmentation January 11, 2001 Drill -

Report On ERO Off-Hours Augmentation December 9, 2001 Drill -

Report On ERO Off-Hours Augmentation December 18, 2001 Drill -

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies Public Notice of October 23, 2001, Off Site Agency Meeting

Generic Corporate Performance Areas - January - June, 2001 Emergency Preparedness NOA-BW-01-1Q Nuc. Oversight Continuous Assessment January - March, Report Braidwood Gen. Station 2001 Braidwood Overall Program Self Evaluation January - June, 2001 EP.1 and EP.2 CR A2000-01013 Insufficient Oversite of Respiratory Prog. in March 8, 2000 IMD CR A2000-01257 Critical Path Workers Denied Respirators March 19, 2000 Due to Process Failure CR A2001-00094 Common Cause Identified For Off Hours Call January 11, 2001 Out Failures CR A2001-00445 ARs/HSKeeping Requests Not Completed February 8, 2001 For Greater Than Six Months CR A2001-00478 N. O. Identified EP Is Not Initiating ARs For December 17, 2000 GSEP Revisions CR A2001-00491 ERDS System Problems During Quarterly February 15, 2001 Test CR A2001-00870 Loss of ENS Phone Line in the Control Room March 23, 2001 CR A2001-00479 NOS Continuous Assessment Program Does March 23, 2001 Not Meet the Requirements of 10 CFR 50.54(t)

CR A2001-01201 Injury While Working In Crosstown Area April 24, 2001 CR A2001-01479 ERDS Quarterly Test Unable to Transmit May 17, 2001 Data to NRC Via the U2 Modem CR A2001-01662 Lack of Comms. Ability In GSEP Assembly June 6, 2001 Area CR A2001-01926 GSEP Van Failed During Integrated Drill June 27, 2001 CR A2001-02080 Shutdown of Multiple TSC Computer Room July 16, 2001 Servers CR A2001-02185 NRC Question of Dedicated Phone Line in July 26, 2001 Control Room CR A2001-02289 Monthly Comm. Drill EOF GSEP Radio Not August 6, 2001 Able to Communicate

CR A2001-02300 EP Equipment Problems Observed During August 8, 2001 Annual Exercise CR A2001-02306 Demo. Criteria Evaluated As Unsatisfactory August 8, 2001 During Annual EP Drill CR 00076166 Safety Issue: P.A. System Cannot Be Heard September 4, 2001 In The CSRs CR 00028481 MMD Personnel Not Mask Fit Qualified November 10, 2001 AR 00035774 Braidwood Plant Support 4Q 2000 Observations AR 00036196 Plant Support FOs For NOA BW-00-4Q Assessment AR 36187 AR 00041561 Braidwood Plant Support 1Q 2001 Observations AR 00076875 Plant Support FOs For NOA-BW-01-4Q Assessment Ar 76870 AR 00082115 EP Improvement Items Ident. From October EP Activities 21

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