IR 05000254/1999017

Insp Repts 50-254/99-17 & 50-265/99-17 on 990823-27.No Violations Noted.Major Areas Inspected:Pilot Baseline Insp for Annual Review of Changes to Safety Analysis Rept & Biennial Review Permanent Plant Mod

ML20212C705
Person / Time
Site:	Quad Cities
Issue date:	09/15/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212C699	List: IR 05000254/1999017 ML20212C696
References
50-254-99-17, 50-265-99-17, NUDOCS 9909220117
Download: ML20212C705 (13)

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U.S. NUCLEAR REGULATORY COMMISSION REGIONlli Docket Nos: 50-254;50-265 License Nos: DPR-29; DPR-30 Report No: 50-254/99017(DRS); 50-265/99017(DRS)

Licensee: Commonwealth Edison Company I

Facility: Quad Cities Nuclear Power Station Units 1 and 2 Location: 22710 206th Avenue North

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Cordova,IL 61242 Dates: August 23 - August 27,1999 Inspectors: David Butler, Reactor Inspector Patricia Lougheed, Reactor Inspector Roger Mendez, Reactor inspector

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Approved by: John M. Jacobson, Chief, Mechanical Engineering Branch j Division of Reactor Safety I

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9909220117 990915 PDR ADOCK 05000254 G PDR

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SUMMARY OF FINDINGS Quad Cities Nuclear Power Station, Units 1 & 2 NRC Inspection Report 50-254/99017(DRS); 50-265/99017(DRS)

This report covers the pilot baseline inspections for the annual review of changes to the safety l analysis report and the biennial review permanent plant modification '

Inspection findings were assessed according to potential risk significance, and were assigned colors of GREEN, WHITE, YELLOW, or RED. GREEN findings are indicative of issues that, while not necessarily desirable, represent little risk to safety. WHITE findings would indicate issues with some increased risk to safety, and which may require additional NRC inspection YELLOW findings would be indicative of more serious issues with higher potential risk to safe performance and would require the NRC to take additional actions. RED findings represent an unacceptable loss of margin to safety and would result in the NRC taking significant actions that could include ordering the plant shut down. No individual finding by itself would be indicative of I I

either acceptable or unacceptable performance. The findings, considered in total with other inspection findings and performance indicators, will be used to determine overall plant performanc Cornerstone: Mitigating Systems Green: The inspectors ideniitiea that tha Unit 1 post modification test for a design change package on the fuel transfer pump was not retained by the licensee. The licensee had retained the Unit 2 test and had signature evidence that the Unit 1 test was performe Cornerstone: Barrier Integrity No findings were identified in this are I l

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Report Details j REACTOR SAFETY Cornerstones: Mitigating Systems and Barrier Integrity 1R02 Chanaes to License Conditions and Safety Analysis Report (IP 71111, Attachment 2)

.1 Review of 50.59s Evaluations for Chances to the Safety Analysis Report ) Insoection Scooe 1 I

The inspectors reviewed five evaluations done pursuant to 10 CFR 50.59, one of which '

pertained to the barrier integrity comerstone. All five evaluations related to changes to the updated final safety analysis report. The inspectors also reviewed three changes to the updated final safety analysis report where the licensee had determined that a 50.59 evaluation was not necessary. In regard to the three changes where no 50.59 j evaluation was performed, the inspectors verified that the changes were minor editorial clarifications that did not meet the threshold of a " change to the facility as described in the safety analysis report." For the 50.59 evaluations, the inspectors confirmed that prior NRC approval was not required for any of the change I Observations and Findinas No findings were identified in this are .2 (Closed) URI 50-254/98201-18: 50-265/98201-18: Updated Final Safety Analysis Report Discrepancies. This unresolved item was previously closed in Inspection Report 98019 with the exception of one item dealing with Tables 8.3-2 and 8.3-3 regarding emergency diesel generator loading. The licensee acknowledged that the information in Tables 8.3-2 and 8.3-3 had changed, that the emergency diesel generator loading information was design basis information, and that an updated final safety analysis report revision was necessary. The cognizant regulatory assurance engineer revised the unresolved item action tracking item (000156) to track this issue. The licensee also reviewed other Section 8 updated final safety analysis report tables to determine if those tables contained design basis information. In one case, the licensee discovered that the information presented in the updated final safety analysis report table was not available in the load tracking database, despite a note that referenced the reader to that database. Therefore, the licensee expanded the action tracking iterr *o address whether these tables needed to remain in the updated final safety analysis report and whether the reference to the load tracking database was necessary in all cases. The inspectors reviewed the current load tracking database and determined that the emergency diesel generators loadings were within their design basis value Therefore, the inspectors had no further technical question CFR 50.34(b)(2) requires, in part, that the final safety analysis report contain a description and analysis of the structures, systems and components of the facility, with !

emphasis upon performance requirements, the bases upon which the requirements I were established, and the evaluations required to show that the safety functions will be I accomplished. 10 CFR 50.71(e) requires, in part, that each licensee periodically update the final safety analysis report to assure that the information included in the final safety analysis report contains the latest material develope I J

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The loading on the emergency diesel generators following a loss of coolant accident and/or a loss of offsite power are performance requirements that are necessary to understand the emergency diesel generator system design and safety evaluatio Therefore, they are required to be in the updated final safety analysis report by 10 CFR 50.34(b)(2). The failure to update the design performance requirements for the emergency diesel generator loadings is a violation of 10 CFR 50.71(e). However, this item was identified as part of a larger unresolved item for which enforcement discretion was already granted (VIO 50-245/264-98201-07). Therefore, it will be considered another example of that violation and no separate enforcement action will be take .3 (Closed) Unresoh/ed item (URI) 50-254/97013-02: 50-265/97013-02: Updated Final Safety Analysis Report Discrepancy Regarding the Residual Heat Removal Service Water Pumps. In 1997, the NRC identified a discrepancy among the pump surveillance test requirements, the design basis document and the updated final safety analysis report regarding the required residual heat removal service water pressure. The inspectors confirmed that the pumps were capable of meeting their design function of removing heat from the residual heat removal system fol!owing a design basis acciden Therefore, the inspectors had no further technical question ,

The inspectors also reviewed the approved updated final safety analysis report change (UFSAR-97-R5-016) and associated 10 CFR 50.59 evaluation. The actual change was to add a footnote denoting the information as being the " original manufacturer's )

specification of the size of the pumps chosen." The wording implied that the pumps were no longer capable of meeting their original design specification or functio Additionally, the updated final safety analysis report did not address the pumps functional requirement (to provide water to the residual heat removal heat exchanger at a specified flow and pressure) which was different than the pump design informatio Therefore, the licensee revised the planned corrective actions for problem identification form Q1999-02345 to ensure that the design basis of the pumps was adequately captured in the updated final safety analysis report. This item is close R17 Permanent Plant Modifications (IP 71111, Attachment 17)

,1 Review of Recent Plant Modifications Inspection Scoce The inspectors reviewed seven plant modifications which were installed since August 1998. The packages were chosen based upon their affecting systems that had high Maintenance Rule safety significance or high risk significance in the licensee's Individual Plant Evaluation. Five of the modifications involved changes to mitigating systems, while the last two affected barrier integrity. The inspectors reviewed the modifications to confirm that the changes did not affect any systems' safety functio Design and testing aspects were verified to ensure the functionality of the mcdification, it's associated system, and any support systems. Walkdowns were conducted to ensure proper installation of the modifications, j Observations and Findinos I The inspectors identified that the Unit 1 post modification test for a design change package on the fuel transfer pump was not retained by the licensee. Retention of the ,

post modification test would have provided evidence that the test was adequately performed. The licensee had retained the Unit 2 test and signature evidence that the Unit 1 test was performe The post modification test to demonstrate operability of the fuel transfer pump was outlined in the design change acceptance testing summary and was to be performed in accordance with surveillance procedure OCOP 4100-16, " Manually Filling the Design Fire Pump Day Tank," Revision 2. However, the procedure was not considered a quality document and, therefore, the completed test was not retained for the life of the plant 1 when the design change package was microfilmed. The licensee subsequently produced a copy of the signed cover sheet and the work request form that required the test. In addition, the system engineer stated that the test was adequately performe This assured the inspectors that the test was performed. The licensee committed to review the requirements for retaining post modification test records for surveillances that were not required to be retaine The inspectors verified that there were adequate assurances that the fuel transfer pump would operate as modified. The inspectors performed a Phase I screening of this issue under the significance determination process and the issue screened out as " Green" 4 OTHER ACTIVITIES 40A4 Other(IP 93902)

.1 (Closed) URI 50-254/97022-02: 50-265/97022-02: Breaker Coordiration Issues. The inspectors reviewed the Station's position paper (White Paper) prepared for this item and the 250 Vdc system licensing basis. The White Paper provided additional information demonstrating that partial breaker coordination existed. The lack of breaker coordination occurred in the breakers instantaneous tripping region. This mis-coordination would only result from faults that occurred at the breaker load side terminsis or by a cable fault within close proximity of the load breaker. Since cables are highly reliable and a failure at the breaker load terminals is highly unlikely, the most likely fault condition would be at the load. Due to cable length, load fault currents would be limited allowing the load breaker to clear the fault without tripping the upstream feed breaker. In addition, the licensee indicated that breaker coordination was a design consideration and that they attempted to optimize breaker coordination during the design process. Therefore, the inspectors determined that there was no technical concem with the licensee's approach. This item is considered closed.

.2 (Closed) Violation 50-254/98019-04: 50-265/98019-04: Inadequate Corrective Actio In Inspection Report 50-254/265-98019, the inspectors noted that the licensee had taken adequate corrective actions to the violation and that no response was necessar Therefore, this violation is closed.

.3 (Closed) Violation 50-254/98019-05: 50-265/98019-05: Inadequate Design Control. In inspection Report 50-254/265-98019, the inspectors noted that the licensee had taken adequate corrective actions to the violation and that no response was necessar Therefore, this violation is close u

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40A5 Manaaement Meetinas i

.1 Exit Meetina Summary The inspector presented the inspection results to members of licensee management in an exit rneeting on August 27,1999. The licensee acknowledged the information and findings presented. No proprietary information was identifie i l

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PARTIAL LIST OF PERSONS CONTACTED Licensee G. Barnes, Station Manager -

J. Dimmette, Site Vice President M. Mcdonald, Operations Manager C. Peterson, Regulatory Assurance D. Wozniak, Engineering Manager NRC J. Caldwell, Deputy Regional Administrator, Region lli L. Collins, Resident in.spector J. Jacobson, Chief, Mechanical Engineering Branch, DRS S. Reynolds, Deputy Division Director, Division of Reactor Safety M. Ring, Chief, Branch 1, Division of Reactor Projects INSPECTION PROCEDURES USED IP 71111.02 (draft) Changes to License Conditions and Safety Analysis Report IP 71111.17 (draft) Permanent Plant Modifications IP 93902 Followup - Engineering ITEMS OPENED, CLOSED AND DISCUSSED

. Opened None Closed (254/265) .

97013-02 URI Updated Final Safety Analysis Report Discrepancy for Residual Heat Removal Service Water Pumps: Two issues 97022-02 URI Breaker Coordination issues 98201-18 URI Updated Final Safety Analysis Report Discrepancies ,

98019-04 VIO Inadequate Corrective Action 98019-05 VIO Inadequate Design Control Discussed None

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LIST OF DOCUMENTS REVIEWED The following is a list of licensee documents reviewed during the inspection, including documents prepared by others for the licensee. Inclusion on this list does not imply that NRC inspectors reviewed the documents in their entirety, but, rather that selected sections or portions of the documents were evaluated as part of the overallinspection effor ,

Calculations ATD-0057 Evaluation of Bore Diameter of Unit 2 Restricting Orifices RO 2-3924 and 2-3925, Revision 0 QC-270-C-024 Cable Tray Loading Calculation, Revision 7 ODC-5200-S-0732 Seismic Qualification of CP 2940 Switch Addition to Diesel Generator Panel, Revit, ion 0 QDC-6600-S-0722 Seismic Qualifications of TD2 Time Delay Relay and Mounting of Enclosure and Conduit, Revision 0 Drawings 4E-1350C Schematic Control Diagram of the Diesel Fuel Oil Transfer Pump 1 and 2

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Feed Controls, Revisions K & M 4E-1351 A Schematic Control Diagram of the Engine Control and Generator

. Excitation for the Standby Diesel Generator 1/2, Revisions AG, AK, & AL 4E-1358A Schematic Diagram Electro-Hydraulic Control System, Revision K 4E-2358B EHC Alarm and Trip Schematic, Revision K 4E-2629 Turbine EHC Pressure Switches, Revision K 4E-2637 EHC Cabinet No. 902-31 Alarm and Trip Wiring Diagram, Revision D 4E-2655J . Wiring Diagram of the 4160V Switchgear Bus 23-1 Cubicle 8 Revision J 4E-6820H EHC Power-Load Unbalance Demodulator Schematic, Revision G 4E-6823M EHC Power-Load Unbalance Circuit Schematic, Revision H

- 4E-6872F Wiring Diagram of the Exterior Cable Tray Layout Station Blackout Sections and Details, Revision C 4E-7871C Wiring Diagram of the Station Blackout 4160 Switchgear Bus 71 Cubicle 3, Revision A M-4A-1,2,3,& 4 Environmental Zone Maps M-36 Diagram of Core Spray Piping, Revision AU M 37 Diagram of Residual Heat Removal Service Water Piping, Revision AP M-39-1,2,3 Diagram of Residual Heat Removal Piping, Revisions BC, AW, & B M-43 Diagram of Reactor Building Equipment Drains, Revision BJ M-46-1,2 Diagram of High Pressure Coolant Injection Piping, Revisions BN & F I M-47-1 Diagram of Reactor Water Cleanup Piping, Revision -

M-50-1 Diagram of Reactor Core Isolation Cooling Piping, Revision BA M-75-2 Diagram of Reactor Building Closed Cooling Water Piping, Revision B M-77-2 Diagram of Nuclear Boiler and Reactor Recirculating Piping, Revision AN M-725-1,2,3 Diagram of Control Room Heating, Ventilation, and Air-Conditioning System, Revisions K, K, & E M-2022- Turbine Control Diagram Units 1 and 2, Revision A Electrical Standards ,

EM-29105 600 Volt Power Cable for Nuclear Generating Stations, September 1,1992 . .

EM-29115 600 Volt Control Cable for Nuclear Generating Stations, June 1,1994

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EM-29116 SKV Ethylene Propylene Insulated Chlorosulfonated Polyethylene Jacketed Power Cable for Stations and Substations, December 14,1990 EQ-GEN 029 3M Scotch Tape Splices, Revision 4 N-C-0008 Cable Pulling Guidelines, Revision 5 N-EM-0035 Cable Standards, Revision 6 N-EM-0048 Low Voltage Tapes, Revision 2 Engineering Raquests ER9805134 Revise General Electric Specification (GEK) No.11367C Steam Turbine Instruction Manual for Electrical Alarm and Trip System (Pressure Switches PS-101 A and 101B)

ER9900536 Revise Procedure QCIPM 5610-31. Revision 3,"EHC Low Hydraulic Pressure Turbine Trip Functional Test" ER9805001 Revise Genral Electric Specification GEK No.11367C Steam Turbine i instruction Manual for Electrical Alarm and Trip System (High Heae:

Temperature Bypass above 30 Percent Pressure)

ER9804426 Revise Procedure No. QOA 5600-03, Revision 5," Turbine Hood Spray Regulation Valve Failure" ER9804425 Revise Procedure No. QOA 5600-04, Revision 12, " Loss of Turbine Generator" General Electric (GE) Specifications GEH 3626 Pressure Switch Model CR127A GEK 11365A Electrohydraulic Contros GEK 11354A Power / Load Unbalance Circuit and Relays GE Technical Service Letters 1212-2 Plant SCRAM Frequency Reduction Features for BWR and PWR Nuclear Turbines with MKl and MKil EHC Controls, January 27,1997 Miscellaneous Control Room Habitability Study, Revision 2,6/14/1982 Licensed Operator Continuing Training, " Mods and Lessons Learned 99-1," January 12,1999 Modifications DCP 9700346 Modify Turbine Trip Logic (EHC Low Power) to Reduce the Probability of a Spurious Trip DCP 9700351 Enlarge Bore Sizes of Reducing Orifices 2-3924 and 2-3925 3 DCP 9700366 Modify Turbine Trip Logic from Low Pressure Turbine Exhaust i Hood High Temperature to Bypass the Trip Above 30% )

Turbine / Generator Power DCP 9800226 Replace Emergency Diesel Generator Time Delay Relay TD2 i DCP 9800233 Splice Station Blackout Diesel Control Cables and Reroute Power Cable to Switchgear 23-1 DCP 9800235 Modify the Emergency Diesel Generator Fuel Transfer Pump Logic  !

DCP 9900009 Modify the 24 Volt Circuit Powering the Unit 2 Scram Discharge )

Volume Instruments )

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Nuclear Design Information Transmittals ODC-99-071 Design input - ECCS Room Coolers Part Evaluations NEP-18-04 Equipment Dynamic Qualification Test Review of Time Delay Relays Problem identification Forms Q1997-04066 High Pressure Coolant Injection High Steam Flow Switch Logic and updated final safety analysis report Wording,10/26/1997 Q1999-02345 Residual Heat Removal Service Water Pump Testing and Safety Margin of Pumps,7/19/1999 Procedures .

ECTP-19 Control Circuits, Revision 2 ECTP 24 Operational Analysis Department Electrical Construction Test Procedure for Modifications at Nuclear Stations, Revision 2 NEP 04-01 Plant Modifications, Revision 6 NEP-04-05 Design Change Acceptance Testing Criteria, Revision 0 -

NEP 14-03 Control and Tracking of Electrical Load Changes, Revision 1 NSWP E-02 Electrical Cable Termination and Inspection, Revision 5-1 1 PMID 154086 VC2A15 (A15-VC2) Voltage Comparator Calibration l

QCAP 0200-15 High Risk Activity Mitigation Plan, Revision 9 )

QCAP 1100-13 Procedure Field Change Request, Revision 8 QCEM 700-1 Cable Pulling Procedure, Revision 0 ]

QCEPM 0700-18 Calibration of Diesel Generator Time Delay Relays, Revision 7 QCOP 4100-16 Manually Filling the Diesel Fire Pump Day Tank, Revision 2 QCOP 5750-09 Control Room Ventilation System, Revision 15 QCOS 0010-07 Equipment External Leak Test, Revision 1 QCOS-0201-12 Class One ASME Section XI Post-replacement Pressure Test at Power i Operation, Revision 0 (also completed procedure 3/1/1999) l QCOS 1000-04 Quarterly Residual Heat Removal Service Water Pump Operability Test, l Revision 12 QCOS 5750-09 ECCS Room and DGCWP Cubicle Cooler Monthly Surveillance, Revision 15 QCOS 6600-03 Diesel Fuel Oil Transfer Pump Monthly Operability, Revision 6 !

QCOS 6620-01 SBO DG 1(2) Quarterly Load Test, Revision 3  ;

QCIP 0100-05 Instrument Maintenance Department Administrative Drift Limits j Guidance, Revision 2  ;

QCIPM 5610-30 EHC Low Hydraulic Pressure Turbine Trip Functional Test, Revision 3 QCIPM 5610-33 Turbine Exhaust hood High Temperature Turbine Trip Functional Test, .

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QCTS 0220-02 Unit 1 and 2 24/48 Vdc Battery Performance Test, Revision 4 QCTS 0220-05 Unit 1 and 2 24/48 Vdc A Battery Service Test, Revision 2 QCTS 0220-08 Unit 1 and 2 24/48 Vdc B Battery Service Test, Revision 1 OlP 0100-18 Refuel Outage Balance of Plant Calibration Schedule, Revision 12 OOA 900-7 C-4 901-7 (902-7) Row C Annunciator Procedures, Revision 2 OOS 0005-S01 Operations Department Weekly Summary of Daily Surveillance, l Revision 93 SPP-VT-2-1 VT-2 Visual inspection Performed for Section XI, Revision 7

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QC Engineering Transmittal Letter e Addition of a Nominal 3 Second Time Delay to EHC Low Pressure Turbine Trip Logic, September 9,1998 System Planning Operating Guides 1-1 Generating Stations Operating Voltage Levels, Revision 2 1-1-A Operating Nuclear Stations at Reduced Excitation Levels, Revision 2 10 CFR 50.59 Evaluations and Screenings SE-97-151 Clarify updated final safety analysis report Description of the Residual Heat Removal Service Water pumps in Table 9.2-1,11/12/1997 SE-98-088 - Remove and Install New Type of Time Delay Relays in the Unit 1, Unit 2 and 1/2 Diesel Rooms SE-98-098 Modify the Unit 1 Emergency Diesel Generator Fuel Transfer Pump Circuit to Allow Operation at Both Diesel Rooms SE-98-099 Modify Turbine Trip Logic from Low Pressure Turbine Exhaust Hood High Temperature to Bypass the Trip Above 30% Turbine / Generator Power SE-98-100 Modify Turbine Trip Logic (EHC Low Power) to Reduce the Probability of a Spurious Trip SE-98-108 Modify Power Supply to the Scram Discharge Volume Level Switches SE-98-110 Splice Station Blackout Control Cables and Reroute Power Cable Between 4 kV Switchgear 71 and 4 kV Switchgear 23-1 SE-98-151 _ Revise updated final safety analysis report Discussion on the Control Room Emergency Ventilation System,12/2/1998 SE-99-015 Revise Updated Final Safety Analysis Report on Minimum System Voltages and Location of Infonnation,2/26/1999 SE-99-016 Revise Updated Final Safety Analysis Report Section 5.3.2.2 to Allow ASME Section XI Pressure Testing of Non-Welded Components Using Nuclear Heat,2/25/1999 -

SE-99-031 Correct Equipment Configuration Description Errors for the Station Blackout Diesel Generator System,6/3/1999 SE-99-056 Revise Updated Final Safety Analysis Report Table 6.2-7 to Reflect Correct Primary Containment isolation Valves, 7/21/1999 SS-F-99-157 Enlarge Bore Sizes Of Reducing Orifices 2-3924 and 2-3925 SS-H-99-45 Revise Updated Final Safety Analysis Report Section 5.3.2.2 to Allow Pressure Testing at Power,3/29/1999 SS-U-98-10 Correct Updated Final Safety Analysis Report Typographier.1 Error on Product Name/ Number,2/11/1999 )

SS-U-9811 Minor Wording Changes to Updated Final Safety Anatysis Report Section i 9.4.5.B, Page 9.4-8,11/26/1998 SS-U-9902 Clarification of High Pressure Coolant injection Steam Flow Switch Logic, 3/9/1999'

Technical Specifications 3. Ultimate Heat Sink

, Updated Final Safety Analysis Report Change Packages 97-RS-016 Clarify Updated Final Safety Analysis Report Description of the Residua!

Heat Removal Service Water Fi ups in Table 9.2-1,11/12/1997 97-R5-101 Minor Wording Changes to Updated Final Safety Analysis Report Section 9.4.5.B, Page 9.4-8,11/26/1998

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97-R5-103 Revise Updated Final Safety Analysis report to Show Technical Specification Allowed Range on Control Room Ventilation System Air intake Value and Discuss Necessary Manual Operations,12/2/1998 97-R5-115 Correct Vendor Product Name/ Number,2/13/1999 97-R5118 Revise Updated Final Safety Analysis Report Section 5.3.2.2 to Allow Pressure Testing at Power,3/29/1999 97-R5-119 Revise Updated Final Safety Analysis Report on Minimum System Voltages and Location of Inforniation,4/8/1999 97-R5-121 Clarification of High Pressure Coolant injection Steam Flow,3/11/1999 97-R5-130 - Revise Updated Final Safety Analysis Report Sections 8.3.1.9.4.3, 8.3.1.9.4.4 & 8.3.1.9 to Correct Station Blackout Diesel information, 6/7/1999 97-R6-005 Revise Updated Final Safety Analysis Report Table 6.2-7 on Containment isolation Valves,7/23/1999 Updated Final Safety Analysis Report Sections Section 6.3.2. Core Spray Subsystem Interfaces with Other ECCS Subsystems Section 6.3.2. Subsystem Characteristics Section 6.3.2. LPCI Subsystem Interfaces with Other ECCS Subsystems Section 6.3.2. Subsystem Characteristics Section 6.3.2. HPCI Subsystem Interfaces with Other ECCS Subsystems Section 6.3.2. Subsystem Characteristics Section Control Room Heating Ventilation and Air Conditioning Section Reactor Protection System Section 7.2. Single Failure Criteria Section 8. ' DC Power Systems Section 8. AC Power Systems Section 9. Diesel Generator Cooling Water System Section 1 Tarbine-Generator Section 15.2. Load Rejection With Bypass Section 15. Turbine Trip Section 15.6.5. Control Room Dose Rates Vendor Manuals Barton Model 227A Differential Pressure Indicator White Papers DG99-000135 Conduct of ASME Section XI Testing Following Replacements with Core Criticality at Boiling Water Reactors,2/19/1999 254-100-96-01004 120VAC,125VDC and 250VDC System Fuse / Breaker Coordination Work Requests 980028462 01 Modify EHC Low Pressure Turbine Trip Logic ECTP #19 980044304 01 Perform Load Test for 24/48 Volt Battery 980075137 03 Perform Pre-fabrication Shop Work Required for Installation of New TD2 Relay 980083679 01 - Low Pressure Turbine High Temperature Bypass Above 30% Power Per the Dee!g:. Change Package 980084210 01- Manua!!y Fill the Diesel Fire Pump Day Tank 980084652 06 Determinate / Terminate and install Cables at Switchgear 71 in Station Blackout Building

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f 980087963 02 Manually Fill the Diesel Fire Pump Day Tank 980087964 01 Modify Emergency Diesel Generator Fuel Oil Transfer Pump Logic 990005621 02 Revise Power Feed to Scram Discharge Volume Level Instruments l

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