ML20197G499
| ML20197G499 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/23/1998 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20197G504 | List: |
| References | |
| 2574, NUDOCS 9812070274 | |
| Download: ML20197G499 (23) | |
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Davis-Besse Nuclear Power Statron 5501 North State Route 2 m
Oak Harbor. Ohio 43449-9760 John K. M 419 249 2300 Vc2 President-Nuclear raw 419 321.g337 Docket Number 50-346 License Number NPF-3 l
Serial Number 2574 November 23, 1998 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555-0001
Subject:
Revision 21 to the Davis-Besse Nuclear Power Station (DBNPS) Unit 1 Updated Safety Analysis Report (USAR)
Ladies and Gentlemen:
1 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 21 to the DBNPS Updated Safety Analysis Report (USAR).
Revision 21 to the USAR reflects facility changes implemented between June 2,1996 and May 23,1998. This USAR revision includes facility changes that occurred during operating fuel Cycle 11 and the Eleventh Refueling Outage (11RFO) which concluded on May 23,1998.
Revision 21 to the USAR also updates the Fire Hazards Analysis Report (FHAR), which is incorporated by reference into USAR Section 9.5.1, Fire Protection Program. This update transmits Revision 17 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 is incorporated by reference into 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.
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Docket Number 50-346 l
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Serial Number 2574 l
Page 2 This submittal also reports, in accordance with 10 CFR 50.4(b)(7), changes to the DBNPS Quality Assurance Program completed under 10 CFR 50.54(a)(3). Attachment 4 provides a brief summary of the changes made in this revision.
The DBNPS Cycle 11/11RFO USAR Update Program was modeled after the previous cycle's update program. Prior to the submittal of this revision, the USAR was reviewed by the DBNPS staff that are responsible for the information contained within their assigned USAR sections.
The Cycle 11/11RFO USAR reviews were performed with increased emphasis on topical / system reviews of the USAR by the cognizant units. During this review by the cognizant units, specific attention was directed to the review of operational and design aspects of systems and components described throughout the USAR to reasonably assure the text, tables and figures of the USAR accurately depict the facility and its operation.
In addition to the cognizant unit reviews, selected sections of the USAR were reviewed by Operations Senior Reactor Operator (SRO) licensed individuals. During this review, the selected USAR sections were reviewed for conformance with the as-built facility, operating procedures and drawings.
Several items were identified during these reviews 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 DBNPS staff performing 10 CFR 50.59 safety review / evaluations.
The DBNPS staff conducted detailed USAR reviews in preparing USAR Revision 20 (Cycle 10/10RFO) and USAR Revision 21 to provide reasonable assurance that the facility as described in the USAR is consistent with the as-built facility, its operation, and procedures. Identified discrepancies have been reviewed as potential changes to the facility under the DBNPS procedure NG-EN-00304," Safety Review and Evaluation". No unreviewed safety questions have been identified.
Information cor.tained in these revisions to the USAR, FIIAR and TRM is up-to-date as of May 23,1998 in accordance with the requirements of 10 CFR 50.71 (e)(4). Please insert the Revision 21 material, dated November 1998, into the USAR and FIIAR per the attached Listing of Effective Pages. At this time, Toledo Edison is also transmitting an information copy of the TRM to be located with the information copy of the USAR.
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Dockct Number 50-346 License Number NPF-3 l
Serial Number 2574 Page 3 Should you have any questions or require additional information, please contact Mr. James L.
Freels, Manager - Regulatory Affairs, at (419) 321-8466.
Very truly yours, JMM/JCS/laj Enclosure Attachments l
cc: J. L. Caldwell,(Acting) Regional Administrator, NRC Region III A. G. Hanst,. DB-1 NRC/NRR Project Manager, w/o attachment S. J. Campbell, DB-1 NRC Senior Resident Inspector Utility Radiological Safety Board, w/o attachment
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c Docket Number 50-346 l-License Number NPF-3 Serial Number 2574 Enclosure i
SUBMITTAL OF REVISION 21 l
TO i
i THE DAVIS-BESSE UPDATED SAFETY ANALYSIS REPORT l
l FOR l
DAVIS-BESSE NUCLEAR POWER STATION l
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UNIT NO.1 l
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l Enclosed are the original and 10 copies of Revision 21 to the Davis-Besse Nuclear Power Station, Unit Number 1, Updated Safety Analysis Report.
r I, John K. Wood, state that (1) I am Vice President - Nuclear of the Centerior Sen ice Company, (2) I am duly authorized to execute and file this certification on behalf of the l
Toledo Edison Company and The Cleveland Electric Illuminating Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.
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By:
J. K. Wo/d, Vice President - Nuclear L
I Affirmed and subscribed before me this 23rd day of November, 1998.
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l Docket Number 50-346 License Number NPF-3 Serial Number 2574 l -
Page1of11 l
SUMMARY
OF USAR REVISION 21 MAJOR CHANGES Section 1 Clarifications were made throughout Section 1.2, General Station l
Description, to more accurately reflect system descriptions contained in other l
sections of the USAR.
Section 11.1, General (Identification of Agents and Contractors), was revised ta reflect the merger of the Centerior Energy Corporation (CEC) and Ohio Edison and the formation of a new holding company, the FirstEnergy Corporation. Toledo Edison Company (TE), Cleveland Electric Illuminating Company and Centerior Service Company (CSC) are subsidiaries of FirstEnergy.
Section 1.5.1, Reactor / Fuel Vendor Reports, was modified to include reference to topical report BAW2303P, OTSG Repair Roll Qualification l
Report. This report was used as the technical basis for Technical Specification Amendment 220, which allowed DBNPS to implement an alternate method (reroll) for Steam Generator tube repair.
Two drawings were added to the Controlled Drawing listing in Section 1.5, j
Material Incorporated By Reference. These drawings are continuation sheets j
of drawings currently listed in the USAR. A drawing was also added to this section which depicts the Auxiliary Feedwater Pump Turbine (AFPT) Main Steam minimum flow lines.
l Section 2 Section 2.1.2.1, Exclusion Area Control, and Figure 1.2-12, Site Plan, were revised to clarify the site boundary for exclusion area control. This section was also revised fo show a modified flow path to the Toussaint River that occurred when the dike along the river was rebuilt.
Section 2.2, Nearby Industrial, Transportation, and Military Facilities, was revised to include the results of new studies that were conducted for Control Room habitability.
The Barometric Pressure and Solar Incidence parameters were removed from Table 2.3-8, Meteorological Sensor Locations and Specifications, since DBNPS is not committed or required to monitor these parameters.
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4 Dockc1 Number 50-346 License Number NPF-3 Serial Number 2574 -
Page 2 of11 Clarifications were made in Section 2 to more accurately reflect the physical site description.
Section 3 Section 3.6.2.7.1.5, Main Steam to the Auxiliary Feed Pump Turbines, was revised to reflect a modification that installed Auxiliary Feedwater Pump Turbine (AFPT) Main Steam minimum flow lines. These lines run from the AFPT Main Steam supply lines to feedwater heater E6-2.
Section 3.6.2.7.2.13, Circulating Water System, was revised to reflect the installation oflevel switches in the condenser pit to detect flooding in the condenser pit area. These switches will provide an automatic trip of the circulating pumps and the closure of their discharge isolation valves, thus isolating the break from the circulatin;; water canal.
I Several clarifications were made to Section 3.7, Seismic Design, to reflect the plant and other sections of the USAR.
Section 3.8.2.1.9, Post-Operational Testing and Inspection, was revised to reference the DBNPS coramitment to ASME Section XI,1992 Edition,1992 Addenda, for the inservice inspection of the Containment Vessel.
Section 3.9.2.11.3, Decay Heat Coolers, was revised to reflect a seismic reanalysis of Decay Heat Coolers and the attached piping systems. This j
reanalysis is a result of the Seismic Qualification Utility Group Program i
review at DBNPS.
l A new code case was added to Table 3.9-10, Code Cases. Code Case N-496, Helical-Coil Threaded Inserts,Section XI, Division 1, was added for a modification to the Main Steam Isolation Valve (MS-100) body to bonnet stud holes.
Section 3.11.1, Equipment Qualification Program, was revised to reflect the results of analysis for the conversion to a 24 month fuel cycle.
Clarifications were made throughout Section 3 and Appendix 3D to more accurately reflect system descriptions and testing contained in other sections of the USAR.
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i Docket Number 50-346 License Number NPF-3 Serial Number 2574 Page 3 of 11 Section 4 Section 4.3.5.3, Control Rod Drive Control System Instrumentation (CRDCS), was revised to remove a statement regarding actions taken when i
the Control Rod Drive Position Indication meters are inoperable. These changes were made to accurately reflect the requirements of the DBNPS Technical Specifications.
Core Operational Transients, Section 4.4.2.9, was revised to reflect i
Amendment 222 of the Technical Specifications. This Amendment modified the actions taken for reactor coolant system (RCS) degraded flow.
Appendix 4B, Reload Report, was updated to reflect Cycle 12 fuel loading.
Other minor clarifications were made in Section 4 to more accurately reflect l
system descriptions contained in other sections of the USAR.
Section 5 A note was added to Table 5.1-4, Pressurizer Design Data, to clarify the essential pressurizer heater capacity requirements.
Table 5.1-5, Steam Generator Design Data, was revised to show the correct main feedwater nozzle diameter.
Table 5.1-7, Reactor Coolant System Piping Design Data, was revised to reflect the DBNPS RCS volume and the RCS piping dry weight.
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l Section 5.2.5, Inservice Inspection Program, was revised to reflect DBNPS's commitment to ASME Section XI,1992 Edition,1992 Addenda for the l
inservice inspection of the Containment Vessel.
A new ASME Code Case was added to Table 5.2-2, Code Case Interpretations. Code Case N-389, Alternative Rules for Repairs, Replacement, or Modifications,Section XI, Division I was added for the fusion welding of the steam generator tube plugs. This table was also revised to add N-474-1, Design Stress Intensity and Yield Strength Values for UNS N06690 with a Minimum Specified Yield Strength of 35 ksi, Class 1 ComponentsSection III, Division 1, for steam generator tube repairs.
Section 5.5.1, Reactor Coolant Pumps, was modified to reflect a t
modification which added enclosures to the RCP motor lube oil system for 10 CFR 50 Appendix R commitments.
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Page 4 of 11 The use of reroll for steam generator tube repairs was added to Section 5.5.2, Steam Generators. This change reflects Technical Specification Amendment 220.
The capacities for the Reactor Coolant Drain Tank and the Pressurizer Quench Tank were revised to reflect the nominal capacity.
Section 6 Section 6.2.3, Containment Vessel Air Purification and Cleanup Systems, was :,ised to reflect Technical Specification Amendment 221. This Amendment deleted the requirements for the SFAS containment radiation monitors, added a requirement that the containment purge supply and exhaust isolation valves remain closed during Modes 1 through 4 and deleted the option to use SFAS containment radiation monitors to provide automatic containment isolation during refueling operations. This section was also revised to reflect Amendment 217, which extended the surveillance intervals for the Emergency Ventilation System to 24 months. This affected DBNPS commitment to Regulatory Guide 1.52, which is discussed in this USAR Section.
A discussion was added to Section 6.2.4, Containment Vessel Isolation Systems, to indicate that the hydrogen purge outlet line can also be used to vent containment. This Section was also revised to reflect modifications completed in response to Generic Letter 96-06, Assurance of Equipment Operebility and Containment Integrity During Design-Basis Accident Conditions. These modifications installed bypass check valves around several containment isolation valves inside containment and also installed a relief valve for a penetration.
Tables 6.2-25 and 6.2-26, Containment Vessel Isolation System Number 1 and 2, respectively, were revised to reflect Technical Specification Amendment 218. Amendment 218 extended SFAS surveillance to 24 months and revised SFAS RCS Low Pressure setpoints based upon the instrument drift study.
Section 6.3.4, ECCS Test and Inspections, was changed to reflect revised acceptance criteria for Decay Heat Valve Pit Leakage.
Section 6.3.5, ECCS Instrument Application, was changed to reflect Technical Specification Amendment 221, which deleted the requirements for the SFAS containment radiation monitors, as in USAR Section 6.2.3 above.
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Docket Number 50-346 License Number NPF-3 Serial Number 2574 l
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A new Section, Section 6.4, IIabitability Systems, was added to the USAR.
This section summarized DBNPS Control Room Habitability studies and procedures for control room personnel protection in the event of radiation or toxic gas releases.
Other minor clarifications were made in Section 6 to more accurately reflect system descriptions contained in other sections of the USAR.
Section 7 Section 7.2, Reactor Protection System (RPS), Section 7.3, Safety Features Actuation System (SFAS), and Section 7.4, Systems Required for Safe Shutdown, were modified to clarify how the DBNPS controls access to safety system setpoints, calibrations and test points of associated safety systems.
Section 7.3, Safety Features Actuation System (SFAS), was modified considerably to reflect Technical Specification Amendment 221. This Amendment deleted the requirements for the SFAS containment radiation monitors, added a requirement that the containment purge supply and exhaust isolation valves remain closed during Modes I through 4 and deleted the option to use SFAS containment radiation monitors to provide automatic containment isolation during refueling operations. This section was also revised to reflect Technical Specification Amendment 218. Amendment 218 extended SFAS surveillances to 24 months and revised SFAS RCS Low Pressure setpoints based upon the instrument drift study. USAR Section 7.3 was revised to reflect the change in the SFAS RCS Low Pressure and RCS Low-Low Pressure trip setpoints.
As a result of reviews performed in response to Generic Letter 96-01, Testing of Safety-Related Circuits, Section 7.3.2.6, Compliance with AEC Safety Guide 22, was changed. Containment Emergency Sump Isolation Valves and the Borated Water Storage Tank (BWST) Isolation Valves were previously exempted from testing during power operations. After the review conducted for Generic Letter 96-01, these valves will now be tested during power operations.
Section 7.4.1.3.5, Steam and Feedwater Line Rupture Control System (SFRCS) Interlocks, was modified to reflect the revised interlock circuitry between the startup of the Auxiliary Feedwater System and the Decay Heat Removal System operation. This circuitry was revised by a modification l
performed in a previous fuel cycle.
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Docket Number 50-346 License Number NPF-3 Sedal Number 2574 Page 6 of 11 Section 7.4.1.3.10, SFRCS Design Bases, was revised to show an increase in the SFRCS Steam Generator high level setpoint. This section was also revised to reflect Technical Specification Amendment 218, which revi;ed the SFRCS Steam Generator Low Level Allowable Value.
Section 7.4.2.3.3, Compliance with AEC Safety Guide 22, was revised based upon reviews conducted for Generic Letter 96-01. Prior to this review, the SFRCS signal to the Main Steam Isolation Valve bypass valve was not tested during power operations. This signal will now be tested in all required Modes of operation.
Section 7.6.1.1, Normal Decay Heat Removal Valve Control System, was modified to reflect Technical Specification Amendment 218. The specific change revised the allowable RCS pressure value for the SFAS Decay Heat Isolation Valves (DH11 and DH12) and Pressurizer Heater Interlock. This section was also modified to remove the specific response times for Decay Heat Isolation Valves DH11 and DH12.
Several changes were made throughout Section 7 to make the text more accurately reflect other sections of the USAR, Technical Specifications and the system operation.
Section 8 The grid descriptions and stability analyses were updated in Section 8.1.1, Utility Grid, and Section 8.2.1.1, Offsite Power System Reliability Considerations.
l Section 8.3.2.2.1.2, Station Batteries, was revised to incorporate load changes made to the 125 VDC station batteries since the last USAR revision.
A clarification was made to Section 8.3.2.1.3, Battery Chargers, regarding the capacity of the battery chargers.
Several changes were made throughout Section 8 to make the text more accurately reflect other sections and figures of the USAR, and the system operation.
Section 9 Additional Service Water and Component Cooling Water valves were added to the listing of valves that were purchased utilizing the provisions of Generic Letter 89-09, ASME Section III Component Replacements.
Docket Number 50-346 License Number NPF-3 Serial Number 2574 l
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l In Section 9.1.3, Spent Fuel Pool Cooling and Cleanup System, the descriptions of the heat loads were changed to reflect the analyzed values.
The text was also modified to clarify what normal and abnormal refueling activities will not cause the decay heat load in the SFP to exceed the maximum e.nalyzed heat loads for normal and abnormal refueling activities.
This clarUication also included analyses conducted for the conversion to a 24 month fuel cycle.
A new section was also added to Section 9.1.3 to discuss the use of a portable filtration unit. Another change made to this section pertained to revised procedural actions taken upon the receipt of an alarm for the Spent Fuel Pool demineralizer filter.
Section 9.1.4.2.2, liandling Equipment, was revised to describe inanual handling of control rod assemblies in the Spent Fuel Pool.
A new section, Section 9.1.5, Control of Heavy Loads, was added to provide a description of the DBNPS program for controlling heavy loads in accordance with the commitments to NUREG-0612.
Information was added to Section 9.2.1, Service Water System, to accurately show the effects of using raw lake water, from the forebay, to feed the Steam Generators to cool the plant following a seismic event.
Section 9.2.2, Component Cooling Water System, was revised to reflect a modification that deleted the CCW pump low flow and high temperature trip ftmetions. The CCW Single Failure Analysis Table was also revised to add new containment isolation valves that were added as part of the modifications completed in response to Generic Letter 96-06.
Section 9.2.7, Auxiliary Feedwater System, was revised to reflect a modificc. ion that installed Auxiliary Feedwater Pump Turbine (AFPT) Main Steam minimum flow lines to increase the reliability of check valves MS734 and MS735.
The Makeup Tank Low Pressure Alarm setpoint was revised in Section l
9.3.4, Makeup and Purification System.
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Seria! 's nmber 2574 Page 8 of11 l
l As part of the Source Term Reduction Program, the filter size for the l
Purification Demineralizer Filter and the Makeup Filter were decreased.
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These filters are listed in Table 9.3-8, Makeup and Purification System Component Data.
The letdown cooler description in Table 9.3-5, Malfunction Analysis of Makeup and Purification System, was clarified to reflect the plant configuration and modifications that were completed m previous years.
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In response to LER 96-007, Control Room Emergency Ventilation System (CREVS) Design Bases Calculation Error, Section 9.4, Air Conditioning, l
Heating, Cooling, and Ventilating Systems, was revised. The description of the CREVS was revised to allow operation of the CIEVS in the pressurization mode immediately following a LOCA. A discussion was also added to this section regarding the allowable openings in the Control Room l
pressure boundary during maintenance activities.
Section 9.5.3.1, Normal Station and Security Lighting, was modified to provide cable separation criteria for free aired electrical cables supplying fluorescent lighting fixtures.
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Other minor clarifications were made in Section 9 to more accurately reflect l
system descriptions contained in other sections of the USAR.
Section 10 Section 10.3.4, Main Steam Inspection and Tests, was modified to correct a discrepancy regarding the testing of the Main Steam non-return check valves and the Main Feedwater Control <alves.
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Other minor clarifications were made in Section 10 to more accurately reflect actual plant configuration, system descriptions contained in other sections of the USAR and modifications that were completed in previcus years.
Section i1 Sections 11.2, Liquid Waste Systems, was revised to more accurately reflect the operation of the radwaste demineralizers.
Table 11.2-1, Clean Liquid Radioactive Waste System Equipment List, was I
modified to reflect a decrease in the filter size of the Primary Demineralizer Filters. This change is part of the Source Term Reduction Program.
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i Docket Number 50-346 License Number NPF-3 Serial Number 2574 l
Page 9 of 11 Table 11.2-2, Miscellaneous Liquid Radioactive Waste System Equipment List, was modified to revise the tank capacity for the Miscellaneous Waste I
Evaporator Storage and the Miscellaneous Waste Monitor Tanks to depict their nominal values. This table was also changed to revise the filter rating for the Miscellaneous Waste Monitor Tank Filter. The descriptions of the Demineralizer components were also revised to reflect a modification to the Demineralizer.
Several sections within 11.3, Gaseous Waste System, were revised to depict l
a revised valve line up for the Bope Acid Evaporators.
Section 11.4.4, Process and Effluent Radiological Monitoring Systems Calibration and Maintenance, was revised to reflect Technical Specification Amendment 218 which revised the calibration frequency for radiation monitors within containment to 24 months as part of the conversion to a 24 month fuel cycle.
Table 11.4-1, Liquid, Gas, and Airbome Radiation Monitors, was revised to correct detector sensitivity values, delete unnecessary background information, and correct other minor discrepancies within the table.
A discussion was added to Section 11.5.5, Storage Facilities, to allow additional operations in the Low Level Radwaste Storage Facility, such as opening of dry active waste (DAW) containers, loading SeaLand boxes for shipment, and minor tool and equipment repair.
Section 12 Several changes were made to Section 12, Radiation Protection, to update the method and procedures for designatiag radiologically restricted areas and the requirements for entry into these areas.
Section 12.1.4, Area Monitoring, was revised to reflect an NRC-approved exemption from the provisions of 10 CFR 70.24, Criticality Monitors, for the new and spent fuel pool areas. This section was also revised to reflect Technical Specification Amendment 221. This Amendment deleted the requirements fo*: the SFAS containment radiation monitors, added a requirement that the containment purge supply and exhaust isolation valves remain closed during Modes I through 4 and deleted the option to use SFAS containment radiation monitors to provide automatic containment isolation during refueling operations.
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Docket Number 50-346 License Number NPF-3 Serial Number 2574 Page 10 of 11 3
Section 13 Section 13.1, Organizational Structure, was revised to reflect the merger of the Centerior Energy Corporation (CEC) and Ohio Edison and the formation of a new holding company, the FirstEnergy Corporation. The Toledo Edison Company (TE), the Cleveland Electric Illuminating Company and the Centerior Service Company (CSC) are subsidiaries of FirstEnergy.
Other organization changes were made to this section. Changes made include; the creation of Supervisor-Outage Work Control Management, in the Operations Section, the creation of Supervisor-Outage Management in the Outage Management Section, and the deletion of Superintendent-Planning, in the Maintenance Section.
Realignment of responsibilities also occurred in this section. The industrial health and safety function (OSHA activities) was moved from the Manager-Regulatory Affairs to the Manager-Nuclear Safety and Inspections, the non-l l
radiological environmental compliance function (EPA activities such as hazardous and mixed waste) was transferred from the Manager-Regulatory Affairs to the Manager-Radiation Protection.
Section 15 Several accident analyses in Chapter 15 were revised to reflect Technical Specification 218 which revised the SFAS RCS Low Pressure Trip analytical j
i setpoint. Section 15.3.1, Loss of Reactor Coolant from Small Ruptured l
Pipes or from Cracks in Large Pipes Which Actuates Emergency Core L
Cooling; Section 15.4.2, Steam Generator Tube Rupture; Section 15.4.4, Steam Line Break; and Section 15.4.5, Break in Instrument Lines or Lines from Primary Systems that Penetrate Containment, were revised based upon analyses performed to support Amendment 218.
In support of the conversion to a 24 month fuel cycle, several acci' ents were d
reanalyzed. Sections reanalyzed include; Section 15.4.3, Control Pod Assembly Ejection Accident; Section 15.4.6, Major Rupture of Pipes Containing Reactor Coolant Up To and Including Double-Ended Rupture of the Largest Pipe in the Reactor Coolant System (Loss-of-Coolant Accident);
and Section 15.4.7, Fuel Handling Accident. Appendix 15A, Radiation l
Sources, was also revised to add a discussion regarding the source term for l
extended cycles and to add a comparison table of core fission product l
inventory for a 24 month fuel cycle with source terms from TID 14844, l
Calculation of Distance Factors for Power and Test Reactor Sites, and USAR l
Table 15A-2, Total Core Fission Product Inventory in Fuel and Fuel Rod Gaps.
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l Page 11 of 11 A new analysis was added to Section 15.4.6, evaluating the impact of starting the Control Room Emergency Ventilation System (CREVS)in the pressurization mode immediately following a maximum hypothetical accident. LER 96-007, Control Room Emergency Ventilation System (CREVS) Design Bases Calculation Error, discussed the operation of the CREVS in the pressurization mode immediately following a LOCA.
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ATTACHMENT 2
SUMMARY
OF FIIAR REVISION 17 MAJOR CHANGES Section 1 Incorporated the NRC-approved exemption for the Reactor Coolant Pump Motors Oil Collection System.
Section 2 Updated several references relating to Information Notice (IN) 92-18,
" Potential For Loss Of Remote Shutdown Capability During A Control Room Fire."
Section 3 Made changes relating to the Containment Air Cooler 1-3 not being credited for safe shutdown. Other changes were made based on resolution ofIN 92-18.
Section 4 Made various changes to update combustible loading values and to j
incorporate plant modifications, drawing changes and Potential Condition i
Adverse to Quality Reports (PCAQRs). The changes include revised circuit numbers, deleting sub-components, and revising the associated notes.
Section 7 Added wording to reflect NRC-approved exemption fer the Reactor Coolant Pump Motors Oil Collection System.
Section 8 Incorporated changes relating to the surveillance frequencies based on the i
24-month fuel cycle. Changes include compensatory measures for inaccessible areas. Other minor changes were also made to this Section.
Section 9 Clarified fire alarm response and other minor wording changes.
Appendix A Made various minor 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 minor changes due to the incorporation of plant modifications, drawing changes and PCAQRs. The types of changes include revised raceway identifications and circuit numbers.
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Appendix B-2 Made various minor changes due to the incorporation of plant i
modifications, drawing changes and PCAQRs. The types of changes include revised raceway identifications and circuit numbers.
Appendix C-1 Made various minor 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 minor 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.
3 Appendix C-3 Made various minor changes due to the incorporation of plant modifications, drawing changes and PCAQRs. The types of changes include changing load descriptions and circuit numbers.
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Docket Numbcr 50-346 License Number NPF-3 Serial Number 2574 Attaclunent 3 Page1ofI 1
ATTACHMENT 3 l
SUMMARY
OF TECHNICAL REQUIREMENTS MANUAL CHANGES Revision 4 Amendment 212 to the Technical Specifications removed the specific overtime limits and working hours from TS 6.2.3, and relocated them to the TRM, Section 5.2.3, Facility Staff Overtime.
Revision 5 This revision changed TRM section 5.2.3, Facility Staff Overtime, to allow for up to 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> work shifts under a rotating work week schedule, with a nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> work week, for the facility staff.
Revision 6 This revision extended the surveillance frequency requirements for the Reactor Coolant System Vents (SR 4.4.11) and seismic sensors inside containment (SR i
4.3.3.1.c) from 18 months to at least once per Refueling Interval. The definition of a Refueling Interval was also added in accordance with Technical Specification Amendment 213.
Revision 7 Technical Specification Amendment 224 relocated TS Section 3/4.9.5, Refueling Operation - Communications, and associated bases to the TRM. The definition of Core Alteration was also added to the TRM.
Revision 8 This revision reflects Technical Specification Amendment 225 which revised TS Section 3/4.3.1.1, Reactor Protection System Instrumentation; Section 3/4.3.2.1, Safety Features Actuation System Instrumentation; Section 3/4.3.2.2, Steam and Feedwater Rupture Control System Instrumentation, and the associated bases. The tables of response time limits were relocated to the TRM.
This revision also added definitions and bases for these instrumentation j
systems.
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Docket Number 50-346 License Number NPF-3 Serial Number 2574 Page 1 of 5' ATTACilMENT 4 CliANGES TO QUALITY ASSURANCE PROGRAM
- 1. Reduction in commitment changes made to Section 17.2:
Sections 17.2.1.4,17.2.15.2,17.2.16.1, and 17.2.16.2 were revised to allow replacement of the Nuclear Assurance Department's Nuclear Safety and Inspection Section in-line l
close-out review of corrective action documents other than audit findings with line j
management's review of the proposed corrective actions for adequacy. The change also replaced the commitment for the Nuclear Assurance Department to perform follow-up reviews and audits of all significant conditions adverse to quality other than audit findings with a commitment for reporting of th9 corrective actions taken to appropriate levels of management. These changes were submitted to the NRC for their review prior to implementation via Toledo Edison letter Serial 2509, dated February 16,1998. These changes were reviewed and approved by the NRC on July 15,1998 (NRC TAC MA0988 and MA2152, and TE Log Number 5307).
- 2. The organization described under Section 17.2.1, Organization, underwent several changes.
These changes were determined not to be reductions in commitments in 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 April,1997, the DBNPS Nuclear Group Departments were restructured from the existing three departments to four, with the associated realignment of activities to accommodate this new structure. This organizational change created a new Director of Nuclear Support Services to assume some non-quality assurance functions from the Director of Nuclear Assurance. Reporting to the VP Nuclear are: (1) the Plant Manager, responsible for Operations, Maintenance, Radiation Protection; (2) the Director-Engineering and Services, responsible for Design Basis Engineering, Plant Engineering, Regulatory Affairs, DB Business Services, and DB Supply;(3) the Director - Nuclear Assurance, responsible for Quality Assessment and Nuclear Safety and Inspections; and (4) the Director - Nuclear Support Services, responsible for Nuclear Training, Security, and Quality Services. The Manager-Quality Assessment continues to have direct access to the Vice-President Nuclear and to the Chairman, if the need for such access occurs.
The Manager-Quality Assessment continues to have sufficient authority, organizational freedom and independence from cost and schedule for maintaining the independence of quality assurance activities.
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Docket Number 50-346 License Number NPF-3 Serial Number 2574 Page 2 of 5 in December of 1997, the procurement engineering function was transferred from DB Supply to Design Basis Engineering. Design Basis Engineering will have procurement engineering responsibilities, including the establishment of technical and quality requirements in, and review / approval of, procurement documents supplied to vendors and related technical evaluation activities (e.g., commercial grade dedication and the selection of source verification assigrunents). This consolidation of engineering functions (design and procurement) streamlined engineering functions into a more effective organization, thus improving communication, work flow and effectiveness.
In January,1998, the reporting responsibility of the Manager-DB Business Services moved from the Director of Engineering and Services to the Vice President-Nuclear.
The Manager-DB Business Services responsibilities continue to be activities associated with budget and cost control and nuclear projects. The programs and functions accomplished by the DB Business Services organization do not implement quality assurance functional requirements. This change did not eliminate any activities previously performed.
In July of 1998, the industrial health and safety function (OSIIA activities) was moved from the Manager-Regulatory Affairs to the Manager-Nuclear Safety and Inspections, and the non-radiological environmental compliance function (EPA activities such as hazardous and mixed waste) was transferred from the Manager-Regulatory Affairs to the Manager-Radiation Protection. The reorganization of these non - Appendix B related activities does not affect Appendix B program commitments.
In July of 1998, the Manager-DB Supply's activities (material purchasing and materials management) were transferred from the Director-Engineering and Services to the Director Nuclear Assurance. DB Supply is responsible for purchasing, warehousing, material management, and coordination of related service activities (except nuclear fuel procurement) provided by the Nuclear Group and the Centerior Service Company.
Reorganization of these activities brings similarity in organizations between the Perry and Davis-Besse Nuclear organizations.
The responsibility for the trending and analysis program was clarified in the responsibility section of 17.2 by clearly assigning it to the Director-Nuclear Assurance consistent with the existing program described in USAR 17.2.15.4. The trending and analysis program relies on input from both the Manage -Nuclear Safety and Inspections and Manager-Quality Assessment, and, therefore, ov( tall responsibility is best suited at the director level (Director-Nuclear Assurance).
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Docket Number 50-346 License Number NPF-3 i
Serial Number 2574 Pag'c 3 of 5' The organizational changes described above do not reduce the efTectiveness 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 (USAR Section 17.2).
As such, these changes do not reduce commitments to the previously approved Quality Assurance Program. 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 contim'c to satisfy the requirements of 10CFR50, Appendix B.
- 3. Other non-reduction changes made to Section 17.2:
Section 17.2.1, Organization, and Section 17.2.11.3, Procedures, are revised to indicate Station Review Board (SRB) activities are no longer controlled by a " Charter", but by the DBNPS Technical Specifications and implementing procedures. Changes to these documents are controlled by regulation and USAR Section 17.2.5, Instructions, Procedures and Drawings. Additionally, the SRB no longer reviews test procedures and test results. The current SRB responsibilities are described in Technical Specifications 6.5.1.6 and 6.5.1.7. This change was approved by License Amendment 109, dated March 9,1988.
Section 17.2.15.1, Nonconformance Identification, was revised to clarify supplier nonconformance report processing. This clarification is made to make the 17.2.15.1 discussion consistent with ANSI N45.2-1977, Quality Assurance Program Requirements for Nuclear Power Plants, and ANSI /ANS 3.2-1982, Administrative Controls and l
Quality Assurance for Nuclear Power Plants, as committed to in USAR Table 17.2-1, Applicable NRC Regulatory Guides. ANSI Standards, and Industry Codes. These standards require supplier submittal, and purchaser approval, of supplier l
nonconformance reports with recommended dispositions of"use-as-is" or " repair".
l Section 17.2.15.2, Review and Evaluation (of nonconforming materials, parts, or components), describes the if tiation and review of DBNPS nonconformance reports j
(Potential Condition Adverse to Quality Report (PCAQR)). This section was modified to l
permit any cognizant supervisor to review the PCAQR for completeness, or if the l
initiator is a supervisor or above, no further management review is required. The requirement for the initiator's supervisor's review of a PCAQR represented an unnecessary restriction which could cause delays in PCAQR identification and review, 4
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Docket Number 50-346 License Number NPF-3 Serial Number 2574 Page 4 of 5 and potential failure to comply with 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors.
Section 17.2.18.7, Surveillance, was revised to allow surveillance findings to be documented and corrected by use of a generically described " deficiency document" such as a surveillance finding report or nonconformance report (PCAQR). This change provides flexibility in the methodology for documenting and correcting deficiencies identified during surveillance activities. The DBNPS surveillance process supplements the formal QA audit process. The surveillance process does not require the application of ANSI N45.2.12-1977, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants, for audit finding identification, reporting and correction. The documentation, tracking and correction of deficiencies identified during surveillances will continue to be accomplished as committed to in USAR 17.2.18.7.
Section 17.2.4, Procurement Document Control, and Section 17.2.7, Control of Purchased Material, Equipment and Services, was revised to reflect the reorganization of the pmcurement engineers from DB Supply to Design Basis Engineering as described above.
Section 17.2.14, Inspection, Test, and Operating Status, and Table 17.2-1, Applicable NRC Regulatory Guides, ANSI Standards, and Industry Codes, were revised to clarify that the material receipt shipping damage inspection by warehouse personnel (DB Supply) is not an inspection activity performed by a Regulatory Guide (9/80),
Qualification of Nuclear Plant Inspection, Examination and Testing Personnel, qualified inspector.
l Section 17.2.15.1, Nonconformance Identificmion, was revised to eliminate the reference to specific QA " Hold Tags" and permit use of generic equipment status tags for identification ofinstalled and non-installed hardware. Under the change, identification and segregation of nonconforming items will continue with a generically titled equipment status tag. The change does not alter the nonconforming equipment corrective action dispositions or allow use of nonconforming hardware which could affect USAR described equipment or system operation. The change does not affect the safety function of any SSCs and does not affect the operation cf any plant system or reduce any commitments to 10 CFR 50 Appendix B, or the Quality Assurance Program.
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i Docket Number 50-346 License Number NPF-3 Serial Number 2574 Page 5 of 5 The following commitments were made and reflected in USAR Table 17.2, Applicable NRC Regulatory Guides, ANSI Standards, and Industry Codes:
The DBNPS position for ASME Section XI,1986 Edition, is revised to clarify that this edition of the Code is applicable to those components subject to the rules of IWB, IWC, or IWD. This is not a change to present requirements or commitments.
These components have been and will remain under the requirement of the 1986 Edition, No Addenda for the remainder of the second ten year inspection interval.
ASME Section XI,1992 Edition,1992 Addenda is added to the list of applicable Codes and Standards. This change updates the Inservice Inspection commitments to include the rules ofIWE for the inspection of the Containment Vessel and its 1
penetrations. This change is a result of new inspection rules for the Containment Vessel which were not previously in effect. These new rules were promulgated j
through an amendment to 10 CFR 50.55a which was published in the August 8,1996 Federal Register. This change has no effect on present quality assurance requirements as repairs / replacements will be made using the 1986 edition which is the Code of reference for this ten year inspection interval.
Regulatory Guide 1.78 (6/74), Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical j
Release, was added to the table. This change supports the control room habitability evaluations performed in accordance with this regulatory guide as discussed in NUREG 0737, Clarification of TMI Action Plan.
The above non-reduction changes do not reduce the effectiveness of any programs, do not i
reduce any oversight or reviews, do not eliminate any activities and do not add any non-quality assurance functions to the Quality Assessment Section. The requirement of 10 CFR 50, Appendix B and the DBNPS-committed Regulatory Guides and ANS standards are still satisfied.