ML20132C666
| ML20132C666 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/02/1996 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20132C671 | List: |
| References | |
| 2423, NUDOCS 9612180435 | |
| Download: ML20132C666 (18) | |
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4 CENTERDOR ENERGY 5501 N. State Route 2 419-249-2300 John K. Wood Oak Harbor OH 43449 FAX: 419-321-8337 Vce Prescent - Nuclear Davis-Besse Docket Number 50-346 License Number NPF-3 Serial Number 2423 December 2, 1996 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C.
20555-0001
Subject:
Revision 20 to the Davis-Besse Nuclear Power Station (DBNPS)
Unit 1 Updated Safety Analysis Report (USAR)
Ladies and Gentlemen:
The Toledo Edison Company hereby submits, pursuant to the requirements of 10 CFR 50.71(e) and 10 CFR 50.4(b)(6), one (1) original plus ten (10) copies of Revision 20 to the DBNPS Updated Safety Analysis Report (USAR).
Revision 20 to the USAR reflects facility changes implemented between November 16, 1994 and June 2, 1996. This USAR revision includes facility changes that occurred during operating fuel Cycle 10 and the Tenth Refuel-ing Outage (RFO) which concluded on June 2, 1996. Revision 20 to the USA 7 also updates the Fire Hazards Analysis Report (FHAR) which is incorporate by reference into USAR Section 9.5.1, Fire Protection Program. This update transmits Revision 16 to the FHAR.
This submittal also includes the DBNPS Technical Requirements Manual (TRM).
The TRM contains requirements that have been relocated from the Operating License, Appendix A, Technical Specifications, in accordance with NRC-approved License Amendments. The TRM in incorporated by refc.:ence into the USAR Section 1.5.5, Davis-Besse Controlled Documents.
A summary of the major changes made in the USAR, FHAR and TRM can be found in Attachments 1, 2 and 3, respectively.
This submittal also reports changes to the DBNPS Quality Assurance Program l
In accordance with 10 CFR 50.54(a). provides a brief summary of the changes made in this revision.
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' Licznza Number NPF-3 Serial Number 2423 Page 2 The Cycle 10/10RFO USAR Update Program was modeled after the previous cycle's update program. Prior to the submittal of this USAR revision, it was reviewed by DBNPS personnel that are responsible for the information contained within their assigned USAR section. The Cycle 10/10RFO reviews j
were performed with increased emphasis by management on the detai2 ed review of each USAR section by these responsible " cognizant units".
During this review by the cognizant units, specific attention was provided to the review of operational and design aspects of systems and components described in the USAR to reasonably assure the text, tables and figures of the USAR accurately depict the facility and its operation.
Several items were identified during this review that are undergoing further evaluation for incorporation, if appropriate, into the USAR.
As these items are resolved, any resulting USAR changes will be processed as described in DBNPS Procedure NG-NS-00806, " Preparation and Control of USAR Changes," and will be available for reference and use by personnel performing 10 CFR 50.59 safety review / evaluations.
Information contained in these revisions to the USAR, FHAR and TRM is up-to-date as of June 2, 1996 in accordance with the requirements of 10 CFR 50.71 (e)(4).
Please insert the Revision 20 material, dated December 1996, into the USAR and FHAR per the attached Listing of Effective Pages. At this time, Toledo Edison is also transmitting En information copy of the TRM to be located with the information copy of the USAR.
Should you have any questions or require additional information, please contact Mr. James L.
Freels, Manager - Regulatory Ait' airs, at (419) 321-8466.
Very truly yours, JMM/laj Enclosure Attachments cc:
A. B. Beach, Regional Administrator, NRC Region III A. G.
Hansen, DB-1 NRC/NRR Project Manager w/o attachment S.
Stasek, DB-1 NRC Senior Resident Inspector Utility Radiological Safety Board w/o attachment 1
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Dockst Number 50-346
- Licznce Number NPF-3 Sarial Number 2423 Enclosure SUBMITTAL OF REVISION 20 TO THE DAVIS-BESSE UPDATED SAFETY ANALYSIS REPORT FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Enclosed are the original and 10 copies of Revision 20 to the Davis-Besse Nuclear Power Station, Unit Number 1, Updated Safety Analysis Report.
By:
J. K.
od, Vice President - Nuclear Sworn to and subscribed before me this 2nd day of December, 1996.
Y Notary Publ'ic, State of Ohio 1
LAURA A. JEtettSON Notary Public, State of Ohio My Commission Empires 8 15-2001
Dockst Number 50-346 Licznsa Number NPF-3
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SUMMARY
OF USAR REVISION 20 MAJOR CHANGES Section 1 Clarifications were made throughout Section 1.2, General Station Description, to more accurately reflect system descriptions contained in other sections of the USAR.
Section 1.2.7, Fuel Handling and Storage, was updated to reflect the addition of the Dry Fuel Storage Facility.
i Several aanges were made to the referenced drawings in Section 1.5, Material Incorporated By Reference. These changes reflect modifications that were made in previous years. Also in Section 1
1.5, two listed drawings that were unique to the USAR were replaced with station controlled drawings.
Section 2 Section 2.1, Geography and Demography, was revised to update area population estimates. Current population estimates are contained in the Davis-Besse Evacuation Time Estimates.
Section 2.2, Nearby Industrial, Transportation, and Military Facilities, and 2.3, Meteorology, were revised to add a brief discussion of the effects, on the operation of the station and the meteorological towers, of a new water tower that was constructed adjacent to the site.
Section 2.2.3.6, On-Site Facilities, was updated to reflect the 2,000 gallon Station Blackout Diesel Generator Fuel Oil Storage Tank.
Section 2.2 and Appendix 2A were revised to reflect the cessation of anti-aircraft training firing at a nearby military j
training installation and updates the discussion of ordnance i
test firing.
Section 2.4, Hydrclogy, was revised to reflect the current plant configuration of identifying all electrical duct banks that enter safety related buildings below elevation 575 feet as having waterproof membrane in lieu of identifying duct banks at elevation 577' 10" as being protected by waterproof membrane.
The reference to certain site elevations, and marsh acreage descriptions were revised for consistency in Section 2.3, Meteorology, and Section 2.4, Hydrology.
Sections 2.2, 2.4, and 2.5, Geology and Seismology, were revised to update agricultural and industrial adjacent land usage.
Surface and groundwater usage was also updated to reflect post-construction information.
Dockst Number 50-346 Licsnsa Numbcr NPF-3 Sgrial Numb 3r 2423 Atttchm:nt 1 Page 2 of 10 Section 3 Numerous changes were made throughout Section 3 to clarify the description of the plant design relative to flooding and High Energy Line Break analysis.
Added a discussion of the Dry Fuel Storage Facility (DFSF) to Section 3.8, Design of Seismic Class I and Class II Structures.
Revised Section 3.8 to reflect the removal of the Annulus sand pocket from the containment vessel base embedment and applied a protective coating on the steel containment vessel.
Section 3.9, Mechanical Systems and Components, was revised to reflect a revision to the Main Steam Safety Valve seismic qualification.
Section 3.9 was also revised to reflect the seismic qualification of the replacement valves for decay heat valves j
DH13A, DH13B, DH14A and DH14B.
Revised Section 3.10, Seismic Design of Category I j
Instrumentation and Electrical Equipment, to add a statement j
which clarifies that the qualification programs described are for the original complete piece of equipment, not replacement i
parts. Replacement parts would be qualified in accordance with the appropriate standard, at least IEEE 344-1971, and to the acceleration levels required for the application.
l Section 3.10 was also revised to reflect the qualification of the new inverters installed in the station.
Section 3.11, Environmer tal Design of Mechanical and Electrical Equipment, and Section 3D.1.23, Criteria 27 - Combined Reactivity Control Systems Capability, were revised to reflect the increase in boron concentration in the Borated Water Storage Tank (BWST) to reflect Operating License (OL) Amendment 207.
Clarifications were made throughout Section 3 and Appendix 3D to more accurately reflect system descriptions and testing contained in other sections of the USAR.
Section 4 Added reference to the end of cycle T reduction, in Section 4.1, Summary Description, and Table 4 4, Thermal-Hydraulic l
Core Design Summary.
Updated Appendix 4B, Reload Report, to reflect Cycle 11 fuel loading.
Section 5 Several subsections in Section 5, Reactor Coolant System, were revised to include a reference to the end of cycle T e
reduction.
i Several tables were changed to reflect a modification to the Pressurizer Pilot Operated Relief Valve (PORV) to alleviate the i
potential for pressure locking.
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Page 3 of 10 Table 5.1-5, Steam Generator Design Data, was changed to add the hydrostatic test limits, which was relocated from Technical Specifications per OL Amendment 204.
Table 5.1-8, Transient Cycles - 40 Year Design Life, was modified to add several test transients that were relocated from Technical Specifications per OL Amendment 204.
Section 5 was also revised to reflect OL Amendment 199 which extended the pressure-temperature (P-T) limits from 10 effective full power years (EFPY) to 21 EFPY.
l Section 5.2.2.3, Overpressure Protection, was revised to provide a more complete and accurate description of the Decay Heat isolation valve and pressurizer interlocks.
1 Section 5.2.3.3, Material Compatibility With Insulation and Environment, was revised to add 'Nukon' as an insulating material used on reactor coolant pressure boundary piping.
Revised Section 5.2.3.4, Reactor Coolant Additives, to include hydrogen peroxide as a chemical that may be added to the reactor 4
coolant during shutdown to initiate forced oxidation for source term reduction.
Section 5.4.2, General Description, of the Reactor Vessel and Appurtenances was modified to eliminate reference to using the pressure tap in the annulus between the reactor pressure vessel head to flange o-rings for hydrotesting the outer o-ring seal after head closure.
Section 5.5.1, Reactor Coolant Pumps, was revised to clarify l
that Reactor Coolant Pump oil drip pans have a common drain line to the oil drain collection tanks vice indiviiual lines.
The description of the Steam Generators in Section 5.5.2 was revised to reflect the operation of the Steam Generators with fouling consistent with OL Amendment 192.
i Added a description of Steam Generator tube pulling in Section j
5.5.2.
j Revised Section 5 to reflect the current core limits for the Cycle 11 Core.
Section 6 Section 6.2.1.3.2, Containment Pressure Transient Analysis Break Spectrum, was revised to clarify that the design allowable external pressure differential for service levels A and B of ASME Code, is 0.67 psi and 0.8 for service level C.
Section 6.2.1.4, Testing and Inspection, was revised to reflect OL Amendment 205, which adopts the. provisions of 10CFR50, Appendix J, Option B, for Type A cantainment leakage testing.
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Section 6.2.2.6.3, Inspection and Test, for the Containment i
Vessel Emergency Sump was revised to reflect that the emergency sump isolation valves are not required to be leak tested per 10CFR50, Appendix J.
Revised several Sections to reflect OL Amendment 207 which increased the boron concentration of tha BWST and Core Flooding Tanks from 2100 ppm to 2600 ppm.
Section 6.2.3, Containment Vessel Air Purification and Cleanup Systems, was revised to reflect OL Amendment 209 which clarified f
the testing requirement and updated the regulatory and industry guidance references for charcoal adsorber units.
Revised Section 6.2.3 to reflect the peak annulus pressure of j
0.8 psig.
Revised the description of the Hydrogen Purge System charcoal filter testing to make it consistent with OL Amendment 155, which clarified the testing requirements for High Efficiency Particulate Filters and charcoal adsorber.
Table 6.2-23, Containment Vessel Isolation Valve Arrangements, was revised to reflect a modification which removed relief f
valves that were installed at each Containment Air Cooler Service Water return line and replaced them with pressure taps.
i Section 6.3.1.4, System Short-and Long-Term Capability, and i
Table 6.3-6, Single Failure Analysis - ECCS, were revised to reflect an action taken to address 10CFR50 Appendix R concerns regarding hot shorts. Specifically, Motor-Operated Valve HP31 has its power removed by changing its breaker from normally closed to normally open.
In Section 6.3.3.1.2, Results of Analysis (Large Break), two paragraph were added. The information contained in these paragraphs were originally submitted to the NRC, as a response to an NRC question, during initial licensing of the DBNPS.
1 The first paragraph summarizes sensitivity studies which were performed for axial flux peaks at different core elevations.
The second paragraph added information regarding the conse-quences of a LOCA during startup or shutdown while the Core Flooding Tanks are isolated.
Information was also added in Section 6.3.3.1.2, regarding Peak Cladding Temperature acceptance criteria.
In Section 6.3.3.2.4, Minimum Conditions of ECCS, the volume of the Core Flooding Tanks was revised to reflect the volume used in the latest LOCA analysis.
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- Snrial Numbsr 2423 Page 5 of 10 Section 6.3.3.2.5, Provisions to Protect the ECCS, was revised to include information originally submitted to the NRC during initial licensing, regarding the systems ability-to maintain water in the ECCS lines.
In Section 6.3.4, Test and Inspections, and Table 6.3-5, Relief Valves in ECCS, the manufacturer specific capacity of relief valves CF7A and CF7B was replaced with design basis requirements.
i A modification which replaced relief valve DH1509 was also reflected in Section 6.3.4 and Table 6.3-5.
Table 6.3-6, Single Failure Analysis - Emergency Core Cooling
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Systen was revised to indicate that valves DH81 and DH82 preve s backflow to the BWST if a single failure of valve DH7A or DH7B should occur. This information was originally submitted to the NRC during initial licensing.
i Table 6.3-9, ECCS Component Testing Procedure and Frequency, was deleted since the information is addressed in the USAR text.
Section 7 Section 7.2, Reactor Protection System (RPS), was revised to make the text more accurately reflect the USAR figures, Technical Specifications and the system operation.
Section 7.3, Safety Features Actuation System (SFAS), was revised to reflect a modification which removed the digital
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indicator originally used for calibration of the SFAS trip bistables.
l Compliance with AEC Safety Guide 22, Section 7.3.2.6, and Table 7.3-2, Periodic Test on SFAS and Actuated Equipment, were revised to make the USAR description accurately reflect the channel functional testing that is performed in accordr ce with Regulatory Guide 1.22.
Table 7.3-1, Sequence Logic, was deleted since this information is also depicted on USAR Figure 7.3-1.
Section 7.4.1.3, Steam and Feedwater Line Rupture Control System (SFRCS), was revised to make the text more accurately reflect the USAR figures, Technical Specifications and the system operation.
Section 7.4.1.3.10, Design Basis, was also revised to reflect a setpoint change to the SFRCS high level trip.
l Anticipatory Reactor Trip System (ARTS), Section 7.4.1.4, was revised to make the text more accurately reflect the USAR figures, Technical Specifications and the system operation.
Dockst Number 50-346 Lic&nta Number NPF-3
~ Sarial Number 2423 Page 6 of 10 Section 7.4.1.4 was also re rised to reflect an action that allows the resetting of the second Main Feedwater (MFW) pump turbine in Mode 1 while maintaining ARTS trip capability for loss of MFW pumps.
Section 7.6.1.1, Normal Decay Heat Removal Valve Control System, was revised to provide a more complete and accurate description of the Decay Heat isolation valve and pressurizer interlock.s.
Section 7.7.1.4, Turbine Generator Electro-Hydraulic Controls (EHC), was revised to reflect procedure changes that increased the power limitation for testing the Main Turbine Stop and Control Valve testing to less than 93 percent power and approximately 96 percent power respectively.
Station Computer System, Section 7.10, was modified to accurately depict the plant computer resolution time of 5 milliseconds for recording sequence-of-events points.
Section 7.13.3.12, Steam Generator Start-up Range Level l
Indicators, was revised to reflect the recalibration of the SFRCS Steam Generator Startup Range level instruments.
Section 8 Section 8.3.1.1.3, 4160 Volt Auxiliary System, was modified to 1
reflect a plant change to the interlocks that prevent more than one component cooling pump and one service water pump from automatically connecting to either emergency diesel generator at the same time.
The discussion of the Emergency Diesel Generators was modified to update, correct, and reorganize information.
Information was added which discusses the major components of the exhaust system and operation of the turbocharger.
Revised Section 8.3.1.1.6, 120V AC Auxiliary System, to reflect the replacement of remaining Cyberex inverters with Solidstate Controls Inc. inverters.
Information was added to Section 8.3.1.2.14, Cables and Raceway Functions, to clarify color coding for cable trays, wireways and conduits.
Section 8.3.2, DC Power, was modified to provide clarification regarding essential rectifier output voltage and flow of DC current to the essential inverters and added a description of eight indicating lights mounted on the main control board.
Table 8.3-2, Administratively Controlled Circuits (AC), was revised to reflect a modification which added the capability of supplying containment lighting loads from 480VAC non-1E sources.
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Licznsa Numb 2r NPF-3 S3 rial Numbsr 2423 Page 7 of 10 Table 8.3-4, Environmental Requirements for Class IE Cable, was revised to reflect new analysis performed to support increasing the boron concentration in the BWST.
Section 9 Added an additional valve to dua listing of Service Water valves that were upgraded utilizing the provisions of Generic Letter 89-09, ASME Section III Component Replacements.
Several discussions of the Dry Fuel Storage Facility were added to Section 9.1.2, Spent Fuel Storage.
Corrected the dimensions of the opening between the spent fval pool and the cask pit area in Section 9.1.2.
In Section 9.1.3, Spent Fuel Pool Cooling and Cleanup System, the discussions of the spent fuel pool cleanup system and the use of the Decay Heat Removal system for spent fuel pool cooling were clarified.
Section 9.1.4.2.2, Handling Equipment, and Section 9.1.4.2.3, Loading and Removing Fuel, were rewritten to more accurately depict the fuel handling equipment and their fuel handling capabilities and refueling activities.
A paragraph was added to Section 9.1.4.2.3 to discuss fuel repairs.
Section 9.1.4.4, Safety Provisions, was modified to reflect OL Amendment 207 which increased boron concentration. The revision to this Section ensures the refueling canal and storage pool boron concentration will be high enough to ensure < 0.95 keff*
l Information was added to Section 9.2.1, Service Water System, to allow bypassing service water through the spare Component Cooling Water heat exchanger during cold weather operation.
Section 9.2.2, Component Cooling Water System, was revised to remove superfluous information and to add additional information on the reactor coolant pump bearing internal heat exchangers overpressure protection.
Section 9.2.5, Ultimate Heat Sink, was revised to clarify the discussion of DBNPS compliance with Safety Guide 27.
Section 9.2.6, Condensate Storage Facilities, was revised to more accurately reflect the system design and operation.
Section 9.3.1, Station and Instrument Air System, was also revised to more accurately reflect the system design and operation.
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Dockst Number 50-346 Licznto Numb:r NPF-3 Srriel Numbar 2423 Attachm:nt 1 Page 8 of 10 Table 9.3-2, Station and Instrument Air Control Room Alarm i
Setpoint, and Section 9.3.1.6.1, Alarm Setpoints, were deleted since setpoints are more efficiently controlled by other plant programs.
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Section 9.3.2.1, Process Sampling Systems, was revised to i
clarify secondary sampling locations.
Section 9.3.3, Equipment and Floor Drainage System, was i
rewritten to more accurately reflect the design and flow of the station drains.
j In Section 9.3.3.2, Post-LOCA Sump pH Control, a change was made l
to reflect the addition of new Trisodium Phosphate baskets to support the increase in boron concentration.
Makeup and Purification System, Section 9.3.4, was revised to allow, at reduced RCS pressure, all three parallel letdown valves be opened at the same time to allow letdown flow be j
maintained at 140 gallons per minute.
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Section 9.3.4 was also modified to state that Purification Demineralizer number three can be used as either a mixed or cation resin bed, i
Section 9.3.6, Chemical Addition System, was revised to add hydrogen peroxide as a chemical that may be added to the reactor coolant to achieve forced oxidation after shutdown for source term reduction.
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A procedure change that allows concentrated boric acid from the l
Boric Acid Addition Tanks to be pumped through existing lines j
that do not have heat tracing to the Clean Waster Receiver Tank, was added to Section 9.3.6.
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Section 9.3.6.3.2, Malfunction Analysis, was revised to reflect the new Chemical Addition Malfunction analysis which was performed to support the increase in the minimum boron concentration in the BWST.
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Tables 9.3-4, Reactor Coolant Quality and 9.3-5, Steam Generator Feedwater Quality, were revised to incorporate current industry guidelines.
l Table 9.3-13, Chemical Addition System rquipment Data, was i
revised to accurately reflect the dec1gn of the equipment.
In Section 9.4, Air Conditioning, Heating, Cooling, and Ventilating Systems, the description of the Control Room Emergency Ventilation System charcoal filters was revised to make it consistent with OL Amendment 155, which updated the references and clarified the testing requirements for High Efficiency Particulate Filters and charcoal adsorbers.
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i Low Voltage Switchgear Rooms, Section 9.4.2.1.2.2, was revised to include a paragraph discussing the ventilation of the electrical isolation rooms located within each low voltage j
switchgear room.
Section 9.4.3, Radwaste Area, was revised to reflect a modification which abandoned an auxiliary building non-essential i
humidifier.
I Information was added to Section 9.5.3.2, AC Emergency Lighting, L
to reflect a modification which added the capability of supplying containment lighting loads from 480VAC non-lE sources.
l Section 9.5.3.1, Normal Station and Security Lighting, was I
l revised to clarify switchyard, security and hand-held lighting i
systems used at the plant.
l Information in Section 9.5.7, Diesel Generator Lubrication l
System, was relocated to Section 8.3.1.1.4.1, Emergency Diesel Generators, to provide a more complete description of the EDGs f
in one section of the USAR.
Section 10 Discussions of Turbine overspeed setpoints were clarified in Section 10.2.4.2, Turbine Overspeed Control.
A paragraph was added to Section 10.2.5, Turbine Missile Protection, discussing the use of probabilistic calculations for determining the likelihood of generating a turbine missile.
A reference to an inaccurate design code was removed in Section 10.4.1, Main Condenser.
Sections 10.4.1, Main Condenser, and 10.4.6, Condensate Demineralizer System, were clarified to accurately discuss the operation, monitoring and chemistry of the condensate.
Section 11 Sections 11.1, Source Terms, 11.2, Liquid Waste Systems, 11.3, Gaseous Waste System and 11.5, Solid Waste System were revised to denote material which is now considered to be historical 1
design information, and to better reflect current plant l
operating characteristics.
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Table 11.2-1, Table 11.2-2, and Table 11.3-1 which list j
radioactive processing equipment, were revised to accurately reflect the design of the equipment.
Section 12 Added a statement to clarify what is represented by the data contained in the radiation zone Figures 12.1-1 through 12.1-9.
i In Section 12.1.3.4, Radioactivity Stored Outside, a statement j
was added regarding the Dry Fuel Storage Facility (DFSF).
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Dockst Number 50-346 Licensa Numbar NPF-3 Ssrial Number 2423 Page 10 of 10 Section 12.3.2.2.2, Counting Equipment for Radioactivity Measurements, was revised to reflect the use of new counting equipment.
Section 13 Information contained in Section 13.7, Security, was removed and replaced with reference to the Security Plans.
Section 15 Added reference to a computer code used in an accident analyses, to Table 15.1-1, Digital Computer Programs / Analog Simulations.
The Pressurizer level was added to Table 15.1-2, Parameters Applicable to All Accidents in the Accident Analysis.
Section.15.2.8, Loss of Normal Feedwater, was revised to include reference to the analysis and results which allowed the pressurizer normal level to be increased to 220 inches.
Table 15.4.3-2, Nominal Values of Input Parameters for Control Rod Assembly Ejection Accident Analysis, was modified to include reference to the end of cycle T reduction.
ave Section 15.4.6.5, Effects of Engineered Safety Features Leakage During the Maximum Hypothetical Accident, was revised to reflect OL Amendment 195 which removed the ECCS and Containment Spray systems specific leak testing requirements for piping located outside of containment.
A postulated dry fuel storage cask drop accident was added to Section 15.4.7, Fuel-Handling Accident.
A new section was added to Section 15.4.7 to reflect OL Amendment 202 which allows both doors of the containment personnel air lock to be open during core alterations or
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movement of irradiated fuel within containment.
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Dockst Number 50-346 Licanza Numb 3r NPF-3 Ssrial Number 2423 Page 1 of 1 ATTACHMENT 2
SUMMARY
OF FHAR REVISION 16 MAJOR CHANGES i
I Section 2.
Updated several references relating to Information Notice (IN) 92-18, Potential For Loss Of Remote Shutdown capability During A Control Room Fire, and Dry Fual Storage.
Section 3 Made changes relating ~to the Seal Injection, Seal Return and Seal Cooling based on resolution of IN 92-18.
Changed wording l
concerning the increase in boron concentration and to reflect the Station Blackout Diesel Generator.
Section 4 Made various changes due to IN 92-18 and the incorporation of plant modifications, drawing changes and Potential Condition Adverse to Quality Reports (PCAQRs). The changes include reflecting revised circuit numbers, deleting sub-components, and revising the associated notes.
Section 8 Incorporated changes relating to the surveillance frequencies based on the reviews of past performances and made other minor changes.
Appendix A Made various small changes due to the incorporation of plant modifications, drawing changes and PCAQRs. The type of changes included adding new circuit numbers to reflect revised cable routings.
Appendix B-1 Made various small changes due to the incorporation of plant modifications, drawing changes and PCAQRs. The type of changes include revised raceway identifications and circuit numbers.
Appendix B-2 Made various small changes due to the. incorporation of plant modifications, drawing changes and PCAQRs. The type.of changes j
include revised raceway identifications and circuit numbers.
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Appendix C-1 Made various small changes due to the incorporation of plant modifications, drawing changes and PCAQRs. The type of changes included adding new circuit numbers to reflect revised cable routings.
Appendix C-2 Made various small changes due to the incorporation of plant modifications, drawing changes and PCAQRs. The type of changes included adding new circuit numbers to reflect revised cable routings.
Appendix C-3 Made various small changes due to the incorporation of plant modifications, drawing changes and PCAORs. The type of changes include changing load descriptions and circuit numbers.
4 Appendix D Incorporated changes relating to the surveillance frequencies based on the reviews of past performances and made other minor changes.
Dochet Number 50-346 Lic*nse Number NPF-3 Ssrial Number 2423
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SUMMARY
OF TRM MAJOR CHANGES The Davis-Besse Technical Requirements Manual (TRM) was created in February 1996. The first items relocated to the TRM were removed from the Technical Specifications per OL Amendment 201.
Sections 3/4.3.3.3, Seismic Instrumenta-tion, 3/4.3.3.4, Meteorological Instrumentation, and 3/4.4.11, Reactor Coolant System Vents and their associated bases were transferred to the TRM.
The requirements were relocated in their entirety, except for the special reporting requirements for inoperable seismic and meteorological monitoring instrumentation.
Amendment 206, which allowed a seven day outage time for the Emergency Diesel Generators (EDG), was the basis for the second addition to the TRM.
Section 3/4.8.1, A.C. Sources - Operating, was created to verify that the alternate AC power source will be capable of being connected to the safety bus associated with an inoperable EDG, and will verify this capability of being able to connect to the safety bus every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
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Dock 2t Numb:r 50-346 Licanze Number NPF-3 Serial Numbsr 2423 Page 1 of 3 ATTACHMENT 4 CHANGES TO QUALITY ASSURANCE PROGRAM The organization described under Section 17.2.1, Organization, underwent numerous changes. These changes were determined not to be reductions in commitment to the Quality Assurance Program. The following is a description of the changes made to USAR Section 17.2.1.4, Toledo Edison Nuclear Group In June of 1995, Design Engineering and Plant Engineering were consolidated into a single section, Plant Engineering. This consolidation stremmlined the Engineering organization, realigned functional groups into a more effective organization and reduced inter-departmental barriers, thus improving communication, work flow and effectiveness.
In January of 1996, the responsibility for advising cognizant management of Nuclear Quality Assurance effectiveness was transferred from the Director -
Nuclear Assurance to the Manager - Quality Assessment.
1 In March of 1996, responsibility for budgeting and cost control and nuclear i
project management was transferred from the Manager - Nuclear Support to a new position of Manager - DB Business Services. This change facilitates the creation of a Centerior Supply Department. This change did not eliminate i
any activities previously performed and did not introduce issues regarding organization independence.
1 In May of 1996, the responsibility for quality assurance review of procedures was transferred to the Manager - Quality Assessment from the Director - Nuclear Assurance.
In June of 1996, the Manager - Nuclear Support was renamed Manager - DB Supply. At this time the descriptions of DBNPS corporate and non-DBNPS corporate hierarchy positions were updated to reflect the formation of the Centerior Energy Power Generation Group. Corporate groups providing services to DBNPS were eliminated and the functions were reflected under the responsibilities of the appropriate DBNPS management described in USAR 17.2.
l Any safety-related or quality related services provided to DBNPS continued j
to be provided by secunded or matrixed personnel directly reporting to the l
DBNPS Nuclear Group or by contracted organizations controlled under the l
DBNPS' procurement program as vendors.
In July of 1996, the Engineering Department was restructured. The Nuclear and Plant Engineering Sections were realigned into Design Basis Engineering and Plant Engineering, respectively. Design Basis Engineering is responsible for design basis functions such as facility design, safety analysis, nuclear fuel, PRA analysis, nuclear and simulator engineering and site computer systems. Plant Engineering is responsible for day-to-day direct plant support functions to ensure optimum system performance and l
Dockat Numbor 50 346 Lican2s Numb 0r NPF-3 Serial Number 2423 Page 2 of 3 reliability such as thermal performance monitoring, ISI and IST activities, MOV analysis, ASME Code functions, modification and special testing, predictive maintenance, chemistry analysis, etc.
Plant Engineering is also responsible for the FHAR.
The organizational changes described above do not reduce the effectiveness of any program, do not reduce oversight or reviews, do not eliminate activities commitments or functions, and do not add non-quality assurance functions to the Quality Assessment Section previously described in the Quality Assurance Program Description. As such, these changes do not reduce commitments to the previously approved Quality Assurance Program Description. These changes do not alter or negate DBNPS's commitment to provide and describe the organizations responsible for establishment and implementation of the Nuclear Quality Assurance Program.
The above changes continue to satisfy the requirements of 10CFR50, Appendix B.
Other non-reduction changes made to Section 17.2 include:
The following sections were revised to add reference and discussions related to the implementation of dry spent fuel storage activities licensed under 10CFR72 through the provisions of the DBNPS 10CFR50, Appendix B Quality Assurance Program.
Section Title l
17.2.2 Quality Acsurance Program, 17.2.3 Design Control, 17.2.5 Instruction, Procedures and Drawings, 17.2.6 Document Control, 17.2.15 Nonconforming Materials, Parts, or Components, 17.2.16 Corrective Action and 17.2.17 Quality Assurance Records.
The specific changes include:
1.
Introduce dry spent fuel storage as an activity that is subject to the provisions of the DBNPS 10CFR50, Appendix B Quality Assurance Program.
2.
Discuss the safety review and evaluation requirements as related to dry spent fuel storage in accordance with 10CFR72.48.
3.
Discuss the dry cask storage ver. dor's graded quality classification system as it is implemented through the existing DBNPS 10CFR50, Appendix B quality classification system.
4.
Include dry spent fuel storage documentation and records requirements.
5.
Discuss nonconformance reporting and immed.iate notification requirements for dry spent fuel storage activities.
The addition of references and discussions relatta to dry spent fuel storage activities within the USAR is intended to satisfy the requirement for utilizing a Quality Assurance Program that meets the criteria of 10CFR72, Subpart G.
The j
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Page 3 of 3 existing text of the USAR that satisfies 10CFR50, Appendix B requirements remains unchanged.
Section 17.2.1, organization, and Section 17.2.2.5, Program Review, were revised to modify the basis of the Vice President - Nuclear's review of Nuclear Quality Assurance Program effectiveness from periodic (or annual) to continuous. This change more accurately depicts the process used for this review. The processes I
used for this review include: program document approvals, receipt of audit l'
reports and other status reports, minutes of meetings with the CNRB and the I
Manager - Quality Assessment, documented involvement in resolving significant adverse conditions and resulting follow-up actions. This change continues to satisfy the requirements of 10CFR50, Appendix B.
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