IR 05000445/2002008

IR 05000445-02-008, IR 05000446-02-008, on 10/18 - 12/06/2002, Comanche Peak Units 1 & 2

ML030100394
Person / Time
Site:	Comanche Peak
Issue date:	01/10/2003
From:	Marschall C NRC/RGN-IV/DRP/RPB-C
To:	Terry C TXU Energy
References
IR-02-008
Download: ML030100394 (23)

Text

ary 10, 2003

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 and 2 - INSPECTION REPORT 50-445/02-08; 50-446/02-08

Dear Mr. Terry:

On December 6, 2002, the NRC completed an inspection at your Comanche Peak Steam Electric Station, Units 1 and 2. The enclosed report documents the inspection findings, which were discussed on December 6, 2002 with Mr. Lance Terry, Senior Vice President and Principal Nuclear Officer, and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.

Based on the results of this inspection, the NRC has identified one violation of regulatory requirements that was evaluated under the risk significance determination process using the Significance Determination Process described in NRC Inspection Manual Chapter 0609. The NRC concluded from this process that the issue has very low safety significance (Green) and no immediate safety impact. Because of the very low safety significance and because the licensee took immediate and effective action to correct the problem, the violation is being treated as a noncited violation, consistent with Section VI.A.1 of the Enforcement Policy. If you deny the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Comanche Peak Steam Electric Station.

TXU Energy -2-In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response will be made 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/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety Dockets: 50-445; 50-446 Licenses: NPF-87; NPF-89

Enclosure:

NRC Inspection Report 50-445/02-03; 50-446/02-03

REGION IV==

Dockets: 50-445; 50-446 Licenses: NPF-87; NPF-89 Report No.: 50-445/02-03; 50-446/02-03 Licensee: TXU Energy Facility: Comanche Peak Steam Electric Station, Units 1 and 2 Location: FM-56 Glen Rose, Texas Dates: October 18 through December 6, 2002 Team Leader W. McNeill, Senior Reactor Inspector Engineering and Maintenance Branch Inspectors: L. Ellershaw, Senior Reactor Inspector Engineering and Maintenance Branch P. Goldberg, Senior Reactor Inspector Engineering and Maintenance Branch J. Mateychick, Reactor Inspector Engineering and Maintenance Branch J. Melfi, Reactor Inspector Engineering and Maintenance Branch Accompanying J. Leivo, Contractor Beckman and Associates Personnel:

Approved By: Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety

-2-SUMMARY OF FINDINGS IR 05000445-02-08, IR 05000446-02-08; TXU Energy; on 11/18/2002-12/06/2002; Comanche Peak Steam Electric Station; Units 1 and 2. Safety system design and performance capability.

The NRC conducted an inspection with a team of five regional inspectors, and one contractor.

The inspection identified one Green noncited violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 00609 "Significance Determination Process". Findings for which the SDP does not apply are indicated by "No Color" or by the severity level of the applicable violation. The NRC described its program for overseeing the safe operation of commercial nuclear power reactors in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Mitigating Systems The inspectors identified one finding, which was a violation of NRC regulatory requirements. The inspectors found that the licensee had failed to fully and routinely test the control circuits for the residual heat removal system crosstie valves (two per unit), which are opened from the control room to provide suction to the charging and safety injection pumps during intermediate pressure cold leg recirculation following a loss-of-coolant accident. During the inspection, to address the inspectors concerns, the licensee performed special tests, which revealed that a limit switch for one interlock for a Unit 1 valve failed to close as required, and wiring connections for another interlock on a Unit 2 valve were loose. The licensee determined that the remaining parts of the degraded interlock circuits were intact, and concluded that these as-found conditions would not have prevented the operator from opening the valves for the recirculation mode. Despite the problems encountered, the system and its trains would have performed their safety function with the proper valve line up.

The inspectors concluded that failure to routinely test these circuits and detect these failures was a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Criterion XI requires a licensee establish a test program to assure identification and performance of all testing required to demonstrate that systems and components will perform satisfactorily in service. The inspectors considered the finding greater than minor because the lack of testing affected the reliability of a mitigating system. The inspectors considered the risk significance to be green because there was not an actual loss of a train of risk significant equipment. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy (50-445;446/0208-01). This violation is in the licensees corrective action program as SmartForms 2002-004158, 2002-004227, and 2002-004228 (Section 1R21.6.b).

Report Details 1. REACTOR SAFETY Introduction The NRC performed an inspection to verify that the licensee adequately preserved the facility safety system design and performance capability and that the licensee preserved the initial design in subsequent modifications of the systems selected for review. The scope of the review also included any necessary nonsafety-related structures, systems, and components that provided functions to support safety functions. The inspection effort also reviewed the licensees programs and methods for monitoring the capability of the selected systems to perform the current design basis functions. This inspection verified aspects of the initiating events, mitigating systems, and barrier cornerstones.

The licensee based the probabilistic risk assessment model for the Comanche Peak Steam Electric Station on the capability of the as-built safety systems to perform its intended safety functions successfully. The inspectors determined the area and scope of the inspection by reviewing the licensees probabilistic risk analysis models to identify the most risk significant systems, structures, and components according to their ranking and potential contribution to dominant accident sequences and/or initiators. The inspectors also used a deterministic effort in the selection process by considering recent inspection history, recent problem area history, and all modifications developed and implemented.

The inspectors reviewed in detail the safety injection and the Class 1E 480Vac systems.

The primary review prompted parallel review and examination of support systems, such as, electrical power, instrumentation, and related structures and components.

The inspectors assessed the adequacy of calculations, analyses, engineering processes, and engineering and operating practices that used by the licensee to support the performance of the safety systems selected for review and the necessary support systems during normal, abnormal, and accident conditions. Acceptance criteria utilized by the NRC inspection team included NRC regulations, the technical specifications, applicable sections of the Final Safety Analysis Report, applicable industry codes and standards, as well as, industry initiatives implemented by the licensees programs.

1R02 Evaluations of Changes, Tests, or Experiments (71111.02)

a. Inspection Scope The inspectors reviewed five licensee-performed 10 CFR 50.59 evaluations to verify that the licensee had appropriately considered the conditions under which the licensee may make changes to the facility or procedures or conduct tests or experiments without prior NRC approval.

The inspectors reviewed an additional 14 licensee-performed 10 CFR 50.59 screenings, in which the licensee determined that evaluations were not required to ensure that the licensees exclusion of a full evaluation was consistent with the requirements of 10 CFR 50.59.

-2-The inspectors reviewed and evaluated the most recent licensee 10 CFR 50.59 program audit to determine whether the licensee conducted sufficient in-depth analyses of their program to allow for the identification and subsequent resolution of problems or deficiencies.

b. Findings No findings of significance were identified.

1R21 Safety System Design and Performance Capability (71111.21)

.1 System Requirements a. Inspection Scope The inspectors inspected the following attributes of the safety injection and Class 1E 480Vac systems: (1) process medium (water, steam, and air), (2) energy sources, (3)

control systems, and (4) equipment protection. The inspectors examined the procedural instructions to verify instructions were consistent with actions required to meet, prevent, and/or mitigate design basis accidents. The inspectors also considered requirements and commitments identified in the Final Safety Analysis Report, technical specifications, design basis documents, and plant drawings.

b. Findings No findings of significance were identified.

.2 System Condition and Capability a. Inspection Scope The inspectors reviewed the periodic testing procedures for the safety injection and Class 1E 480Vac systems to verify that the licensee adequately designed the systems.

The inspectors also reviewed the systems operations by conducting system walkdowns; reviewing normal, abnormal, and emergency operating procedures; and reviewing the Final Safety Analysis Report, technical specifications, design calculations, drawings, and procedures.

b. Findings No findings of significance were identified.

-3-

.3 Identification and Resolution of Problems a. Inspection Scope The inspectors examined a sample of problems identified by the licensee in the corrective action program to evaluate the effectiveness of corrective actions related to design issues. The sample included open and closed condition reports for the past three years that identified issues affecting the selected systems. Older condition reports that were identified while performing other areas of the inspection were also reviewed.

b. Findings No findings of significance were identified.

.4 System Walkdowns a. Inspection Scope The inspectors performed walkdowns of the accessible portions of the safety injection, Class 1E 480Vac systems, and required support systems. The inspectors focused on the installation and configuration of switchgear, motor control centers, manual transfer switches, field cabling, raceways, piping, components, and instruments. During the walkdowns, the inspectors assessed:

• The placement of protective barriers and systems,

• The susceptibility to flooding, fire, or environmental conditions,

• The physical separation of trains and the provisions for seismic concerns,

• Accessibility and lighting for any required local operator action,

• The material condition and preservation of systems and equipment, and

• The conformance of the currently-installed system configurations to the design and licensing bases.

b. Findings No findings of significance were identified.

.5 Design Review a. Inspection Scope The inspectors reviewed the current as-built instrument and control, electrical, and mechanical design of the safety injection and Class 1E 480Vac systems. These reviews

-4-included an examination of design assumptions, calculations, required system thermal-hydraulic performance, electrical power system performance, protective relaying, control logic, and instrument setpoints and uncertainties. The inspectors also performed selected single-failure evaluations of individual components and circuits to determine the effects of such failures on the capability of the systems to perform their design safety functions.

The inspectors inspected calculations, drawings, specifications, vendor documents, Final Safety Analysis Report, technical specifications, emergency operating procedures, and temporary and permanent modifications.

b. Findings No findings of significance were identified.

.6 Safety System Inspection and Testing a. Inspection Scope The inspectors reviewed the program and procedures for testing and inspecting selected components in the safety injection and Class 1E 480Vac systems. The review included the results of surveillance tests required by the technical specifications and selective review of Class 1E control circuits for testability.

b. Findings (1) Introduction The inspectors identified a finding concerning the licensees failure to routinely test certain interlocks, which, if failed, could prevent the control room operators from manually opening the residual heat removal crosstie isolation valves.

(2) Description Design Basis Document DBD-ME-261, "Safety Injection System," Section 5.2.3, "Cold Leg Recirculation" describes the design basis for operation of the residual heat removal crosstie isolation valves and their interlocks. Plant operators must manually align the valves for the recirulation mode. The design of the interlocks prevents manual opening of the residual heat removal crosstie valves. This assures no open flow path from the reactor coolant system to the reactor water storage tank via the charging or safety injection pump minimum flow lines during the switchover from injection mode to cold leg recirculation mode. It also assures suction from containment sump during the switchover.

-5-The inspectors asked the engineering staff to provide the procedures used to test the interlocks. In response, the engineering staff identified that the only testing of these interlocks occurred during preoperational testing about 10 years ago. The engineering staff also identified that subsequent ASME inservice testing for stroke times exercised the circuits, but these tests did not verify proper operation of the interlocks. The inspectors determined that parallel circuits could mask test results without special provisions, such as lifting leads and/or repositioning valves associated with the interlocks.

To address the inspectors concern about the absence of testing, the engineering staff initiated SmartForm-2002-004158 and performed tests of the interlocks for the residual heat removal crosstie valves. These tests revealed unsatisfactory conditions. First, a limit switch for one safety injection minimum flow interlock for Unit 1 failed to close as required. Second, before the test, the test technicians found a loose wiring connection on a Unit 2 residual heat removal crosstie valve. The failed limit switch for the safety injection pump minimum flow valve would not have prevented opening of the residual heat removal crosstie valves. The inspectors considered the identification of the failure of the limit switch to be a test failure, but did not consider the loose wire a test failure because it did not affect component function. However, both problems demonstrated the need for testing.

(3) Analysis The inspectors determined that the significance of this finding affecting the reactor safety cornerstone was GREEN. The inspectors considered the finding greater than minor because of the potential for an undetected failure leading to a system failure. The inspectors considered this violation to be Green because there was not an actual loss of a train of risk significant equipment.

(4) Enforcement The inspectors concluded that failure to routinely test these circuits and to detect failures was a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Criterion XI requires a licensee to establish a test program to assure identification and performance of all testing required to demonstrate that systems and components will perform satisfactorily in service. Without comprehensive and routine testing of these interlocks, the inspectors concluded that over time additional undetected failures could occur and might prevent remotely opening of the valves. The failures could prevent alignment of the charging pumps and safety injection pumps for cold leg recirculation. The licensee entered this violation in its corrective action program as SmartForms 2002-004158, 2002-004227 and 2002-004228. The inspectors treated this violation as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy (50-445;446/0208-01).

-6-4. OTHER ACTIVITIES (OA)

4OA6 Management Meetings Exit Meeting Summary The team leader presented the inspection results to Mr. Lance Terry, Senior Vice President and Principal Nuclear Officer, and other members of licensee management at the conclusion of the onsite inspection on December 6, 2002.

At the conclusion of this meeting, the team leader asked the licensees management whether any materials examined during the inspection should be considered proprietary.

No proprietary information was identified.

ENCLOSURE KEY POINTS OF CONTACT Licensee M. Blevins, Deputy to the Senior Vice President S. Ellis, Operations Manager R. Flores, Deputy to the Vice President for Engineering T. Hope, Regulatory Performance Manager J. Kelley, Vice President, Nuclear Engineering and Support S. Lakdawala, Engineering Programs Manager D. Moore, Plant Manager D. Reimer, Technical Support Manager L. Terry, Senior Vice President and Principal Nuclear Officer R. Walker, Regulatory Affairs Manager NRC D. Allen, Senior Resident Inspector A. Sanchez, Resident Inspector ITEMS OPENED AND CLOSED Opened and Closed 50-445/0208-01 NCV Failure to test interlock circuits for residual heat removal system cross-tie valves (Section 1R21.6).

LIST OF DOCUMENTS REVIEWED The following documents were selected and reviewed by the team to accomplish the objectives and scope of the inspection.

CALCULATIONS Number Title Revision CN-FSE-99-53 Comanche Peak ECCS Testing Criteria - Expanded 0 Kmin/Kmax Criteria Band for SI Pumps EE-MCC- MCC and Distribution Panel Methodology 5 METHODOLOGY EE-1E-1EB1 480 Vac Switchgear CP1-EPSWEB-01 (1EB1) Bus Based 0 Calculation

-2-Number Title Revision EE-1E-1EB1-1 480 Vac Motor Control Center CP1-EPMCEB-01 (1EB1-1) 1 Bus Based Calculation EE-1E-1EB2 480 Vac Switchgear CP1-EPSWEB-02 (1EB2) Bus Based 0 Calculation EE-1E-1EB2-1 480 Vac Motor Control Center CP1-EPMCEB-02 (1EB2-1) 1 Bus Based Calculation EE-1E-1EB3 480 Vac Switchgear CP1-EPSWEB-03 (1EB3) Bus Based 0 Calculation EE-1E-1EB3-1 480 VAC Motor Control Center CP1-EPMCEB-03 (1EB3- 1 1) Bus Based Calculation EE-1E-1EB3-1 480 Vac Motor Control Center CP1-EPMCEB-03 (1EB3-1) 1 Bus Based Calculation EE-1E-1EB3-2 480 Vac Motor Control Center CP1-EPMCEB-05 (1EB3-2) 1 Bus Based Calculation EE-1E-1EB3-3 480 Vac Motor Control Center CP1-EPMCEB-07 (1EB3- 1 3) Bus Based Calculation EE-1E-1EB3-4 480 Vac Motor Control Center CP1-EPMCEB-09 (1EB3-4) 1 Bus Based Calculation EE-1E-1EB4 480 Vac Switchgear CP1-EPSWEB-04 (1EB4) Bus Based 0 Calculation EE-1E-1EB4-1 480 Vac Motor Control Center CP1-EPMCEB-04 (1EB4-1) 1 Bus Based Calculation EE-1E-1EB4-2 480 Vac Motor Control Center CP1-EPMCEB-06 (1EB4-2) 1 Bus Based Calculation EE-1E-1EB4-4 480 VAC Motor Control Center CP1-EPMCEB-10 (1EB4- 1 4) Bus Based Calculation EE-1E-1ED1 125 Vdc Switchboard CP1-EPSWED-01 (1ED1-1) Bus 0 Based Calculation EE-1E-1ED1-2 125 Vdc Distribution Panel CP1-ECDPED-03 (1ED1-2) 1 Bus Based Calculation EE-1E-XEB1-1 480 Vac Motor Control Center CPX-EPMCEB-07 (XEB1- 0 1) Bus Based Calculation

-3-Number Title Revision EE-1E-XEB1-2 480 Vac Motor Control Center CPX-EPMCEB-01 (XEB1- 0 2) Bus Based Calculation EE-1E-XEB1-3 480 Vac Motor Control Center CPX-EPMCEB-03 (XEB3- 0 2) Bus Based Calculation EE-CA-0008-169 Coordination Study - 480 V Class 1E Unitized MCC 3 Buses EE-CA-0008-871 Protective Relay Settings for Safeguard Buses 5 Over/Under Voltage Relays and Associated Time Delay Relays EE-SC-U1-1E Unit 1 Class 1E System Short Circuit 0 EE-VP-U1-1E Unit 1 Class 1E System Voltage Profile 0 IC-CA-0232-5158 Instrument Uncertainties and Indicator Loop Accuracy for 0 RWST Level Instrumentation Loops L-0930 through L-0933 GENX-185 Aircraft Cable Seismic Restraint 1 FSE/SS-TBX-1634 TBX/TCX Revised ECCS Data/Test Criteria FSE/SS-TBX-471 Develop a TBX Safeguards Model to Reproduce Maximum Safeguards Data FSE/SS-TBX-1615 TBX ECCS Performance ME-CA-0000-1093 Design Margin Review Calculations for MOV 1-8804A, 1-Attachment J 8804B, 2-8804A and 2-8804B ME-CA-0000-1093 Design Data for CPSES Unit 1, 2, and Common Safety 8 Related MOVs within the Scope of GL-89-10 [for MOVs 1-8801A/B; 1-8809A/B; 1-8811A/B; 1-8924; 1-LCV-112D/E]

ME-CA-0000-4070 Equipment Qualification Total Integrated Dose to ABB 0 Relays in Switchgear Located in Rooms 1-083, 2-083, 1-103, and 2-103 R&R-PN-011 SI System Notebook Safety Injection 1 R&R-PN-017 EP System Notebook Electric Power 2 RXE-TA-CP2/0-007 Containment Pressure Trip Setpoints 1 16345-ME(B)-130 Component Cooling Water Surge Tank Pressure 3 16345-ME(B)-346 SI Pump NPSH During Cold Leg Recirculation 0

-4-Number Title Revision 16345-ME(B)-389 RWST Setpoints, Volume Requirements, and Time 7 Depletion Analysis [I&C interface review]

16346-ME(B)-078 Component Cooling Water Surge Tank Venting 1 2-ME-0260 Determine NPSH Available of SI Pump During Cold Leg 0 Recirculation, and Compare with NPSH Required DESIGN BASIS DOCUMENTS Number Description Revision DBD-CS-068 Non-ASME Pipe Stress Analysis and Support Design 3 DBD-EE-041 480V and 120V AC Electrical Power System 1 DBD-EE-051 Protection Philosophy 15 DBD-EE-052 Cable Philosophy and Sizing Criteria 10 DBD-ME-261 Safety Injection System 10 DRAWINGS Drawing Number Title Revision A1-0503 Primary Plant - Unit 1 Containment and Safeguard CP-2 Buildings Plans at El 852-6" and 860-0" BRP-CS-2-AB-092 Pipe Support Location Isometric - Chemical and Volume CP-23 Control and Boron Thermal Regen CS-2-076-405- Large Bore Pipe Support CP-2 A42R E1-0001 Plant One Line Diagram - Units 1 and 2 CP-22 E1-0001, sht A Plant One Line Diagram - Unit 1 and Common- Distribution CP-17 Panels E1-0005 480v Auxiliaries One Line Diagram Safeguard Buses CP-23 E1-0007 Safeguard and Auxiliary Buildings Safeguard 480V MCCs CP-31 One Line Diagram E1-0009 Containment and Diesel Generator Safeguard 480V MCCs CP-23 One Line Diagram CP-36 E1-0010 Common Auxiliary and Control Bldgs Safeguard 480V MCCs One Line Diagram E1-0014 Service Water Intake Structure and Diesel Generator CP-26

-5-Drawing Number Title Revision Safeguard 480V One Line Diagram E1-0031, sht 45 Schematic Diagram, Safety Injection Pump 11 CP-6 E1-0031, sht 47 Schematic Diagram, Safety Injection Pump 12 CP-6 E1-0031, sht 53 Schematic Diagram, Centrifugal Charging Pump 11 CP-6 E1-0031, sht 55 Schematic Diagram, Centrifugal Charging Pump 11 CP-8 E1-0061, sht 17 Schematic Diagram, 1-LCV-115E CP-4 E1-0061, sht 94 Schematic Diagram, 1-8511A CP-8 E1-0061, sht 95 Schematic Diagram, 1-8511B CP-7 E1-0061, sht 96 Schematic Diagram, 1-8512A CP-6 E1-0061, sht 97 Schematic Diagram, 1-8512B CP-7 E1-0062, sht 5 Schematic Diagram, 1-8801A CP-5 E1-0062, sht 6 Schematic Diagram, 1-8801B CP-6 E1-0062, sht 7 Schematic Diagram, 1-8802A CP-4 E1-0062, sht 8 Schematic Diagram, 1-8802B CP-4 E1-0062, sht 11 Schematic Diagram, 1-8804A CP-5 E1-0062, sht 11A Schematic Diagram, 1-8804A CP-4 E1-0062, sht 12 Schematic Diagram, 1-8804B CP-8 E1-0062, sht 12A Schematic Diagram, 1-8804B CP-2 E1-0062, sht 16 Motor Operated Valve 1-8808A Accumulator Isolation CP-5 Valve E1-0062, sht 22 Schematic Diagram, 1-8811A CP-7 E1-0062, sht 023 Schematic Diagram, 1-8811B CP-8 E1-0062, sht 26 Schematic Diagram, 1-8813 CP-5 E1-0062, sht 27 Schematic Diagram, 1-8814A CP-7 E1-0062, sht 59 Schematic Diagram, 1-8814B CP-8 E1-0062, sht 76 Auxiliary Relays 2/1 8808AX and 2/1 8808CX CP-4 E1-0063, sht 1 Schematic Diagram, 1-8701A CP-6

-6-Drawing Number Title Revision E1-0063, sht 2 Schematic Diagram, 1-8701B CP-6 E1-0063, sht 3 Schematic Diagram, 1-8702A CP-6 E1-0063, sht 4 Schematic Diagram, 1-8702B CP-7 E1-0064, sht 11 Nitrogen Operated Valve 1-PCV-0455A - Pressurizer CP-6 Power Relief Valve E1-0064, sht 12 Nitrogen Operated Valve 1-PCV-0456 - Pressurizer Power CP-6 Relief Valve E1-0067, sht 23 Diesel Generator Air Compressor 1, Tag CP1-MECAED-01 CP-8 E1-0067, sht 39A Diesel Generator Miscellaneous Switches CP-1 E1-0067, sht 39A Diesel Generator Miscellaneous Switches CP-1 E1-0076, sht 38 Annunciator Lamp Cabinet 1-ALB-4C Schematic Diagram CP-2 E2-0001 Plant One Line Diagram CP-3 E2-0005 480v One Line Diagram Safeguard Buses CP-12 E2-0007, sht C Safeguard and Auxiliary Buildings Safeguard 480V MCCs CP-23 One Line Diagram E2-0024, sht 3C Plant Support 480 VAC Distribution Panels One Line CP-10 Diagram E2-0008, Sht A Containment & Circulating Water Intake Structure Normal CP-12 480V MCCs One Line Diagram E2-0009 Containment and Diesel Generator Safeguard 480V MCC's CP-13 One Line Diagram E2-0014 Service Water Intake Structure and Diesel Generator CP-13 Safeguard 480V MCC's One Line Diagram E2-0024, sht 3D Plant Support 480 VAC Distribution Panels One Line CP-17 Diagram M1-0215, sht D Flow Diagram Starting Air Piping, CP1-MEDGEE-01 CP-22 M1-0260 Flow Diagram, Residual Heat Removal System CP-30 M1-0261, sht 1 Flow Diagram, Safety Injection System CP-20 M1-0262, sht 2 Flow Diagram, Safety Injection System CP-22 M1-0263, sht B Flow Diagram, Safety Injection System CP-15 M1-0263, sht 5 Flow Diagram, Safety Injection System CP-13

-7-Drawing Number Title Revision M2-0229 Flow Diagram, Component Cooling Water System CP-17 M2-0229, sht A Flow Diagram, Component Cooling Water System CP-12 M2-0229, sht B Flow Diagram, Component Cooling Water System CP-14 M2-0230 Flow Diagram, Component Cooling Water System CP-14 M2-0230, sht A Flow Diagram, Component Cooling Water System CP-6 M2-0231 Flow Diagram, Component Cooling Water System CP-14 M2-0231, sht A Flow Diagram, Component Cooling Water System CP-154 M2-0260 Flow Diagram, Residual Heat Removal System CP-16 M2-0263 Safety Injection System CP-10 VL0805 Nozzle Type Relief Valve CP-0 VL-93-119 Nozzle Type Relief Valve CP-B 212B7150, sht 15 MCC Space Heaters 5 2323-A1-0501 Primary Plant - Unit 2 Containment and Safeguard Building CP-1 Plans El. 808-0" and 810-6" 2323-A1-0502 Primary Plant - Unit 1 Containment and Safeguard CP-1 Buildings - Plans at El. 831-6" and 832-6" 2323-A1-0503 Primary Plant - Unit 1 Containment and Safeguard CP-2 Buildings - Plans at El. 852-6" and 860-0" 2323-A1-0507 Primary Plant Auxiliary and Electrical Control Building Floor CP-1 Plan El. 778 and 790-6" 2323-A1-0508 Primary Plant Auxiliary and Electrical Control Building Floor CP-1 Plans El. 807-0" and 810-6" 2323-A1-0513 Electrical and Control Building Floor Misc. Plans and CP-1 Details 2323-A2-0500 Primary Plant - Unit 2 Containment and Safeguard 5 Buildings Plans at El. 773 and 790-6" 10 CFR 50.59 Evaluations 59EV-2001-000751-01-00 59EV-2002-000587-01-00

-8-59EV-2002-001634-01-00 59EV-2002-002151-01-00 59EV-2002-002189-01-00 ENGINEERING EVALUATIONS Number EVAL-1999-002570-01-00 EVAL-1999-00303-01-00 Evaluate Current IST Test Data for SIP 1-02 Pursuant to the CPSES IST Program EVAL-2001-001-60-01-00 Evaluate IST Data for SI Pump 1-01 MODIFICATIONS DM # 98-061, Convert TM 2-96-008 to a Permanent Installation (SI Relief System), Revision 0 TM # 94-008, Gagging of Component Cooling Water Thermal Relief Valves, Revision 0 DCN-8573, Revise the DBD and Flow Diagrams as Shown on the Attached Pages, Revision 0 ONE FORMS 92-1129 94-0990 97-1073 97-1390 98-0440 92-733 96-0408 97-1129 97-1417 98-0452 94-0889 97-0814 97-1344 97-1556 98-0628 PROCEDURES Number Title Revision ECE-2.13-06 Comanche Peak Engineering Reports 3 ECE-2.25 Equipment Qualification Program EDCN #6 ECE-5.01 Design Control Program 4 ECE-5.01-01 Design Basis Documents 7 ECE-5.01-03 Design Change Notices and Related Process Documentation 8 ECE-5.02 Specifications 8 ECE-5.03 Calculations 12 ECE-5.05 Design Drawings and Special Documents 8 EOS-1.3B Transfer to Cold Leg Recirculation 1

-9-Number Title Revision INC-7532A Sensor Response Time Test - Containment Pressure, 5 Channel 0934 IST-1301 Inservice Testing of Motor Operated Valves 1 MSM-GO-0205 Freeze Seal Formation and Maintenance 2 MSM-GO-0204 Safety Valve and Relief Valve Bench Testing 5 MSM-GO-0203 Flange Alignment and Fastener Torque Data 3 MSM-G0-0208 Gagging of Valves 0 MSM-CO-8871 Crosby Safety Valve Maintenance 1 OPT-201A Charging System 12 OPT-432B Train A Safety Injection Test 3 OPT-508B CVCS Section XI Valves 8 OPT-510B SI Section XI Valves 7 OPT-512B RHR and SI Subsystem Valve Test 8 OPT-513A SI Pump Performance and Flow Balancing 0 OPT-523A CCP Performance and Flow Balancing 0 OPT-204A SI System, Surveillance Test 10 PPT-SO-6004 Motor Operated Rising Stem Valve Risk-Informed IST Testing 2 SOP-103A Chemical and Volume Control System 13 SOP-201A Safety Injection System 11 STA-606 Control of Maintenance and Work Activities 26 STA-422 Processing of SmartForms 18 STA-421 Initiation of SmartForms 10 STA-702 Surveillance Program 16 STA-707 10 CFR 50.59 Reviews 16 WCI-606 Work Control Process 4

-10-TECHNICAL EVALUATIONS TE-SE-91-393 TE-92-765 TE-94-000853 TE-96-391 TRAINING MANUALS Number Title Revision OP12.GFE.CZF Breakers, Relays and Disconnects 1 OP51.SYS.S11 Emergency Core Cooling System 0 OP51.SYS.AC2 6.9Kv and 480 V Electrical Distribution 3 10 CFR 50.59 SCREENINGS 59SC-1999-001397-01-02 59SC-2001-001972-01-00 59SC-2002-001675-01-00 59SC-1999-001458-01-00 59SC-2001-002388-01-01 59SC-2002-002274-01-00 59SC-2000-001749-01-00 59SC-2001-002523-01-01 59SC-2002-003436-01-00 59SC-2001-000751-02-00 59SC-2002-000194-01-00 59SC-2002-003523-01-00 59SC-2001-000774-01-00 59SC-2002-001336-01-00 SMART FORMS SMF-1999-000078-00 SMF-2001-000741-00 SMF-2002-000532-00 SMF-1999-000541-00 SMF-2001-001016-00 SMF-2002-000596-00 SMF-1999-000836-00 SMF-2001-001703-00 SMF-2002-000754-00 SMF-1999-000849-00 SMF-2001-001787-00 SMF-2002-000870-00 SMF-1999-002570-00 SMF-2001-002265-00 SMF-2002-001257-00 SMF-1999-002629-00 SMF-2001-002353-00 SMF-2002-001405-01*

SMF-1999-002699-00 SMF-2001-002388-00 SMF-2002-001423-01*

SMF-1999-002718-00 SMF-2002-000143-00 SMF-2002-001613-00 SMF-1999-002727-00 SMF-2002-000277-00 SMF-2002-001623-00 SMF-2000-000014-00 SMF-2002-000325-00 SMF-2002-001645-00 SMF-2000-000170-00 SMF-2002-000346-00 SMF-2002-002060-00 SMF-2000-000940-00 SMF-2002-000354-00 SMF-2002-002859-00 SMF-2000-001555-00 SMF-2002-000361-00 SMF-2002-002911-00 SMF-2000-002066-00 SMF-2002-000371-00 SMF-2002-003000-00 SMF-2000-002123-00 SMF-2002-000372-00 SMF-2002-003210-00 SMF-2000-002675-00 SMF-2002-000373-00 SMF-2002-003515-00 SMF-2000-002699-00 SMF-2002-000383-00 SMF-2002-003598-00 SMF-2000-003275-00 SMF-2002-000395-00 SMF-2002-003614-00

-11-SMF-2002-004054-00* SMF-2002-004158-03* SMF-2002-004204-00*

SMF-2002-004058-00* SMF-2002-004174-00* SMF-2002-004206-00*

SMF-2002-004066-00* SMF-2002-004188-00* SMF-2002-004212-00*

SMF-2002-004071-00* SMF-2002-004189-00* SMF-2002-004227-01*

SMF-2002-004073-00* SMF-2002-004201-00* SMF-2002-004228-01*

SMF-2002-004083-00*

• Initiated during inspection to address team concern.

WORK ORDERS 1-93-037693-00 4-02-140349-00 5-99-504602-AA 5-01-504600-AA 1-93-041620-00 4-02-142159-00 5-99-504613-AA 5-01-504614-AA 1-94-071706-00 4-02-142231-00 5-99-504614-AA 5-01-504749-AA 1-94-071738-00 4-02-142232-00 5-99-504722-AB 5-01-505629-AD 1-94-072806-00 4-02-145112-00 5-99-504747-AA 5-01-505635-AE 1-94-079478-00 5-97-504603-AA 5-99-504748-AB 5-01-505638-AD 1-96-102163-00 5-97-504612-AA 5-99-504750-AB 5-01-505635-AA 1-96-102171-00 5-97-504614-AA 5-99-504751-AB 5-02-504753-AB 1-97-114467-00 5-97-504744-AA 5-99-504752-AB 5-02-504780-AA 1-99-125392-00 5-98-504601-AA 5-99-504772-AB 5-02-505628-AA 1-99-125393-00 5-98-504602-AA 5-99-504773-AB 5-02-505628-AB 3-01-318089-01 5-98-504703-AA 5-99-504777-AB 5-02-505628-AF 3-01-318093-01 5-98-504743-AA 5-99-504778-AB 5-02-505629-AA 4-95-087689-00 5-98-504744-AA 5-99-505039-AA 5-02-505629-AC 4-95-091214-00 5-98-504745-AA 5-99-505039-AB 5-02-505635-AB 4-96-100840-00 5-98-504774-AA 5-00-505053-AA 5-02-505635-AC 4-96-103024-00 5-98-504775-AA 5-00-505053-AB 5-02-505638-AA 4-96-103025-00 5-98-504776-AA 5-00-505142-AA 5-02-505638-AB 4-97-111427-00 5-98-504780-AA 5-00-505143-AA 5-02-505638-AC 4-98-117632-00 5-98-504781-AA Westinghouse Electric Corporation Letters WPT-11222 April 10,1989 WPT-12421 January 4, 1990 WPT-13885 September 11,1991 WPT-13964 September 25, 1991 WPT-15211 August 5, 1993 WPT-15992 April 7, 1999 WPT-15995 April 14, 1999 MISCELLANEOUS DOCUMENTS Final Safety Analysis Report, Amendment 98, Supplement 1 Technical Specifications, Amendment 100

-12-CPSES 10 CFR 50.59 Resource Manual, Revision 1 CPSES Inservice Testing Plan for Pumps and Valves, Revision 17 TXU Office Memorandum CPSES-9403542, April 19, 1994 VTMR-001-806-002, American Standard, Heat Transfer Division, Heat Exchanger Specification Sheet for CCP Oil Cooler and Safety Injection Pump Oil Cooler.

TU Electric Assessment Report for INPO SOER 97-01, Potential Loss of High Pressure Injection and Charging Capability from Gas Intrusion, Recommendation 1 [level sensing devices], May 28, 1998 TM# 2-96-008, Temporary Pressure Relief Device/System to Maintain the Safety Injection System Below the Normal System Relief Valve Set Point, Revision 3