ML20083L797
| ML20083L797 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/15/1994 |
| From: | Stetz J CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2297, NUDOCS 9505180608 | |
| Download: ML20083L797 (169) | |
Text
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f 300 m h ue John P. Stet Toledo, OH 43652-0001 Vice President Nuclear 419 249 2300 payi3_m May 15, 1995 i
Docket Number 50-346 License Number NPF-3 Serial Number 2297 United States Nuclear Regulatory Commission Document Control Desk Vashington, DC 20555 Gentlemen:
Subj ect :
10 CFR 50.59 Report of Facility Changes, Tests and Experiments i
The Toledo Edison Company hereby subm.ts, pursuant to 10 CFR 50.59 (b)(2), the 10 CFR 50.59 report of facility changes, tests and experiments for Davis-Besse Nuclear Power Station, Unit 1.
Those changes, tests and experiments identified via the safety review process during the reporting period of May 1, 1993 through November 15, 1994 are attached. This report includes facility changes that occurred during Cycle 9 and the Ninth Refueling Outage which concluded on November 15, 1994. Attachment 1 provides an executive summary of those changes, tests and experiments contained in the attachment. The attached safety evaluation summaries do not involve an unreviewed safety question.
1 If you have any further questions concerning this matter, please j
contact Mr. V. T. O'Connor, Manager - Regulatory Affairs at (419) 249-2366.
Very truly yours, hs:A W JMH/lk Attachments cc L. L. Gundrum, NRC Project Manager J. B. Martin, Regional Administrator, NRC Region III S. Stasek, DB-1 NRC Senior Resident Inspector J
Utility Radiological Safety Board IT operating componies:
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Docket Number 50-346
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ATTACHMENT I 10 CFR 50.59
SUMMARY
SHEET Number Title DCR 92-0013 Isolation of Auxiliary Steam Desuperheating DCR 93-0005 Miscellaneous Drawing Update DCR 93-0019 Miscellaneous Drawing Update DCR 93-0024 Revise Motor-Operated Valve Motor Horsepower Ratings DCR 93-0029 0 Boundaries for the ECCS Room Sump Pumps DCR 93-0031 Delete USAR Section 9.3.3.3, Equipment and Floor Drainage System-Setpoints DCR 93-0040 As-Built Routing of Conduit Number 48257A DCR 93-0049 Change in Position of Service Water Valves to Turbine Plant Cooling Vater Heat Exchanger DCR 93-0050 Update Depiction of PPF Alternate Feed DCR 93-0053 Emergency Ventilation System Configuration Issues DCR 93-0057 Emergency Ventilation System Configuration Issues DCR 93-0060 Closure of SW325 and Opening of SW1368 During Normal Power Operation DCR 93-0063 Emergency Ventilation System Configuration Issues DCR 93-0068 Revise Main Steam Valve Positions DCR 93-0069 Containment Gas Analyzing System Configuration Management Issues DCR 93-0070 Revise MU23 and MU323 Valve Positions DCR 93-0071 Revise Makeup and Purification System Drawings DCM ?3-0072 Revise Component Cooling Water Valve Positions DCh 93 -0073 Closure of RC46, Reactor "0" Ring Drain to Normal Sump Valve DCR 94-0004_
Isolation of the Demineralized Vater Supply to the Primary Water Storage Tank DCR 94-0007 Hydrogen Storage Cylinders Valve Positions DCR 94-0008 Revise Vacuum System Valve Positions DCR 94-0011 Revise PV87 Valve Position DCR 94-0017 Revise Main Steam Valve Positions DCR 94-0019 Revise Position of Nitrogen Supply to the Makeup Tank Valve DCR 94-0022 Revise Steam Generator Vet Layup Recirculation Valve Positions DCR 94-0029 Revise Position of AS271 and AS272 DCR 94-0030 Revise Main Steam Valve Positions DCR 94-0031 Resolution of Configuration Concerns for the Station and Instrument Air System Revise Low Pressure Extraction Steam System Drawings DCR 94-0032 DCR 94-0037 Control Room Emergency Ventilation System Configuration Issues DCR 94-0041 Revise Domestic Vater Piping and Valve Configuration DCR 94-0045 Revision of Plant Drawings Per EN-DP-1030 Drawing Guidelines DCR 94-0050 Revise Extraction Steam System Drawings DCR 94-0064 Condensate Configuration Control Issues DCR 94-0065 Condensate Demineralizer Configuration Management Changes
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- Dockat Numb:r 50-346 Licensa Nuxbsr NPF-3 Serial Number 2297 Page 2 of 4 Number Title DCR 94-0070 Revise Turbine Plant Cooling Vater System Drawings DCR 94-0072 Resolution of Configuration Concerns for the Station and Instrument Air Systems
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DCR'94-0073 Configuration Concerns for the Makeup Vater Treatment System DCR 94-0074 Revise Auxiliary Steam Valve Positions DCR 94-0080 Configuration Concerns for the Primary and Demineralized Water Systems DCR 94-0082 Configuration Concerns for the Makeup Vater Treatment System DCR 94-0083 Throttling CCV and TPCV Outlet Isolation Valves DCR 94-0088 Revise Station Drains and Sumps System Drawings DCR 94-0089 Revise Valve Positions on Auxiliary Boiler Drawings DCR 94-0095 Auxiliary Radioactive and Fuel Handling Ventilation System Configuration Issues DCR 94-0098 Gaseous Radvaste Configuration Management Issues DCR 94-0105 Change Normal Position for Valve VT105 DCR 94-0115 Correct SFAS Output Module Power Sources DCR 94-0119 Hiscellaneous Reactor Protection System Related Drawing Corrections DCR.94-0125 Addition of Valve AS575 to System Drawings DCR 95-0006 Change Normal Position of Valves SV5421, SV5422, SV5424 and SV5425 from " Closed " to "Open" DCR 95-0008 Containment Drain Header Isolation Valves DCR 95-0014 Revise Position of Nitrogen Cover Gas Supply Valves to the Demineralized Vater Storage Tank 1-2 FPR 89-0289 Replacement of C0H-5 Relay in AD113 FPR 91-0055 Configuration Concerns Regarding Drains FPR 92-0008 Revised Hain Steam to the Turbine Generator Pipe Break Analysis FPR 92-0040 Supply Vater to Lime Feed System FPR 92-0075 Revise Control Logic For AF3870 and AF3872 FPR 92-1236 Replacement of General Electric Type 4701 Transducers FPR 93-0957 Removal of Main Steam Safety Valve Position Monitors FPR 93-0958 Removal of Hain Steam Safety Valve Position Monitors FPR 93-4738 Replacement of 90% Type-27D Undervoltage Relays FPR 94-0140 SFAS Power Supply Replacement FPR 94-0390 Install Dessicant Dryer in Compressor C11-1 Loadless Start Line FPR 94-0463 Fuel Handling Communications FPR 94-0861 Abandon Reactor Vessel Head Accelerometers ZE-8907 and ZE-8908 H0D 88-0215 Revision of Power Supply to C5752 MOD 89-0036 Revision of Power Supply to Various Computer Equipment H0D 89-0099 Removal of Pipe Collars From the Pressurizer Surge Line Vhip Restraints H0D 90-0046 Replacement of RC10 MOD 90-0073 Decontamination Showers in Room 417 i
H0D 91-0004 Install New Sample Line to Radiation Monitor RE8432 i
- l Dockst Numbsr 50-346 License Number NPF-3 Serial Number 2297 Page 3 of 4 Number Title MOD 91-0030 Spent Fuel Handling Access Hodification H0D 91-0040 Fuel Handling Equipmeat Enhancements H0D 91-0043 Elimination of Normally Energized Agastat Series 7000 Relays in Environmentally Qualified Circuits H0D 91-0046 New Central Fire Alarm System H0D 92-0003 Provide Backup Exciter Field Breaker Trip Circuit H0D 92-0006 Hakeup Water Treatment Chlorination System NOD 92-0046 Victoreen Process Radiation Monitor Digital Upgrade MOD 92-0070 Increase Motor Size for Motor Operated Valve HS106 H00 92-0071 Modification of Door 517 H0D 93-0011 Replacement of Valves SV6406 and SV6407 H0D 93-0026 Abandon Domestic Water Heater in Place MOD 93-0043 Delete Annunciator 9-3-B, Station Vater Pre-Treatment System Trouble MOD 93-0060 Replace Motor Operators for Valves HS603 and HS611 MOD 93-0064 9 RF0 Fuel Repairs H0D 93-0071 Fused Safety Switch for the Switchyard 125V DC Distribution Panels DA1 and DB1 MOD 94-0005 Replace Motor Operators on AF3870 and AF3872 H0D 94-0014 AFPT 1 Governor Control Circuit Appendix R Isolation H0D 94-0022 Increase Motor Sizes and Change Circuitry for Valves AF599 and AF608 PCAO 92-0030 Use-As-Is Disposition of Discrepant Fuses PCA0 93-0287 Use-As-Is Disposition for HS734 and MS735 PCAO 93-0552 Operation With Only Three Turbine Bypass Valves Immediately Available PCAO 94-0065 Capability of AF3870/AF3872 to Isolate Auxiliary Feedvater Flow PCAO 94-0108 Cai-Tronic Power Supply PCAO 94-0459 HSIV Bypass Valves, MS100-1 and MS101-1 PCAO 94-1288 ASME Certification for Hot Leg Level Monitoring System Component SCC 89-0529 Installation of Unions and Isolation Valves for the Turbine Plant Cooling Vater Sample Coolers SCC 90-3037 Replacement of Globe Valve FV174 vith Three Restricting Orifices SCR 92-5006 Raising the SFRCS High Level Trip Setpoint SCR 92-5007 ICS Steam Generator High Level Limiter Setpoints and NNI Steam Generator Operate Range High Level Alarm Setpoints SCR 93-5004 Update Lov Voltage Setpoint Information SCR 93-5012 Auxiliary Feedvater Pump High Speed Stop Setting SCR 93-5014 Provide Breaker Trip Setpoint for Replacement Lighting Breaker SCR 94-5006 Reduction of Emergency Diesel Generator Jacket Vater Lov Temperature Alarm Setpoint SE 93-0053 Changing Shift Supervisor Hanning Requirements SE 94-0017 Outside Storage of Radioactive Material SE 94-0041 Line-up of CD75, CD188 and CD189 SE 94-0048 ICS Loss-of-Hain Feedvater-Pump Runback Limit
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Dockat Nu bar 50-346 Licensa Nurbtr NPF-3 Serial Number 2297 Page 4 of 4 Number Title SE 94-0061 Cycle 9 Licensed Length Extension and Addendum to the Reload Report SE 94-0062 Testing Main Steam Safety Valves During Modes 1 and 3 SE 94-0067 Modify Operational Restrictions on Spent Fuel Pool Eme'rgency Ventilation System SE 94-0071 Cycle 10 Reload and Core Operating Limits Reports SE 95-0006 Combining Maintenance Superintendent Responsibilities SE 95-0011 Reorganization of Planning, Scheduling and Chemistry Responsibilities TM 93-0030 Temporary Smoke Detectors in Room 114 UCN 87-012 Detergent Vaste Drain Tank Pump Classification UCN 91-066 Deletion of Soldering as a Special Process UCN 93-039 Revision of USAR Section 7.13, Post Accident Honitoring System UCN 93-059 Switchyard Battery Size UCN 93-062 Change in the Distribution of External Supplier Audits UCN 93-074 Clarification of Emergency Conditions for Decay Heat Removal Cooler Characteristics UCN 93-075 Effluent Radiation Monitor Setpoints UCN 93-077 Generic Letter 89-10 Program Calculations UCN 93-079 ECCS Room Cooler Required Service Vater Flov UCN 94-007 Containment Air Cooler Service Water Valve Open with Fan Off UCN 94-029 Remove the Setpoint for the Reactor Coolant Drain Tank Pressure Regulator UCN 94-040 Zone-Loaded Fuel Storage UCN 94-082 Grout Penetration Seal Inspections UCN 94-088 Implementation of License Amendment 186 UCN 94-100 Miscellaneous FHAR Changes UCN 94-122 Vater Treatment Building Fire Detectors UCN 94-155 Revision of USAR Chapters 11 and 12 UCN 95-006 Changes to the DBNPS Nuclear Quality Assurance Program Site Organization UCN 95-007 Changes to the DBNPS Nuclear Quality Assurance Program Site Organization UCN 95-018 Revision of USAR Discussion of the Battery Room Ventilation' UCN 95-025 Update Description of the Solid Radvaste Handling System j
UCN 95-027 Revision to USAR Section 9.5.2, Communication Systems l
i SAFETY EVALUATION
SUMMARY
j FOR UCN 95-027 (SE 95-0024)
TITLE:
Revision to USAR Section 9.5.2, Communication Systems CHANGE:
Removed specific details concerning radio console locations.
In addition, this change deleted the UHF radio system mode of communication " remote to Station Superintendent" that no longer exists and documented the prior removal of a fixed security UHF radio transceiver in the Davis-Besse Administration Building (DBAB). The console location was part of the specific details previously given in the USAR but was removed.
REASON FOR CHANGE:
This change revised USAR Section 9.5.2, Communication Systems, to reflect what i
exists in the field.
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SAFETY EVALUATION
SUMMARY
The Sound-Powered Emergency Address System is unaffected by this change.
The removal of the DBAB Security Transceiver had no effect on the ability of the Security Radio Communications System to perform its intended function. The security radio communications are carried out via the use of existing portable radios and other existing fixed transceivers.
The deletion of the " Remote to Station Superintendent" mode of operation has no affect on the ability of the UHF radio communications to perform its intended function. This mode of operation is carried out via the use of existing portable radios and the radio repeater system.
The USAR Communication Systems description requirements are still being maintained in the USAR, and there are no physical plant changes being made, therefore, there are no adverse effects on safety.
SAFETY EVALUATION
SUMMARY
TITLE:
Isolation of Auxiliary Steam Desuperheating CHANGE:
Revised USAR Figures to depict the isolation of the Auxiliary Steam Desuperheating water and revised the High Energy Line Break (HELB) analysis for the auxiliary steam lines.
REASON FOR CHANGE:
It was identified that the isolation of the Auxiliary Steam system desuperheating water resulted in steam temperatures in excess of the rating of the piping and components in the auxiliary steam system.
Corrective action for this issue included rerating of the affected piping and components, reanalysis for high energy line beaks, re-evaluation of pipe stress analysis, updating hanger / pipe support calculations and verification of equipment nozzle loads.
SAFETY EVALUATION
SUMMARY
The change to the system lineup to normally operate with the desuperheating water isolated results in higher temperature steam in the auxiliary steam system. Reanalysis has shown that the piping, flanges, fittings, valves and various heat exchanges are acceptable for use at the new pressure / temperature conditions.
The elevated temperatures caused by the changed system lineup necessitated the re-analysis of the affected piping involving thirty one stress problems.
The increased thermal expansion resulted in increased piping stresses and j
support loads. This required design changes to alleviate piping stresses and i
the increased support loadings.
Increasing the auxiliary steam temperature affects the heat transfer capability of the affected heat exchangers.
Superheated steam causes the overall heat transfer coefficient to decrease and therefore degrades the performance of heat exchangers.
Degraded heat exchanger performance could affect the ability of the borated water storage tank heat exchanger and the fire water storage tank heat exchanger to maintain water temperature at design conditions. However, l
heat exchanger sizing was performed conservatively and winter operation at extreme conditions has demonstrated that the heat exchangers utilizing steam i
without desuperheating are capable of performing adequately.
USAR section 3.6.2.7.1.9 discusses postulated high energy line breaks (HELBs) in the auxiliary steam system. This USAR section is being revised. The revised analysis does not take credit for desuperheating water. The revised analysis also assumes steam blowdown continuing for 30 minutes prior to being
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diagnosed and isolated by plant operators. This time period is customary for analysis when no direct control room annunciator is available and operator action is required outside the control room.
Vithout desuperheating, the increase in blovdown enthalpy required a change in the position of AS379 from normally open to normally closed.
The position of AS379 on drawings, in procedures and in the USAR is revised.
Postulated breaks / cracks on this normally unused line vould involve relatively high mass flow rates.
Because it is nov normally isolated, this line is no longer considered to be a high energy line.
For lines which remain in service, with auxiliary steam routed in proximity to equipment which would be required to function in the event of a crack or a break, the increase in predicted compartment temperatures was calculated.
Due to both the relatively lov steam mass flow rates and the essential ventilation systems in the intake structure, equipment in the intake structure vill not be adversely impacted.
Predicted temperature increases in some areas of the auxiliary building are substantially higher without desuperheating vater.
However, in all but four rooms the resultant temperatures remain bounded by the temperature from current or previously postulated breaks or cracks in other system piping.
In rooms 209, 221, 227 and 304 equipment peak qualification temperatures are greater than the peak temperature of the postulated profile with margins and the thermal exposure experienced during the test is greater than that of the HELB plus seven days post-HELB operability.
Based on the above, it is concluded that the proposed change is safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
Miscellaneous Drawing Update CHANGE:
Change the load description for transformer XDF6, which is shown on drawings El Sheet 1 and E3.
These drawings are depicted in the USAR as Figures 8.3-1 and 8.3-4, respectively.
REASON FOR CHANGE:
XDF6 was installed to power Service Building 1, and was for some time used to power Service Building 2.
The USAR load description for this transformer simply says " Service Bldg",
i.e.,
apparently only Service Building 1 was intended.
Service Building 1 was destroyed several years ago, and Service Building 2 is currently supplied f rom of f site power (Projects Center / Service Building 3 feed). XDF6 is now only used when temporary power is needed in the north yard.
This DCR contains Document Change Notices (DCNs) which revise the designation for this transformer from service building to temporary loads.
SAFETY EVALUATION
SUMMARY
This change will have no effects on safety.
XDF6 is not important to safety, l
and changing the load description for :(DF6 will ecuse no adverse interaction with equipment which is important te safety.
In addition, a note has been added to E3 to require that XDF6 be loaded in accordance with the National Electric Code, thus ensuring that XDF6 will not be loaded beyond its design capabilities.
DBP 5302FFF/6
SAFETY EVALUATION
SUMMARY
TITLE:
Miscellaneous Drawing Update.
1 CHANGE:
Change the power supply for the Personnel Processing Facility (PPF) which is shown on drawing E4 Sheet 5, and depicted in the USAR as Figure 8.3-18.
REASON FOR CHANGE:
The normal supply to the PPF is from bus C2, via disconnect DSC205 and transformer X3050.
During cycle 8R, an alternate feed was established between j
the Outage Support Substation (OSS) and a new transfer switch (DSC3050) at the secondary of X3050.
The alternate feed, installed under Service Request 93-0230, was intended to be temporary but is being made permanent.
The OSS is fed from offsite sources shown on E2018, and X3050 is shown on E4 Sheet 5.
DSC3050 is capable of connecting the PPF to either the OSS or to X3050, but can not connect the OSS to the secondary of X3050.
SAFETY EVALUATION
SUMMARY
This change will have no effects on safety.
The PPF power supply is not important to safety, and the addition of an alternate connection to an offsite source will not involve any interaction with any other system which is important to safety.
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DBP 5302FFF/8 i
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SAFETY EVALUATION
SUMMARY
FOR DCR 93-0024 (SE 93-0041 R01)
TITLE:
Revise Motor-Operated Valve Motor Horsepower Ratings CHANGE:
As-build the Motor-operated Valve horsepower ratings shown on drawing E-6, sheets 1 and 2.
Drawing E-6, sheets 1 and 2, are USAR Figures 8.3-23 and 8.3-24.
REASON FOR CHANGE:
The horsepower ratings given on the single line diagrams are inconsistent with the actual horsepower ratings as shown in the Motor Operated Valve Data Packages and verified by walkdown.
This Document Change Request changes drawing E-6, sheets 1 and 2, to correct these inconsistencies.
SAFETY EVALUATION
SUMMARY
There is no effect on safety as a result of as-building the motor horsepower ratings on drawings E-6 sheets 1 and 2.
Horsepower ratings are used to determine the locked rotor current for calculation C-EE-006.01-026.
Calculation C-EE-006.01-026 calculates the terminal voltage for each valve. This value is used to verify that terminal voltage available is not less than the minimum voltage required to achieve the maximum selected target thrust.
The maximum selected target thrust is calculated per the Target Thrust Calculation Procedure using motor foot-pound ratings which were correct.
The horsepower ratings used in calculation C-EE-006.01-026 for all but two motors were the as-built hosepower ratings not the horsepower ratings from drawing E-6 sheets 1 and 2.
For motors MVMU12A, and MVMU40 the horsepower ratings used in calculation C-EE-006.01-026 were larger than the as-built ratings. When the as-built ratings are used for these motors the voltage drop is less.
Therefore, these motors still have a terminal voltage greater than that required to achieve the maximum selected target thrust.
Consequently, there is no affect on the conclusion of the target thrust calculations, thus the valves will perform their safety functions.
The as-built horsepower ratings have been evaluated for effects en breaker i
sizing, cable sizing and motor starter sizing.
The existing circuit design capacities are adequate for the as-built horsepower ratings.
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DBP 5302DDDD/27
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SAFETY EVALUATION
SUMMARY
FOR DCR 93-0029 (SE 94-0013, R.01) i l
TITLE:
0 Boundaries for the ECCS Room Sump Pumps J
CHANGE:
Revise ECCS room drawings to designate the "O boundary" at the functional limit and to show the boundary for structural analysis as "S/I".
REASON FOR CHANGE:
Presently the P&ID, the Quality classification List (OCL) and the system isometric drawings do not show the same 0 boundary locations. The reason for the inconsistencies is that the boundary for structural analysis, which is designated as seismic category I ("S/I"), extends beyond the functional "0"
boundary to the nearest anchor point. This boundary has been referred to as the physical limit of a O boundary and normally is the 0 boundary shown on isometric and hanger location drawings. The functional 0 boundary is based on i
the functional requirements of the components and is the boundary normally l
shown on the P&ID.
Currently drawings are being revised to designate the 0 boundary at the functional limit and to show the boundary for structural analysis as S/I.
1 SAFETY EVALUATION
SUMMARY
DCR 93-0029 is intended to establish 0 boundaries that are based on functional rather than structural requirements. The functional 0 boundary is at the sump pump discharge check valves. This boundary is consistent with USAR sections 3.6.2, 6.3.2, and 9.3.3 which state that the ECCS sump pumps and piping are designated as seismic class I and the pumps are 0-listed.
All piping in the ECCS rooms and to the first anchor beyond the rooms are i
seismic category I. The seismic designation for this piping provides assurance that the pressure boundary will be maintained in the ECCS rooms, and water vill l
be removed in the event of flooding.
The instrument tubing and isolation valves, connected to the system upstream of the pump discharge check valves are non-0 which is consistent with the rules for the quality classification of instrument tubing. The tubing connects a O system that is not closed to instrumentation with no active function, therefore the instrument tubing has no active 0 function. The instrument tubing is designed to Seismic Category I requirements to maintain pressure boundry.
Based on the above discussion it is concluded that the proposed change is safe.
T' SAFETY EVALUATION
SUMMARY
TITLE:
Delete USAR Section 9.3.3.3, Equipment and Floor Drainage System-Setpoints CHANGE:
Delete USAR Section 9.3.3.3, " Equipment and Floor Drainage System - Setpoints",
which shows sump pump level setpoints for the Water Treatment Building Sump and the Unit Lube 011 Storage Room Sump.
REASON FOR CHANGE:
This USAR section is not needed as these setpoints are not safety-related and are maintained by other controlled design documents such as Operational Schematics.
SAFETY EVALUATION
SUMMARY
The Water Treatment Building's permanent sump pumps were removed and replaced with portable pumps.
This change revised USAR Figure 9.3-4 to remove the permanent putaps and also abandoned Sump Level Switches LS4604 and LSHH4604, but did not remove the associated setpoints listed in USAR section 9.3.3.3.
The Unit Lube Oil Storage Room Sump Pump setpoint for Pump P137-1 was revised per a Setpoint Study but the USAR was not revised when this study was completed.
As these sumps do not drain safety related compartments, there is no safety effect adverse or otherwise.
Therefore the proposed action is safe.
DBP 5302DDDD/3
SAFETY EVALUATION
SUMMARY
TITLE:
As-Built Routing of Conduit Number 48257A CHANGE:
Revise Electrical Raceway Drawings to reflect the as-built routing of conduit-48257A. Additionally, the Davis-Besse Fire Hazard Analysis Report (FHAR) is being revised as this report tracks cables and raceways of Appendix R safe shutdown components by plant fire areas.
REASON FOR CHANGE:
Conduit 48257A was located in Room 314 and in the cable chases of Rooms 314 and j
115 not in Rooms 123, 234. 227, 221 and 209 as was shown on raceway drawings
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E-335 and E336.
SAFETY EVALUATION
SUMMARY
The current Appendix R Safe Shutdown analysis for Fire Area A includes cable number 1PBE1146A (in conduit 48257A) as this cable and conduit are correctly listed in Room 124 which is included in Fire Area A.
Room 238 is assigned to Plant Fire Area F.
The current Appendix R Safe Shutdown analysis does not account for cable 1PBE1146A in Fire Area F.
However, valve AF3869 which is normally closed is not required to operate (open) to mitigate the consequences of a fire in this area.
Thus, a fire induced fault on cable 1PBE1146A will not disable equipment required to achieve safe shutdown for a fire in this area.
The environments in Room 314 and the cable chases in Rooms 314 and 115 are influenced by several HELB incidents.
The HELB's which impact Room 314 and the cable chases more severely and require that AF3869 fulfill these conditions are a steam generator blowdown in Room 236 and a main feedwater line break in Room 314.
Hitigating the consequences of a steam generator blowdown in Room 236 requires AF3869 to go open.
Thus, power cable 1PBE1146A must be capable of supplying voltage and electric current to the valve operator.
The highest expected temperature in these areas is approximately 2320F.
Cable 1PBE1146A is capable of functioning well above 2320F.
Spurious operation of AF3869 is not possible as a result of subjecting its power cable to the main feedwater line break in Room 314 because a hot short on this cable is not credible because conduit 48257A separates cable IPBE1146A from other voltage sources. Moreover, the highest expected temperature in these rooms due to this line break is approximately 3240F which remains below the qualification temperature for this cable type.
Based on the above discussions it is concluded that the as-built routing of conduit 48257A is safe.
DBP 5302DDDD/2
SAFETY EVALUATION
SUMMARY
TITLE:
Change in Position of Service Water Isolation Valves to Turbine Plant Cooling Vater Heat Exchanger CHANGE:
Provide for the changing of the valve line-up for the Service Vater Isolation Valves on Turbine Plant Cooling Vater (TPCV) Heat Exchanger 1-3 to show SV56 normally closed and SV53 normally open.
REASON FOR CHANGE:
Maintaining SV53 open and SV56 closed when the TPCV Heat Exchanger is in standby vill keep it full of water.
This vill help prevent a waterhammer when the heat exchanger is placed in service.
SAFETY EVALUATION
SUMMARY
The portion of the Service Vater System that this valve line-up change affects is classified as non safety related and is seismic category II.
It is analyzed for a line rupture per USAR paragraph 3.6.2.7.2.16.
The line-up change is to the standby cooler and was done to reduce the possibility of a water hammer when placing the standby cooler in operation.
This is accomplished by shutting the outlet isolation valve while leaving the inlet isolation valve open so that the cooler remains full of water.
A rupture in the standby cooler and the associated piping to the downstream isolation valve is bounded by the USAR description referenced above for a rupture both upstream and downstream of the cooler temperature control valve.
Thus, there is no effect on plant safety due to the valve line-up change.
Based on the above considerations, changing the valve line-up for the Service Water isolation valves on the TPCV Heat Exchanger 1-3 is safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
Update Depiction of PPF Alternate Feed CHANGE:
Change the power supply for the Personnel Processing Facility (PPF) which is shown on drawing E-4 Sheet 5, and depicted in the Updated Safety Analysis Report (USAR) as' Figure 8.3-18.
REASON FOR CHANGE:
The normal supply to the PPF is from bus C2, via disconnect DSC205 and transformer X3050.
During cycle 8R, an alternate feed was established between the Outage Support Substation (OSS) and the secondary of X3050, via a new disconnect switch (DSC3050).
SAFETY EVALUATION
SUMMARY
This change will have no effect on safety.
The PPF power supply is not important to safety, and a revision to the alternate offsite source will not involve any interaction with any other system which is important to safety.
The alternate power supply is normally disconnected from the normal PPF feed.
as well as isolated via DSC3050.
The design of alternate power supply takes into consideration the ampacity of the distribution components and the selectivity of the protective devices (i.e.,
inses and relays) to ensure that the design does not adversely affect the nonessential distribution system's reliability.
The design also ensures that the alternate feed will not allow a backfeeding of nonessential 4.16 KV bus C2 via transformer X3050.
The design of the alternate feed to the PPF via the Outage Support Substation is safe.
DBP 5302DDDD/11 i
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f SAFETY EVALUATION
SUMMARY
FOR l
DCR 93-0053, DCR 93-0057, DCR 93-0063 (SE 94-0005) i TITLE:
l Emergency Ventilation System Configuration Issues l
CHANGE:
l Revise USAR Figure 9.4-11 to depict several differential pressure tap connection arrangements sharing taps to reflect actual plant cond.t, on.
The deficiencies being corrected include placing volume control dampers CV69 and CV73 in the proper relative position in the ductwork and adding the normal positions of the ventilation control dampers in the EVS System.
REASON FOR CHANGE:
To increase the level of detail shown on the controlled drawings and USAR figure 9.4-11 for the Emergency Ventilation System (EVS) and to correct minor configuration concerns with the EVS controlled drawings and USAR figure 9.4-11.
SAFETY EVALUATION
SUMMARY
The use of a common pressure tap for the identified pressure differential indicators (PDIs) has no impact on safety.
The PDIs do not perform a safety j
function.
A common tap also does not affect the normal function since there is no significant flow rate through these taps. The PDIs provide local indication only.
The correction of the location of the volume control dampers CV69 and CV73 does l
not affect the way the system is to be operated or the analysis of the systems l
ability to maintain a negative pressure or affect the filter efficiency.
l Notation of the normal position of the air operated control dampers only adds l
additional information and does not affect the safety function or operation of l
the EVS.
Based on the above discussions it is concluded that the proposed changes are l
safe.
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SAFETY EVALUATION
SUMMARY
FOR DCR 93-0060 (SE 93-0069, R.01)
TITLE:
Closure of SV325 and Opening of SV1368 During Normal Power Operation CHANGE:
USAR Figure 9.2-1 currently depicts the service water inlet motor operated valve (HOV) SV1368 for the spare Containment Air Cooler (CAC) 1-3 as closed and the manual outlet valve SV325 as open. The described plant activity is to open SV1368 and close SV325 during normal power operation.
Additionally, the FHAR must be revised to reflect the normally open position of SV1368.
REASON FOR CHANGE:
Current operating practice for the spare CAC is to maintain open the air operated outlet temperature control valve and close the inlet MOV. During the eighth refueling outage, when the inlet H0V was stroked for a Surveillance Test, a water hammer occurred and resulted in the rupture of one of the tube bundles on the spare CAC. The water hammer event was evaluated and, it was decided to return inlet H0V SV1368 to the open position as was originally designed, and to close manual outlet valve SV325 to ensure that the spare CAC is isolated from the operating service water loop. This lineup will maintain the spare CAC full and pressurized thereby reducing the potential water hammer when it is placed in service.
SAFETY EVALUATION
SUMMARY
The spare CAC is not required to mitigate any postulated accident. As long as the manual outlet valve (s) are closed, service water to the spare CAC will remain isolated. Therefore, the potential for one service water pump feeding two CACs will not exist.
This method of isolating the spare CAC simplifies the FHAR analysis due to the elimination of a spurious opening concern regarding the CAC inlet / outlet isolation valves which would have resulted in feeding two CACs off of one operating service water train. This simplification vill eliminate required operator actions to close the inlet valve to isolate the standby CAC.
The service water air operated outlet valve (SV1358) for the standby CAC is closed during normal operation. Therefore, closure of CAC 1-3 manual service water outlet valve SW325 results in the isolation of piping between SV1358 and SV325 and SV331. This piping is located in Hechanical Penetration Room #4.
Depending on the initial temperature of the water trapped in this section of piping, the temperature would increase / decrease to approximately room temperature with a resultant change in internal pressure.
Conservatively assuming the initial conditions of the trapped water and a final temperature equal to the maximum design ventilation conditions for room 314, the internal pressure increase, based on no seat, packing or other leakage, could cause a failure of the isolated piping. This failure is of no safety significance since the spare CAC is not required to perform any safety or normal operational function and SV1358 vill remain closed thereby preventing a possible diversion of water from the operating CAC. Additionally, assuming a LOCA with a loss of off-site power (which results in the normal air supply to SV1358 being unavailable), subsequent to an undetected pipe rupture, a safety related accumulator tank would provide the air pressure required to maintain SV1358 closed thereby preventing diversion of water from the operating CAC.
Rupture of the isolated section of piping is not likely because the pressure vill be relieved via any one of the following leakage paths: leakage by the valve seat (s), leakage by the valve packing (s), and leakage at the flange joints at the manual valves.
Therefore it is concluded that this plant activity results in no adverse affects of a safety significance or otherwise to the plant during normal operation or shutdown.
SAFETY EVALUATION
SUMMARY
TITLE:
Revise Main Steam Valve Positions CHANGE:
USAR Figure 10.3-1 was revised to show that the Steam Generator Annulus Drain Valves HS859, MS860, MS883, and MS884 are normally closed and to show that the Hain Steam Isolation Valves MS100 and MS101 are normally open.
REASON FOR CHANGE:
Revise drawings to ensure that the positions for valves in the Hain Steam System are consistent in the drawings, procedures and the USAR.
SAFETY EVALUATION
SUMMARY
MS859, MS860, MS883 and MS884, the steam generator annulus Drain Valves, were shown as normally open. These valves are closed for a normal operating lineup in accordance with procedure DB-0P-06900, therefore the drawings were revised to show the valves in the correct position per the procedure. Changing the normal position of these valves indicated on the drawings vill have no effect on actual plant configuration or procedures.
HS100 and MS101 are balanced stop valves that are open for normal plant operation. A convention did not exist for showing the position of balanced stop valves, therefore the position was not previously indicated on the drawings. A new convention has been developed and the normal position of these valves vill be indicated on affected drawings. The addition of a normal position for these valves on the drawings vill have no effect on the plant configuration or plant procedures.
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SAFETY EVALUATION
SUMMARY
TITLE:
Containment Gas Analyzing System Configuration Management Issues CHANGE:
Revised USAR Figure 9.4-11A to close valves CV103, CV543, CV5027C, CV5028B and to open valves CV104 and CV544.
REASON FOR CHANGE:
DCR 93-0069 was written to correct configuration management concerns with USAR Figure 9.4-11A, Containment Gas Analyzing System, and the system drawings.
These changes will bring these documents into conformance with applicable plant procedure.
SAFETY EVALUATION
SUMMARY
The revised valve position depiction on USAR Figure 9.4-llA does not have an impact on the function of the Containment Gas Analyzing System. The normal sample path for the containment atmosphere radiation monitors was established to reduce the moisture build-up in the monitors and thus improve system reliability.
The changes in the valve positions of CV103, CV104, CV543 and CV544 aligns the radiation monitor suction to a lover elevation in containment reducing the amount of moisture that accumulates in the monitor. The revised normal position for the Hydrogen Analyzer solenoid valves (5027C and 5028B) is not a field change. The valves are interlocked with their respective pump and open only when the pump is energized. The new depiction for these valves reflects the philosophy of showing valve positions in the normal plant line-up.
This change brings the valves positions into conformance with the operating procedure.
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SAFETY EVALUATION
SUMMARY
TITLE:
Revise HU23 and MU323 Valve Fositions CHANGE:
1 Revise drawings to ensure that HU23 & HU323 valve positions are consistent in system drawings, procedures and USAR.
REASON FOR CHANGE:
l This change ensures that the position for valves HU23 and MU323 in the Chemical Addition System are consistent in the drawings, procedures, USAR and in the plant.
SAFETY EVALUATION
SUMMARY
l MU323 serves as the isolation inlet valve for the Boric Acid Addition Tank and MU23 serves as the control valve from the BA Pumps discharge. These valves do not perform any nuclear safety related function, l
The proposed change to the USAR Figure 9.3-18, to depict HU23 & MU323 in the
.)
closed position, vill not adversely impact the USAR Chapter 15 accident 1
analysis or other USAR analysis. The valves meet all previously specified Design Requirements. The changing of valve position from open to normally closed on the system drawings and USAR figure.411 make the drawings, 1
procedures and USAR figure consistent.
l Based on the above discussion it is concluded that the proposed drawing change is safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
Revise Makeup and Purification System Drawings CHANGE:
Revise USAR Figure 9.3-16 to show needle valves MU6423A, MU230, MU231, MU232, and MU233 as throttled and to annotate that the internals of the flovmeter in the bypass line around MU32 have been removed.
REASON FOR CHANGE:
Revise documents to ensure that the positions for valves in the Makeup system are shown consistent in drawings, procedures and the USAR.
SAFETY EVALUATION
SUMMARY
USAR Figure 9.3-16 currently depict MU230, MU231, MU232 and MU233 as fully open needle valves. The original drawing convention depicted all fully open or throttled needle valves as fully open. Revising these documents to depict these valves as throttled is an enhancement and is consistent with the current operating procedures.
Since the function of these valves is not affected, there is no effect on safety.
USAR Figure 9.3-16 currently depict valve MU6423A as a closed needle valve.
Revising these documents to depict this valve as throttled makes these documents consistent with current operating procedures.
Since this change does i
not affect the function of the valve, there is no effect on safety.
Since the normal position and function of these valves is not changed, the reliability of the make-up system is not effected.
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l SAFETY EVALUATION
SUMMARY
TITLE:
Revise Component Cooling Vater Valve Positions CHANGE:
Revise USAR Figure 9.2-2 to show globe valves CC670, CC672 and CC84 as throttled, and to show valves CC544, CC545, CC546, CC547, CC60 and CC61 as closed.
REASON FOR CHANGE:
l Revise documents to ensure that the positions for valves in the Component j
Cooling Water system are shown consistent in drawings, procedures and the USAR.
1 SAFETY EVALUATION
SUMMARY
1 USAR Figure 9.2-2 currently shows valves CC670, CC672 and CC84 as open. The original drawing convention showed all fully open and throttled globe valves as open. Revising documents to show CC670 and CC672 as throttled is an
)
enhancement and is consistent with the operating procedures.
Since the normal operation and function of these valves is not affected, there is no effect on l
- safety, l
w USAR Figure 9.2-2 currently shows valves CC544, CC545, CC546, CC547, CC60 and CC61 as open.
Revising documents to show these valves closed vill ensure consistency with the operating procedures and plant configuration. Closing these valves does not adversely impact CCV flow to essential components and does not impact the function of any other system. Therefore, there is no effect on safety.
Since the normal position and function of valves CC84, CC670, and CC672 is not changed, the change in drawing convention for these valves does not affect CCW l
system reliability. Closing valves CC60, CC61, and CC544-547 to isolate unused j
portions of the CCV system has negligible effect on system reliability.
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SAFETY EVALUATION
SUMMARY
TITLE:
Closure of RC46, Reactor "0" Ring Drain to Normal Sump Valve CHANGE:
Changed USAR Figure 5.1-2 to add a note stating "RC 46 may be closed if evidence of Inner "0" ring leakage exists."
REASON FOR CHANGE:
The valve's position was shown in the open position in the USAR, but there were specific allowances in Operating Procedure DB-0P-06900 to close the valve in the event that inner "0" ring leakage is detected.
SAFETY EVALUATION
SUMMARY
The actions described in the operating procedures i.e., the allowance to close valve RC 46, affects the functions of RCS Pressure boundary, by altering the pressure area for the Reactor Head enclosure from that analyzed originally.
Additionally, this change could affect the clamp force provided by the RV Head on the plenum assembly for the function of maintaining the alignment features of Reactor Internals. These changes result from the original Stress Report assuming that pressure boundary area was defined by the inner 0 ring.
By closure of valve RC 46 the pressure area becomes defined by the outer 0 ring.
I The affect of valve RC 46 closure has been evaluated and found to be acceptable with the requirements of the Certified Design Specification for the Reactor Vessel.
Furthermore these changes do not result in changes to the RV as described in the USAR, i.e.,
the Vessel is designed and fabricated in accordance with the requirements of the 1968 Edition, Summer 68 addenda of ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.
Based on the above, closure of RC 46 vill not have an adverse affect on the functions of the RV head closure.
Additionally, it is noted that closure of valve RC 46 has no affect on the USAR defined capabilities for the pressure tap in the annulus region between the inner and outer "0"
- rings, i.e.,
monitor seal leakage and hydrotest of the outer seal after closure.
Based on the above discussion the allowance to close valve RC 46 on the evidence of leakage at the inner "0" ring as provided is safe.
SAFETY EVALUATION
SUMMARY
TITLE:
Isolation of the Demineralized Vater Supply to the Primary Vater Storage Tank CHANGE:
Isolate the Demineralized Vater supply to the Primary Water Storage Tank and eliminate the automatic level control capability by the closure of DV6820 sanual isolation valves DV525 and DV526.
REASON FOR CHANGE:
The Primary Vater Storage Tank (PVST) has an air operated automatic level control valve, DV6820, which is actuated by level switches in the PVST.
Based on the limited, intermittent use of the PVST the automatic level control is unnecessary.
SAFETY EVALUATION
SUMMARY
The Primary Vater Storage System (PVSS), this portion of the Demineralized Vater Storage System (DVSS), and specifically valves DV6820, DV525 and DV526 serve no function important to safe plant operation. Therefore, closing valves DV625 and DV626 vill have no adverse affect on plant safety.
Closing valves DV625 and DV626 vill isolate the Demineralized Vater supply to the PVST automatic level control valve.
Because of the limited use of the PVSS, maintaining the water level by operator action vill not degrade the j
reliability of the PVSS. Additionally, isolating the automatic level control 4
capability vill not affect the normal operating functions of the PVSS or the DVSS.
Based on the above discussion, it is concluded that the proposed change is safe.
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P SAFETY EVALUATION
SUMMARY
TITLE:
Hydrogen Storage Cylinders Valve Positions
-CHANGE:
-The Hydrogen (H ) storage cylinders in the H supply system to the make-up tank' 2
2 have had the isolation valves open on three cylir.ders to maintain stendby status while the other three cylinders are isolated.
This change evaluates the safety effects of isolating all six cylinders as well as closing the in-line isolation valve G203.
REASON FOR CHANGE:
Revised drawings to ensure that the position of the Hydrogen cylinder isolation valves are consistent in the drawings, procedures, USAR and in the plant.
SAFETY EVALUATION
SUMMARY
The isolation of the H cylinders by closing the manual cylinder isolation 7
valves and the manual In-line isolation valve will provide two additional isolation boundaries between the H Supply System and the Makeup Tank (MUT).
2 The valves will require manual operator action when supplying H to the MUT.
2 Procedures control the process of H, addition to the MUT and requires manual operator action to unisolate the H Supply System prior to the addition'and to 2
restore after the addition.
Closing these valves will not cause any adverse overpressurizatfi.n conditions, since the piping is designed to withstand the maximum hydrogen pressure delivery. Additionally, the piping trapped between the cylinder isolation valves and the in-line manual valve G203 will'not experience significant pressurization due to the ambient temperature changes because of the compressible nature of the H '2 Based on the above discussion it is concluded that the proposed changes are safe.
DBP S302FFF/41
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l SAFETY EVALUATION
SUMMARY
FOR j
TITLE:
Revise Vacuum System Valve Positions CHANGE:
Revise USAR Figure 10.4-1 and 10.4-11 to show that the steam jet air ejector (SJAE) after condenser drain fluid is routed through normally open valve VS20, SJAE after condenser drain valve, to the condensate pumps seal vater drain tank.
Revise USAR Figure 10.4-1 to show that steam trap 27 has been removed and condenser penetration 72 is capped.
Revise USAR Section 10.4.2.2 to delete the statement that one element of the steam jet air ejector is normally in operation at any time. This statement does not reflect the actual lineup which normally uses both second stage jets in varm weather.
REASON FOR CHANGE:
i Revise drawings to ensure that the poritions for valves in the Vacuum System are consistent in the drawings, procedures and the USAR.
SAFETY EVALUATION
SUMMARY
i l
The new drain line from VS20 to the condensate seal water drain tank is 1
adequately sized to ensure the SJAE after condenser vill not be flooded.
Therefore changing the route of the aftercondenser drains vill not affect condenser vacuum.
The effect of the increased flow through the one inch line from the snal water drain tank to the condenser was analyzed.
It was determined that the line can pass the additional flow without causing high tank levels in the seal water drain tank.
The SJAE aftercondenser drain and the seal water drain tank overflow both l
discharge to the vest condenser sump. Rerouting the drains to the seal vater j
drain tank does not change the flows into any area. Therefore this change vill i
not affect flooding analysis.
Based on the above discussions it is concluded that the proposed changes are j
safe.
SAFETY EVALUATION
SUMMARY
TITLES-Revise PV87 Valve Position CHANGE:
Revise drawings to ensure that PU87 valve position is consistent in design drawings, procedures and USAR.
REASON FOR CHANGE:
To correct Configuration Management issues for the Primary Vater system.
{
SAFETY EVALUATION
SUMMARY
Primary Water (PV) valve PU87 is the manual isolation valve for the PU supply to the Solid Radwaste Drumming station.
The Radwaste Drumming Station has been already abandoned in place.- Therefore changing the position of PU87 from open to close vill not adversely impact'the operation of Primary Water System.. Changing the valve position from open to i
normally closed on the USAR Figure and system drawings vill ensure'the USAR,
~
drawings and procedures are consistent.
Based on the above, the proposed drawing change is safe.
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$f SAFETY EVALUATION SUNNARY FOR t
TITLE:
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' Revise Main Steam Valve Positions CHANGE:
Revise USAR Figure 10.3-1 to indicate valves MS2564 and MS2567 are normally closed.
I REASON FOR CHANGES-Revise drawings to ensure that the positions for valves MS2564 and MS2567 in the Main Steam System are consistent in the drawings, procedures, USAR and the plant.
SAFETY EVALUATION
SUMMARY
These valves are root valves for test points. These valves are also in the bypass line around MS706 (MS707) on the 6" sain steam line to Main Feedvater Pump Turbines. The functions of the valves are to isolate the test points and close the bypass lines.
l HS2564 and MS2567 are shown as normally open on USAR Figure 10.3-1.
These.
valves are closed in the field.
Changing the normal position of these valves from open to close vill have no effect on main steam system functions because the Test Points PP2564(PP2567) are already isolated by the normally closed valves MS2564A(MS2567A) which are downstream of the valves MS2564(MS2567). Also the bypass lines are already closed by the normally closed valves MS2565(MS2568) which are upstream of MS2564(MS2567). Therefore.this change will i
have no effect on safety. These valves do not perform any safety function.
Based on the above discussions it is concluded that the proposed changes are safe.
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d SAFETY EVALUATION
SUMMARY
FOR DCR 94-0019 (SE 94-0018) i TITLE:
Revise Position of Nitrogen Supply to the Makeup Tank Valve CHANGE:
?
Revise USAR Figure 9.3-16 to show that manual diaphram valve NN105, the isolation valve upstream of makeup tank nitrogen supply solenoid valve MU53, is normally closed.
REASON FOR CHANGE:
A procedure change altered the normal position of NN105 from open to' closed in order to prevent leakage from the makeup tank to the nitrogen system. This change revises the USAR Figure to make it consistent with the operating procedure.
SAFETY EVALUATION
SUMMARY
[
The nitrogen system can be used to maintain the total gas pressure in the makeup tank by supplying nitrogen as a substitute for hydrogen should the hydrogen partial pressure be too high. If the makeup tank total pressure is at its minimum and the dissolved hydrogen in the RCS is still higher than desired, nitrogen may be used to allow further hydrogen reduction.
Nitrogen is added per DB-OP-06033 by a batch addition process using solenoid valve MU53 to control the addition. NNJ05 is in the nitrogen supply line upstream of MU53 and must be open to permit nitrogen addition.
All functions of the nitrogen addition system can be performed with NN105 normally closed. When no nitrogen addition to the makeup tank is required MU53 is closed, therefore maintaining NN105 closed provides a backup isolation valve. NN105 can be manually opened as required to permit nitrogen addition to the makeup tank without having any effect on the functions of the nitrogen system.
Based on the above discussion it is concluded that the proposed change is safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
Revise Steam Generator Vet Layup Recirculation Valve Positions CHANGE:
Revise USAR Figure 10.2-1 to show that the Steam Generator Vet Layup Pump Isolation Valves MS 216, MS 226, MS 219 and MS 220 are normally closed.
k REASON FOR CHANGE:
Revise drawings to ensure that the positions for valves in the Steam Generator Vet Layup Recirculation System are consistent in the drawings, procedures and the USAR.
SAFETY EVALUATION
SUMMARY
The position of valves MS 216, MS 226, MS 219 and MS 220 has no effect on the interface with the Steam Generator System or the Auxiliary Feedvater System since there are other normally closed valves both upstream and downstream of these valves. These valves also have no effect on the ability to isolate the containment penetration in the line connecting the S. team Generator Vet Layup Recirculation System with the Steam Generator. They are manual valves physically remote from the containment penetration.
Based on the above discussions it is concluded that the proposed changes are safe.
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e SAFETY EVALUATION
SUMMARY
.l TITLE:
Revise Position of' AS271 and AS272
=;
CHANGE:
To show the normal position of AS271 and AS272 as closed. These valves isolate l
the Auxiliary Steam supply lines to'the Main Feed Pump Turbines-1-1'and 1-2, respectively, located just downstream of the common Auxiliary Steam supply header isolation valve AS270.
REASON FOR CHANGE:
l Revise applicable drawings to establish consistency in the normal positions for valves represented in procedures, drawings and the USAR for the Auxiliary Steam System.
SAFETY EVALUATION
SUMMARY
)
'AS271 and AS272 are isolation valves in the Auxiliary Steam supply lines to the l
Main Feed Pump Turbines (MPPTs) 1-1 and 1-2.
These valves are located just downstream of the common Auxiliary Steam supply line isolation valve AS270.
During normal plant operation these valves vould be opened when Auxiliary Steam is used as the supply when placing the NFPTs in standby mode; or closed when steam supply is not from the Auxiliary Steam system or when transferring steam supply for the MFPTs from the Auxiliary Steam header to the Main Steam header.
AS270 (which is located just upstream of AS271 and AS272 in the common Auxiliary Steam supply line) is the primary valve used when isolating Auxiliary Steam as the steam supply to the MFPTs. Closure of AS271 and AS272 vill only result in providing double isolation of the MFPTs from the Auxiliary Steam t
system. These valves do not perform any. safety related function.
Based on the above discussion it is concluded that the proposed changes are safe.
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.ac SAFETY EVALUATION
SUMMARY
FOR l
DCR 94-0030 (SE 94-0022, R.01)
TITLE:
Revise Main Steam Valve Positions i
~
CHANGE:
Revise USAR Figure-10.3-1 to' indicate valves MS700 and MS703 are normally
-closed.
REASON FOR CHANGE:
P Revise drawings to ensure that the positions for valves MS 700 and MS 703 in the Main Steam System are consistent in the drawings, procedures, USAR and the plant.
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SAFETY EVALUATION
SUMMARY
The valves MS700 (MS703) are 1 1/2" manual gate valves in the Main Steam Line
- 2 (#1) warmup drain line to condenser. These valves are upstream of the containment isolation valves MS375 (MS394). The functions of MS700 (MS703) are l
to isolate these containment isolation valves and form a system pressure boundary.
HS700 and MS703 are shown as normally open on USAR Figure 10.3-1..
These valves f
are closed in the field per DB-0P-06900, Plant Heatup Procedure.
Changing the normal position of these valves from open to'close vill not impact.any main steam system function because the downstream containment isolation valves MS375 (MS394) are already in the closed position. The closure of these valves will-relocate the high energy line break'to MS700 (MS703).from MS375 (MS394)..The
-l environmental affects of the break at the nev location remain unchanged because the size of the break is identical to the previous break. An Engineering Inspection' Team (EIT) review was performed in Room 601 and 602 to determine the-impact of break at the new location. The EIT inspection revealed that there are no new hazards created in Room 601. However the pipe whip in Room 602 could damage the air tubing from Atmospheric Vent Valve (AVV) (ICS1A) air volume tank (T143-2) to the valve operator. But the AVV is not required to j
mitigate the consequences of this break. Therefore this change will have no effect on safety.
Based on the above discussions it is concluded that the proposed changes are safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
1 Resolution of Configuration Concerns for the Station and Instrument Air System i
CHANGE:
The following changes are being made to USAR Figure.9.3-1: a) Remove the non-standard reference to Design Drawing OS-19A Sheet 2.
b) Add a standard note referencing typical valve prefix of IA.
c) Add note for valve 283 indicating installation of flow balancing orifice. d) Delete.the note references and the depiction of valves 282, 37, 16 and Station Air compressor 1-2 discharge check valve.
e) Delete 2 instrument root valves for the controls of Air Dryers 1-3 and 1-4.
f) Show the correct depiction of the purge orifice for dryers 1-3 and 1-4.
g) Show IA369 in its normally throttled i
position.
l REASON FOR CHANGE:
i Recent valkdowns of the Station and Instrument air system have identified some 1
inconsistencies between the Design Drawings, such as the P& ids and the Operational Schematics (OSs); the USAR System Functional Drawings and the Operational Procedures. The purpose of this DCR is to correct these deficiencies and to bring these drawings into agreement with each other and their applicable drawing standards.
SAFETY EVALUATION
SUMMARY
i Requested USAR Figure changes a), b), d), and e) represent changes to bring the USAR System Functional Drawings into compliance with the guidelines for these drawings. Change c) represents a design change that has already been reviewed under a previously issued design package. Change f) represents a drawing error that occurred when the USAR Figure was first created.
Change g) reflects that i
purge flow is normally maintained by throttling IA369.
The revised USAR Figure brings USAR Figure 9.3-1 into agreement with the operational procedure, the P& ids and the Operational Schematics.
None of these drawing changes impact the safety functions of'the Station and. Instrument Air Systems described above or in Section 9.3.1 of the USAR. The Engineered Safety g
Features do not rely on the supply of Station and Instrument Air for its operation.
Where air pressure is required for valve operation, safety related air accumulators are provided.
As discussed above, these changes do not adversely affect plant safety, nor does it create an unreviewed safety question.
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SAFETY EVALUATION
SUMMARY
TITLE:
Revise Low Pressure Extraction Steam System Drawings CHANGE:
Revise the Low Pressure Extraction Steam System Controlled Drawings and USAR Figure 10.4-9 to reflect the throttled position of Lov Pressure Feedvater (LPFV) Heater shell side vent valves ES20, ES22, ES26 and ES27.
REASON FOR CHANGE:
These changes are being made so that the design drawings accurately reflect the normal valve positions as identified in the system operating procedure.
SAFETY EVALUATION
SUMMARY
The function of shell side vent valves ES20, ES22, ES26 and ES27 is to vent air and other non-condensable vapors from the LPFV Heaters to the condenser to optimize efficiency of the LPFV Heaters.
The Extraction Steam System, the LPFV Heaters and specifically the shell side vent valves ES20, ES22, ES26 and ES27 serve no function important to safe plant operation. Therefore, throttling the shell side vent valves ES20, ES22, ES26 and ES27 vill have no adverse affect on plant safety.
Throttling of the shell side vent valves is required to optimize the performance of the LPFV Heaters.
Excessive venting to the condenser vill result in extraction steam bypassing the LPFV Heater with a corresponding reduction in plant efficiency.
Insufficient venting vill result in build-up of air and other non-condensables in the LPFV Heaters with a corresponding reduction in plant efficiency.
Based on the above discussion, it is concluded that the proposed change is safe.
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SAFETY EVALUATION
SUMMARY
FOR DCR 94-0037 (SE 94-0029) 1 TITLE:
Control Room Emergency Ventilation System (CREVS) Configuration Issues CHANGE:
The dampers shown in the discharge of CREVS fans C21-1 and C21-2 are not installed in the field.
REASON FOR CHANGE:
i To correct minor configuration concerns with the CREVS drawings, USAR Figure 9.4-1 and USAR Figure 9.4-5.
SAFETY EVALUATION
SUMMARY
The volume control dampers and motor air operated control dampers provide a method to balance and to control air flow in the system to the required design flows.
The CREVS existing volume control dampers are sufficient to control the air flow rate and system balance. The volume control dampers shown at the fan discharge are unnecessary and would, if installed actually reduce system performance by adding additional flow resistance.
Removal of the damper symbol from the drawings vill not affect the CREVS reliability or performance.
Removal of the dampers from the drawings vill bring the drawings into agreement with the as-built plant arrangement and thus improve plant operation.
Based on the above discussions it is concluded that the proposed changes are safe.
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SAFETY EVALUATION
SUMMARY
'FOR DCR 94-0041 -(SE 94-0063)
TITLE:
Revise Domestic Water Piping and Valve Configuration
't CHANGE:
Revise USAR Figure 9.2-5 to show revised' location of piping tie-ins and valves normally closed.
e REASON FOR CHANGE:
A Revise drawings to ensure that piping tie-in points and valve position is consistent in design drawings,' procedures and USAR.
SAFETY EVALUATION
SUMMARY
The Domestic Vater Chlorine Analyzer System does not perform any nuclear-safety-related or important to safety function. The Domestic Water system i
provides fill or makeup to domestic user equipment.
The Domestic water system components do not perform or affect any safety related function. Therefore changing the position piping tie-ins'or the position of valves from open to close vill not adversely. impact the operation of Water Treatment System. The change to the USAR Figure 9.2-5' correctly depicts the' position piping tie-ins or the position of valves from open to close. The changing of the position piping tie-ins or the position of valves from open to close on the Operational Schematic and P&ID make the drawings, USAR and procedure DB-CH-06007 consistent.
Based on the above discussion it is concluded that the proposed changes are safe.
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SAFETY EVALUATION
SUMMARY
I TITLES-Revision Of Plant Drawings Per EM-DP-1030 Drawing Guidelines-CHANGE:
Delete all normally closed (NC) from plant one-line drawings and add normally p
open (NO) to' plant one-line drawings as necessary to incorporate the-drawing guidelines in procedure EN-DP-1030, " Project Drawings". Normally closed-1:
breakers vill not be labeled and normally open breakers will be labeled with "NO"s per EN-DP-1030. This safety evaluation is being written to revise BE210 from "NC" to "N0" on drawing E-1 sheet 2.
l REASON FOR CHANGE:
Currently, breaker position on plant one-line drawings are not always marked.
During preparation of this DCR only one breaker had a definitively marked position revised to the opposite position. BE210 is currently shown on E-1 sheet 2 as "NC", normally closed.
SAFETY EVALUATION
SUMMARY
The revision of breaker BE210 from "NC" to "N0" on drawing E-1 sheet 2 does not affect the safety function of the 480 Volts Nonessential AC System because there is no change in the load on PE210 which could-affect the supplying of l
power to its various loads. Breaker BE210 supplies power to the. Polar Crane.
The drawing guidelines contained in procedure EN-DP-1030, " Project Drawings",
state that the drawings vill be shown with the plant at 100% power. The Polar l
Crane is located in the Containment Building and is used only during outage conditions.
At 100% power, The breaker BE210 is in the open position. The changes involved are safe.
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SAFETY EVALUATION
SUMMARY
FOR DCR 94-0050 (SE 94-0047) j TITLE:
Revise Extraction Steam System Dravings l
CHANGE:
Revise USAR Figure 10.4-9 to change the positions indicated for valves ES-9, j
ES-10, ES-24 and E-25 from "open" to " throttled".
REASON FOR CHANGE:
Revise drawings to change the designation of valve positions (or the Extraction Steam System to be consistent with the actual valve positions in the Plant.
SAFETY EVALUATION
SUMMARY
The function of these valves is to provide packing leakof f is >1ation or control 1
for non-return valves ES298A, ES298B, ES325A and ES325B. These components have no function important to safety.
The indicated position of the valves listed above has no ef fe:t on safety because they have no safety function. The throttled position of these valves indicates that the packing leakoff rate is controlled. This tas no impact on the function of the Extraction Steam System. Therefore, it is concluded that the proposed changes are safe.
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F SAFETY EVALUATION
SUMMARY
TITLE:
Condensate Configuration Control Issues.
l CHANGE:
The following is a list of changes made by DCR 94-0064; 1)' Changed valve position from open to closed'for valves AS 1 and AS 7, Main Condenser Hotwell Heating Control Inlet Isolation Valves, on USAR Figure 10.1-2,
- 2) Changed position of valve AS 8, Main Condenser 1-1 Hotvell Heating Control Valve Outlet Isolation Valve, from open to closed on USAR Figure 10.4-11,
- 3) Changed the position of valve CD 592, Hotvell Cleanup Drain Throttle Valve, from throttled to closed on USAR Figure 10.4-11,
- 4) Changed valve position of valves FV 114 and FW 116, Condenser Recirculating Vet Layup Pump Inlet and Outlet Isolation Valves, from open to closed on USAR Figure 10.4-12.
REASON FOR CHANGE:
Revised drawings to ensure that the positions for valves in the condensate System are consistent in the drawings, procedures and the USAR.
l SAFETY EVALUATION
SUMMARY
The change in position for these valves is to the normally maintained position of the valve for which they provide isolation function, and to the closed position for the pump isolation valves, when the pump is not normally in service. These changes do not create any new flow paths or defeat any automatic or control functions described or assumed in the USAR. Based on the i
above the position changes for the above listed components are safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
Condensate Demineralizer' Configuration Management Changes CHANGE:
Revise USAR Figure 10.4-8 to show SS1, SS2, SS3, SS4 and CD4A as open.
REASON FOR CHANGE:
To correct Configuration Management issues for the Condensate Polishing Demineralizer System.
SAFETY EVALUATION
SUMMARY
Sample valves SS1, SS2, SS3 and SS4 are normally open to maintain a continuous sample of the polisher effluent. CD4A is open or closed to control flow from condensate Polisher Demineralizer 1-1.
This change is.made to show'the sampling valves in the correct normal position of open.
It also opens CD4A so that the normal line-up of three polishing.
units in operation is reflected in the USAR figure 10.4-8.
None of the changes affect or modify any safety function.
Based on the above, the proposed changes are safe.
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SAFETY EVALUATION
SUMMARY
FOR DCR 94-0070 (SE.94-0046) 1 TITLE:
Revise Turbine Plant Cooling Vater System Drawings i
CHANGE:
Revise USAR Figure 3.6-24 to change the positions for valves CV-87, CV-91, CV-97, CV-98, CV-176 and CV-209 from "open" to " throttled" and the positions indicated for CV-115 and CV-127 from " closed" to "open".
REASON FOR CHANGE:
Revise drawings to change the designation of valve positions for the Turbine Plant Cooling Vater System (TPCW) to be consistent with the actual valve positions in the Plant.
SAFETY EVALUATION
SUMMARY
The function of the valves listed above is to provide a means of controlling the cooling water to the following heat exchangers:
- 1) Turbine Generator Lube 011 2) Hain Feedpump Lube 011: 3) Generator Stator; 4) Isophase Bus to remove heat.
The indicated position of the valves listed above has no effect on safety.
Whether a valve is considered "open" or " throttled" has no effect on any-function as long as there is adequate flow to provide cooling.- These throttled valves in the plant do provide' adequate cooling.
For the valves previously depicted as " closed", this depiction indicated that two of four heat exchangers were in standby when in fact all four heat exchangers are in service. Again, there.is no effect on TPCV system function because there is adequate TPCW available to provide the cooling to all four heat exchangers. Therefore, it is concluded that the proposed changes are safe.
SAFETY EVALUATION
SUMMARY
- TITLE:
Resolution of Configuration Concerns'For The Station And Instrument Air System CHANGE:
Revised USAR Figure 9.3-1 tos a) Nove the location of the interstage temperature device from the top of the Emergency Instrument Air Compressor (EIAC) and Station Air Compressor (SAC) 1-1 intercoolers; b) Show SA54, the temporary air compressor to Station Air Receiver 1-1 isolation valve, open; c) Add the SA prefix to valves 800 and 801; d) Revise valve 603 to reflect its proper designation as SA605; e) Show valve SA640, a typical service station root isolation valve, open; f) Add fail open designations to solenoid valves SA6508, SA6509, and SA6510.
REASON FOR CHANGE:
Recent valkdowns of the Station and Instrument air system have identified some-inconsistencies between the Design Drawings, the USAR System Functional Drawings and the Operational Procedures. The purpose of this DCR is to correct these deficiencies and to bring these drawings into agreement with each other.
SAFETY EVALUATION
SUMMARY
Requested USAR Figure 9.3-1 changes.
Item a) identifies the location of the interstage temperature device for the EIAC and SAC 1-1.
The device senses the LP exhaust temperature.
Item b) reflects the desired valve lineup to ease the transfer for supplying the station air receivers with the EIAC.
Item c) identifies valves 800 and 801 with their system prefix.
Item d) is correcting a typographical error on valve SA605.
It is listed as 603.
In order to maintain air pressure at the service stations throughout the plant, the air line root isolation valves are being opened.
Item e) is a typical depiction of these service stations and therefore SA640 should be shown open. Pressurizing-up to the connection valve decreases the probability that a remote root valve vill be unisolated while the connection valve is open.
Item f) clearly identifies the failed position of these drain valves.
Failing open provides positive indication of an equipment problem to plant operators.
None of the above changes impact the safety functions of the Station and Instrument Air Systems described above or in Section 9.3.1 of the USAR. The Engineered Safety Features do not rely on the supply of Station and Instrument Air for operation. Vhere air pressure is required for valve operation, safety related air accumulators are provided.
Based on the above discussions, it is concluded that these changes are safe.
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SAFETY EVALUATION
SUMMARY
]
TITLE:
Configuration Concerns for the Makeup Water Treatment System CHANGE:
i Revised USAR Figure 9.2-3'by deleting valve WT6721, Acid flow control valve and' added valves WT6720, WT472 and WT473. Regeneration Acid flow control valves.
REASON FOR CHANGE:
Revised documents to ensure that valve positiens and labels.in the Makeup Water Treatment System are consistent between drawings, procedures and the USAR.
SAFETY EVALUATION
SUMMARY
The proposed valve position and configuration changes will not adversely impact-the operation of the Makeup Water Treatment System.
The proposed change to USAR Figure 9.2-3 to depict correct valve position and configuration changes.
.will not adversely impact the USAR Chapter 15, Accident Analysis, or other USAR analysis.
The valves meet all previously specified design requirements.
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. Based on the above discussion it is concluded that the proposed activity is safe.
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DBP $302FFF/38
j SAFETY EVALUATION
SUMMARY
TITLE:
Revise Auxiliary Steam Valve Positions CHANGE:
Revise USAR Figure 10.1-2 to show that AS268 and AS229, the flash tank pump discharge header throttle valves, are normally in the throttled position.
USAR' figure 10.1-2A is revised to shows that AS468, the 10 psig condensate tank recirculation throttle valve, is normally in the throttled position; that AS150 and AS290, the neutralizing tank heat exchanger steam control valve isolation valves, are normally closed; that AS379, auxiliary steam to the miscellaneous vaste evaporator and the degassifier, is normally closed.
USAR'Ffgure 9.2-5 is revised to show that AS126, the auxiliary steam to the domestic water heater isolation, is normally closed.
REASON FOR CHANGE:
Revise applicable drawings to ensure that the positions for valves in the auxiliary steam system are consistent in the procedures, drawings and the USAR.
SAFETY EVALUATION
SUMMARY
Throttling AS268, AS229 and AS468 vill ensure the flash tank pumps and the 10 psig condensate tank pumps normally operate at an appropriate discharge The recirculation flow rate of these pumps does not affect equipment pressure.
safety functions, therefore throttling these valves vill have no effect on safety.
The neutralizing tank heat exchanger is no longer used and performs no safety function. Therefore, closing AS150 and AS290 to isolate steam to the heater will have no effect on safety.
The domestic vater heater is not used and performs no safety function.
Therefore closing AS126 to isolate stem to.the heater vill have no effect on safety.
The miscellaneous vaste evaporator and the degassifier are not used and perform no safety function. Therefore, closing AS379 to isolate steam to these components vill have no effect on safety.
Based on the above discussions it is concluded that the proposed changes are
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safe.
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SAFETY EVALUATION
SUMMARY
TITLE:
Configuration Concerns for the Primary and Demineralized Vater Systems CHANGE:
Revise USAR Figure 9.2-4A to:
a) Show the Component Cooling Vater (CCV) Surge tank fill line valve PV36 closed.
b) Show the pressurizer quench tank fill line valve PV56 closed.
c) Show the Spent Resin Storage Tank supply line valve PV3 closed. d)
Show the Primary Water Storage tank heater outlet isolation valve PV 48 closed. e)
Show the Degasifier supply line valve DV115 closed.
Delete reference to Misc Vaste Evaporator Drain Pumps.
f)
Show the Boric Acid Evaporator supply line valves DV128 and DV129 closed. g)
Show the Primary Water Storage Tank Heater inlet isolation valve PV46 open.
h)
Shov the Moisture Separator Reheater Drain (HSRD) supply line isolation valve DV28 open.
1)
Add additional valves DV29 and DV251 to the office b1dg demin water supply line. j)
Add Sample Panel supply isolation valve SS164.
k)
Show the proper identification for valves NN6340 and NN6341.
REASON FOR CHANGE:
Recent walkdowns of the Primary and Demineralized Vater systems have identified some inconsistencies between the Design Drawings, such as tha P& ids and the Operational Schematics (OSs); the USAR System Functional Drawings; and the Operational Procedures. The purpose of this DCR is to correct these deficiencies and to bring these drawings into agreement with each other and their applicable drawing standards.
SAFETY EVALUATION
SUMMARY
Requested USAR Figure change a) involves closing a valve in an already isolated line. Another manual valve be opened prior to flow being initiated to the component serviced. Therefore, there is no functional change in system operation.
Item b) involves the manual isolation of the makeup line to the pressurizer quench tank.
USAR Section 5.1.7 indicates that Pressurizer Quench Tank is designed for automatic water level control.
If a lov vater level condition is encountered the system is designed to automatically add primary water.
Isolation of PV56 vill require manual operation to restore a pressurizer quench tank lov level.
Manual isolation is preferred to eliminate leakage past the control valve PV225B vhich would cause an unaccounted for rise in the pressurizer quench tank vater level.
Since automatic level control is not a function important to safe operation, manual isolation of PV56 is acceptable.
Components associated with the balance of items (c through k) perform no safety related functions.
Detailed Operating configurations are not described in the USAR text.
Thus, the changes are related to configuration control only and do not affect any functions described in the USAR.
SAFETY EVALUATION
SUMMARY
TITLE:
i Configuration Concerns for the Makeup Vater Treatment System CHANGE:
The following changes are being made to USAR Figure.9.2-3:
a)
Show VT19, Chlorine Detention Tank outlet Cross Connect Valve, closed.
1 b)
Show VT197, Circulation Pumps Gland Cooling Supply line isolation valve closed.
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REASON FOR CHANGE:
Recent walkdowns of the Makeup Vater Treatment System have identified some inconsistencies between the Design Drawings, such as the P& ids and the Operational Schematics (oss); the USAR System Functional Drawings and the Operational Procedures. The purpose of this DCR is to correct these deficiencies and to bring these drawings into agreement with each other and their applicable drawing standards.
SAFETY EVALUATION
SUMMARY
The function of valve VT19 is to isolate flow from a chlorine detention tank to' the main circulating water pumps. The function of VT197 is to isolate clearvell transfer pump discharge flow to the main circulating water pumps.'
Requested USAR Figure 9.2-3 change a) shows the normal discharge flow path of the two chlorine detention tanks separated.
Change b) reflects that the clearvell transfer pumps are no longer required for providing seal water to the main circulating water pumps. Neither change affects any function important to safe plant operation.
Based on the above discussions, it is concluded that these changes are safe.
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t SAFETY' EVALUATION
SUMMARY
.i FOR l
TITLE:
Throttling Component Cooling Vater and Turbine Plant Cooling Water Outlet Isolation Valves CHANGE:
Revised the Service Water System (SV) and Component Cooling Water System (CCV) design drawings, including USAR figures, to reflect the normally " throttled" position of the Component Cooling Water Heat Exchanger. Outlet Isolation Valves (SV36, SW37, SW38), the Turbine Plant Cooling Water Heat Exchanger Outlet Isolation Valves (SV54, SV55, SV56), and the Makeup Pump (MUP) CCW Outlet Valves From Lube 011 Coolers (CCl29, CCl30).
REASON FOR CHANGE:
l These changes were made so that the drawings accurately reflect the normal-valve positions as identified in system operating procedures.
SAFETY EVALUATION
SUMMARY
Throttling valves SW36, SV37 and SV38 is necessary to ensure'that the Service Vater System can supply adequate cooling water to the other safety related heat j
exchangers under design basis accident conditions.
Failure to throttle these valves would result in excessive flow through the CCW Heat Exchangers which could impact the capability of the other safety related heat exchangers cooled by Service Water.
Depicting these valves as " throttled" has no effect on plant safety.
Throttling valves CC129 and CC130 is performed to minimize flow to the MUP Lube Oil Coolers to prevent overcooling of the pump and gear lube oil.
Depicting i
these valves as " throttled" has no effect on plant safety.
Throttling valves SV54, SV55 and SV56 is required to ensure sufficient j
backpressure is maintained on the SV pump supplying the non-essential header.
)
This action serves to prevent a SW lov header pressure trip from occurring due to excessive SV flow through these heat exchangers. These valves are installed in the portion of the SV system that performs no function important to safety i
and therefore this activity has no effect on safety.
Based on the above discussion, the above changes are safe.
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SAFETY EVALUATION
SUMMARY
TITLEt
- Revise Station Drains and Sumps System Drawings CHANGES-The following is being deleted from USAR Figure 9.3-4: equipment that is abandoned in place; valve VM39, es is already shown on USAR figure 11.2-3; the recirculation line for the Marsh Transformer Vault Sump Pump discharge since it does not exists the portion of the figure that shows Sampling Panel C3401 as a source to the Condensate Pit (West) since other sumps are not shown on the figure.
USAR Figure is also revised to add the Station Blackout Diesel Generator Building sump and associated equipment and to revise the description of the discharge from the Condensate Pit Sumps from "to Condensate Demin Hold-up Tank" to "to Settling Basin No. 1 via Condensate Demin System". This reflects the fact that the normal flow path is to the Settling Basin.
REASON 'CHL CHANGE:
Drawings associated with the Station Drains and Sumps System were revised to reflect the configuration existing in the station, to provide consistency in information presented in figures, and to identify previously abandoned equipment as abandoned in place.
SAFETY EVALUATION
SUMMARY
None of the affected components have any function important to safety. Also,
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the changes being made via this activity do not impact the portions of this system used to perform functions important to safety.
Therefore, it is concluded that the proposed activity is safe.
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-y SAFETY EVALUATION
SUMMARY
FOR-
TITLE:
' Revise Valve Positions on Auxiliary Boiler Drawings CHANGE:
l The following changes in indicated valve positions are being made to the USAR Figure 10.1-2: CD-19 and CD-21 from "open" to " closed"; AS-1666A, CD-1666B,
~
' CD-1666C, SS-800 and SS-801 from " closed" to "open"; AS-14 from "open" to
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" throttled".
REASON FOR CHANGE:
Revise drawings associated with the Auxiliary Boiler to change the indicated r
positions for various valves to match plant procedures.
St."ETY EVALUATION SUMMAR*i The Auxiliary Boiler system's function is to provide rteam for heating and.
process needs during plant shutdown periods and to provide steam for plant startup and low load operation.
Changing the indicated positions of valves CD-19, CD-21, CD-1666B, CD-16660, has no effect on safety since none of these valves are in.the portion of the system containing high energy fluid; therefore, they have no effect on the system's ability to control which sections of piping are classified as high energy lines. The flow path downstream of S5-800 and SS-801 is 3/8 inch tubing and is physically distant from any Seismic I equipment used for safe shurdown therefore it is outside the scope of high energy line analysis and changing the normal position for SS-800 and SS-801 has no effect on safety.
Steam would be present downstream of AS-14 whether it is "open" or." throttled", therefore changing the indicated position from "open" to " throttled" has no effect on safety. AS-1666A modulates in response to Demerator Storage. Tank level. -The USAR figure is changed solely to meet the convention of depicting modulating valves as "open", therefore there is no impact on the function important to safety.
It is concluded that this activity has no effect on any equipment important to safety.
Based on the above discussions it is concluded that the proposed changes are safe.
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SAFETY EVALUATION
SUMMARY
l FOR
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TITLE:
Auxiliary Radioactive and Fuel Handling Ventilation System Configuration Issues CHANGE:
Revise USAR Figure 9.4-9 to add equipment detail, clarify air paths and correct nomenclature.
REASON FOR CHANGE:
To clarify the air flow paths, correct nomenclature and add equipment details to update, to the latest configuration control guidance, the Operational Schematic, P & ID, and USAR figure 9.4 9 for the Auxiliary Building Radioactive Ventilation Systems.
SAFETY EVALUATION
SUMMARY
Changes to the Fuel Handling Area ventilation drawings included adding details of the supply to Room 401 not previously shown. This ventilation path is part of the initial plant design, however it was not correctly represented on the referenced drawings. This ventilation duct was shown exhausting into the adjacent room (Room 300). This change has no effect on any analysis or safety function, as the actual air flows from Room 401 out to Room 300 and is then l
exhausted as previously shown. The negative pressure in the area is unaffected
(
and there is no affect on the ability for the EVS to take a suction.on the Fuel Handling Area in case of a fuel handling accident.
Changes to the Radvaste Area Ventilation Drawings were to clarify some of the l
ductless air flov (transfer) paths, add duct connections to the Clean Vaste Monitor Tank filter area, to the Demin Filter Casks, and to the Seal Injection Filter casks, and to add a volume control damper to the exhaust from Room 107.
l The room names for Rooms 417, 417A and 419 in the Access Control Area were revised and an air duct was added supplying corridor 404. These changes have l
no effect on the ability to maintain the Radvaste Area at a negative pressure with respect to clean areas of the Auxiliary building, nor do they affect being able to line up the EVS to ventilate these areas.
Changes to the Make-up Pump Room Air Conditioning included updating details to l
the drawings from walkdowns such as deleting a non-existant filter and bypass i-valve at the compressor suction line, removing the symbol for a liquid line l
trap, removing a non-existant bypass line around the filter and numbering and L
showing the proper location of the manual system isolation valves. These changes have no effect on how the system is operated or on its design capacity.
Based on the above discussions it is concluded that the proposed changes are safe.
L
SAFETY EVALUATION
SUMMARY
FOR DCR 94-0098 (SE 94-0081, Rev. 01)
. TITLE:
Gaseous Radwaste Configuration Management Issues CHANGE:
Revised plant drawings to show the valve positions matching current valve line-ups based on the Waste Gas Compressor No. 3 (WGC #1) lined up to the.
Waste Gas Surge Tank and WGC #2 lined up to the Nitrogen (cover gas) Header, no release is in progress.
REASON FOR CHANGE:
Revised USAR Functional Drawings, and system drawings for the Gaseous Radwaste System to depict them in the same line-ups.
SAFETY EVALUATION
SUMMARY
The valve positions of WG2860, WG2862, WG1803, WG1811, WG1823, WG1826, WG1837 and WG1840 are consistent with the described line-up with the Waste Gas compressors not running.
No safety functions are affected.
Isolation of WG122 provides additional leakage. protection for.the pressure regulatot WG1849.
It is opened when the regulator is placed in service and it performs no safety function.
WG127, WG128 WG129, WG133 WG137, WG140 and WG141 are closed when not making a controlled release.
These valves isolate the moisture traps and RE1822A and RE1822B on the Waste Gas Discharge line.
The isolation of these valves prevents leakage of radioactive gases when the line is not in use.
The system functions are shown in an acceptable line-up and actual valve position will be controlled by procedures.
The Radiation Monitors are required to be placed in service prior to making a release. The valves perform no safety ~ function.
Revising the position of NN78, NN79, NN117 and NN158 to normally closed does not affect safety because this has no effect on the operation of the Nitrogen System containment isolation valves.
No accidents described in the USAR can be initiated by the regulators isolated by these valves.
Closing of the above valves help limit the generation of excessive Waste Gas by leakage through the Pressure Regulators if they were left in service.
The nitrogen blanket pressure is maintained manually.
The acceptability of this method has already been evaluated in Safety Evaluation SE90-0058.
The reliability of the Nitrogen System or the Gaseous Radwaste System is not affected by this change.
-f DBP 5302FFF/39
)
SAFETY EVALUATION
SUMMARY
TITLE:
Change Normal Position for Valve VT105 CHANGE:
Revise drawings' associated with the Makeup Vater Treatment System to change the normal position for valve VT105, Fire Water Storage Fill Valve, from "open" to
" closed".
The FHAR vas also revised to indicated that the Fire Vater Storage Tank (FWST) cannot be filled within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
REASON FOR CHANGE:
VTIO5 was closed to eliminate the concern of exceeding EPA chlorine limits in the Clearvell when the FWST pump was operating.
SAFETY EVALUATION
SUMMARY
This activity affects the means of providing a water supply to the FVST and therefore indirectly affects the means of providing fire suppression capability. The function of valves VT105 and VT1051 is to control the flow of water to the FVST.
Since valve VT105 vill be normally closed the flow path from the clearvell to the FWST will normally be isolated.
VT1051 vill therefore no longer provide automatic level control for the FWST. When the level in the Fire Water Storage Tank reaches its low level limit valve VT1051 vill automatically align to supply it with water, but water vill not flow until WT105 is manually opened.
This activity does not compromise safety because manual initiation of flov to the Fire Vater Storage Tank is adequate.
Procedure DB-OP-03007, Hiscellaneous Instrument Daily Checks, requires that the FWST level be recorded on a daily basis. There is also an alarm in the control room to alert the operating staff if the tank level falls below 33 ft - 0 inches.
Procedure DB-0P-02009, Plant Services Alarm Panel 9 Annunciators, vill be changed concurrent with this DCR to alert operating personnel of the need to manually open and close VT105.
This ensures the ability to maintain the FVST at a level above 30 ft - 0 inches as required by the Fire Hazards Analysis Report.
The FHAR was also revised to identify that the FWST can not be refilled within eight hours. This deviation does not impact the fire suppression system's function since the diesel fire pump has an unlimited source of water.
Based on the above discussions it is concluded that the proposed change is safe.
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SAFETY EVALUATION
SUMMARY
FOR' DCR 94-0115 (SE 94-0078).
' TITLE:
Correct SFAS Output' Module Fover Sources CHANGE:
Correct'the SFAS vendor drawings to show the as-built configuration.
REASON FOR CHANGE:
The Safety Features Actuation System (SFAS) has output modules which provide the 2-out-of-4 logic and the relay actuations (de-energizations) which in turn actuate the required equipment. There are three (3) 15 VDC power supplies in-each channel which provide power to these output modules. The SFAS vendor drawings depict which power supply provides power to each output module. The-power supplies shown for output modules L272 and L274, for channels 2 and 4
.respectively, vere incorrect.
SAFETY EVALUATION
SUMMARY
The correction of the channel 2 and 4 dravings vill make the power supply distribution identical to the channel 1 and 3 distribution and drawings. As further evidence, there has been no indication of overload of the power supply
~
during approximately 20 years of operation with the present power distribution. Therefore, the reliability of the actuation' equipment vill remain the same.
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SAFETY EVALUATION
SUMMARY
FOR DCR 94-0119 (SE.94-0082)1 TITLE:
Miscellaneous. Reactor Protection System Related Drawing Corrections CHANGE:
Revised RPS drawings which are incorporated into the USAR by reference. These changes reflect as-built, as-designed information regarding the Reactor Protection System (RPS).
l REASON FOR CHANGE:
This update is in part remedial action for PCA0Rs 94-0543, 94-0656, 94-0964 and 94-1059.
In addition, changes of a minor nature are also being made to RPS schematic drawings.
SAFETY EVALUATION
SUMMARY
PCAO 94-0543:
One of three RPS Channel 1, Reactor Coolant Flow Loop "A"' Square Root extractor module test jacks were shown on drawing M-536-43. - This DCR adds the test jacks to the drawing. As the test jacks are used for testing and monitoring which is not a safety function of the RPS, no safety function is affected.
PCAO 94-0656:
Isolation Amplifiers for Ion Chamber Linear Amplifiers were shown inconsistently on drawings M-536-29,39,49,59, and M-720I..This DCR correctly shows the Ion Chamber Linear Amplifiers on the drawings. As the inolation amplifiers are used for signal testing, monitoring and buffering, which is not a safety function of the RPS, no safety function is affected.
PCAQ 94-0964:
RPS Reactor Coolant Flow Instrument String Shields Grounded in two places.
FCR 78-0525A made changes-to RPS schematics that should have resulted in a single grounding point for these shields. This DCR correctly removes these deleted grounding points from drawings M-536-38, 48, 58, 67.
Correctly depicting an instrument shield ground has no adverse effect on any safety function of the RPS, therefore no safety function is affected.
i PCA0 94-1059:
The Auctioneered Nuclear Instrumentation power signal input to RPS shield ground was shown incorrectly on M-536-48.
This signal is routed through RPS to the Start Up Test Panel.
Correctly depicting the shield ground has no adverse effect on safety.
Minor changes were made to the description of the Power /R.C. pumps trip output to show that it goes to the plant computer.
As this output has no safety function, there is no adverse effect on safety.
These drawing changes have no effect directly or indirectly on the safety functions of the RPS or on the ability of the RPS to perform its safety functions. Therefore there is no adverse effect on safety as a result of these drawing changes.
i-1 SAFETY EVALUATION
SUMMARY
i FOR DCR 94-0125 (SE 94-0087).
l l
TITLE:
Addition of Valve AS575 to System Drawings CHANGE:
Added valve AS575, Condensate return header isolation valve, in the open position downstream of ST104 to USAR Figure 10.1-2A and other system drawing.
REASON FOR CHANGE:
Revised system drawings and USAR Figure 10.1-2A to reflect what exists in the field.
SAFETY EVALUATION
SUMMARY
The addition of AS575 to USAR figure 10.1-2A in the open position reflects the "As Built" condition of the plant. This valve does not provide any safety function. The function of AS575 is to maintain system pressure boundary and to isolate components for maintenance. These functions are unaffected. AS575 in either the open or closed position vill not affect the auxiliary steam header performance. Based on the above, the addition of valve AS575 in the open position in the USAR figure to reflect the plant configuration is safe.
SAFETY EVALUATION
SUMMARY
TITLE:
Change Normal Position of Valves SV5421,SV5422, SV5424 and SV5425 from " Closed" to "Open" CilANGE:
Change the normal positions of valves SV5421, SV5422, SV5424 and SU5425 from
" closed" to "open".
Breakers BE1135, BE1136, BF1165 and BF1166 which supply power to the motor operators for the above valves vill be opened to ensure that the valves remain open. These valves are in the discharge lines for ECCS Room Coolers 1, 2, 4 and 5.
ECCS Room Cooler 3 does not have a motor operated valve in its discharge line.
REASON FOR CHANGE:
To ensure adequate flov to the ECCS Room Coolers is available for SFAS operation and to enhance flushing of the ECCS piping during normal operation.
SAFETY EVALUATION
SUMMARY
Changing the normal position of SV5421, SU5422, SV5424 and SV5425 does not impact the ability of the room coolers to maintain a suitable room temperature.
Placing these valves in the "open" position ensures that they vill be in their design basis position. Calculation C-ME-11.01-130 Revision 01 has been generated to demonstrate that adequate flow is available for all the safety related heat exchangers supplied by the Service Vater System for both normal and emergency operations.
Calculation C-ME-011.05-001 rev. O established a maximum flow rate through the coolers based on vinter operating conditions (and " clean" piping) to determine the potential for tube erosion. The corresponding flow rate under "vinter" operation, which varies based on total system demand, could be as high as 250 GFM. This flow rate results in a nominal tube velocity in excess of 13 ft/see which exceeds the manufacturers recommendation of 8 ft/sec, creating potential for tube erosion. Tube failure due to flow induced vibration effects is not considered credible based on the coil design.
Tube erosion is not expected to result in catastrophic failure of a tube based on the ductile characteristics of the tube material.
Additionally, failure of multiple tubes simultaneously is not expected based on the varying exposure to erosion.
Tube erosion vould ultimately result in minor leakage which would be detected based on ECCS room sump activity and by operator observation of water on the floor of the ECCS room. Operations personnel vould then take actions to initiate maintenance of the affected cooler. Thermal performance of the cooler vill not be impacted because leakage would not be significant. Therefore, tube leakage would not render the cooler incapable of performing its intended function. Based on the above, the potential for tube erosion is considered a maintenance issue which does not affect plant safety.
Opening the breakers will remove power from the indicating lights. This is not a concern since valve position can still be checked by methods such as observation of the valve stem or manual operation of the valve operator.
Additionally, because this actielty is not being performed to design against a single failure in the electrical system, the guidance contained in ICSB 18, which requires operable, redundant, status indication in the control room, does not apply.
Based on the above discursion, the proposed change'is safe.
SAFETY EVALUATION
SUMMARY
)
i l
TITLE:
Containment Drain Header Isolation Valves CHANGE:
Changed the valve position fot valves RC1773 A and RC1773 B, Containment drain j
header isolation valves, from normally open to normally closed. The actual'
)
normal position of valve RC1773 B is " closed".
The actual normal position of valve RC1773 A vill be changed to " closed."
REASON FOR CHANGE:
Revised drawings to ensure that the positions for valves RC1773 A and RC1773 B are consistent in the drawings, procedures and the USAR.
SAFETY EVALUATION
SUMMARY
RC1773 A and RC1773 B are the containment isolation valves in the drain line to the Reactor Coolant Drain Tank. These valves are not part of the RCS pressure boundary.
Since the affected components are containment isolation valves their safety position is to be closed. The change in the valves position has no effect on safety since the valves are in their safety position (i.e, closed).
l Vhen RC1773 A and B are closed, their associated solenoid valves are de-energized therefore the solenoid valve life is not adversely affected by changing the position of RC1773 A to closed.
Based on the above, it is concluded the proposed activity is safe.
1 1
r
SAFETY EVALUATION
SUMMARY
TITLE:
Revise Position of Nitrogen Cover Gas Supply Valves to the Demineralized Water Storage Tank 1-2 CHANGE:
t Revised station drawings and USAR Figure 9.2-4A to show that manual isolation valves NN6336, NN6337, NN6338, NN6339, and NN6341, Nitrogen sparging valves to Demineralized Vater Storage Tank, are normally closed.
REASON FOR CHANGE:
System walkdown for the Demineralized Vater Supply System found that the above valve positions shown on drawings need to be revised in order to show that the Nitrogen from the Nitrogen Bottles is normally not lined up to the tank.
SAFETY EVALUATION
SUMMARY
)
The nitrogen sparging headers in the tank can be used to provide a nitrogen
" bubble cloud" in the Demineralized Vater Storage Tank to reduce the level of oxygen when necessary. This is a secondary method for oxygen control. The primary method is to remove the oxygen at the Water Treatment System Vacuum Degasifier and then to minimize absorption by use of the floating roof in-the tank.
Showing the Nitrogen valves to the Demineralized Vater Storage Tank closed on the referenced drawings puts them in conformance with the original design in FCR 78-0048.
Based on the above discussion it is concluded that the proposed change is safe.
j 1
l SAFETY EVALUATION
SUMMARY
FOR
TITLE:
Replacement of COM-5 Relay.in AD113 CHANGE:
Replaced a Westinghouse COM-5 relay in Unit 13 of the 4.16 kV essential switchgear D1.
The associated switchgear, AD113, provides power to Component Cooling Water Pump 1-2.
REASON FOR CHANGE:
Class IE Westinghouse COM-5 relays are no longer available, therefore an Asea Brown Boveri (ABB) COM-5 equivalent is to be used as a replacement.
SAFETY EVALUATION
SUMMARY
The proposed replacement will have no adverse effect on safety.
The Material Engineering Evaluation provided with this FPR shows that the COM-5 relay is a functionally equivalent and seismically acceptable replacement for the Westinghouse COM-5 relay. Only the SSC-T unit of the ABB COM-5 overcurrent sensing elements is different from the corresponding units in the installed Westinghouse COM-5 relay.
Inadvertent actuation of that unit can not trip AD113 prematurely, and failure to actuate will not preclude a trip, since the device contains an alternate instantaneous trip device.
The SSC-T-unit interacts with the rest of the COM-5 device through a current transformer and a set of contacts in essentially the same way that the Westinghouse equivalent (ITH) unit interacts. Also, the alternate instantaneous trip device curve does not cross (and therefore coordinates with) the curve for the nearest upstream overcurrent protection device.
The purpose of the COM-5 relay is twofold.
One, the relay limits overcurrent damage to the individual motor load and to the downstream distribution components.
Secondly, the relay establishes a protective zone which prevents a downstream overcurrent f rom disabling otherwise unaffected equipment. An analysis performed and documented within the body of previous Safety Evaluation concluded that these two goals of the COM-5 relay where accomplished by the new ABB COM-5 equivalent replacement.
After reviewing the scope of work as well as the above effects on safety review engineering concludes that the proposed COM-5 equivalent replacement is safe.
r DBP 3302FFF/25
SAFETY EVALUATION
SUMMARY
FOR FPR 91-0055-001 (SE 94-0011)
TITLE:
Configuration Concerns Regarding Drains CHANGE:
Revise USAR Figure 11.2-3 to show drains from Aux Bldg Area 7 El. 603'-0" in a single line with reference only to piping plan drawing H-190.
H-190 shows many tie-ins from many sources, such as sinks, floor drains and wash facility.
REASON FOR CHANGE:
To ensure that depiction of non-0 Chemistry Lab floor and sink drains and Detergent Wash Facility drains are consistent in design drawings, procedures and USAR.
SAFETY EVALUATION
SUMMARY
The proposed drawing change, to show the drains by reference to the piping plan, vill not adversely impact the operation of Hiscellaneous Liquid Radvaste System. The proposed change to the USAR Figure 11.2-3, to depict the drains similar to design drawings, vill not adversely impact any USAR analysis. The drains meet all previously specified Design Requirements. The changing of drains to reference the piping plan drawing on USAR figure 11.2-3 vill make the drawings and USAR figure consistent and more accurate.
Based on the above discussion it is concluded that the proposed drawing change is safe.
i SAFETY EVALUATION
SUMMARY
FOR FPR 92-0008-002 (SE 94-0032, R 01)
TITLE:
Revised Main Steam to the Turbine Generator Pipe Break Analysis CHANGE:
j This change revised Stress Problems 10A and 10B, revised the pipe break analysis contained in the revised stress analysis and modified hangers EBB 1-H1 and EBB 1-H3.
USAR Figures 3.6-10 and 3.6-11 are also changed to reflect the line breaks determined in stress problems 10A and 10B are in accordance with i
the requirements defined in NRC Standard Review Plan 3.6.2, Branch Technical Position MEB 3-1.
REASON FOR CHANGE:
A proposed modification required a new pipe break analysis. The pipe break analysis also results in a revised stress analysis. The postulated break locations in the revised analysis are determined in accordance with NRC Standard Review Plan 3.6.2, Branch Technical Position MEB 3-1 as permitted by Generic Letter 87-11.
SAFETY EVALUATION
SUMMARY
The function of secondary pressure boundary is unchanged by the deletion of the postulated piping breaks.
Piping stresses for the modified piping system have been verified to be within the code stress allovables and all supports, restraints and anchors have been analyzed for the new loads.
Implementation of this change has no affect on the function of containment isolation, ie. the attributes of the system which provide containment isolation function are unaltered by the scope of work.
Implementation of this change has no affect on the function of providing a release path to atmosphere for decay heat removal, ie. the attributes of the system which provide the release path to atmosphere are unaltered by the scope of work.
It is further noted that implementation of this change does not affect either the normal or post accident environment for Rooms 601 and 602, nor does it install any new equipment that could create a hazard for existing hardware.
Based on the above discussion implementation as described in FPR 92-0008-002 have no affect on safety for the affected structures.
i SAFETY EVALUATION
SUMMARY
FOR FFR 92-0040-003 (SE 94-0009)
TITLE:
Supply Vater to Lime Feed System CHANGE:
1 Installed a larger supply line to the lime feed mixing tank in the Vater Treatment Facility. To facilitate installation, an isolation valve was installed at the source of the header and the individual lime tank isolation i
valves was relocated.
j REASON FOR CHANGE:
The water supply line to the lime feed mixing tank in the water treatment system does not supply adequate flow.
SAFETY EVALUATION
SUMMARY
The portion of the Makeup Water Treatment System (MVTS) which is routed through Seismic I structures is analyzed to determine the environmental effects of a postulated rupture on safe shutdown.
Since the piping changes addressed in this change are in the non seismic water treatment building, there is no effect on the hazards analysis.
The net effect of these changes is to improve the reliability of the MVTS by upgrading the Lime Feed Equipment.
1 l
SAFETY EVALUATION
SUMMARY
FOR FFR 92-0075-29 and FFR 92-0075-30 (SE 94-0010, R.01)
TITLE:
Revise Control Logic For AF3870 and AF3872 CHANGE:
Change the closing direction control logic of AF3870 and AF3872 so that they will be controlled for the entire closing stroke by the limit switch.
REASON FOR CHANGE:
The new control scheme for AF3870 and AF3872 has the torque switch bypassed to ensure that actuator motor operation vill not be interrupted by torque switch actuation at any point of the valve closing sequence. The limit switch will be set to stop the motor when the valve is fully closed.
SAFETY EVALUATION
SUMMARY
The changes in the control scheme enhance the reliability of the closing function of these valves for the following reasons: (1) Actuator stall torque capability is available during the full valve closing stroke; (2) Inadvertent torque switch actuation during any portion of the valve close stroke is eliminated; (3) Factors involving torque switch setpoint accuracy including stem nut friction, rate of loading and torque switch setting accuracy are inconsequential.
In the event of an isolable steam line break that initiates an SFRCS low pressure signal, AF3870 (AF3872) vill receive a closing signal followed by an opening signal af ter the ' low pressure trip clears.
For this condition motor burnout while closing would prevent the subsequent opening of the valve.
If the single failure of AFV pump 1-2 (AFV pump 1-1) is assumed and the operator motor of AF3870 (AF3872) has failed during valve closing, the result would be the loss of all AFV flow to the steam generator. Calculations show that AF3870 and AF3872 would be capable of closing without damage to the operator motors for the condition of an isolable steam line break with a single failure of the opposite AFV train at reduced voltage conditions, therefore subsequent opening operations vill not be affected.
When closing during design conditions, with the steam generator fully depressurized, the valve continuous service rating and the seismic limit vould be exceeded. However, a one time overthrust at 250 percent of the rating is allowed by the actuator manufacturer. The seismic limit vould only be exceeded during the valve closing stroke at design differential pressure. Since a seismic event is not required to be postulated simultaneously with an initiating event that would depressurize a steam generator, exceeding the seismic limit during valve stroke vill not cause component failure.
Based on the above discussions it is concluded that the proposed changes are safe.
SAFETY EVALUATION
SUMMARY
FOR FPR 92-1236-901 (SE 93-0019)
TITLE:
Replacement of General Electric Type 4701 Transducers CHANGE:
Safety Evaluation 92-0054 was written to evaluate the effect of replacing GE type 4701 watt and volt transducers on the diesel generators, using functionally equivalent transducers made by Rochester Instrument Systems, Incorporated.
This safety evaluation extends SE 92-0054 to include other type 4701 transducers located in the 4.16 kV, 13.8 kV, 25kV, and 345 kV Electric Metering Systems.
FPR 92-1236-901 includes transducers JT4232, JT4332, JT6025D, JT6025E, and JT6025F, which are associated with the RCPs and Bus B Startup and Auxiliary transformers, but the scope of this safety evaluation includes all GE type 4701 transducers on site.
REASON FOR CHANCE:
GE type 4701 watt, var, volt, and current transducers are used in the plant electrical distribution systems to provide signals for indication of electrical parameters.
The original plant design included 30 or more such transducers.
These transducers are becoming unreliable, and must be replaced.
Replacement is undesirable, because of cost and reliability.
SAFETY EVALUATION
SUMMARY
The proposed replacement transducers will be commercially dedicated in accordance with EN-DP-00070 (Procurement) and EN-DP-01023 (Material Engineering Evaluation (MEE)).
Each MEE ensures adequate reliability, seismic qualification, and the overall suitability of the replacement transducers.
The replacement transducers are functionally equivalent, except for a slight difference in dielectric insulation strength.
However, the dielectric strength of the replacement transducers is over twenty times the nominal operating voltage of these devices, and therefore is acceptable.
This change affects only the transducer manufacturer and the transducer terminal numbers, and does not affect any function of the transducers.
The replacement of these transducers will not affect the operation of any associated circuit, and will not introduce any new failure modes or any possibility for common mode failure, and the single failure criteria analyzed in the USAR are ut.af fected.
DBP 5302DDDD/22 1
SAFETY EVALUATION
SUMMARY
FOR FPR 93-0957-901 and FPR 93-0958-901 (SE 94-0068)
TITLE:
Removal of Main Steam Safety Valve Position Monitors CHANGE:
Remove the linear variable differential transducers (LVDTs) that monitor the position of eight of the eighteen Main Steam Safety Valves (MSSVs).
REASON FOR CHANGE:
This change is desired as the MSSV position monitoring is not used and the LVDTs are a burden as they must be removed for MSSV testing and maintenance.
SAFETY EVALUATION
SUMMARY
Removing ths MSSV position monitoring LVDTs does not affect the capability of the MSSVs to perform their safety function. The MSSVs are safety related ASME Section III, Class 2 safety valves.
These valves provide over-pressure protection for the Main Steam System and are capable of dissipating all of the energy generated at the reactor over-power trip setting.
Credit is taken for the functioning of the these valves in the accident analyses contained in USAR Chapter 15.
These LVDTs are installed on only eight of eighteen MSSVs. When installed they are designed to not interfere with MSSV operation nor affect the MSSV setpoint.
The LVDTs are externally attached and form no part of any pressure boundary.
Therefore the removal of the LVDTs vill not affect the ability of the MSSVs to perform d:eir safety function.
In addition, the LVDTs are not required to monitor the operation of the MSSVs following an accident as the LVDTs are not considered post-accident monitoring instrumentation.
SAFETY EVALUATION
SUMMARY
FOR FPR 93-4738-501 (SE 93-0045)
TITLE:
Replacement of 90% Type-27D Undervoltage Relays CHANGE:
Replaced the existing Type-27D undervoltage relays with Asea Brown Boveri Type-27N undervoltage relays in order to correct for the design problems imposed by the increased contact restoration time.
R,EASON FOR CHANGE:
Identified problems associated with the 90% degraded grid undervoltage scheme falsely actuating upon restoration of direct current (DC) following any power interruption were identified.
The analysis performed determined that newly manufactured Type-27D relays take considerably longer to reset their contacts
(~100 milliseconds) upon restoration of DC power than those of the relays originally installed in the degraded grid scheme (<10 milliseconds). This increased time is sufficient to energize the 90% relay's auxiliary relays (55-65 milliseconds required) and deenergize the bus.
This false trip results in an unnecessary power interruption to the 4.16 KV essential bus and an unnecessary challenge to the start scheme for the affected busses respective i
j SAFETY EVALUATION
SUMMARY
There are no adverse effects on safety imposed by the replacement of the Type-27D undervoltage relays with Type-27N undervoltage relays for the degraded grid (90%) undervoltage scheme.
Selecting Class IE Type-27N relays as replacements will allow the degraded grid undervoltage scheme to function exactly the same as it does with the presently employed Type-27D relays.
No changes to the setpoint selection or relaying logic will be required in order to implement this change.
The only difference is that the new relay's output l
contacts will not change state upon loss of DC control power.
Replacement of the Type-27D relays with Type-27N relays is considered a significant enhancement to the degraded grid undervoltage scheme.
The Type-27N relays are recognized by the manufacturer as being more accurate than the Type-27D relays.
The Type-27N relay also has a much lower pickup to dropout ratio than the Type-27D relay.
Pickup to Dropout ratio is a characteristic which defines at what point the undervoltage relay will reset after a momentary dip in sensing voltage.
The lower pickup to dropout ratio means that the new relays will greatly enhance system reliability by reducing the chance for an undesired 1
separation from the offsite power source following large meter starts at low grid voltage.
The analyses performed by Engineering has concluded that the proposed change is safe.
The replacement of the 902 undervoltage relays with a newer model will not result in any adverse effects on safety.
DBP 5302FFF/24
g-SAFETY EVALUATION SUHHARY FOR FPR 94-0140-901 (SE 94-0065)
TITLE:
SFAS Power Supply Replacement CHANGE:
Replaced SFAS -15 VDC power supply.
REASON FOR CHANGE:
The previous SFAS -15 VDC power supply was no longer manufactured and was replaced.
SAFETY EVALUATION
SUMMARY
The power supply will function equivalent to the previous power supply.
The power supply has been acceptably reviewed for operating characteristics and seismic capabilities.
The seismic qualification of the SFAS cabinets remains unchanged.
This ensures that no new hazards will be introduced by the installation of the power supply.
The reliability of the actuation equipment remains the same.
The MOD does not alter the initiators of the equipment malfunctions previously evaluated.
The SFAS actuated equipment important to safety has not been adversely affected by this modification.
DBP 5302FFF/4
SAFETY EVALUATION
SUMMARY
FOR FPR 94-0390-901 (SE 94-0052)
I TITLE:
Install Desiccant Dryer in Compressor Cll-1 Loadless Start Line CHANGE:
Install a desiccant dryer in the loadless start line to Emergency Diesel Generator Air Compressor Cll-1.
l REASON FOR CHANGE:
I The desiccant dryer is installed to provide a means of removing moisture from the air being applied to the loadless start device. This moisture is causing failures of air compressor internal parts.
1 SAFETY EVALUATION
SUMMARY
The portion of the EDG Air Start System affected by this activity is not required for the system to perform its safety function.
The ability to charge the air receivers from minimum to maximum pressure in not more than 30 minutes is not affected. Also, the flow rate in the loadless start line is not impacted because the pressure drop through this filter is negligible.
In the event of a failure of the pressure boundary in this portion of the i
system, the automatic isolation valve would isolate the failure from the l
essential portion of the system.
In the event of a failure that would prevent l
passage of air to the loadless start device, the compressor would start in a loaded condition and the air compressor's lov oil pressure protection circuit would not be functional, but the air compressor vould still provide vir to the air receivers. The air compressor vould be capable of providing the same volumetric flow rate if a dryer failure disabled the loadless start device.
Therefore, the air compressors as well as the essential portion of the EDG Air Start System would still perform their functions.
Based on the above discussion it is concluded that the above modification is j
safe.
j
o i
P SAFETY EVALUATION SUHHARY FOR FPR 94-0463-901 (SE 94-0053)
TITLE:
Fuel Handling Communications CHANGE:
Connect existing spare GAI-Tronics wires to audio connectors at the Control Room and Fuel Handling areas for use during fuel movement. Outlets will be provided at the fuel handling areas to plug in the local radio system base units.
REASON FOR CHANGE:
To facilitate communications between the fuel and spent fuel handling operators and the control room.
SAFETY EVALUATION
SUMMARY
Normally each GAI-Tronics cable contains two #14 AWG pairs which provide power from either YAU11 or YBU11 through fuses at U500.
Since each GAI-Tronics station actually only uses power from one source or the other, there is usually a spare #14 AWG pair available on each GAI-Tronics cable. With the-exception of HSE 5750, all of the GAI Tronics stations affected by this FPR receive power from YAU11.
The FPR transfers HSE 5750 to YAU13, inconsequential 1y increasing the load on YAU11 and the associated power fuse at U500. After this change is made, the "B" pair will be available for use at HSG5714, and throughout the fuel handling circuit.
The B pair will then be terminated at audio connectors located at GAI-Tronics stations in the control room, and near the Spent Fuel Pool and the Refueling Canal.
Since only spare conductors will be used for the communications signal, the existing system will continue to operate as 1
designed.
l Two outlets will be provided for powering the local base units.
These outlets will be connected to the A channel, and will therefore increase the load on the A channel.
However, the increase in load represents only a minute fraction of the load capability of the (3 Ampere) fuse at U500.
It is recognized, however, that plant personnel may attempt to power other loads from these outlets.
To prevent this from happening, a warning label will be placed at the outlets.
i An electromagnetic interference study was performed in July 1990.
This led to j
the establishment of restrictions on radio use in certain areas.
The Spent Fuel Pool and Refueling Canal areas do not contain equipment which would be susceptible to radio interference, and are not subject to radio restrictions.
l a
DBP 5302FFF/3 i
I
SAFETY EVALUATION
SUMMARY
FOR FPR 94-0861-901 (SE 94-0072)
TITLE:
Abandon Reactor Vessel Head Accelerometers ZE-8907 and ZE-8908 CHANGE:
Abandon the accelerometers that monitor the top of the reactor vessel for the Vibration and Loose Parts Monitoring System (V&LPHS).
REASON FOR CHANGE:
This change is desired because the sensors are hard to calibrate and place a burden on the field for calibration.
SAFETY EVALUATION
SUMMARY
Removing the accelerometers, located on top of the vessel, from service does not affect the ability of the V&LPH System to perform its function.
The most likely place for loose parts to collect is in the bottom of the vessel and the top of the steam generators, both of these positions will still be monitored.
The abandonment of these accelerometers does not affect the Reactor Coolant System's ability to perform its functions as no breach of the actual coolant system vill take place as a result of this change. These accelerometers are mounted on top of the vessel and do not penetrate the vessel. The V&LPH System is not required to help shutdown the plant in an accident scenario.
i
SAFETY EVALUATION
SUMMARY
FOR MOD 88-0215-02 (SE 93-0048).
TITLEr Revision of Power Supply to C5752 CilANGE :
Replaced the Local Analog Input Panel (LAIP) C5752 and changed the power supply to this cabinet from YBU34 to transfer switch YATS4601.
REASON FOR CHANGE:
This modification is part of the plant computer upgrade.
This modification replaced obsolete equipment and increased system reliability.
SAFETY EVALUATION
SUMMARY
Changing the power supply to C5752 will not affect the safety function of the 120 Volts Nonessential Instrumentation AC System.
YAU has sufficient capacity to supply the additional load required by C5752.
The loading on YBU will decrease or remain the same, depending on the availability of YAU.
The disconnect switches / fuses, cables associated with circuits YAUO7 and YBUO4, and transfer switch YATS4601 were properly sized per FCR 84-0083 in anticipation of the additional load required by C5752.
The individual feeder to C5752 will also be sized and fused properly for the load.
YBU34 will be spared and therefore, is no longer an active circuit.
Based upon the above evaluation, the changing of the power supply to C5752 is safe.
DBP 5302FFF/35
i SAFETY EVALUATION
SUMMARY
TITLE:
Revision of Power Supply to Various Computer Equipment CHANGE:
Provided two sources of uninterruptible power to C5772F and the plant computer disk drives. This modification also revised several other plant computer equipment power supplies.
REASON FOR CHANGE:
This modification allows C5772F and the plant computer disk drives to function if either YAU or YBU power fails.
Previously, the system was not usable if YAU or YBU failed.
SAFETY EVALUATION
SUMMARY
The changing of the power supplies to the various plant computer equipment will j
not affect the safety function of the 120 Volts Nonessential Instrumentation AC System. YAU and YBU have sufficient capacity to supply the additional load required by the changes in this modification.
The cables, the uninterruptible power source, and the transfer switch YATS5701 associated with circuits YBU32 and YAU48 are properly sized ~for the load required by C5772F and the plant computer disk drives.
The cables associated with circuits YAU16, YAU17, YAU18, and YBU12 are properly sized for the additional load being added by this modification.
Circuits YBU32, YAU48, YBU14, YAU16, YAU17, YAU18, and YBU12 are not safety-related circuits but are associated circuits.
The power supply changes associated with MOD 89-0036 do not affect the coordination between these circuits and their upstream fuses.
The load on YBU14 will be reduced by MOD 89-0036 and the coordination between YBU14 and its upstream fuse will not be affected.
The UPS system will add a load of 1467 BTU /HR to the Computer Room Air.
Conditioning Unit which is a part of the Control Room Normal Ventilation j
System. The Computer Room Air Conditioning Unit S77 has sufficient capacity to handle this additional load.
The Control Room Emergency Ventilation System Load Shedding Procedure will be revised to add an additional step to turn off the UPS system in a high temperature situation to prevent overloading the Control Roem Emergency Ventilation System.
DDP 5302FFF/37 3
1 SAFETY EVALUATION
SUMMARY
FOR MOD 89-0099-01 (SE 91-0013, R02)
TITLE:
]
Removal of Pipe Collars from the Pressurizer Surge Line Whip Restraints CHANGE-Removed the pipe collars from the Pressurizer Surge Line Whip Restraints. The pipe collars were removed in conjunction with the Surge Line reanalysis.
The pipe will have insulation installed in place of the removed pipe collars.
The USAR was revised to reflect that the Pressurizer Surge Line Whip Restraints are abandoned in place, and that based on MEB 3-1, Revision 2, criteria breaks are postulated to occur anywhere on this piping.
REASON FOR CHANGE:
The Surge Line has been reanalyzed considering the effects of thermal stratification, to determine the postulated pipe break locations.
This reanalysis shows that numerous postulated pipe break locations must be evaluated.
SAFETY EVALUATION
SUMMARY
Based on the analysis performed by Babcock and Wilcox for the Bulletin 88-11 (Thermal Stratification), there are numerous points on the Pressurizer Surge Line that must be postulated as break locations.
In order to ensure a complete review of these postulated breaks the Pressurizer Surge Line has been reviewed considering that a break could occur anywhere on this piping.
PCAQ 88-0915 identified that the existing Surge Line Whip Restraints were not adequate to resist the forces associated with a " Break Anywhere Criteria". A thorough review of potential targets was required to ensure that no safety related items would be affected. This review was performed assuming the absence of any pipe whip restraints.
The direction of pipe whip and size of the jet spray were determined for the break locations.
Field walkdowns were conducted to identify and evaluate items that would be impacted. Also, the secondary shield walls have previously been evaluated for pipe whip loadings, and found to be adequate.
No safety related systems, components, or structures were identified that would be adversely affected.
Therefore, the existing whip restraints are not required to mitigate the effects of a Surge Line Pipe Break and they may be abandoned in place.
The potential for a section of pipe breaking and creating a missile has been evaluated.
The surge line piping is A376 Type 316 stainless steel.
The Design Criteria Manual, states that stainless steel piping will whip 1800, impact itself, and stop.
No missiles will be generated, and no safety related equipment or components will be impacted.
Since the Pressurizer Surge Line Whip Restraints are no longer required, the pipe collars may be renoved without affecting any safety related equipment, components, or structures.
The USAR was revised to reflect that the Pressurizer Surge Line Whip Restraints are abandoned in place and to identify that pipe breaks are postulated to occur anywhere.
DBP 5302FFF/15
SAFETY EVALUATION
SUMMARY
FOR i
H0D 90-0046 (SE 93-0051, R.01)
TITLE:
Replacement of RC10 CHANGE:
Replace valve RC10.
REASON FOR CHANGE:
Motor Operated Valve (H0V) HU2B failed during the shutdown from power following the sixth operating cycle, H0V RC11 failed during the return to power following the sixth Refueling Outage, H0V HU1A failed and its vedge was replaced during the 8th refueling outage. All valves suffered a stem to disc separation. Two additional valves, RC10 and MU1B have been identified as being subject to the same type of failure. This Hod was written to provide an approved replacement valve for these solid vedge Motor operated gate valves.
- SAFETY EVALUATION
SUMMARY
The function of RC10 is to close to direct auxiliary spray flow to the pressurizer and to isolate a stuck open spray valve. The only safety function of RC10 is to maintain Reactor Coolant system Pressure Boundary integrity.
Isolation is not a safety function for RC10.
Motor Operated Valve RC10 is subjected to large temperature transients during hot plant operations. The replacement valve has a flexible wedge which is less susceptible to thermal binding caused by these types of transients. The replacement valve meets or exceeds all design requirements of the existing valve.
The weight and the center of gravity associated with the replacement valve and operator are different than the existing valve.
Analysis shows the allovable pipe stress vill not be exceeded. The qualified seismic loading of the replacement valves exceeds the requirements for the original valves.
The nominal horsepover rating of the actuator motor is 0.66 HP.
The rating of the new motor is 1.0 HP.
This slight load increase is within the capability of the electrical distribution system.
The replacement valve is being provided with a stem manufactured from SA638 Grade 660 stainless steel. This material is made specifically for use in higher temperature applications.
Its mechanical properties meet the manufacturers and code design requirements. The metallurgical properties of the grade 660 material is preferred over the grade 630 material for design temperatures exceeding 600'F.
The phenomenon of pressure locking associated with flexible wedge gate valves is avoided by allowing the bonnet to vent upstream when the valve is closed through a hole in the valve disc. The new valve vill continue to operate in the same manner as the existing valve.
1
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7 SAFETY EVALUATION
SUMMARY
FOR l
MOD 90-0073 (SE.94-0028) 1 TITLE:
Decontamination Showers in Room 417 CHANGE:
Install decontamination showers and head / face sinks in Room 417.
j REASON FOR CHANGE:
To provide separate f acilities for male and female personnel, a permanent privacy partition will be installed across the room between the existing door
-openings in the north wall.
This division of the room creates two rooms, designated Room 417 and Room 417A.
This room creation requires additional fire detection capabilities to Fire Detection Zone FDZ-412A.
SAFETY EVALUATION
SUMMARY
The requirements of Section 4.6.CC.3, Fire Propagation Control, of the Fire Hazards Analysis Report identifies that a fire detection device be present1 n 1
each room of fire area CC of the Auxiliary Building, floor' elevation 603'-0'.
Currently, there is one fire detection device in Room 417.
The division of this room into two rooms necessitates that this modification install a fire detection device in Room 417A and relocate the existing smoke detector within.
Room 417 to provide the required coverage.
The division of the present' Room 417 into Rooms 417 and 417A has no effect on.
plant safety.
The relocation of the existing fire detector in Room 417 and the i
addition of a new fire' detector to Detection Zone FDZ-412A for Room 417A will provide the required level of fire detection.
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'DBP 5302FFF/5
e SAFETY EVALUATION
SUMMARY
FOR Mod 91-0004 (SE 91-0050)
TITLE:
Install New Sample Line to Radiation Monitor RE8432 CHANGE:
Install a new sample line designed to reduce head loss and with provision for cleaning in the event of future fouling.
The existing swiple line isolation valve will be capped and abandoned in place. Additionally, tne sample line pleing downstream of the first isolation valve will be downgraded from ASME Sect ion III Class 3 to ANSI B31.1.
Four unused purge solenoid valves and two maintenance isolation valves are permanently removed by this modification.
The flow from the RE 8432 will be directed to the floor drain utilizing a hose approximately 5 feet long.
REASON FOR CHANGE:
Sample flow through radiation monitor RE8432 is inadequate.
Low flow results from low pressure in the service water return header and fouling of the sample piping.
SAFETY EVALUATION
SUMMARY
This modification will not affect the service water (SW) safety function of cooling engineered safety features components because the sample system is installed in the return header and does not interfere with the return flow.
All the piping installed under this modification is seismically supported. A non seismic flexible drain hose is installed at the discharge of RE8432 for housekeeping.
In the event of failure, sample flow will continue to the floor drain with negligible change in. flow rate.
Therefore, this modification does not create a flooding hazard.
The portion of the Service Water System required for emergency operation was designed to ASME Section III Class 3 Seismic Category I requirements.
The piping system changes required to implement this modification were also designed to these requirements.
Removal of maintenance isolation valves SW8432E/F and the four purge solenoid valves will increase the reliability of the sampling system by streamlining the sample pipe.
SW8432 will provide adequate isolation.
Therefore, this change does not effect safety.
DBP 5302DDDD/15
~
The sample piping for RE8432 is currently designated as ASME Section III, Class 3.
The piping downstream of the new isolation valve SW8432 is not required for emergency operation and therefore this designation is unnecessarily restrictive.
Analysis and support of this piping as Seismic Category I to the extent necessary to ensure the functional operability of the Q SW return header will allow downgrading of this piping to ANSI B31.1 without any effect on safety.
Due to the short length of this piping, this criteria has been met by seismically supporting the entire length.
This modification will increase the reliability of the Process Radiation Honitoring System by providing an improved sample line for RE8432 which will restore it to functional service.
DBP 5302DDDD/16
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SAFETY EVALUATION
SUMMARY
FOR-MOD 91-0030 (SE 93-0070) f 1
5 TITLE:
I
' Spent' Fuel Handling Access Modification CHANGE:
Added an auxiliary monorail, hoist with manually activated fuel grapple, and-
- walkway to the existing Spent Fuel Handling Bridge.
REASON FOR CHANGE:
When the Spent Fuel Pool was re-racked in 1979, the Spent Fuel Handling Bridge configuration was not modified to access approximately 64 storage cells in the south end of the pool. 'Use of these storage locations is necessary to maintain the capability of fully off-loading the reactor core during the Ninth Refueling Outage.
SAFETY EVALUATION
SUMMARY
The effects on safety will be examined for the following structures, systems, and components:
Spent Fuel Handling Bridge, Spent Fuel Storage Racks, and the Auxiliary Building.
SPENT FUEL HANDLING BRIDGE:
The bridge was designed for seismic forces to ensure structural integrity over i
spent fuel assemblies.
The new monorail and walkway structures have been a
designed for the applicable loading combinations, including seismic.
The bridge structure has also been reviewed to verify its adequacy for these new l
loads.
The new hoist is equipped with both an electric and a load brake.
The load brake ensures safe fuel handling in the event power to the hoist is lost.
The:
hoist will also have manual operation capabilities that would enable fuel' assemblies to be raised or lowered manually.
The electric hoist will.be.
powered from a new 480v source on the bridge.
To ensure that the-existing non-safety related power circuits are not adversely impacted by this added load, procedures will be revised to require that the bridge not be moved electrically while a fuel assembly is being raised or lowered. Added protection is provided by a 30A breaker between the new hoist and the existing l
power supply.
This procedural requirement will also serve to protect the fuel.
assemblies while their being raised or lowered.
The new fuel grapple is a manually actuated device designed to handle lifting i
fuel and maintain a minimum factor of safety of 5.
A ccmbination of procedural controls and design features are being used to provide assurance against a-fuel handling accident equal to the existing' fuel handling grapple / machine.
To provide assurance that a fuel assembly cannot be inadvertently disengaged while supported from the hoist, the grapple fingers are equipped with a mechanical interlock. This interlock on the new grapple is equal to the mechanical i
DBP 5302FFF/27 1
- - - -., _ -. ~
interlock on the existing fuel handling grapple.
Procedures will be revised to prohibit electrically moving the bridge and/or hoist while a fuel assembly is being raised or lowered.
The engaged / disengaged status of the grapple is indicated on its' top face.
The procedure change and the visual indication of the grapples position perform the same functions and are equal to the electrical interlocks in the existing bridge.
SPENT FUEL STORAGE RACKS:
The auxiliary monorail hoist will have a manually set load limit on the hoist pendant controls that will prevent a net uplift load of greater than 500 lbs.
This is in accordance with the existing storage rack design, as detailed in USAR Section 3.8.1.1.4.c.
The Spent Fuel Handling Procedure has been revised to administratively require overload monitoring; as is presently done with the bridge fuel hoist.
The analysis of the racks for the fuel assembly drop accident will not be affected by this modification.
The maximum weight lifted over spent fuel assemblies during the implementation will be limited to 2430 lbs. in accordance with existing Technical Specification 3/4.9.7.
Procedures provide the administrative controls for lifts over the Spent Fuel Pool, and provides for the safety of the Spent Fuel Assemblies.
AUXILIARY BUILDING:
The bridge monorail, hoist, and fuel grapple have been designed to permit safe handling of spent fuel assemblies while maintaining the minimum water cover of 9 feet over the active fuel.
The minimum water cover is ensured by providing a fixed length rod, between the hoist hook and the fuel grapple.
This rod is sufficiently long to ensure minimum water cover when the hoist hook has been fully raised.
Procedures will be revised to require that only one spent fuel assembly be handled at a time in the spent fuel pool.
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DBP 5302FFF/28
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. SAFETY EVALUATION
SUMMARY
1 FOR Mod 91-0040 (SE 93-0038, R02)
TITLE:
I Fuel Handling Equipment Enhancements CHANGE:
Modification' 91-0040 implemented the following
- 1) Relocation of the cable reel for the Fuel Storage Handling Bridge (FSHB) from the north side of the north beam of the FSHB to the south side of that beam; 2) Modification of the manual bridge rolling device on the Main Fuel Handling Bridge (MFHB) so that the bridge can be operated manually from the side of the refueling canal:
- 3) Installation of bridge wheel covers on the MFHB, the Auxiliary Fuel Handling Bridge (AFHB) and the FSHB 4) Configuration of the FSHB hoist for two speed operation.
REASON FOR CHANGE:
Changes resolved personnel safety problems, reduced maintenance and operation time, and replaced obsolete and unreliable parts.
SAFETY EVALUATION
SUMMARY
1)
The reel relocation does not change any bridge interlock or operating-function. The new reel mounting will have no deleterious ~ effects on any of the. bridge's functions, and will not affect the bridge's structural-integrity.
Therefore, the reel relocation will not affect any function important to safety.
No new components are added by this modification, and the operation of the bridge and the reel-(other than location) are unchanged, reliability of the bridge is unaffected.
2)
The manual bridge rolling device modification does not change any interlock or operating function of the bridge. Manual bridge rolling can already be done, and the bridge is presently operated manually.
This modification will merely allow manual bridge operations to be performed more safely and easily.
This modification will involve installing mounting brackets, and extension drive shaft, and a brake release shaft on one of the MFHB beams. These new parts will not affect the bridge's ability to perform any of its existing functions.
The bridge's structural integrity will not be impaired.
3)
The bridge wheel covers will cover the wheels to provide a safer walkway.
They will not affect any interlock or operating function of the bridge.
l DBP 5302DDDD/17
The wheel covers will be-stainless steel plates which will be mounted to prevent personnel from touching a wheel while the bridge is operating.
The covers will not affect the bridge's ability to perform any of its functions.
The bridge's structural integrity will not be impaired.
4)
The hoist on the FSHB is equipped with a two speed motor. A maximum hoist speed of approximately 20 feet per minute and a low speed of approximately 5 or 6 feet per minute would be available, except that the. hoist fast speed has never been used.
Two speed operation was part of the original design capabilities of the bridge, and the Cycle 9 Refueling Specification considers the possibility of using fast speed. Hoist speed is not mentioned in USAR 9.1, Fuel Storage and Handling, or USAR 15.4.7, Fuel Handling Accident. Although included for information, hoist speed was not used in our 1983 response to NUREG 0612, Control of' Heavy Loads at Nuclear Power Plants.
There will be no effect on hazards or seismic design.
Based on the above discussion it is concluded that the proposed modification is safe.
DBP 5302DDDD/18
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11 SAFETY EVALUATION
SUMMARY
FOR I
/ TITLE'r-Elimination of Normally Energized Agastat Series 7000 Relays in Environmentally 4
Qualified Circuits I
CHANGE:
Resolved an Environmental. Qualification (EQ) problem by eliminating the use of normally energized Agastat Series 7000 relays'in the control circuits for Auxiliary Feedwater Pump Turbine (AFPT) Isolation Valves MS106, MS106A, MS107 I
ar.d MS107A. Additionally, addressed an EQ problem with the Static-0-Ring-(SOR)
L pressure switches in the AFPT rooms by installing supplemental fuses in the q
isolation valves' control circuits.
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REASON FOR CHANGE:
.i To demonstrate compliance with 10CFR50.49, four Series 7000 relays were tested in the energized state. EQ testing of these energized relays was abandoned in
/une 1991 as three of the four test specimens exhibited coil failure during the thermal aging portion of the test.
SAFETY i" VALUATION
SUMMARY
Since normally energized operation precipitated the coil failures observed in the relay test specimens, replacing PSL4930X1, PSL4930X2, PSL4931X1 and PSL4931X2 with normally deenergized EQ relays will enhance the relays'
+
reliability in maintain electrical circuit integrity.
Changing the connections 1
at PSL4930A, PSL4930B, PSL4931A and PSL4931B from normally open to normally closed will not affect the function of these switches.
Using normally deenergized relays to protect equipment is accepted practice.
Thus, changing PSL4930X1, PSL4930X2, PSL4931X1 and PSL4931X2 from normally energized to normally deenergized will cause these relays to function like
- 1 other equipment-protecting relays. Moreover, the risk of losing an operating AFP as a result of a relay coil failure will be eliminated, l
Use of the existing reset pushbuttons, in lieu of relays D135-33X and BF1124-33X, to block and enable the AFPT steam line pressure switches will be
)
administratively controlled. These pushbuttons are presently qualified for.use i
in AFPT isolation valve circuits.
In the normal state, the pushbuttons will enable the AFPT steam line pressure switches.
Since these pushbuttons are
" spring return to normal", inadvertent blocking due to incorrect switch position is very remote.
Use of the reset pushbuttons te block the AFFT steam l
line pressure switches will not adversely affect mitigation of events by these i
valves.
+
DBP 5302FFF/9
The supplemental fuses are seismically and environmentally qualified for their plant locations.
These fuses have been properly sized to supply the current required by the pressure switch branch circuits and to clear postulated ground faults at the SOR pressure switches while leaving the control circuit's main fuse intact.
Thus, the supplemental fuses will not interfere with proper circuit operation and will allow recovery from a room 237 or 238 HELB induced ground fault by preventing the loss of power to the isolation valves' control circuits.
i DBP 5302FFF/10
SAFETY EVALUATION
SUMMARY
TITLE:
New Central Fire Alarm System CHANGE:
Replaced approximately 681 existing detection-devices with new, state-of-the-art Simplex Time Recorder Company (Simplex) detection devices.
It also eliminated or replaced existing and obsolete fire panels, consolidating them into a network of nine new, state-of-the-art Simplex fire control panels.
REASON FOR CHANGE:
Upgrade the Fire Detection System (FDS) and to separate the FDS from the security system.
SAFETY EVALUATION
SUMMARY
1he Simplex FDS uses the latest in electronics technology and is Underwriters Laboratory (UL) listed for several NFPA Codes.
The Simplex FDS is designed and installed to NFPA 72-1990.
It supersedes Toledo Edison's commitment to NFPA 72D-1975.
Testing requirements for the Simplex FDS will be based on NFPA 72E-1990.
This will supersede Toledo Edison's commitment to NFPA 72E-1978.
Approximately 681 of the existing detecting devices (smoke, heat, and flame) are being replace by Simplex or Fenwal detection devices.
Although these detection devices are not required to be environmentally qualified per 10CFR50.49, their specification requires that the new detection devices function under specific environmental conditions.
Tha new Simplex detection devices were swapped ona-for-one and like-for-like with the existing detection devices, except as noted below.
The one-for-one provision maintains the existing margin of " extra" installed detectors in various fire areas.
The detector type is changed from ionization detectors to heat detectors in high radiation areas where the calculated doses during power operations preclude the use of Simplex photoelectric detections.
The detectors type is also changed from ionization detectors to heat detectors in high ambient heat areas where the ambient temp 5rature can reach over 1700F during power operations because Simplex detectors begin to plastica 11y deform above 1650F.
Additionally, smoke detector DS8690X in Room 208, #1 Mechanical Penetration Room, is changed to a heat detector.
This detector is failing frequently due to dust coming from a nearby ventilation duct.
Based on walkdowns of the area, relocation of the detector is not recommended, thus replacement with a heat detector (which is not susceptible to dust) is acceptable.
Simplex photoelectric detectors are used in the remaining areas of Containment and the Containment annulus as these detectors were determined, as described above, to survive radiation exposure better than Simplex ionization detectors.
In that both types of detectors (photoelectric and heat) provide fire detection capability, there is no adverse effect on safety created by using photoelectric or heat detectors in lieu of ionization detectors in the areas described above.
DBP 5302FFF/13
The Simplex Model 4100 series fire control panels also have an internal battery backup which will provide approximately 18-24 hours of power to the fire control panel in the event of a power interruption.
The fire control panels will be connected together via a communications link which will allow bi-directional signal transmission.
Failure of any fire control panel or a break between fire control penel will be detected but will not inhibit system operation. Additionally, each fire control panel is capable of stand alone operation.
With one specific exception, there is no reduction in the existing level of detector supervision.
The one exception is with the replacement of the PORV room ionization detector with a heat detector.
This replacement is necessary based on the high temperatures in the PORV room which preclude the use of any other type of detector.
The supervision for the heat detector will be downgraded from ' Class A' to ' Class B'.
It was determined that downgrading to a ' class B' supervision level was acceptable based on the limited length of circuit that will be downgraded and the fact that the circuit will be enclosed in conduit which will minimize the possibility of a fault on the circuit.
Elimination of the extraneous detection zones has no safety significance as these areas do not contain any equipment necessary to meet 10CFR Part 50 Appendix R safe shutdown requirements.
Deletion of the four annulus dome detectors is justified in that it is unlikely they would detect a fire in the nearest combustibles due to the extreme height and space of these detectors.
The electrical penetrations are provided with their own detection devices immediately above them.
Deletion of the Protectowire linear fire detection system is allowed per T.S.
Amendment 174 Abandonment of duct detectors located in the Turbine Building HVAC, is acceptable in that no credit was taken for these detectors in the evaluation of NFPA 90A.
The reduction in the number of local bells in acceptable in that NFPA 72-3990, Section 9-7.4 states "This section does not require the use of audible alarm signals other than the one at the central supervising station".
The arrangement will have one bell at each of the nine new fire control panels.
Reducing the extraneous bells will reduce the overall complexity of the system as well as simplify maintenance and testing.
The Simplex hardware has been tested for electromagnetic interference (EMI) and tested to demonstrate its susceptibility to radio frequency interference (RFI) either during the factory acceptance tests or during field installation.
Should RFI testing show any unique vulnerability's, these will be compensated for by training and posted signs, as appropriate.
Separation of the FDS from the security system improves the operation and i
maintenance of the two systems over the existing design.
Maintaining the RMS interface will not adversely affect the operation of the new Simplex FDS.
Based on the above evaluation, the new Simplex FDS meets UFPA and regulatory requirements for a FDS.
DBP 5302FFF/14
SAFETY EVALUATION
SUMMARY
TITLE:
Provide Backup Exciter Field Breaker Trip Circuit CHANGE:
Added two Main Generator lockout relays, and added contracts from these relays to the exciter field breaker control circuit.
These changes ensure that a Main Generator lockout signal will also terminate Main Generator excitation.
REASON FOR CHANGE:
Previously, removal of excitation depended solely on proper operation of the Main Generator field breaker.
Failure to remove excitation from the Main Generator following an output fault could result in significantly more damage than would be caused by the fault alone.
The probability of such coincident failures is extremely low; however, major equipment damage would result from such an event.
After the modification, Main Generator lockout will cause a simultaneous trip of the exciter field breaker in addition to the Main Generator field breaker, thereby potentially reducing the extent of equipment damage following a Main Generator output fault.
SAFETY EVALUATION
SUMMARY
MOD 92-0003 has no effect on safety.
New circuits are added, slightly increasing the current loading of four nonessential DC Distribution Panels.
However, the affected panels are designed for the increased load, and have no functions which are important to safety.
In addition, two new relays are added to C5750A. Although connected to the existing generator lockout circuit, these relays have no effect on the existing lockout circuit.
In any event, the generator lockout circuit itself is not important to safety.
The new relays are General Electric type "HFA," a type of relay which as widespread application in plant protective systems.
The new relays are being installed in a seismically mounted cabinet.
Contacts from the two new relays will be used to trip the exciter field breaker.
The two new relays will normally be deenergized.
Therefore. the normal failure mode of the two new relays will be to fail to trip when called upon.
This failure would cause the plant to respond as it currently does without the modification.
Much less likely is a failure which would cause an unwanted plant trip, the exciter field breaker itself is not considered to be important to safety, and has no effect on the capability of the plant to achieve safe shutdown.
Therefore, the change to the exciter field breaker circuit has no effect on safety.
DBP 5302FFF/36
SAFETY EVALUATION
SUMMARY
FOR MOD 92-0006-01 (SE 93-0032)
TITLE:
Makeup Water Treatment Chlorination System CHANGE:
Installed a Makeup Water Treatment Chlorination System to allow more effective control of chlorine levels in the system.
The Makeup Chlorination System now inject Sodium Hypochlorite (Na0C1) via pumps P250-1 or P250-2 from the existing Na0Cl system into the Detention Tanks and the Clarifier Flow Splitter Box.
The-residual chlorine levels were increased from 0.35 mg/l to 1.0 mg/l in the Clearwell.
REASON FOR CHANGE:
On June 29, 1993, the EPA instituted new regulations on potable water treatment.
The new rules provide additional requirements for the filtration and disinfection of potable water. The purpose of the regulations is to guard against water borne illness by Giardia Lamb 11a and viral agents.
The new requirements were developed due to the inability to efficiently detect these pathogens by analytical methods.
SAFETY EVALUATION StMMARY:
The only system with a function important to safety, that is affected by this modification, is the Fire Protection System. However, the only impact to this system is indirect as a result of increasing the residual chlorine levels in the Makeup Water Treatment System.
Since the Makeup Water Treatment System supplies water directly to the Fire Water Storage Tank, a slight increase in the chlorine residual in the system can be expected.
The increased chlorine residual will have no significant impact on water chemistry.
It will also have no significant impact on metal corrosion rates.
Therefore, the function and reliability of the system are not affected.
This would also apply to the other systems directly or indirectly affected by this modification.
Based on this evaluation, the modification to the Makeup Water Treatment System is safe.
DBP 5302FFF/40
SAFETY EVALUATION
SUMMARY
FOR Mod 92-0046 (SE 93-0059)
TITLE:
Victoreen Process Radiation Monitor Digital Upgrade CHANGE:
Upgrade ten (10) Victoreen analog process radiation monitoring channels.
This modification is part of the Ten Year System Structure and Expenditures Plan in which all Victoreen process monitors will be upgraded.
This upgrade consists of the following for each channels replace the existing analog rate meter with a modified universal digital ratemeter.
Replace the existing standard gamma scintillation detector with a modified gamma scintillation detector. Replace the existing junction box with a modified junction box.
REASON FOR CHANGE:
This modification will; a) reduce both the calibration time and frequency of failures and repairs to the subject process radiation monitoring equipment and, b) increase the accuracy and reliability.
SAFETY EVALUATION
SUMMARY
All ratemeters are to be seismically mounted on C5765 in the control room cabinet room.
The new equipment weighs less than the installed equipment.
There will be no seismic impact on cabinet C5765. All new detectors shall be mounted in the existing sample chambers.
The mounting configuration shall be identical to that currently employed by the existing detectors.
Electrically, the new and old monitors are compatible.
The new ratemeters will be easier for operators to use than the existing ones.
This modification will improve accuracy and reliability for the process monitors.
j l
Based on the above, it is determined that the proposed action is safe.
1 DBP 5302DDDD/4
i l
l SAFETY EVALUATION
SUMMARY
FOR H0D 92-0070, Supp. 01 (SE 94-0025, R.01)
TITLE:
Increase Motor Size for Motor Operated Valve HS106 CHANGE:
Replace Main Steam (MS) Valve, MS106 motor operator with a larger sized one.
HS106's motor vill be increased from a 15 foot-pound (ft.-lb.) motor to a 40 ft.-lb. motor. By increasing the motor size of this DC operated valve, adequate output thrust can be achieved while still maintaining the required stroke time.
REASON FOR CHANGE:
Due to Generic Letter 89-10 reviews and calculations, it was identified that the motor operator for HS106 has insufficient margin between the output thrust capability at reduced DC voltage and the thrust required to stroke the valve against the design differential pressure.
SAFETY EVALUATION
SUMMARY
There vill be no adverse affects on the piping which constitutes either the Auxiliary Feedvater System, for which HS106 is part of, or the Main Steam l
System, that MS106 interfaces with.
Hodification 92-0070, supplement 01 only replaces the motor on the valve's motor operator. There vill be no modifications done to the valve disk, body or any other pressure retaining portion of either the Auxiliary Feedvater System or the Main Steam System.
s rchased for this modification vill be environmentally and The new mot u
seismically qualified for its application.
There vill be no adverse effects on the reliability of the DC distribution system due to the increased operating current of the larger sized motor.
The increased motor size vill have no adverse effects on the 125/250 VDC distribution system in providing adequate voltage to either the new valve operator or to any other piece of distribution equipment.
There vill be no adverse affects on the Steam and Feedvater Rupture Control System (SFRCS) as a result of this motor changeout. No changes vill take place which vill affect the control circuits of the motor operated valve.
The motor replacement will not affect the stroke times for HS106 as normally measured. However, when loaded, due to the increased torque capability, the new motor vill not slow down as much as with the smaller existing motor, thus decreasing the time to fully close under load.
There vill be no changes to the actuator's gearing for HS106. Therefore, the valve's normal stroke time vill be unaffected by this modification.
Based on the above review of the effects on safety, the proposed replacement of
)
MS106's DC motor with a larger sized one are considered safe.
1 l
l
SAFETY EVALUATION
SUMMARY
FOR Mod 92-0071 (SE 93-0018)
TITLE:
Modification of Door 517 CHANGE:
The physical modification of Door 517 will include installation of two blow-out fasteners on the west side (RM 501) of the door and removal of the latch and handles.
The blow-out fasteners will secure the door closed until Room 501 pressurizes to approximately 0.2 psig at which point the fasteners will fail and the door will swing open.
REASON FOR CHANGE:
Door 517 is a single door located between Room 501, Radwaste Exhaust Equipment
& Main Station Exhaust Fan Room, and Room 514, Heater Bay Area (623').
This MOD will reduce the blow-out pressure of door 517 to allow venting of a High Energy Line Break (HELB) in Room 500/501 into the Heater Bay / Turbine Building Area.
Reducing the blow-out pressure will ensure that any steam flow which enters the Auxiliary Building Radwaste Area HVAC system is isolated to this level of the Auxiliary Building.
SAFETY EVALUATION
SUMMARY
Modification of Door 517 results in a door configuration that does not comply with the Underwriters Laboratory listing. Therefore a Non-rated Opening Evaluation was performed to evaluate the effects of this MOD on the fire protection capability of this barrier. The evaluation concluded that the overall fire protection capability of this fire barrier is not degraded to an unacceptable level by this MOD.
This modification will, prevent Door 517 from being used for personnel access / egress during the emergency. Therefore, there is no adverse impact on personnel safety.
This MOD affects the method of securing the door thereby potentially affecting its capability as an access control barrier.
However, based on the strength of the fasteners, removal of the handles, and the presence of the existing alarm.
Security has determined that this MOD does not degrade the security barrier and therefore does not impact the Security Plan.
1 The HELB analysis, assumes a blowout pressure of 5 0.25 psid for Door 517.
Therefore, the nominal setpoint of 0.20 psid i 15% is adequate to support this analysis.
The Heater Bay Elev. 623 does not contain any equipment required to mitigate the HELB or any other equipment important to safety, therefore, steam venting through Door 517 will not adversely affect plant shutdown.
DBP 5302DDDD/18
SAFETY EVALUATION
SUMMARY
t FOR H'OD 93-0011 (SE 94-0039)
TITLES-r Replacement of Valves SV6406 and SV6407 CHANGE:
Replaced the existing Target Rock solenoid valves with Anchor Darling air operated ball valves (A0V) which are better suited for the application.
4 REASON FOR CHANGE:
I Install new valves better suited to the application.
The poor performance of the existing valves is attributed to either dirt accumulation or flow induced.
vibrations in the recirculation line.
Solenoid valves are typically sensitive to these factors.
SAFETY EVALUATION
SUMMARY
This Modification installs 'Q' valve assemblies inclusive of accumulators,
{
valve operators, limit switches, solenoid valves and check valves.
?
The components must be functional following a seismic event and loss of offsite power.
The station instrument air system is not designed to withstand a seismic event or a loss of offsite power.
Therefore, since the valves fail open on loss of power and are required to remain closed during feed and bleed I
operations, accumulators are provided, each with a conservative air supply of three hours.
Chemical analysis of the valve seats revealed trace amounts of chloride and fluoride were present.
However, the amounts of chloride and fluoride present in the Viton is determined to not adversely affect the RCS chemistry.
Installation of this Modification will increase the reliability of the valves to open and close due to a different valve operating principle.
The existing valves are solenoid valves which utilize a pilot disc to either block flow or pass flow through the main disc. With this type of configuration there are several areas where dirt can accumulate since the flow through the valve is not a direct passage.
This type of valve also has close manufacturing tolerances.
The difference in the flow coefficients for both valves is considered insignificant due to a 15 stage restriction orifice located just downstream of the valves and designed to limit flow to 35 gpm.
Therefore, a negligible flow change will result.
Dirt accumulation will be minimal since the valve seats both on the ball side and the valve body side are exposed to flow which will wash the dirt from the valve internals. When the valve is closed, system pressure forces the ball into the valve seat providing tight shut off.
The valves / operators are capable of operation at a differential pressure of 2
3960 psid which is greater than the expected differential pressure of 3050 psid.
Therefore, implementation of this Modification is safe.
DBP $302FFF/50 1
4
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SAFETY EVALUATION
SUMMARY
TITLE:
Abandon Domestic Vater Heater in Place CHANGE:
Abandon the domestic water heater located in Turbine Building in place. The associated auxiliary steam (AS) and domestic water (DM) valves are also abandoned in place and AS inlet, DM inlet and outlet to this heater was physically isolated.
REASON FOR CHANGE:
Abandoning this water heater is to prevent the potential contamination of the domestic water system following a once Through Steam Generator tube leak if there is a failure of the DM heater tubes. Abandoning this heater in place removes this heater as a potential source of contamination outside the Auxiliary Building.
SAFETY EVALUATION
SUMMARY
The function of the DM heater is presently being performed by the Auxiliary Domestic Hot Vater Heater. Therefore abandoning DM heater in place vill not impact the domestic water system function.
Domestic water heater vill be abandoned in place and steam and water connections vill be isolated. The auxiliary steam supply will be isolated by closing AS265, removing flexible connection and installing blind flanges. The closed valve AS265 in series with the blind flange vill provide adequate steam isolation. The blind flanges vill comply with the design requirements of line class HBD. The steam traps ST102 and ST76 vill be abandoned in place and inlet and outlet valves of steam traps ST102 and ST76 vill be closed shut. This vill isolate the heater from any of the condensate coming out of steam trap ST10. The auxiliary steam system functions are not affected by closing AS265, the installation of the blind flanges or closing ST102 and ST76 inlet and outlet valves.
The Domestic Hot Vater Recirculation Pump recirculates hot vater from the office and Auxiliary Building hot water Recirculation line back to DM heater depending upon temperature. The heater supply valve DH95F vill be closed and the unions in this DM supply line vill be removed and replaced by caps. Also the union in return line vill be replaced by caps. The closed valve DM95F and the caps vill provide adequate domestic vater isolation. These changes vill not impact any Domestic Vater System function.
Based on the above discussions it is concluded that the proposed modification i
is safe.
l 1
SAFETY EVALUATION
SUMMARY
FOR Mod 93-0043 (SE 94-0020)
TITLE:
Delete Annunciator 9-3-B,
" Station Water Pre-Treatment System Trouble" CHANGE:
Remove control room annunciator window 9-3-B,
" Station Water Pre-Treatment System Trouble" while retaining the corresponding computer input point, Q991,
" Unit Water Treatment System".
All of the inputs to annunciator window 9-3-B are indicated locally in the water treatment buildings at instrument panels.
REASON FOR CHANGE:
This window is frequently in alarm due to normal system operating practices and therefore conveys no meaningful information to the control room operators.
The alarm condition does not require any immediate operator actions, the only action is to notify chemistry of the condition.
In addition the frequent alarm status of this window is distracting and presents an operational nuisance.
SAFETY EVALUATION
SUMMARY
There is no adverse effect on safety as a result of removing annunciator 9-3-B because none of the equipment or systems are safety related.
The deleted alarm is not interconnected with any safety feature and is not required for safe shutdown of the station.
This evaluation is required as USAR figure 9.2-3,
" Functional drawing Makeup Water Treatment System", is annotated showing system parameters that input to annunciator 9-3-B.
4 DBP 5302DDDD/7
SAFETY EVALUATION
SUMMARY
FOR H0D 93-0060 (SE 94-0054)
TITLE:
Replace Motor Operators for Valves HS603 and MS611 CHANGE:
Replaced the operators for valves MS603 and MS611 to ensure that the valves are capable of closing against design differential pressure.
REASON FOR CHANGE:
The Steam Generator (S/G) drainlines were modified previously to permit use of the lines at normal S/G pressure.
As a result of this change, a requirement was added for valves HS603 and MS611 to close against a differential pressure corresponding to full steam generator pressure. The ability of the motor operators to develop sufficient thrust to perform this added function is marginal.
SAFETY EVALUATION
SUMMARY
This modification increases the reliability of operation of HS603 and HS611 by significantly increasing the available thrust to close the valve.
Additionally, installation of the larger motor ensures that under degraded voltage conditions the operator is capable of developing full rated operator thrust.
The change in the control scheme to limit control for closing enhances the reliability of these valves.
Increasing the motor capability vill cause the potential stall thrust to be increased. The torque switch is retained in the closing control circuit to protect against motor stall.
The effect of the additional weight of the replacement operator has been analyzed.
All piping and supports remain within the allovable load limits.
This modification vill reduce the stroke time of HS603 and MS611 by approximately 3 %.
MS603 and HS611 are containment isolation valves. A relatively rapid stroke time vill reduce the amount of radioactivity released through this path. HS603 and MS611 are also required to close to isolate a downstream line break. The reduced stroke time vill slightly reduce the amount of S/G inventory lost in the event of a line break. Therefore, environmental conditions resulting from the break vill not be as severe.
HS603 and MS611 are automatically actuated by SFRCS for a steam line break, feedvater line break, or SG overfill.
SFRCS actuation is not affected by this modification.
c.
There vill be no adverse effects on the reliability of the Lov Voltage System due to the increased. operating current of the larger sized motors.
The various distribution components and cabling which serve MS603 and MS611 have been evaluated with respect to the higher starting and operating currents.
Based on the above discussion,.'it is concluded that the proposed changes are safe.
i
T SAFETY EVALUATION
SUMMARY
TITLE:
9 RF0 Fuel Repairs CHANGE:
Repairs may be accomplished in the Spent Fuel Pool (SFP) by reconstitution or recaging. Both processes require removal of the upper-end-fitting (UEF).
This allows access to defective fuel rods, which can be replaced by other fuel. rods or stainless steel dummy rods.
REASON FOR CHANGE:
During refueling outages, Davis-Becse may perform ultrasonic testing (UT) or visual examination of fuel assemblies that will be reused in the next operating cycle (s).
The objective for such a Post-Irradiation-Examination (PIE) campaign is to identify fuel rods with defective cladding.
Once identified, a decision may be made regarding feasibility of repair of the fuel assemblies.
SAFETY EVALUATION
SUMMARY
This safety evaluation addresses the safety aspects of the actual repair process and how it is controlled with vendor procedures.
It is generic and it applies to fuel assemblies with a reconstitutable, B&W Fuel Company (BWFC)
Mark-B fuel design.
The scope of this evaluation is limited to fuel assemblies with a maximum initial enrichment of 3.8 weight percent (3.8 w/o), and at least 5 GWD/HTU burnup.
Defective fuel rod breakage during extraction is possible due to a rod's weakened condition.
If this were to happen and the rod cannot be fully extracted or the spacer grids are damaged during the repair process, all of the sound rods would be transferred to a new Phoenix cage. A Phoenix cage is a reconstitutable. Mark B design FA, minus its fuel rods.
Procedural controls limit the extraction force applied to a fuel rod to a maximum of 200 lbs to minimize the risk of its breakage. Clad creep down and spacer grid relaxation due to irradiation have contributed to relatively low extraction forces and, therefore, acceptable tensile forces on non-defective fuel rod cladding.
Fuel rod extraction is controlled with the rod withdrawal tool by means of a mechanical hard stop which will limit withdrawal to approximately 8 feed below the water surface.
DDP 5302FFF/45 l
One problem with the irradiated rods' insertion process did occur during a repair campaign at Crystal River-3.
The misdirection of fuel rods to different spacer grid cell locations (a " snaking effect"), resulted in rods protruding from the sides of a fuel assembly.
To preclude that problem, recaging now is performed in a SFP rack cell with a checkerboarding tool.
As part of the continued maintenance process to remove a defective rod from its old cage and store it into a defective rod container or rod storage basket, a
special procedure will be issued to control the process with manual tooling and special rigging to the overhead crane.
The use of an auxiliary fuel handling tool and overhead crane to lift the old cage containing only a defective rod is allowed, as long as rigging with special slings assures that minimum water depth of 8 feet above the defective rod is maintained.
The radiation exposure to the fuel rod handlers will consequently be acceptable.
The mitigation of activity release during a fuel handling accident (FHA) is based on the fact that 23 feet of water is available to reduce the activity released by a factor of 100.
However, during the fuel repair the amount of water depth available for iodine removal in the event of a rod break varies depending on the location of the break. Conservatively, if it is hypothesized that the rod breaks after it has been fully extracted from the fuel assembly, the available water depth could be ~8 feet.
In such a case, the iodine removal by the SFP water would be less than is assumed in the FHA.
However, this postulated rod breakage incident is less severe than the FHA analyzed in the USAR and re-analyzed for a higher fuel enrichment and associated extended burnup limit.
Since the fuel repairs will be performed in the SFP with the Auxiliary Building's ventilation system operational and the Emergency Ventilation System (EVS) operable, any releases that result in EVS actuation will be discharged through 95% efficient charcoal filters to the station vent.
Therefore, releases to the environment resulting from a broken rod will be negligible.
If pellets are released from broken rods, they would be caught on trays that will be installed around the FA.
These trays will limit their dispersal into the SFP, and means exist to retrieve these pellets from the trays.
l SFP teactivity or criticality safety criteria during recaging has also been evaluated.
The recaging evaluation bounds the reconstitution repair process.
The evaluation concluded that recaging of category B type fuel with at least 1
5 bVD/MTU burnup could be performed with more than adequate safety margin.
In conclusion there is, therefore, no effect on safety, since this fuel repair is controlled with proven procedures, performed by BWFC personnel with fuel handling experience, uses tested fuel handling tools, quality components, and repair techniques and hardware further described in the Modification Design Report for Fuel Repairs.
The repaired fuel will meet all of the requirements of the BWFC Field Change Authorization (FCA).
FCAs document compliance with the BWFC Quality Assurance Program for Fuel Design Centre 1.
The modification consisting of repair by recaging and/or reconstitution of fuel defined as category B, based on the above is, therefore, considered safe.
DBP 5302FFF/46
4 SAFETY EVALUATION
SUMMARY
FOR p
MOD 93-0071 (SE 94-0001, R. 02)
TITLE:
Fused Safety Switch for the Switchyard.125V DC Distribution Panels DA1 and DB1 CHANGE:
I Installed a 200 amp fused safety switch between the switchyard 12SV DC panels DA1 and dbl.
This switch connects DA1 and DB1 to permit one battery / battery charger to supply both panels.
The' review of USAR Section 8.2.1.3 determined that statements concerning switchyard battery capacity and DC circuit redundancy were' misleading.
Specifically, the USAR stated, "Either battery is of a capacity capable of supplying all of the DC requirements of the switchyard in the event of a loss
-of the other."
This statement has been clarified.
REASON FOR CHANGE:
The fused safety switch allows all the fault' detection and fault clearing-functions to remain in service during the battery / battery charger replacement and during future battery maintenance and testing activities. Allowing the redundant fault detection and fault clearing circuits to remain operational improves the switchyard safety.
The chance of losing a non-redundant' fault detection or fault clearing function (which would result from deenergizing DA1 l
or DB1) is more likely than the chance of losing the " connected" DC system.
i SAFETY EVALUATION SUMHARY:
Closing the fused safety switch to connect panels DA1 and DB1 during battery / battery charger maintenance and testing will eliminate the physical and electrical separation of the existing switchyard battery systems; however, as explained below, this closed switch will not lessen the Station's overall compliance with GDC 17.
Physical and electrical separation exists so as to minimize the chance of f
simultaneous failure of two separate circuits from the transmission network to the standby power distribution system and to. assure that loss of one preferred system circuit will not cause or result in loss of the redundant counterpart.
The Davis-Besse switchyard battery systems and the switchyard DC control systems interact with the transmission network through the 345kV breakers.
Since there is no 345kV breaker in the path between the Ohio Edison i
transmission line and startup transformer 02, a DC system power failure (i.e.,
l overvoltage transients) which may cause the static relays to send erroneous trip signals to the 345kV breakers would not affect this preferred power system path This DC power system failure will not cause spuricus opening of the Ohio Edison line airbreak switches because these switches are not controlled by relays.
Thus, a power failure of the " connected" DC system (i.e.,
safety switch closed) will not increase the chance of a simultaneous failure of the DBP 5302FFF/42 i
preferred power system. Additionally, a loss of switchyard DC power will not cause the loss of any circuit from the transmission network to the standby power distribution system because switchyard equipment will not change state should DC power be lost.
With DA1 and DB1 connected to the same 250 A-h battery, sufficient terminal voltage may not be available to reliably reclose 2 breakers.
However, this evaluation maintains that it is better to lose the auto-reclose feature rather than redundant fault detection and fault clearing functions.
Complete DC circuit redundancy was never a design feature of the Davis-Besse switchyard.
Complete DC circuit redundancy implies that all DC circuits on one battery have functional counterparts supplied by the other battery.
DC circuits which involve fault detection and fault clearing are redundant.
This redundancy exists to ensure that large and destructive fault currents are cleared before significant equipment damage occurs.
With the exception of G Bus differential, redundant DC feeds were never available or planned for the remaining loads listed in that section.
Note, G Bus differential is a backup to the overall generator differential relaying which is powered by the Station DC System.
Thus, in a sense, G Bus differential is redundant.
Since DC circuit redundancy is an important feature in the switchyard fault redundancy exists for transmission line relays, for J Bus differential relays, for breaker trip coils and for blocking carrier pilots on the Bayshore and Lemoyne lines.
GDC 17 does not require that switchyard DC circuits be completely redundant.
GDC 17 states. "...Each of these circuits [ circuits frem the transmission network to the onsite electric distribution system) shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit...."
The existing switchyard design fulfills this requirement. Although 34SkV breaker closing circuits are not redundant, given a loss of onsite AC (Generator breakers trip) combined with a single from an unfaulted transmission line to a startup transformer without needing to close a circuit breaker.
It should be understood that a loss of switchyard DC will not cause any equipment to change state.
DBP 5302FFF/43
]
'c 4
SAFETY EVALUATION
SUMMARY
FOR MOD 94-0005 (SE 94-0051) i ll
. TITLE:
Replace Motor Operators on AF3870 and.AF3872 CHANGE:
Replace the existing motor on AF3870 with a 40 ft-lb DC motor and the motor on
-AF3872 with a 40 ft-lb AC motor. Also replace the SMB-00 operators with larger SMB-0 operators to increase the rated thrust.
REASON FOR CHANGE:
AF3870 and AF3872 vere originally designed to open with a limiting reverse differential pressure of 1180 psig from the steam generators.
Subsequent modifications to the AFV system have changed the normal position of AF3870 and AF3872 to open, and' require them to close against AFV pump discharge pressure.
These changes have resulted in problems with the motor capability and actuator rating.
SAFETY EVALUATION
SUMMARY
The proposed change to the auxilia y feedvater system vill ensure AF3870 and AF3872 can close at design differential pressure conditions with reduced-voltage. Calculations have been completed showing that with the new SMB-0-40
-]
operators installed on both valves are capable of stroking for the differential pressure conditions corresponding to all conditions, including single failures.~
These calculations also.show that the operator thrust and torque ratings may be exceeded if the valves.are closed against the pump head developed by an auxiliary feedvater pump that is operating at a speed in excess of the high speed stop. However, for this condition the operator one time overthrust rating would not be exceeded and the operator, although stressed, vould not be damaged. For all other closing conditions the torque and thrust ratings of the new operators will not be exceeded.
Increasing the motor. capability will cause the-stall thrust developed by the i
actuator to increase. If excessive, the thrust could cause the failure of a i
valve pressure retaining component. For AF3870 and AF3872 the stall thrust j
produced during closing with the larger motor installed is less than the valve one time thrust limit. Therefore, a motor stall during closing vill not cause any valve components to fail. During opening if the valve is stuck and the torque switch protection fails or is bypassed, the motor stall thrust could.
cause a failure of the stem to vedge connection, however no pressure retaining component would be challenged.
The seismic qualification of'AF3870 and AF3872 with the new operators was evaluated. The thrust range for limit switch setting vill limit the maximum stem thrust that occurs after limit switch trip to less than the seismic limit.
The seismic. thrust limit could.be exceeded during the valve stroke at high differential' pressure conditions, however a seismic event is not required to be postulated simultaneously with a initiating event that would depressurize a steam generator.
The effect of the additional vei *
.f the larger replacement operator has been
. analyzed. All piping and suppo' remain within the allowable load limits.
There vill be no adverse effects on the reliability of the DC Distribution
' System or the Lov Voltage System due to the increased operating current of the larger sized motors for AF3870 and AF3872.
The increased motor size vill have no adverse effects on the 125/250 VDC distribution system in providing adequate voltage to either the new valve operator or to any other piece of distribution system.
The increased motor size for AF3872 vill have no adverse effects on the low voltage AC System in providing adequate voltage to the new valve. operator or to other equipment.
The stroke time for these valves vill be decreased slightly due to the.
reduction of the operator overall actuator ratio.
The current.overall actuator ratio is 55.8 and the new overall actuator ratio will be 54.8.
This will cause
.a 1.8% reduction in the valve stroke time which corresponds to about.5 seconds. The stroke time requirement for AF3870 and AF3872 is not specifically listed in the USAR, however the stroke time of AF3870 and AF3872 is related to Auxiliary Feedvater System response time. The AFV SFRCS response time is specified as a' maximum of 40 seconds in Technical Specification Table 3.3-13.
Since AF3870 and AF3872 have a stroke time of approximately 30 seconds a.5 second reduction is negligible and vill have no effect on the overall AFV system response.
Based on the above discussion it is concluded that the proposed changes are safe.
i SAFETY EVALUATION
SUMMARY
TITLE:
AFPT 1 Governor Control Circuit Appendix R Isolation CHANGE:
Provide isolation for indicating light ICS038B (Amber light which provides indication that the control power has been transferred from the Control Room to the Auxiliary Shutdown Panel) and indicating lights ICS038E and F (AFPT 1 Governor Control low and high speed stop lights).
REASON FOR CHANGE:
The circuit for indicating lights of Auxiliary Feedwater Pump Turbine (AFPT) 1 Governor Control were not adequately isolated for a fire in the control Room or cable spreading room.
This is contr.ry to what is stated in the Fire Hazard Analysis Report (FHAR) for these areas.
During a serious control room fire the indicating lights circuit has the potential for affecting the controls of AFPT 1.
SAFETY EVALUATION
SUMMARY
The changes proposed by this modification will not affect the safety functions of the AFWS or the AFPT 1.
Implementing the changes proposed by this modification will not create new hazards for any plant equipment. On the contrary implementation of this modification will ensure that AFPT 1, during a serious control room fire, will be available to perform its intended function.
DBP 5302FFF/29
l
)
SAFETY. EVALUATION
SUMMARY
FOR MOD 94-0022 (94-0050)
TITLE:
Increase Motor Size and Change Circuitry for Valves AF599 and AF608 CHANGE s -
Replaced the existing 40 ft.-lb. motors on AF599 and AF608 with 60 ft.-lb.
motors'and changed the closing control logic of valves AF599 and AF608 so that these valves will be controlled for the entire closing stroke by the limit l
switch.
REASON FOR CHANGE:
1 L
Testing and evaluation, as a result of Nuclear Regulatory Commission (NRC) i Generic Letter 89-10, has determined that the valves will not stroke against a 1600 psid differential and their capability is marginal at the Auxiliary Feedwater Pump shutoff head.
SAFETY EVALUATION
SUMMARY
The proposed changes to the Auxiliary Feedwater System will ensure AF599 and AF608 can close at design differential pressure conditions with reduced voltage. Calculations have been completed showing.that with the new 60.ft-lb motors installed, both valves are capable of stroking for the differential pressure conditions corresponding to the events requiring operation of these valves while taking into consideration single failures.
Changes in the control scheme for the valve's operator such that it utilizes limit control upon closing will have no adverse effects, as it enhances the reliability of the valve.
There will be no adverse affects on the Auxiliary Feedwater Piping that AF599 and AF608 interfaces with.
This MOD only replaces the motor on the valve operator and changes the closing control logic.
There will be no adverse effects on the reliability of the Low Voltage System due to the increased operating current of the larger sized motor.
The various distribution components and cabling which serve AF599 and AF608 have been analyzed with respect to the higher starting and operating currents.
All cabling and other distribution components are adequately sized to prevent overheating.
The motor replacement and the change of the operation te limit control upon closing will not affect the stroke times for AF599 and AF608 as normally measured.
However, when loaded, due to the increased torque capability, the new motor will not slow down as much as with the smaller existing motor, thus decreasing the time to fully close under load.
There will be no changes to the actuator gearing for AF599 and AF608.
Therefore, the valve's normal stroke time will be unaffected by this modification.
DBP 5302FFF/22
The new motors purchased for this modification are environmentally and seismically qualified for their application.
Valve seismic qualification and pipe stress calculations performed indicate that the new motors will have no adverse affects on the piping and valve.
The existing supports and motor attachment are sized to take the increased loadings imposed by the new motors and still maintain functionality during a seismic event.
Based on the above review of the effects on safety, the proposed replacement of the motors on AF599 and AF608 with larger sized motors and the proposed change to the close logic to limit control are considered safe.
DBP 5302FFF/23
SAFETY EVALUATION
SUMMARY
FOR PCI,Q 92-0030 (SE 93-0029)
TITLE:
Use-As-Is Disposition of Discrepant Fuses 1
CHANGE:
Evaluate the effects of use-as-is disposition of discrepant fuses on plant
-safety.
REASON FOR CHANGE:
A walkdown of the fuses in the plant was performed. Differences between the j
field and the drawings were noted and evaluated. Action taken was to either correct the field, or correct the drawings or leave the discrepant fuse installed and not change the drawing.
If a discrepant fuse is ever removed from the circuit it should be replaced with the fuse on the drawing E2014. A reference to PCAQR has been added to the drawing E2014 to explain the difference for each discrepant fuse.
SAFETY EVALUATION
SUMMARY
The differences in the fuses were either in the current carrying capacity of the fuses or in the type of fuses.
The time-current characteristic curves of the installed fuses were compared with that of the upstream fuses and concluded that adequate coordination exists.
For all of these fuses, coordination downstream is not a concern since there are no protective devices downstream.
Also, the load on the circuit is within the current and voltage ratings of the installed fuse.
Since the fuses can withstand the circuit loads and are coordinated with the upstream fuses, the equipment will be operational and the discrepancy will have no effect on the safety of the plant.
e DEP 5302FFF/54
r-o SAFETY EVALUATION
SUMMARY
FOR PCAQ 93-0287 (SE 93-0027)
TITLE:
Use-As-Is Disposition for MS734 and MS735 CHANGE:
S!nce it could not be confirmed that the valves MS734 and MS735 could close from a fully open position, it was proposed to use the valves as-is and manually close the valves after events known to push the disk in the open direction, e.g.,
test runs of the auxiliary feed pumps.
This assures that the projected disk area is large enough that the valves will close to prevent reverse flow in the unlikely event of a steam line break upstream of a valve.
REASON FOR CHANGE:
During a steam line break event upstream of MS734 and MS735, auxiliary feed pump turbine main steam line crossover check valves, the valve is designed to close and ensure the integrity of the undamaged train.
Because of a external arm, the new valves require packing.
Following tightening of the packing sufficient to prevent leakage during hydrostatic testing, it was found that the resultant packing load on both valves required approximately 40 to 45 ft-lbs torque on the valve stem in order to initiate valve motion in either direction.
Scoping analysis using estimated valve dimensions and flow velocities resulting from various line break scenarios were not able to confirm that either of these check valves could close from a fully open position.
The subject check valve disks are capable of retracting well out of the flow stream, which results in a relatively small projected disk area.
SAFETY EVALUATION
SUMMARY
The replacement valves must perform the same function as the original valves.
For High energy line breaks (HELBs) in the Main Steam Supply to the AFPTs upstream of the check valves, the check valves (MS734 or MS735) must close and stay closed.
In addition, the check valves must be capable of opening to supply adequate steam to the auxiliary feedwater pump turbines.
Following sufficient tightening of the packing to prevent leakage during hydrostatic testing, it was found that the resultant packing load on both valves required approximately 40 to 45 ft-lbs torque on the valve stem in order to initiate valve motion in either direction.
The valves were not considered to be damaged.
However, due to the shaft torque, the valves will function differently than the original valves in that the disks will not fall closed by gravity and will require additional force to open.
The maintenance instructions for these valves were modified so that if the packing is tightened, shaft torque will be verified to be within design limits.
In the forward flow direction. if the valve is initially closed, a large differential pressure will initially develop across the full valve disk area to provide a surplus in available opening torque.
It has been shown by field DBP 5302FFF/60
testing that the subject valves will sufficiently open in order to supply steam to the auxiliary feedwater pumps.
Scoping calculations were performed to determine if it could be concluded that the valves would be capable of closing from a fully open position under all postulated requirements.
If the valve is in the fully open position, the projected area of the valve disk in the flow stream is small. Due to the large number of variables involved (i.e., potential changes in packing load, initial valve position, variable flow rates / durations dependent on postulated break size / location) these calculations were not conclusive in providing a definitive result.
However, if the initial position of the valve disk is fully in the flow stream (nearly closed), adequate torque will be developed to close the valve for any postulated break location upstream of the check valves.
For these reasons, it is necessary to manually position MS734 or MS735 to the closed position following any plant evolution which could open the valves.
MS734 and MS735 will open (at least partially) during testing of the Auxiliary Feedwater pumps.
During normal plant operations, the valves will supply a small quantity of steam to makeup for condensation in the stagnant supply lines to the idle AFPTs.
The amount of makeup steam is far below that which would cause significant opening of the valves.
Pressure fluctuations of up to
.5 psid have been observed during plant operation with the original valves in place.
These pressure fluctuations, while capable of overcoming the packing friction on a fully closed valve, ate not expected to significantly open the valves.
Other significant causes of rapid increase in steam generator pressure, such as a main turbine trip, could cause slight opening of the subject check valves. However, due to the limited downstream (dead-end) volume, the valves should not significantly open.
lAlthough a turbine trip may occur following a high energy line break in the main steam piping, slight opening of the subject valves due to an elevated pressure in the main steam piping due to main steam isolation valve or turbine stop valve closure is not a concern.
For breaks upstream of these valvee, there will not be any upstieam pressure to open these valves. With the check valves in the closed position.
these valves will remain closed either due to trapped pressure or due to steam pressure from the unaffected steam generator.
Based on the above, the auxiliary feedwater pump turbine test procedures were modified to require manually repositioning MS734 or MS735 to the closed posit ion whenever the applicable AFPT is run.
A standing order has been established to check the position of the valves following testing.
Other plant operational conditions are not expected to significantly open MS734 or MS735.
1 The administrative controls described above will provide assurance that the valves are always in a nearly closed position during plant operation.
Providing that the subject check valves are initially in a nearly closed position, no USAR analysis was found to be adversely affected.
The AFFTs are only required to operate in response to an emergency condition. An additional accident initiator, such as a HF.LB on the MS to AFPT supply lines. is not required to be postulated while the AFW system is in operation. MS734 or MS735 will remain closed or be capable of closing in the event of other accident conditions, such as a 36" main steam line break or a rain feedvater line break.
DBP 5302FFF/61
SAFETY EVALUATION
SUMMARY
L FOR PCAQ 93-0552 (SE 93-0068)
TITLE:
Operation With only Three Turbine Bypass Valves Immediately Available CHANGE:
To evaluate operation with two turbine bypass valves (one per steam generator) isolated due to seat leakage.
REASON FOR CHANGE:
Turbine Bypass Valve (TBV) SP13A3 was found to be incapable of fully stroking.
The following describes the ramifications and potential effects on plant operation for temporary continued operation in this condition.
SP13A3 is capable of smoothly opening to approximately 50% of full stroke, and exhibits no tendency to stick open.
SAFETY EVALUATION
SUMMARY
The TBVs are used during normal heatup, startup, cooldown, and hot standby of the plant to control main steam pressure.
Three functional TBVs have sufficient capacity to provide for normal heatup, cooldown, and hot standby, but may complicate startup operations.
There is no direct safety function associated with the TBVs.
Normally, six TBVs would be available, having a combined relief capacity of 25%
of rated steam flow. USAR Section 10.2.1 indicates that the control system, in conjunction with the TBVs, is capable of withstanding a turbine trip from 402 power without safety valve action.
The additional reduction from four to three in-service TBVs will have relatively little additional effect.
During a loss of load from 100 percent power. as discussed in USAR section 15.2.7, the TBVs are assumed to operate in conjunction with the steam generator code safety valves.
This characteristic will not qualitatively change.
The basis of the turbine trip arming setpoint is to allow for a turbine trip without reactor trip via ARTS, when the system is below a threshold power level.
This threshold is determined to be a power level where turbine bypass / relief capacity is sufficient to avoid a high RCS pressure trip following a turbine trip.
The current 45 percent power arming setpoint is based on the capacity of a " generic B&W plant."
There is margin available due to the higher total plant-specific relief capacity at Davis-Besse if all components are in service.
The combined relief capacity of the four MSSVs set at '030 usig is approximately 19.8 percent, while the remaining 3 TBVs would be approximately 12,5 percent.
A B&W report indicates that peak RCS pressures will occur at approximately 8 seconds or more.
The AVVs have a stroke time of approximately 10 seconds.
Taking credit for eighty percent of the AVV capacity, the AVVs would add 7.9 percent to the available relief capacity.
Thus, total relief DBP 5302FFF/55
E l
capacity would be 40.2 percent, and the initial reactor power to accommodate a turbine trip without reactor trip would be five percent higher at 45.2 percent.
Therefore, current Technical Specification allowable ARTS arming setpoint of 45 percent power remains justified. Additional margin actually remains because process steam loads were not considered.
Use of the TBVs is also discussed in USAR section 15.4.2, Steam Generator Tube Rupture (SGTR).
The TBVs are described as being used to aid in RCS cooldown efforts by directing steam to the condenser rather than directly to atmosphere.
Operation with only two TBVs might initially result in an increase in actual offsite dose because the affected steam generator might continue to vent directly to atmosphere for a longer period of time.
However, the analyzed radiological consequences would not change because no credit has been taken for retention of fission products by the condenser.
Based on past simulator modeling of the event, with only one TBV and one AVV available per steam generator, isolation of the affected steam generator will still be accomplished within the USAR time frame.
However, the assumed 100 degree per hour cooldown rate might require additional TBV capacity following the first hour.
This would be accomplished by operator action to reactivate the isolated TBVs.
Based on the above considerations, operation with two turbine bypass valves isolated (one per steam generator), and with SP13A3 in service at reduced capacity, is safe.
l DBP 5302FFF/56
SAFETY EVALUATION
SUMMARY
FOR PCAQR 94-0065 (SE 94-0038)
TITLE:
Capability of AF3870/AF3872 to Isolate Auxiliary Feedwater Flow CilANGE :
This evaluation examines the ability of the AFW system to perform its safety functions given the postulated failure of valves AF3869, AF3870, AF3871 and AF3872 to close and an additional system single failure.
REASON FOR CHANGE:
Under certain conditions, AF3870 or AF3872 may fail to fully close due to high differential pressure independent of any system single failures.
The torque switch settings for these valves were based on a 400 lb, differential pressure.
These valves may be required to close against a differential pressure greater than 400 lb.
SAFETY EVALUATION
SUMMARY
The ability of the AFW system to perform its safety functions given the inability of these valves to close and a single failure is evaluated.
AF3870 and AF3872 Double Ended Main Steam Line Break Comparison of the current plant configuration and the B&W reanalysis, done in support of the June 9, 1995 event modifications, demonstrated the following:
- 1) The B&W reanalysis represents a bounding case: 2) For steam line breaks with continued feed to the affected steam generator, the magnitude of AFW flow to the unaffected steam generator is relatively unimportant 3) Some AFW would be added to the unaffected SG via the cross-connect path concurrent with flow to the affected SG.
If this flow were sufficient to recover inventory to the
~130 inch (indicated) control level the flow control valve would then modulate closed /open, controlling level.
This would serve to reduce flow to both SG's, limiting heat transfer via the affected SG. Any RCS heatup would be limited by Tsat of the secondary side in the unaffected SG.
As such, the plant cooldown would be bounded by the B&W analysis and decay heat would also be removed; and
Flow to the affected steam generator would be isolated by operator action to close AF608 or AF599, should AF3870 or AF3872 fail to close.
This action would terminate flow to the affected SG and restore full flow to the unaffected SG from the running AFW pump.
DBP 5302FFF/51
High Energy Line Breaks (Steam Leaks)
For these events. SFRCS actuation on low SG pressure can occur as result of operator mitigation actions.
Breaks at specific locations invoke a procedurally required manual initiation of AFW flow to both SGs, followed by isolation and blowdown of the affected SG.
As a result, a low SG pressure SFRCS actuation on the affected SG will eventually result.
In these cases, however AFW flow is isolated prior to SG blowdown. As such AF3870 or AF3872 would not be required to stroke against a significant differential pressure.
Main Feedwater Line Breaks For main feedwater line breaks downstream of the last check valve, affected SG low pressure will eventually result, aligning top-injected AFW flow to the unaffected SG.
As stated in USAR Section 15.2.8, associated mass and energy releases are bounded by steam line break events.
Review of system performance criteria and applicable plant safety analyses concludes that the interim plant configuration for AF3870 and AF3872 does not render the AFW system unable to perform its safety functions.
AF3869/AF3871 Review of the AFW system determined that these valves are required to close in the event of an isolable steam line break.
For a main steam line break downstream of an MSIV that initiates a low pressure SFRCS signal, AF3869 (AF3871) will initially open. After the line break is isolated the affected SG will repressurize and the low pressure will clear. A subsequent SFRCS initiation (e.g.,
feedwater reverse dp) will realign each AFP to its respective SG, and AF3869 (AF3871) will receive a close signal.
The closing differential for this condition could be greater than the 400 psid torque switch setting.
AF3869 and AF3871 provide a valve position input to the ATW level control system.
This input is controlled by a limit switch contact that is actuated at the valve full.open position.
The limit switch contacts that are actuated at the valve full close position do not provide any input to the level control system.
Therefore, the failure of these valves to fully close will not affect the operation of the AFW level control system.
Analysis indicated that for the worst case closing differential pressure the valves would be substantially throttled or closed.
Therefore, the amount of flow diverted from the intended steam generator would be small.
Also any flow diverted would be going to the opposite steam generator which is not failed. As such, the improper setting of the torque switches for these valves would not affect the system function in the event the valves are required to close after an isolable steam line break.
From a single failure perspective, if AF3869 (AF3271) were to fail to close and the worst case single failure of the ATV train on the
" unaffected" side were to occur, both SGs would be fed following repressurization of the "affected" SG.
All of this water would be used to remove decay heat. Most of the flow would be supplied to the repressurized (affected) SG because the AF3869 (AF3871) cross connect DBP 5302FFF/52
valve to the unaffected SG would be throttled. Level control would be transferred back to the affected SG following repositioning of AF3870/AF3872.
Feed rate and level control in the affected SG would be adequate.
Because both SGs are able to be steamed, the unaffected SG (with a lower feed rate) would potentially have a lesser water level, which is acceptable.
In the event that neither AFW train failed, the AFP on the unaffected side would operate with lesser demand, but with normal (appropriate) level control.
Thus, adequate decay heat removal and AFW system control would be available.
Review of system performance criteria and applicable plant safety analyses conclude that the interim plant configuration for AF3869 and AF3871 does not render the AFW system unable to perform its safety functions.
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DBP 5302FFF/53 i
4 SAFETY EVALUATION
SUMMARY
FOR PCAQ 94-0108, UCN 94-074 (SE 94-0033)
TITLE:
Gai-Tronics Power Supply CHANGE:
1 Clarify language in USAR 9.5.2.2.9 which describes the power supply for the j
Gai-Tronics communication system.
The USAR refers to the Gai-Tronics system as "the main internal system".
The USAR states "Also, the main internal system is 1
supplied by two redundant power feeders from the uninterruptable distribution panels."
This change will revise these words to read, "Also, some sections of the main internal system are supplied by two redundant power feeders from the 1
uninterruptable instrumentation distribution panel. As a minimum, these sections can be found in the Turbine Building, Auxiliary Building, and Containment."
REASON FOR CHANGE:
I The portions of the Gai-Tronics system which are in the operating parts of the plant are powered from YAU and YBU, which are uninterruptable instrumentation distribution panels.
Over time, this system has been extended for non-operating uses, and operating uses which are trivial to safe operation.
Gai-Tronics stations now can be found as far away as the DBAB.
YAU and YBU are not designed to power these additional stations, and even if the panels themselves could power all the stations, it would not be appropriate to extend uninterruptable power to all of the remote locations served by the Gai-Tronics i
system.
SAFETY EVALUATION
SUMMARY
This change clarifles the operating power sources for the Gai-Tronics communications system. The Gai-Tronics communications system is not important to safety, and there is no effect on any safety-related power source.
Therefore, there is no effect on safety.
1 DBP 5302FFF/ll i
i 1
t SAFETY EVALUATION
SUMMARY
FOR PCAQR 94-0459 (SE 94-0044)
TITLE:
l l
MSIV Bypass Valves, MS100-1 and MS101-1 CHANGE:
I l
Changed the test frequency of the Steam and Feedwater Rupture Control System l
(SFRCS) output relays for the Main Steam Isolation Valve (MSIV) bypass valves from monthly to a refueling test.
This testing frequency is consistent with the SFRCS testing frequency of the MSIV.
REASON FOR CHANGE:
Certain output relays in the SFRCS were not being tested with the required frequency.
In order to test such relays, the actuated valve must be placed in its non-actuated position prior to SFRCS testing.
However, the MSIV bypass valves, MS100-1 and MS101-1, are interlocked with the MSIVs. When the MSIVs are open, as during power operation, MS100-1 and MS101-1 are provided with a close signal and cannot be opened without defeating the interlocks.
This configuration makes monthly functional testing of the SFRCS logic which actuates the MSIV bypass valves difficult and significantly increases the risk of inadvertently tripping the SFRCS logic.
This could lead to a trip of the reactor and provoke an unnecessary challenge to plant safety systems.
SAFETY EVALUATION
SUMMARY
The suggested SFRCS testing frequency for these valves would be the same as the SFAS testing frequency when the MSIVs and MSIV bypass valves received an automatic SFAS signal.
1 The inservice testing (IST) program specifies a cold shutdown testing frequency l
i for the MSIV bypass valves.
This ensures that the valves will be stroke tested during cold shutdown in a refueling outage and also during a mid-cycle cold shutdown condition, if time permits. This provides additional assurance that the bypass valves can be manually closed from the control room.
During power operation the MSIV bypass valves are normally in their SFRCS actuated state with the SFRCS output relays already deenergized.
Therefore, monthly functional testing of the SFRCS logic for the MSIV bypass valves which confirms that the SFRCS output relays can go to a de-energized state is unnecessary.
Testing which requires the output relays to be energized may be done during refueling only.
Therefore, given the normally closed and interlocked valve configuration, and the valves being opened for a short duration in Mode 3.
SFRCS testing of the output relays on a refueling frequency is sufficient to verity the accident l
function for the MSIV bypass valves.
This would be consistent with current HSIV SFRCS testing frequency and the previous SFAS testing frequency for these valves.
This also will reduce the possibility of inducing a plant trip when j
testing during power operation.
DBP 5302FFF/26
SAFETY EVALUATION
SUMMARY
FOR PCAO 94-1288 (SE 95-0002)
TITLE:
l ASMB Cetification for Hot Leg Level Monitoring System Component CHANGE:
Allow operation with lack of ASME certification for two 1/2 inch stainless steel tubing tees in the hot leg level monitoring system.
REASON FOR CHANGE:
Two 1/2 inch stainless steel tubing tees installed for part of the hot leg level narrow range monitoring system did not have ASME certification as required.
SAFETY EVALUATION
SUMMARY
The tees are installed for the hot leg level narrow range monitoring system.
The tees provide a tie-in to the 1/2" tubing sensing lines used for the vide range level monitoring system. The failure of these tees vould render both the narrow and vide range systems inoperable. Technical Specification 3.3.3.6 POST-ACCIDENT MONITORING SYSTEM requires one channel of Reactor Coolant Hot Leg Level (Vide Range) operable in modes 1,2, and 3 with a 30 day action statement.
For plants with construction permits issued prior to January 1,1984, the NRC imposes no specific requirements for Quality Class B and C components and systems.
Since Davis-Besse's construction permit was issued prior to January 1, 1984, compliance to ASME Code Section III Class 2 and 3 is not an NRC requirement.
Davis-Besse voluntarily decided to adhere to ASME Class 2 and 3 criteria The system is capable of performing its specified function without a decrease in reliability based on the following a) The Tee's have a pressure rating of 7800 psi; b) A telecon from the manufacture states that there are no differences in the production process, material requirements, or testing requirements for the commercial quality and safety / nuclear quality (0) fittings; c) Dye-penetrant testing was successfully performed as part of the post implementation testing; d) The installed Tee's have been in service since the sixth refueling outage without an occurrence of leaking There is no adverse affect on the capability of the plant to safely shutdown because the capacity of the Make-up system would well exceed the flow of a leak caused by the failure of two 1/2" Tee's.
Since there is no affect on any of the functions important to safety, there is no adverse affect on the design bases, Technical Specification bases, or single failure criteria analyzed in the USAR. There vill be no increase in adverse effects from any hazard as a result of this condition.
Based on this, continued operation does not represent an unsafe condition.
SAFETY EVALUATION
SUMMARY
FOR SCC 89-0529 (SE 89-0244, R.01)
TITLE:
Installation of Unions and Isolation Valves for the Turbine Plant Cooling Vater Sample Coolers CHANGE:
This change added a union and an isolation valve at the outlet TPCV side and a union at the inlet TPCV side of sample coolers S-004A-1 through 3, S-008-1 through 4 and S-014-1 and 2.
REASON FOR CHANGE:
The change facilitates inspection and cleaning of the cooler internals for each cooler without removing additional coolers from service.
With the prior configuration, to take a cooler out of service for coil cleaning (normally performed every two to three years) its inlet isolation valve and valve CV281 had to be closed.
Closing CV281 affects the other coolers since the valve is installed on the coolers common discharge header and serves as the isolation outlet valve for all coolers.
i SAFETY EVALUATION
SUMMARY
The proposed change vill not adversely impact the operation of Turbine Plant Cooling Vater System. The valves and unions required for this modification 3
meet all Design Requirements.
Vith the isolation valve at the outlet side, each cooler may be taken out of service for maintenance while keeping the cooling water flow through the other coolers allowing for sample collection.
The threaded unions vill allow an easy disconnection / connection of the coolers.
The Secondary Sampling System is not affected by the proposed change because the change is outside of the Secondary Sampling System boundaries. The sample i
lines are made of stainless steel tubing and Svagelok fittings that allow easy i
disconnection / connection of the coolers.
Based on the above discussion it is concluded that the proposed modification is safe.
SAFETY EVALUATION
SUMMARY
FOR SCC 90-3037 (SE 90-0086, R01)
TITLE:
Replacement of Globe Valve FW174 with Three Restricting Orifices CHANGE:
Replaced FW174 with three (3) restricting orifices, sized to throttle flow in the test line without damaging cavitation and protect the Motor Driven Feed Pump (HDFP) from runout.
REASON FOR CHANGE:
Globe Valve FW174 was installed under FCR 86-425 to throttle flow in MDFP test line.
During testing, it was determined that FW174 could not throttle the flow without causing excessive vibration due to cavitation.
SAFETY EVALUATION
SUMMARY
This change will not impact the conclusions reached from the 1985 Motor Driven Feedpump Hazard Study.
Specifically, the piping being modified will be capable of withstanding a seismic event of.06g.
There is no safety structure, system. or component affected by the proposed change and it will not degrade any equipment and will not prevent any system from functioning as specified in the USAR.
DBP 5302FFF/44
SAFETY EVALUATION
SUMMARY
TITLE:
l Raising the SFRCS High Level Trip Setpoint CHANGE:
Addresses raising the Steam and Feedwater Rupture Control System (SFRCS) high level trip setpoint to 240" from the previous setpoint of 225".
REASON FOR CHANGE:
i Allows greater operational flexibility and reduces the trip frequency resulting from main feedwater (MFW) upsets without reducing the level of plant safety.
SAFETY EVALUATION
SUMMARY
USAR Chapter 15 addresses overfeed events such as a MFW malfunction and assumes a 152 increase in MFW flow.
For the case at 100Z FW, the USAR transient is terminated by the liquid water filling the steamline and being entrained into the MFW pump turbines and rendering them inoperable.
The analysis does not take credit for the SFRCS high level trip terminating the transient.
The USAR analysis remains bounding and will not be affected by any modification of the SFRCS high level trip setpoint.
At the current setpoint of 225" SU range, the liquid water in the steam generator exceeds 62500 lbm at approximately 215" SU range during an extreme overfeed event.
Based on the current plant design bases, a main steam line l
break (MSLB) is not taken simultaneously with an overfeed event.
Therefore, the increased mass in the steam generator greater than the USAR MSLB analysis is not considered to have an effect on safety during an overfeed event.
Similarly, raising the SFRCS high level trip setpoint will not affect the USAR MSLB analysis.
i The effects on plant operation due to raising the SFRCS high level trip setpoint to 240" SU range will be two fold:
1.
To address continued power operation at high steam generator levels with reduced aspirating steam flow, but at steam generator levels less than the high level trip setpoint, manual action will be available to limit power and steam generator leading to preserve less than 10 degrees F subcooling in the downcemer.
This is consistent with B&W operational guidance to limit stresses in the steam generator " knuckle region" and the lower tube sheet.
2.
To determine the effect on the steam generators and the main steam line of rapidly evolving transients, the SU range was correlated with the amount of superheat at the steam generator exit and carry-over levels from the steam generator during worst case overfeed transients. No specific DBP 5302FFF/18
l l
i requirements exist for the amount of carry-over which is acceptable.
The analysis to support this safety evaluation correlates the SFRCS high level i
setpoint with loss of superheat at the steam generator exit.
This will conservatively limit the amount of carry-over into the main steam line.
The Davis-Besse RELAPS Basedeck was used to perform the analysis with modifications suitable to ensure conservatism for overfeed events.
During an overfeed event,-the cold water in the downcomer cools the outer shell and produces compressive loading on the tubes.
The differential temperature between the steam generator shell and the steam generator tubes is limited to 60 degrees F.
Additional analysis was provided to show that the tube to shell temperature difference is acceptable for the worst case overfeed transients terminated by loss of superheat.
)
DBP 5302FFF/19
SAFETY EVALUATION
SUMMARY
TITLE:
ICS Steam Generator High Level Limiter Setpoints and NNI Steam Generator Operate Range High Level Alarm Setpoints CHANGE:
The setpoint of the Integrated Control System (ICS) Steam Generator (SG) high level limiters, and the setpoint of the Non Nuclear Instrumentation (NNI)
System SG Operate Range (OR) high level alarms were increased from the value of 82.5 percent on the OR to 96 percent on the OR.
REASON F0F CHANGE:
To prevent station power output from being limited because of fouling in the SGs causing water levels in the SG downcomer to increase.
SAFETY EVALUATION
SUMMARY
There is no direct effect on any of SG functions important to safety because they are all inherent to the SG design and these changes are not changing the design of the SG.
Potential adverse indirect effects on the SG tube integrity (reactor coolant pressure boundary) were evaluated as part of SE 92-0014, Addressing raising the SFRCS high level trip to 240".
That evaluation showed that at the proposed high SFRCS high SG level trip of 240" on the Startup Range, the 600F allowable differential temperature between the SG tubes and the SG shell will not be exceeded during overfeed transients.
The level increase proposed in SE 92-0014 bounds the setpoint increase proposed by this evaluation.
Therefore, there is no indirect adverse effect on the SG tube integrity by this proposed change.
There will be no adverse effect.on safety from the proposed change to the NNI SG or high level alarm setpoint because its function is not important to safety but rather is an operational function which can still be performed with the proposed setpoint.
Any effects this proposed change may have on the Main Steam Line Break analysis are bounded by the analysis done for LAR 91-0019 which requests a change to the Technical Specification maximum SG water level. Any effects this proposed change may have on the Excessive Heat Removal due to the Feedwater System Malfunction analysis are bounded by the analysis done for LAR 91-0019 and SE 92-0014.
The probability of malfunction of the ICS SG high level limiter is a function of factors such as its design, installation, maintenance and operating environment.
These factors are not affected by changing the setpoint of the equipment.
The proposed setpoint will also not cause the equipment to operate more frequently than the current setpoint, such that a wear-and-tear malfunction is no more probable.
Therefore, since changing the setpoint does not affect the design, operating environment, installation, etc. the probability of malfunction of the equipment has not changed.
DBP 5302FFF/16
The LOFW accident is a USAR class 1 event. As such it leads to no radioactive release at the exclusion area boundary. Changing the setpoints as proposed will not prevent or degrade any actions described or assumed in the LOFW accident, nor alter any assumptions previously made in evaluating the radiological consequences.
These setpoints play no role, direct or indirect, in mitigating the radiological consequences of any accident as described in the USAR, nor do they have any effect, direct or indirect, on any fission product barrier.
Changing these setpoints will not create any new radioactive release pathways, nor impede any personnel actions needed to mitigate the radiological consequences of any accidents evaluated in the USAR.
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DBP 5302FFF/17
SAFETY EVALUATION
SUMMARY
TITLE:
Update Low Voltage Setpoint Information CHANGE:
Clarified existing relay setting information for the 125/250 VDC Power System and deleted setpoints and part descriptions for the Battery Charger Input Low Voltage Alarm Relays, Battery Charger Output Low Voltage Alarm Relays, DC MCC Under/overvoltage Alarm Relays, and DC Distribution Panel Undervoltage Relays from the USAR.
The setpoints were replaced by references to the controlled documents which list the required setpoints.
REASON FOR CHANGE:
A Request for Assistance identified a number of errors in the documentation of undervoltage relay setting information for 125/250 VDC Power System. Although none of these errors led to an operability concern, this change was needed to correct errors and consolidate information.
SAFETY EVALUATION
SUMMARY
SCR 93-5004 will have no effects on safety.
The settings of the DC MCC Under/overvoltage Alarm Relays are established by Calculation C-EE-002.01-009.
These settings are developed by considering drift, accuracy, precision, M&TE error, and the safety function of the relays.
Therefore, deleting the setpoints of the DC MCC Under/overvoltage relays from the USAR will have no effect on safety.
The setpoints of the Battery Charger Input Undervoltage Alarm Relays, Battery Charger Output Alarm Relays, and DC Distribution Panel Undervoltage Alarm Relays have no safety function. Therefore, deleting the setpoints of these relays from the USAR will have no effect on safety.
Although SCR 93-5004 contains related editorial corrections and minor plant configuration clarifications, there are no other changes which have any effect on plant safety.
DBP 5302FFF/7
l SAFETY EVALUATION SUMHARY FOR 4
TITLE:
Auxiliary Feedwater Pump High Speed Stop Setting CHANGE:
Change AFW pump High Speed stop setting from 3600 rpm to 3700 rpm.
REASON FOR CHANGE:
The normal operating speed of the AFW Pumps is presently controlled by the High Speed Stop (HSS) setting at 3600 rpm.
Setpoint Change Request 93-5012 will change the HSS setting to 3700 rpm to provide a higher AFW pump differential j
pressure.
This will allow adequate margin between the actual AFW pump differential pressure and the acceptance criteria for the AFW pump quarterly surveillance test SAFETY EVALUATION
SUMMARY
Changing the HSS setting to 3700 rpm will ensure adequate margin is available between the actual-differential pressure and the required differential pressure i
for the AFW pump quarterly surveillance test.
Based on pump affinity laws, increasing the high speed stop from 3600 rpm to 3700 rpm will result in J
~70 psid of additional pump head and provide adequate margin for the pump.
Based on a review of engineering documents, this increased discharge pressure of the pump will not adversely affect the piping stresses, due to either dynamic or static loading.
Increasing the HSS setting by 100 rpm does not affect the ability of the AFW System to meet Technical Specification requirements for delivery of 600 gpm AFW flow to the SG in less than 40 seconds.
Although the turbine has always been operated at 3600 rpm and the vendor manual has been changed by Toledo Edison to reflect this, 3710 rpm still is a vendor approved HSS setting which will prevent any pump / turbine damage.
Historical overshoot on AFPT start-up has never exceeded 100 rpm based upon past quarterly and monthly testing.
Therefore. 3700 rpm will not cause the turbine to reach its overspeed trip upon a start-up.
The higher turbine speed will require an insignificant increase in steam flow which will not adversely affect the MS System.
Pursuant to the above, it is concluded that raising the MSS setting by 100 rpm is safe.
DBP 5302DDDD/23
SAFETY EVALUATION
SUMMARY
TITLE:
Provide Breaker Trip Setpoint for Replacement Lighting Breaker CHANGE:
Provide a setpoint for the spare lighting substation breaker, and revise drawing E-4, sheet 4 in order to allow either the existing breaker or the new spare as the main feedwater breaker for either lighting substation.
REASON FOR CHANGE:
In order to allow for increased availability of normal plant lighting during maintenance evolutions on breakers BCE5 and BDF5, a spare main feeder breaker was procured.
The spare breaker will be used in place of either breaker BCE5 or BDF5 in order to maintain operability of the switchgear while the breaker undergoes scheduled maintenance.
This spare breaker, however, contains a 1200 Amp Siemens SA-II, solid-state trip unit.
This new trip unit, although technically acceptable, cannot be set to exactly match the 1300 Amp trip setpoint by virtue of its 1200 Arap base rating. As a result, drawing E-4, sheet 4 will have to be revised to allow either breaker (i.e.,
the 1000 Amp trip unit set at 1300 Amps or the 1200 Amp trip unit set at 1200 Amps) to be used.
SAFETY EVALUATION SUMht.RY:
There are no adverse affects on safety as a result of the proposed use of the spare 480 V breaker with the Siemens trip unit in either lighting substation E5 or F5.
The spare breaker, which will serve as part of the power distribution system for the plant's normal, nonessential lighting, will not affect the operability of either the standby or emergency lighting systems.
The spare breaker with the Siemens trip unit is properly rated for the load of either nonessential substation and has a trip characteristic that closely matches the existing trip characteristic of the SS2 equipped breakers.
Proper setpoint selection will also ensure that the standby and emergency lighting systems are not challenged in the event that illumination is required for safe shutdown and accident mitigation.
There will be no human factors concerns generated by the proposed use of the spare breaker in either location.
The spare breaker will appear functionally identical with the existing breakers to the plant operators.
Based on the above review of the effects on safety, the proposed use of the above described spare breaker as a lighting substatien feeder breaker is considered safe.
DBP 5302FFF/20
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i SAFETY EVALUATION
SUMMARY
FOR SE 93-0053 i
i TITLE:
j Changing Shift Supervisor Manning Requirements CHANGE:
1 Delete the position of Shift Supervisor's Administrative Assistant.
~
REASON FOR CHANGE:
j In order to facilitate reorganization of the Operations Section, DB-OP-00000,
- Conduct of Operations,.will be changed to delete Section 6.4.6 Shift Supervisor's Administrative Assistant Manning. Most activities currently being performed by the Shift Supervisor's Administrative Assistant (SSAA) will continue to be performed by people in positions other than that of Shift Supervisor.
SAFETY EVALUATION
SUMMARY
The changes to DB-0P-00000 have no effect on any structures, systems, and components or their associated safety functions.
The change is administrative in nature and does not affect the operation of any' plant system.
The change to
' DB-0P-00000 deletes the requirement to assign an SSAA to assist the Shift Supervisor in handling administrative duties on day shift seven days a week.
The deletion of the SSAA. requirement is justified as the position of Shift Manager has also been created.
Some of the duties of the Shift Manager are.to perform administrative duties to allow the Shift Supervisor to direct attention to supervising unit operation and to support the Shif t Supervisor in matters pertaining directly to the plant' operation.
Procedure DB-OP-00200,. Shift Manager, delineates the responsibilities and training requirements of the Shift Manager.
The redistribution of activities previously performed by the SSAA will not create more burden for the Shif t Supervisor nor detract from his ability to effectively. maintain control of the facility.
DBP 5302DDDD/1
SAFETY EVALUATION SUM 1ARY FOR SCR 94-5006 (SE 94-0074)
TITLE:
Reduction of Emergency Diesel Generator Jacket Water Low Temperature Alarm Setpoint CHANGE:
Changed the low jacket water temperature alarm setpoint from 100F to 1000F.
Revised USAR Section 8.3.1.1.4.1 by combining two paragraphs written on separate pages.
Both the paragraphs describe the same design feature in minor details of jacket water and lube oil heating subsystem for Emergency Qiesel Generator (EDG) fast start.
This change also deleted the heater capacity (15 KW) and the low jacket water temperature alarm setpoint data from the description.
REASON FOR CHANGE:
Decreasing the low Jacket water tempecature alarm setpoint eliminates nuisance alarms.
Revising the USAR text consolidates information which will help clarify the information presented in the USAR.
SAFETY EVALUATION
SUMMARY
The alarm setpoint should be high enough to ensure that the oil temperature does not go below the minimum prior to the alarm being received. Also the setpoint should be low enough not to cause spurious alarms during heater ON-OFF cycling. A range of values may satisfy this requirement.
Setpoint of 1000F provides enough margin to ensure oil temperature to be above 850F.
Therefore, deletion of this data from the USAR does not alter the main function of the system, and does not affect the plant safety.
The heater capacity of 15 KW is a design detail and is not a limiting design parameter.
It should be sufficient enough to keep the oil and the jacket water at an adequate temperature as suggested by the engine vendor.
The oil temperature is maintained at or above 850F. A slightly varying KW capacity may change the ON-OFF cycle duration and frequency but still be enough to maintain the temperature.
Therefore, the KW capacity of the heater as a design data is not critical to the plant safety.
Therefore, deletion of this data does no.
alter in any way the basic system function. and does not affect the plant safety.
DBP 5302FFF/31
h SAFETY EVALUATION
SUMMARY
FOR SE 94-0017, Rol J
TITLE:
Outside Storage of Radioactive Material CHANGE:
This evaluation provides technical justification for storage of radioactive materials (RAM) outside of the station generating facilities.
REASON FOR CHANGE:
I To allow for storage of RAM outside of the station generating facilities.
l SAFETY EVALUATION
SUMMARY
i RAM evaluated for outside storage includes only contaminated tools and materials for reuse during outage and maintenance activities.
Scenarios which could result in a potential does to the public involve the release of radioactivity from the container.
This most restrictive scenario would be fire.
l Storage of tools, maintenance equipment, lead, powdered resin from condensate polishing demineralizers and Dry Active Waste (DAW) was evaluated for the radiological consequences of a fire.
The conclusion of the evaluation was that
]
the most restrictive material was DAW.
1 i
l The resulting dose rate and highest annual organ dose were 259 mrem / year and
]
4.39 mrem respectively.
This is significantly below a small fraction (10%) of l
that specified in 10 CFR 100.
The calculated values are within 10 CFR 50 Appendix I limits of 1500 mrem / year for radionuclides in particulate form with half-lives greater than 8 days, and 15 mrem to any organ per year.
In summary, it is demonstrated that the resultant dose to a member of the public from a release of activity from stored RAM is a small fraction of the l
applicable limits established in 10 CFR 100.
In order to further minimize the rick and ensure proper controls are in place, activities involving the storage of RAM outside the Station Generating Facility will be subject to the following controls:
1.
All RAM stored outside of the Protected Area will be specifically evaluated and documented for suitability and acceptability.
2.
No corrosive radioactive material or radioactive materials that could
)
experience spontaneous exothermic chemical reactions may be stored outside of the generating facility.
3.
Quarterly inspections will be performed on all outside stored RAM.
Based on the above, the proposed activity is considered safe.
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DBP 5302FFF/32 U
SAFETY EVALUATION
SUMMARY
FOR DB-OP-00016 (SE 94-0041)
TITLE:
Line-up of CD75, CD188, and CD189 CHANGE:
Revised USAR Figure 10.4-11 and system drawings to show condensate valves CD74, CD75, CD188 and CD189 as closed, 4
REASON FOR CHANGE:
Procedure DB-OP-00016, Removal and Restoration of Station Equipment, requires that Safety Reviews be performed on all temporary line-ups greater than 6 months in duration.
The Safety Review performed on the line-up that isolates the Hotwell level control system for Chemistry control purposes, identified this line-up as a change to the facility as described in the USAR.
SAFETY EVALUATION
SUMMARY
USAR Figure 10.4-11 currently show CD74, CD75, CD188, and CD189 as open and providing level control using CD550A and B for the Hotwell.
These valves are closed to prevent undesired transfer of High Oxygen content water to the hotwell or condenser chemicals to the Condensate Storage Tanks (CSTs).
The proposed action of leaving the isolation valves closed reduces the risk of loss of the Condensate Storage Tank (CST) water due to pipe breaks, seismic events, or any other condition that could affect the hotwell.
Therefore the proposed action has no effect on safety.
Closing these valves will improve the reliability of the CSTs since.any malfunction of the hotwell level control valves will have no effect on CST volume.
DBP 5302FFF/12
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SAFETY EVALUATION
SUMMARY
FOR SE 94-0048 TITLE:
ICS Loss-Of-Main Feedwater-Pump Runback Limit CHANGE:
Justify unit operation at above 55 percent power with only one Main Feedwater Pump (HFP) available.
REASON FOR CHANGE:
On some occasions, it becomes necessary to remove one MFP from service for scheduled or emergent maintenance. Under this condition, it is desired to operate the unit at a power level which can safely be accommodated by a single HFP.
One MFP has sufficient capacity to support operation at 65 percent reactor power, while staying within rated speed and horsepower.
SAFETY EVALUATION
SUMMARY
Raising the allowable power level above the percent 55 percent reactor power for operation with only one MFW pump requires consideration of the allowable loading on the MFW pumps / turbines, and flow characteristics of the main feedwater system.
If the ICS runback setpoint is modified, it is of practical necessity'to ensure that no unacceptable adverse effects will result from.
termination of the runback at a higher power level. Alternatively, if the ICS runback function is defeated, then the method of defeating the interlock must be evaluated.
Equipment Capacity:
The turbine vendor indicated.that the last stage blades limit the rated turbine horsepower and speed to the nameplate values with a condenser vacuum of 3"HgA.
Degraded condenser vacuum of up to 5.5"HgA would not have an immediate effect, but might slightly decrease the longevity of the turbine.
Within the rated load and speed, analyses indicate that a power level of 65 percent reactor power is readily attainable.
Both the main and booster feedwater pumps are rated at 15,000 gpm.
Operation at 65 percent reactor power requires flows of almost 15,700 gpm.
While 15,700 gpm is slightly higher than the " rated" pump flow, it remains on both pump curves.
With one main feedwater pump operating, flow in the discharge piping of the MFW pump will be higher than at normal 100 percent power with two MFW pumps operating.
The increase in piping erosion will be proportional to the 1
increase in flow velocity.
Therefore, due to the relatively low percentage of lifetime operation which is anticipated in this condition, there should be no adverse increase in wall thinning of the affected piping. When DBP 5302FFF/57
operating at increased power levels on one MFW pump, the potential increase in erosion levels will be factored into the existing piping erosion prediction program.
Based on the above considerations, it is concluded that the main feedwater system can safely supply sufficient water to support 65 percent reactor power. Additionally, if practical special monitoring of the operating Main and Booster feedwater pumps and turbine will be conducted shortly after operation this condition is commenced.
Plant response to an ICS loss of MFW pump setpoint change from 55 percent to 65 percent:
With the unit operating in full automatic, the ICS attempts to control power at the operator set megawatt demand.
If a main feedwater pump is tripped from a high power level, a runback in megawatt demand at 20 percent per minute is initiated.
When generated megawatts decrease below the runback limit (currently 55 percent), a bistable in the loss of main feedwater pump runback circuitry changes state to terminate the runback.
Past simulations indicate that the remaining MFW pump increases speed and flow very quickly in an attempt to maintain total feedwater flow as close as possible to demand, but the flow capability is limited by available MFW pump capacity.
The flow / power mismatch continues, producing a rapid increase in RCS pressure until the reactor runback is able to re-establish a balance between power and flow.
During the runback, reactor to feedwater cross limits will attempt to hold feedwater flow as high as possible.
This is important because the reactor power runback can not generally follow the full rate of demanded runback. Without the cross limits, the feedwater flow demand reduction could outpace the rate of reactor power decrease.
However, the reactor to feedwater cross limits has a fairly large dead band. As time progresses in the transient, saturated integrals in the main feedwater pump controls will unwind, follow the decreasing demand signal, and allow the pump speed to decrease.
If the runback is terminated earlier (i.e., at 65 percent), the feedwater target runback will end sooner.
Thus, the reactor would potentially " catch-up" with feedwater flow earlier.
Based on the above, although the effect will be small, the chances of a successful runback from high power should slightly improve if the runback setpoint is increased to 65 percent.
Plant Response to Temporary Deactivation of the ICS Loss of MFW Pump Runback:
To prevent the ICS from detecting a MFW pump trip, fuse 4-0-F2 (4-9-FS) may be pulled for MFFT-B (MFPT-A).
This will prevent relays from energizing.
The contacts from these relays go to the rapid feedvater reduction (RFR) circuits, the feedwater pump hand / auto transfer circuits, and to the maximum load limit selection (runback) circuits.
The latter contacts will deactivate the ICS loss of main feedwater pump runback function.
Since this action would be performed below 65 percent, there would be no further need for the runback circuit until the unit is restored to twe MFW pump operation and power is escalated.
DBP 5302FFF/58
i If one of the above fuses is pulled, the RFR circuit will not be capable of detecting when the associated main feedwater pump is tripped.
Thus it will initiate RFR on a reactor trip if the remaining MFP trips.
However, plant response will not be altered since no main feedwater would be available, and since the Anticipatory Reactor Trip System (ARTS) will sense the loss of MFW pumps and the Steam and Feedwater Rupture Control System (SFRCS) will actuate auxiliary feedwater as designed.
If a reactor trip occurs with one MFW pump running, RFR will function normally.
j The feedpump hand / auto circuitry responds to a pump trip by transferring the hand / auto station to manual and running the ICS demand for that pump to its ICS low speed stop.
The proposed actions of pulling fuses 4-0-F2 or 4-0-F5 would prevent this function, but would have no impact on plant operation.
With the current 55 percent setpoint and with active runback circuitry, margin is available to cope with plant upsets which might cause an increase in flow demand.
Actions shall be taken to limit steady state feedwater pump demand to less than or equal to 65 percent when operating with a single main feedwater pump. With such precautions, based on engineering judgment, the probability of a main feedwater pump malfunction is not increased.
j i
Based on the above, there is no adverse nuclear safety impact if the plant is operated at a power level of up to 65 percent reactor power (by heat balance) on one MFW pump.
Furthermore, either of the two described methods of addressing the ICS loss of MFW pump runback circuit are safe.
DBP 5302FFF/59 l
1
1 SAFETY EVALUATION
SUMMARY
FOR SE 94-0061 TITLE:
Cycle 9 Licensed Length Extension and Addendum to the Reload Report CHANGE:
This Addendum appends the original Cycle 9 Reload Report, and addresses evaluations and re-analyses to allow operation beyond the initial 500 EFPD licensed length to a maximum of 505 EPPD.
REASON FOR CHANGE:
The length of Cycle 9 was extended due to Corporate outage requirements.
SAFETY EVALUATION
SUMMARY
The Cycle 9 depletion and design analyses that were generated for the original Cycle 9 licensed length were performed with the NEMO code. The analyses were extended to the new cycle length of 505 EFPDs. The resulting power distribution and fuel assembly / pin burnup distributions were inputs to the mechanical fuel assembly and fuel rod evaluations.
All batches, except for Batch 9B, were evaluated as still meeting design criteria with the cycle length extended to 505 EFPD.
Batch 9B, which originally had been analyzed with the TACO 2 computer code to verify that internal pin pressures would not exceed system pressure, required batch-specific reanalysis with the never TACO 3 code.
Maximum fuel assembly burnuns for all batches were all below the calculated creep collapse burnups.
The extension of Cycle 9 by 5 EPPDs increased the fast' fluence on Control Rod Assembly (CRA) C038 by an !nsignificant amount.
C038 had been scheduled for i
replacement by an Extended Life Control Rod Assembly (ELCRA) since it vould have approached its design life limit. A specific analysis was performed during 8RF0 to justify its continued use during Cycle 9.
A review of that analysis indicates that the acceptance criterion of <0.25% cladding strain vould still be met due to the insignificant higher fluence caused by the Cycle 9 extension to 505 EFPDs.
The nuclear analyses for the extended cycle consisted of calculations and/or evaluation of the following physics parameters:
total, stuck and ejected rod worths, Shut Down Margin, Hot Zero power (HZP) Moderator Temperature Coefficient (MTC), and Hot Full Power (HFP) HTC. There were insignificant effects on the total, stuck and ejected rod worths.
Therefore, due to the conservatisms already included, there are no changes required to the COLR's figures for control rod group position limits.
The minimum shutdown margin (SDM) required is 1% 6 K/K.
At 505 EPPD, i.e.
end-of-cycle 9 (EOC 9), the SDM is projected to be > 1.0 % 6 K/K, and the acceptance criterion is satisfied.
The higher core exposure at 505 EPPD causes the HFP HTC value to become more negative by approximately 0.02 x 10-2 % a K/K/ F.
The corresponding HZP Temperature coefficient pertigent to the Main Steam Line Break (MSLB) Accident 0
was calculated as -2.62 x 10- %
It is still vell below the value assumedintheUSARof-3.1x10pK/K/F.
%K/K/ F. Therefore, for the extended cycle 9, there still vill be a margin of approximately 0.5 x 10-2 % A K/K/ F, which means that the present USAR value is bounding.
Evaluation of accidents other than Loss-Of-Coolant-Accidents (LOCAs) for the Cycle 9 extension included the calculation of the minimum required BWST boron concentration, effect of revised physics parameters with respect to safety analysis inputs, and verification of RPS power / imbalance / flow or Flux /6 Flux / Flow trip setpoints and Protective Limits. There were no effects on previously determined values for Cycle 9.
Haneuvering analyses were also evaluated to verify that the core Protective and Operating Limits vould accommodate operation of Cycle 9 beyond 500 EFPDs.
No changes were necessary in the power distribution COLR limits or LCOs for control rod group positions (rod insertion limits), Axial Power Imbalance or Quadrant Power Tilt Limits.
Extended Cycle 9 analyses also evaluated the effects on off-site dose consequences (thyroid and whole-body) for each Design Basis Accident (DBA).
The extension of the licensed length did not have any effect on the original cycle's evaluation. All of the DBAs are still bounding.
Based on this evaluation of the effects on safety, the proposed action (i.e.
the operation of Cycle 9 to its extended licensed length of 505 EFPDs, in accordance with TS and maintaining the present COLR figures and tables has been determined to be safe.
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SAFETY EVALUATION
SUMMARY
FOR-SE 94-0062 TITLE:
Testing Main Steam Safety Valves during Modes 1 and 3 CIIANGE :
The setpoints of Main Steam Safety Valves (MSSVs) are verified while in place on the main steam lines using a Dresser / Consolidated hydraulic lift assist (Hydroset) device. While this testing has been performed under hot standby (mode 3) conditions in the past, it is desired to also perforte the test under power (mode 1) conditions.
REASON FOR CHANGE:
In order to reduce the time spent in Mode 3, it is desirable to test the MSSVs in Mode 1.
SAFETY EVALUATION
SUMMARY
A momentaty valve lift in mode 3 would have a negligible effect.
- However, because reactor thermal output is low, the plant will undergo a substantially larger plant transient if an MSSV should fail open in mode 3 than in mode 1.
Mode 3 testing conditions would most likely and ideally include steam generator inventory at low level limits. With a failed open MSSV, the transient should proceed as the USAR Class 1 (i.e.. no radioactive release at the exclusion area boundary) " Excessive Load Increase" event which is described in USAR chapter 15.2.11.
At worst, a boil-off of steam generator inventory and the associated cooldown would bounded by USAR chapter 15.4, Main Steam Line Break analysis. Existing procedural guidance provided for mitigation of steam leaks is appropriate.
The existence of a failed valve would be known by both reports from test personnel and audible noise.
Because the operators would be aware of the testing in progress and because the faulted valve would be known, operators would expeditiously perform the prescribed procedural actions.
Sufficient shutdown margin would be available to offset positive reactivity insertion due to the relatively small cooldown.
1 The above discussion indicates that there is a small potential for an MSSV to lift during the testing of the valve. However, because the valve would operate at or below normal design pressure with no damaging imposed load, the valve should close without malfunction. As in mode 3, a momentary lift during mode 1 would have a negligible impact on plant operation.
The transient should, at i
worst, proceed as the USAR Class 1 excessive load increase accident which is described in USAR chapter 15.2.11.
However, an automatic reactor trip and/or 1
RCS cooldown would normally not occur unless initiated by the operator.
In f
general, the minor transient induced by a failed open safety valve would be i
even less severe in mode 1 than in mode 3.
I DBP 5302FFF/21
?
SAFETY EVALUATION FUMMARY FOR SE 94-0067 TITLE:
Modify Operational Restrictions on Spent Fuel Pool Emergency Ventilation System CHANGE:
Evaluate modifications to procedural restrictions that require the containment purge supply and exhaust system to be isolated if the containment hatch is open and fuel handling or heavy loads over the spent fuel pool is to take place.
The restriction would no longer apply in MODE 6 if the SFAS containment radiation detectors function is disabled, REASON FOR CHANGE:
The above restriction was intended to prevent a potentially inappropriate actuation of the SFAS containment radiation monitors following a postulated radiation release in the spent fuel pool area.
This would result in SFAS realigning the EVS from the spent fuel pool area to the containment negative pressure boundary.
This realignment of EVS would not allow a negative pressure boundary to be maintained in the spent fuel pool area and could result in an unfiltered radioactive release.
SAFETY EVALUATION
SUMMARY
License Amendment No. 186 allows the use of containment purge and exhaust system noble gas monitor RE5052C to mitigate the consequences of a fuel handling accident inside the containment in MODE 6.
When using the noble gas monitor, SFAS is not required to be operable in MODE 6.
Therefore, if the function of the containment SFAS radiation monitors is disabled in MODE 6, the possibility of their actuation and the consequent shifting of the EVS to the negative pressure boundary is eliminated.
Since RE5052C is actuated by an air sample drawn from the containment purge exhaust, its actuation by radiation from outside containment is judged as not credible.
Once the SFAS area radiation monitors (if located inside the containment) are de-energized or the SFAS level 1 is bypassed, the concern identified above does not exist.
j If the SFAS radiation monitors are located in the shield building annulus, they will not be actuated by a radioactive release in the spent fuel pool area.
However, if the SFAS radiation monitors are being used inside containment, and the containment equipment hatch is open, SFAS level 1 shall be bypassed and/or the radiation monitor function disabled to alleviate the concern on realignment of EVS.
With is procedural restriction, fuel or heavy loads can be moved over l
the fuel pool with the equipment hatch open and the containment purge and exhaust operating.
Anytime core alternations are being conducted or irradiated fuel is being moved within the containment, the containment equipment hatch must be closed per Technical Specification.
Therefore, bypassing SFAS Level 1 and/or de-energizing the radiation monitors when the equipment hatch is in an open position is not a concern.
DBP 5302FFF/49
w
)
g
' SAFETY EVALUATION
SUMMARY
FOR' SE 94-0071 TITLE:
Cycle 10 Reload and Core Operating Limits Reports.
CHANGE:
Cycle 10 Reload REASON FOR CHANGE:
The Cycle 10 core loading consists of the following: 1) 5 batch 9 fuel-assemblies (FAs) of the Mk-B8A design that will include one repaired batch 9 assembly; 2) 48 batch 10 FAs of the Mk-B8A type;
SAFETY EVALUATION
SUMMARY
There have been no facility modifications that have affected the Cycle 9 RPS setpoints, so it is concluded that maintaining the same setpoints for Cycle 10 would have no effect on safety.
Another change for Cycle 10 is the implementation of Statistical Core Design (SCD). That methodology was approved by the Nuclear Regulatory Commission (NRC) in March of 1994. To protect the core, a Statistical Design Limit (SDL) with a Minimum Departure of Nucleate Boiling Ratio (MDNBR).of 1.313 was determined for the " hot" or limiting fuel rod. The SDL value of 1.313 was approved by the NRC.
Since SCD methodology allows higher peaking, special analyses were performed to determine the maximum allowable design (Radial x Local, FNH) peaking factor, such that the present Variable Low Pressure Trip Setpoint could be maintained.
Those analyses resulted in an allowable increase of the design FNH from 1.714 to 1.795.
The control rod group designations for Cycle 10, as reflected in the design analyses by BWFC, also remain unchanged from those of Cycle 9.
Twelve CRAs 4
were replaced with twelve ELCRAs, to add to the eight installed during 8 RF0, bringing the total to 20.
CRA C-038 vas replaced during 9 RF0 by C-012, a re-insert, which had been justified for an additional cycle of irradiation.
There have been no significant operating anomalies during Cycle 9 vhich would affect safety or fuel performance during Cycle 10.
The new core pattern also minimizes cross-core shuffling so that Cycle 9's Quadrant Fover Tilts vill not be amplified.
In Cycle 9, there was an indication of several fuel defects.
During the ninth i
refueling outage (9RFO), every Cycle 9 FA was ultrasonically tested.
One fuel rod was found defective in each of three FAs.
Two of the FAs had been scheduled for Cycle 10 operation.
Appropriate substitutes were available from discharged assemblies with minimal impact on the Cycle 10 Effective Full Fover Life and Reload Report analyses.
There is no current evidence of a generic fuel failure mechanism that would lead to fuel failures during Cycle 10.
For Cycle 10, with the assemblies substituted during 9RF0, BVFC also evaluated the effects cf operation of one FA 53L) that had been repaired during the spring of 1994 by recaging.
It contains two stainless steel (SS) rods in place of two defective fuel rods. The analyses, with NRC approved methodology, addressed the effect of a slightly increased amount of local (in-assembly) peaking. The maximum increase in the peak pin power throughout Cycle 10's operation is less than 0.1 %.
BVFC has also evaluated the impact of the reuse of the repaired assembly from a Loss-Of-Coolant-Accident (LOCA) analysis standpoint, with the conclusion that the Cycle 10 core is insensitive to the presence of the stainless steel pins in that assembly.
One manufacturing problem persisted during fabrication of batch 12.
An occasional veld undercut during the endcap velding process may have a potential effect on fuel reliability.
An Engineering Report documents that there is no effect on safety but, based on statistics to-date, there is a possible fuel reliability challenge for about two rods per reload batch.
The Mk-B9A fuel rod initial pre-pressurization was maintained at 300 psia. That initial rod pressure was factored into the cladding creep collapse analysis with the NRC approved CROV 7 code, which resulted in an adequate collapse burnup of >55,000 MVD/HTU.
The maximum expected rod burnup in batch 10 is projected to be 53,850 MVD/HTU.
The acceptability of the high rod exposures was independently verified with Davis-Besse fuel analysis methodology by use of the NRC approved ESCORE fuel rod modeling code as well as by performance of Quality Assurance fabrication audits and technical reviews of Reload Report analyses.
Although the Cycle 10 loading was subsequently changed, many of those calculations were evaluated as being bounding and applicable.
As stated before, the B9A fuel rods now reside in a Mark B10A cage. The Mark B10A cage is different from the previous Mark B types.
The performance of Mark B10 Lead Assemblies at the Oconee 1 Unit has been monitored and was documented in BAV-2186, "Foolside Examination of four Mark-B10 Lead Assemblies from Oconee Unit 1, Cycle 14-First FIE Report".
Overall there have been no anomalies reported in fit, form and function.
All of the mechanical design changes were evaluated as not having an effect on the FA's form loss coefficients.
From Thermal-Hydraulics and LOCA reflood standpoints the Mark B10 FA has been treated the same as a Mark B8 design.
USAR Chapter 15 accidents other than LOCA vere also reviewed or re-done because of the previously discussed higher allovable design (Radial x Local) peaking factor of 1.795.
A re-analysis of the Control Rod Ejection Accident (CREA) was performed using USAR methodology inputs and verified with nev (3-D) methodology, by use of the RELAP5 code. The RELAP5 methodology also was within 3% of the USAR's old KAPP code methodology in calculating the resulting effective Full-Power-Seconds (FPS) for the CREA.
Using the acceptance criterion of a fuel enthalpy of 210 cal /gm, the maximum total peak allowed would be 3.43.
That is higher than the design total peak of 2.96, which is the product of the new maximum allovable design Radial x Local peak of 1.795 and the design chopped cosine axial shape peak of 1.65.
Therefore, the CREA re-analysis bounds the design peak assumption.
BVFC also verified that the 45% of fuel rod failures originally calculated as the consequence of the CREA and assumed in the offsite dose calculation is bounding.
All of the Cycle 10 nuclear design analyses were performed with the NRC approved NEM0 code.
BVFC Reload Report analyses were independently reviewed and some parameters were verified for reasone/bleness and acceptability with Davis-Besse reactor physics methodology.
The core loading for Cycle 10, like Cycle 9, is a Very-Low-Leakage (VLL) pattern. The implementation of the VLL reload shuffle scheme for Cycle 10 is expected to have an insignificant impact during all aspects of reactor startup and power operation. An NI re-calibration is performed above 15% of Full Power based on a primary system heat balance. This vill validate the setting of the high flux trip setpoint prior to the continuance of the power escalation.
Because of the longer fuel cycle design of Cycle 10, the core has substantial excess reactivity.
Increased excess reactivity requires a higher Refueling Boron Concentration. The Refueling Boron Concentration analysis includes the f
assumption of misloading of the most reactive FA and up to two out-of-position CRAs.
BVFC, also evaluated Makeup and Purification Malfunctions during refueling operations, since the Boron Dilution Accident (BDA) has a greater effect on Shut Down Margin (SDM) reduction.
The safety analysis criterion requires a minimum SDM of 1% 6 k/k post-BDA, and the SDM was verified to exceed that.
During the Cycle 10 operating cycle, the SDM required is 1% a k/k. That requirement is ensured by compliance with -the rod insertion (Group Rod Position Alarm Setpoints) figures in the COLR. At 520 EFPD, i.e. end-of-Cycle 10, the minimum SDM is projected to be 1.61 % 6 k/k.
The Moderator Temperature Coefficient limit (at Rated Thermal Power) for Cycle 10, gs referenced by TS 3.1.1.3c and specified in the COLR (Table 2) is -3.73 x 10 ' 6 k/k/oF. This is based on the Main Steam Line Break (MSLB) accident analysis, and the HTC limit is a fuel cycle design The predicted end-of-Cycle 10 MTC is predicted to be -3.38 x 10 gonstraint.
6 k/k/'F, which ensures the MSLB accident analysis remains bounding.
Based on this evaluation of the effects on safety, the proposed action (i.e.
implementation of the COLR) and operation of Cycle 10 has been determined to be safe.
a
4 SAFETY EVALUATION
SUMMARY
j FOR SE 95-0006 TITLE:
Combining Maintenance Superintendent Responsibilities CHANGE:
In order to facilitate the reorganization of the Maintenance Section, USAR Section 13, was changed to combine the responsibilities of the " Superintendent
- Maintenance Services" with those of the " Superintendent - Mechanical Maintenance." The responsibilities of the " Superintendent - Electrical Maintenance" were combined with those of the " Superintendent - Instrumentation and control." This results in a single superintendent over the " Mechanical Maintenance Unit" and a single superintendent over the " Electrical and Contro}s Maintenanc'e Unit."
REASON FOR CHANGE:
The combining of the four maintenance units into two is justified as the newly formed units will be able to complete maintenance activities in a more efficient and timely manner due to the reduced number of inter departmental interfaces required to complete the activity.
The two maintenance unit concept will also allow a greater flexibility in the use of manpower within the unit.
SAFETY EVALUATION
SUMMARY
The proposed change to USAR Section 13, has no effect on any structures, systems, and components or their associated safety functions.
The proposed change is administrative in nature, and does not effect the operation of any plant system.
The proposed change to the USAR Section 13, will allow the formal integration of the four maintenance units into two.
The increase in the superintendents work load will be minimal as the combined units will be able to handle the work activities in a more efficient fashion therefore, reducing the superintendents burden of tracking numerous work activities between units.
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SAFETY EVALUATION
SUMMARY
FOR SE 95-0011 TITLE:
Reorganization of Planning, Scheduling, and Chemistry Responsibilities CHANGE:
The responsibilities of the Planning and Scheduling Section;were divided. The Planning activities report to the Maintenance Section while the Scheduling activities report to the Operations Section.
In addition, the Chemistry fur.ctions nov report to the Radiation Protection Section instead of the Operations Section.
REASON FOR CHANGE:
These organizational changes realign the sections into a more effective organization and improve communications, work flow and effectiveness.
j SAFETY EVALUATION
SUMMARY
The organization change has no effect on any structure, system and component or their associated safety functions. The change is administrative in nature and does not affect the operation of any plant system. The change allows a more efficient operation by reducing the number of inter-departmental interfaces required to complete activities.
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SAFETY EVALUATION
SUMMARY
FOR TH 93-0030 (SE 93-0062) i TITLE:
Temporary Smoke Detectors In Room 114 CHANGE:
Temporarily install three Simplex smoke detectors in Room 114 in place of the three PDS detectors currently installed in the room.
Failure of any or all of the temporary Simplex detectors in Room 114 vill not prevent the detectors in Room 115 from responding properly.
REASON FOR CHANCE:
Due to NRC concerns with the operability of the Thermo-Lag fire barrier in Room 114 (Hisc. Vaste Monitor Tank and Pump Room), implementation of compensatory measures is required.
In order to avoid a continuous fire vatch, three Portable Detection System (PDS) detectors were installed and an hourly fire watch established. The temporary detection and the hourly fire watch are necessary until the Thermo-Lag fire barrier concerns are resolved.
In order to eliminate the high cost of maintaining the PDS, it has been determined that the existing Simplex detection system in Room 115 (ECCS Pump Room 1-2) could be extended into Room 114 until the Thermo-Lag concerns are resolved.
SAFETY EVALUATION
SUMMARY
The temporary Simplex smoke detectors perform the same function as the Portable Detection System. The Simplex smoke detectors in conjunction with a roving fire watch provides the appropriate compensatory measures as required by the FHAR.
Modification of the Simplex software for Panels C57968 and C2720 to reflect the addition of the three temporary smoke detectors does not affect the functions of these panels. The viring for the temporary smoke detectors are connected to the Simplex system at detector DS8694H in Room 115. This does not affect the function of detector DS8694H. The viring passes through an existing negative pressure foam seal in the non-rated vall between Rooms 114 and 115.
This routing does not affect any of the other equipment installed in either of the two rooms, a
I SAFETY EVALUATION
SUMMARY
FOR UCN 87-012 (SE 93-0016, R01)
TITLE:
Detergent Waste Drain Tank Pump Classification CHANGE:
Revise USAR Table 11.2-2 to reflect the fact that the design code for the Detergent Waste Drain Tank (DVDT) Pump has been changed from ASME Section III Class 3 to " Manufacturer's Standard" (i.e., conformance to the ASME Code is not required).
REASON FOR CHANGE:
The change from ASME Section III Class 3 to a "non code" pump is consistent with the change to the design basis of the Miscellaneous Liquid Radwaste System from ASME Section III Class 3 to ANSI B31.1 (augmented)
SAFETY EVALUATION
SUMMARY
Use of a "non code" pump will not adversely affect the function of the Miscellaneous Liquid Radwaste System.
The manufacturer's standard pump provides adequate flow at design head loss and is suitable for the liquid radwaste service.
The pump aids only in transferring waste and does not alter the provisions for controlling releases of radioactive liquids.
The radiological controls in this system have been designed and installed in accordance with the criteria in 10 CFR 20 and 10 CFR 50.
Use of manufacturer's standard pump does not change Davis-Besse's conformance to these criteria.
The DVDT pump handles liquids with low levels of radioactivity.
No accident involving this pump would yield uncontrolled release to the environment.
The original pump was capable of supplying 140 gpm at 165 feet of head.
The replacement pump is capable of supplying 140 gpn at 340 feet of head and 330 gpm at 200 feet of head.
Flow is controlled by throttling the pump's recirculation line isolation valve (WM85) to provide the backpressure to achieve the 140 gpm f'.ow rate.
Therefore, the replacement pump is functionally equivalent to the original pump.
The system continues to meet the requirements of Regulatory Guide 1.143 and 1.21 and Standard Review Plan Section 11.5.
The use of a "non code" pump does not affect the size of any potential radiological releases.
Therefore, the limits described in the Offsite Dose Calculation Manual (0DCM) are not impacted.
DBP 5302DDDD/24
SAFETY EVALUATION
SUMMARY
FOR UCN 91-066 (SE 91-0074)
TITLE:
Deletion of Soldering as a Special Process CHANGE:
This UCN corrects contradictory language in the USAR definition of Special Processes.
REASON FOR CHANGE:
The only type of soldering that is done on quality related equipment is the creation of electrical connections.
The quality of an electrical connection can be determined by inspection or test; therefore, electrical soldering is not a special process.
SAFETY EVALUATION
SUMMARY
This change will not have any affect on safety, as the existing inspection and testing requirements for safety related electrical connections are unchanged, and remain sufficient to detect unacceptable soldered connections.
Structural or pipe soldering is not performed on equipment which is important to safety.
This change was approved by the NRC via Log 1-3058, dated July 21, 1994.
i DBP 5302DDDD/16
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?
SAFETY EVALUATION
SUMMARY
FOR UCN 93-039- (SE 95-0032)
TITLE:
I
^
Revision of USAR Section 7.13, Post Accident Monitoring System CHANGE:
Removed extraneous details or vendor specific details of Post Accident Monitoring Equipment and made revisions to Section 7.13 to better clarify the system.
REASON-FOR CHANGE:
USAR Section 7.13 was revised to make the description of the Post Accident Monitoring System (PAMS) more concise and to make it accurately reflect what is
.in the field.
SAFETY EVALUATION
SUMMARY
The information deleted from USAR section 7.13 includes references to vendor specific details of equipment design or operation. Also, references to redundant instrumentation were removed since PAMS is not required to have redundant monitoring of each parameter even though redundant trains (train 1
~
and train 2) are monitored.
Only written text in USAR section 7.13 vas changed. No hardware changes were i
made. This activity simplifies the USAR vithout removing required information.
The USAR Post Accident Monitoring System performance requirements are still being maintained in the USAR and the Post Accident Monitoring System still complies with Reg Guide 1.97, so there are no affects on safety.
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l SAFETY EVALUATION
SUMMARY
FOR UCN 93-059 (SE 93-0050)
TITLE:
Switchyard Battery Size CHANGE:
The Switchyard batteries which were rated at 320 Ampere-hours were replaced with 250 Ampere-hour batteries.
REASON FOR CHANGE:
Switchyard batteries BA and BB were replaced due to age related degradation.
SAFETY EVALUATION
SUMMARY
The calculated 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Ampere-hour battery size is 121 Ampere-hours for BA, and 209 Ampere-hours for BB.
The 250 Ampere-hour battery selected therefore includes an additional margin of 129 Ampere-hours at BA, and 41 Ampere-hours at BB.
The replacement batteries operate on the same principle as the original batteries, but have fewer plates per cell.
Therefore, the probability of failure of each cell is unchanged or reduced.
Since the number of cells is not increased, and the number of cell interconnections is not increased, the probability of battery failure will not be increased.
This change is beneath the level of detail of the Technical Specification bases, and is outside the seismically designated portion of the plant.
Since each battery is capable of supplying all of the 125 VDC requirements of the switchyard, safety will not be compromised during implementation, as long as the batteries are replaced sequentially, i
L DBP 5302DDDD/9
SAFETY EVALUATION
SUMMARY
FOR UCN 93-062 (SE 94-0016)
TITLE:
Change in the Distribution of External Supplier Audits CHANGE:
Removed the requirement for External Supplier Audits to be distributed to the Vice President - Nuclear.
REASON FOR CHANGE:
In an effort to improve the use of resources available at Davis-Besse it was determined that the distribution of the detailed external audit reports to the i
Vice President - Nuclear was not needed.
External audit reports are reviewed by the Company Nuclear Review Board (CNRB).
SAFETY EVALUATION
SUMMARY
b This proposed change regards removing the requirement for distribution of External Supplier Audits to the Vice President - Nuclear - Davis-Besse.
This is an administrative change and there is no affect on any structures, systems or components either directly or indirectly. This change was approved t
by the NRC via Log 1-3122, dated December 19, 1994.
l
SAFETY EVALUATION SUFD4ARY FOR UCN 93-074 (SE 93-0065)
TITLE:
Clarification of Emergency Conditions for Decay Heat Removal Cooler Characteristics CHANGE:
Clarify the " emergency conditions" for USAR Table 6.3-3,
" Decay Heat Removal Cooler Characteristics".
The SFAS level 4 accident condition is actually the bounding heat load case for the DHR Coolers, with an SFAS level 3 condition being less limiting.
REASON FOR CHANGE:
During a review of the USAR it was identified that USAR Table 6.3-3 needed clarification to state what the limiting emergency conditions are for the parameters listed in the table. Table 6.3-3 of the USAR lists the required CCW flow to a single DHR Cooler under " emergency conditions," for which an SFAS level 4 condition is the bounding case. An SFAS level 3 scenario, such as for a small break loss of coolant accident (LOCA), would actually require less CCW flow to the DHR Coolers at the time emergency sump recirculation would be required due to decreased Reactor Coolant System (RCS) heat loads.
SAFETY EVALUATION
SUMMARY
This clarification of emergency conditions is based on previous work performed which relles on the fact that SFAS level 3 scenarios would result in a lower heat load for the DHR Coolers at the time of sump recirculation.
RCS breaks which would only actuate SFAS levels 1, 2, or 3 would be relatively small breaks and would require more time to empty the BWST.
Under SFAS level 3 conditions, CCW loads inside containment would not have been automatically isolated, and therefore, the CCW flow to the DHR Cooler is less than that for SFAS level 4 by ~700 gpm.
For an SFAS level 4 actuation, all non-essential CCW loads are isolated and the DHR Cooler is supplied with the specified mass flow rate of 3,000,000 lbm/hr ('6,000 gpm).
Due to the larger injection flow for
)
breaks resulting in an SFAS level 4 actuation, sump recirculation is begun i
early at relatively high decay heat values.
For an accident condition not progressing beyond SFAS level 3, sump recirculation would not be required until approximately one hour following reactor shutdown, resulting in a 17% reduction in decay heat.
The worst case CCW flow through the DHR Cooler (given an SFAS level 3 condition) results only in approximately a 4% reduction in the value for heat removal capacity as specified in Table 6.3-3 of the USAR.
Therefore, this explanation implicitly allows for a lower flow under SFAS level 3 conditions.
Since the CCW system can provide adequate flow for SFAS level 4, which is more limiting, adequate flow for SFAS level 3 is assured.
It is determined by testing that the CCW flow through the DHR Coolers is adequate.
It is therefore concluded that this added explanation of the limiting case for emergency conditions in Table 6.3-3 of the USAR is safe.
DBF $302DDDD/10 l
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SAFETY EVALUATION
SUMMARY
FOR UCN 93-075 (SE 93-0066)
TITLE:
Effluent Radiation Monitor Setpoints CIIANGE:
Revise the text in USAR Section 11.4 describing the alarm values for process and effluent radiation monitors.
REASON FOR CHANGE:
The previous USAR text does not accurately reflect the setpoint values the radiation detectors are currently adjusted to.
SAFETY EVALUATION
SUMMARY
The proposed administrative changes do not affect any structures, systems or components important to safety.
The proposed text continues to prescribe the required dose controls which result in the dose to individual members of the public being As Low As Reasonably Achievable ("ALARA"), consistent with the regulatory requirements outlined in both 10 CFR 20 and 10 CFR 50.
DBP 5302FFF/30
I l
SAFETY EVALUATION
SUMMARY
FOR UCN 93-077 (SE 93-0072)
TITLE:
Generic Letter 89-10 Program Calculations CHANGE:
Update the USAR description for certain motor operated valves (MOVs).
Additional MOVs have been added to USAR Table 3.9-2, ASME Code Class 2 and 3 Components and Active Components and Valves, for completeness. Additional sections have been added where necessary to describe the analytical method used for these valves.
Calculated stresses, where previously provided, have been removed.
Finally, various editorial corrections have been made to Table 3.9-2.
REASON FOR CHANGE:
In support of Toledo Edison's Generic Letter 89-10 program, calculations were generated to determine the maximum allowable stem thrust to ensure component stresses remain within allovables during a seismic event.
USAR Section 3.9.2.9.2 and Table 3.9-3, Load Combinations and Stress Limits for ASME Code Class 2 and 3 Components, describe the original analytical method used for certain seismic Class I motor operated valves (MOVs) listed in Table 3.9-2.
This change updated the USAR descriptions.
SAFETY EVALUATION
SUMMARY
The original analyses for these Motor Operated Valves demonstrated that component stresses were within allovables based on a nominal stem thrust and a coincident seismic event. The revised analyses were performed to determine the maximum allowable stea thrust coincident with a seismic event. This does not I
increase the effects of a seismic event because the stresses are maintained within allowables. Thtrefore, the revised analyses have no effect on safety.
The USAR description of these analyses has been revised primarily by deleting specific analysis results.
In addition the table of valves has been expanded to included additional valves for completeness. These changes effect only the level of detail in this USAR section and have no effect on the equipment design or function. Therefore, these changes have no effect on safety.
)
The revised analyses are part of a program to optimize motor operated valve I
switch settings.
Determination of the maximum allowable stem thrust vill prevent exceeding equipment limits and vill therefore tend to increase the reliability of the affected equipment.
p 4
SAFETY EVALUATION
SUMMARY
.FOR UCN 93-079 (SE 93-0064)
TITLE:
- ECCS Room Cooler _ Required Service Water Flow CHANGE:
Modify USAR Table 9.2-1,
" Service Water System Design Parameters for Major Equipment", to provide'the manufacturer's design flow rates as nominal values and to allow lower flow rates to be acceptable as long as they are specifically-analyzed by approved calculations.
REASON FOR CHANGE:
The manufacturer's design flow rate is 140 gpm per cooler. The present revision to Table 9.2-1 provides a value of 100 gpm per cooler, which was the previously required flow based on Toledo Edison calculations. As identified in recent Service Water (SW) flow testing, the current SW flow rates for coolers E42-4 and E42-5 are below the specified 100 GPM of Table 9.2-1, but above the most recently analyzed acceptable values.
SAFETY EVALUATION
SUMMARY
It has been documented that the present measured flow rates to ECCS room coolers E42-4 and E42-5 with an SFAS actuation valve lineup are as. low as approximately 50 gpm per cooler. Observations in other piping sections have indicated that the flow reductions are predominantly due to biologically induced corrosion.
Flow trending programs and modifications to allow scheduled cleaning of affected piping sections have been instituted.
A calculation has evaluated service water flow rates of 45 gpm per ECCS room cooler for a single operating cooler and 30 gpm per cooler for two operating coolers.
The results indicate that for coolers which are not fouled on either the water or the air sides, a 45 gpm single cooler SW flow rate is acceptable
]
up to a SW temperature of 75.40F (with the fan on the inoperable cooler de-activated).
If 30 gpm is supplied to both ECCS room coolers in either ECCS room, acceptable heat removal for that room would be ensured up to a service water temperature of approximately 970F, which is well in excess of the technical specification limit.
With respect to potential fouling, it has been determined by external inspection that the heat exchanger fins are clean. Analysis contained in a calculation indicates that even with the reduced measured flow rates. the overall heat transfer coef ficient of clean heat exchangers in over 25 percent higher than the manufacturer's data sheet. An even larger margin is indicated by an analysis using the computer code AIRCOOL, which showed better than the vendor's heat transfer rate with as much as 50% of the heat transfer area blocked.
Finally, ECCS room cooler thermal performance tests, while containing a large uncertainty due to low heat loads under test conditions, have also provided indication that the coolers are performing normally.
Therefore, it is concluded that the present condition is safe and acceptable.
DBP 5302DDDD/5
SAFETY EVALUATION
SUMMARY
FOR UCN 94-007 (SE 94-0008)
TITLE:
Containment Air Cooler Service Water Outlet Valve Open with Fan Off CHANGE:
Revise USAR Section 6.2.2.2.1, and Table 9.2-1 to stipulate that during cold weather, one CAC may be operated with its service water control valve fully open and its fan not running.
REASON FOR CHANGE:
During cold weather with low service water (SW) temperature, the flow demand to SW loads is reduced, especially for the SW pump supplying primary loads.
The low flow has potential to cause an increase in SW pump discharge pressure which is high enough to lift the 6 inch service water relief valves, SW-3962 or SW-3963.
It is possible to stop the fan on a CAC with the SW outlet valve not modulating, but positioned fully open.
This will increase SW flow and not increase CTMT air cooling to excessive levels.
The other operating CAC will continue to maintain normal CTMT air temperature.
SAFETY EVALUATION
SUMMARY
If the CAC SW outlet control valve is fully opened on the SW loop aligned to primary (CCW) loads, past operation has indicated that the expected affected CAC flow will be approximately 2100 gpm.
The design CAC flow, in accordance with USAR Table 9.2-1 is 540 gpm (normal) or 1600 gpm (LOCA).
Thus, the anticipated flow will be higher than normal. However, the average tube velocity will be approximately 7.5 ft/sec at 2100 gpm.
This remains at a reasonable tube side velocity, which will not significantly increase tube erosion.
In addition, the CACs have been operated for prolonged periods with the discharge valve fully opened in the pasti no adverse effects with respect to tube wear have been found.
In order to meet design and accident analysis (LOCA) requirements, the CACs must respond to a Safety Features Actuation System (SFAS) level 2 trip.
If an SFAS level 2 actuation occurs, the fans on both train 1 and 2 CACs must trip, and then restart in low speed.
In addition, the SW outlet control valves must open.
This will occur regardless of whether or not the outlet SW control valve has been opened.
Because the service water control valve would already be in the failed position (fully open) the valve would not need to reposition in response to the SFAS signal.
Thus, accident response would not be affected.
The penetrations associated with the CACs must also meet the requirements of General Design Criteria (GDC) 57, for systems which are closed inside containment.
The CAC outlet SW control valves, SW-1356, SW-1357, and SW-1358, provide this function.
If the fan on any CAC is first stopped, then these valves may be closed from the control room.
This function is not affected by the subject USAR change as described.
l DBP 5302DDDD/6 1
i
During normal operation, containment air temperature is required by Technical Specification to be less than 1200F, based on the average of the inlet temperatures of the operating CACs.
The surveillance for this specification requires verification of this temperature at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With only one CAC fan in service, the temperature used to verify the surveillance will simply be the indicated temperature of the operating CAC.
It has been shown in past operation, that one CAC has sufficient capacity to maintain normal containment temperature during cold weather (Winter) conditions.
The dominant uncontrolled variable affecting CAC capability is SW temperature.
Because SW temperature chagnes relatively slowly, the technical specification surveillance will be sufficient to determine when two CAC operation must be resumed.
In order to meet design and accident analysis (LOCA) requirements, the CACs must respond to a Safety Features Actuation System (SFAS) level 2 trip.
If an SFAS level 2 actuation occurs, the fans on both train 1 and 2 CACs must trip, and then restart in low speed.
In addition, the SW outlet control valves must open.
This provides sufficient air flow through the cooler (without excessive motor load) in the relatively dense post-LOCA containment environment and provides maximum service water flow.
Elementary wiring diagrams show that if the fan is topped on a selected CAC by use of the control room switch, the fan will restart in low speed upon receipt of an SFAS level 2 signal.
This will occur regardless of whether or not the outlet SW control valve has been opened.
Because the service water control valve would already be in the failed position (fully open) the valve would not need to reposition in response to the SFAS signal.
Thus, accident response would not be affected.
As discussed above, the containment penetrations associated with the CACs are required to open in response to a LOCA or other containment pressurization transient.
However, the penetrations associated with the CACs must also meet the requirements of General design criteria (GDC) 57, for systems which are closed inside containment.
GDC 57 requires that a penetration isolation valve be capable of being remotely isolated.
The CAC outlet SW control valves, SW-1356, SW-1356, and SW-1358, provide this function.
If the fan on any CAC is first stopped, then these valves may be closed from the control room.
This function is not affected by the subject USAR change as described.
DBP 5302DDDD/9
SAFETY EVALUATION
SUMMARY
FOR UCN 94-029 (SE 94-0027)
TITLE:
Remove the Setpoint for the. Reactor Coolant Drain Tank Pressure Regulator CHANGE:
This USAR Change Notice removes the setpoint value for the nitrogen pressure regulator'to the Reactor Coolant Drain Tank (RCDT).
REASON FOR CHANGE:
A Potential Condition Adverse to Quality (PCAQ) identified that the setpoint for the nitrogen pressure regulator to the RCDT was incorrect.
Instead of revising the valve it was determined to remove the setpoint from the USAR.
SAFETY EVALUATION
SUMMARY
The RCDT continues to be supplied with nitrogen as required to perform its intended function. A previous safety evaluation' justified the isolation of PCV 1776 during normal operation.
No safety functions were affected by operating the RCDT with its nitrogen supply PCV 1776 normally. isolated and manually placed in service as required.
PCV 1776,.RCDT Nitrogen Supply Pressure Regulator, is the only pressure regulator listed in the USAR with'its setpoint identified.
Removal of the PCV 1776 setpoint value of 1.5 psig from the USAR will have tua effect on plant safety.
Plant setpoints are controlled by which lists the setpoint of PCV 1776 as 1.5 psig. Changes to the PCV 1776 setpoint requires a setpoint change request be submitted and reviewed.
This review directs a safety review and follow-on safety evaluation if required, be performed.
These reviews will ensure the setpoint of PCV 1776 is adequately controlled.
Removing the setpoint listing for PCV 1776 from the USAR will have no effect on plant safety.
The RCDT will continue to perform its intended function.
Therefore, this activity is considered safe.
DBP 5302DDDD/16
l SAFETY EVALUATION
SUMMARY
l FOR UCN 94-040 (SE.94-0026)
TITLE:
Zone-Loaded Fuel Storage CHANGE:
The zone loading for Batch 12 involves reducing the fuel enrichment in a total of 24 fuel rods, 5 in each corner of the fuel assembly and 4 immediately surrounding the center instrument tube.
REASON FOR CHANGE:
The purpose of this is to permit the storage of zone-loaded fuel assemblies (fuel assemblies containing fuel rods with multiple enrichments) in the new fuel storage racks (NFSR) and the spent fuel storage racks (SFSR).
SAFETY EVALUATION
SUMMARY
Since varying the enrichments in individual fuel rods within a fuel assembly does not change the mass of the fuel rods or the entire fuel assembly, the seismic functions provided by the NFSR and the SFSR will not be affected by zone-loading the fuel assemblies.
Additionally, fission product inventories are not significantly affected by initial enrichment.
Therefore, zone-loading has no significant effect upon the fission product inventories in the fuel assemblies.
Thus, the fuel handling accidents described in the USAR are still bounding.
The only postulated safety effect would be on criticality in the NFSR and SFSR.
For the NFSR, it had to be determined that a fuel assembly containing a uniform fuel enrichment of 5.0 weight-percent (wt!) uranium-235 (U-235) would bound any zone-loaded fuel assembly.
Calculations were performed for both flooded conditions and optimally moderated, or " mist", conditions.
These calculations clearly demonstrated a reduced reactivity for the zone-loaded cases versus the uniform-loaded cases, with reactivity trending lower as the low enrichment zone was reduced in enrichment.
These results are expected, since these cases merely involve the reduction of the total amount of fissile material in the problem.
Further, the trend clearly indicates that even larger enrichment
" splits" would be bounded.
Therefore, for the 24 rod zone loading proposed for Batch 12, the criticality analysis that currently forms the basis for the USAR remains bounding, and loading such assemblies, including Batch 12, in the NFSR is safe.
For the SFSR, fuel assemblies are normally divided into three categories, in accordance with Technical Specification (TS) Limiting Condition for Operation (LCO) 3.9.13.
Forty separate comparisons of reactivities between uniform enrichment assemblies and zone-loaded assemblies were made, using varying combinations of average enrichment, enrichment " split", and assembly burnup.
In all cases, the reactivity differences between the uniform assemblies and the zone-loaded assemblies were very small, with the largest positive difference being +0.00022 Ak.
This difference is sufficiently small that it is statistically insignificant.
Therefore, it can be concluded that, for 24 rod zone-loaded fuel assemblies of the type proposed for Batch 12, fuel assembly categorization in accordance with TS LCO 3.9.13 can be performed using the assembly-average enrichment (numerical average of low and high enrichments, weighted by the number of rods of each enrichment).
Thus, placing zone-loaded assemblies of this type, including Batch 12, in the SFSR is safe.
DBP 5302DDDD/22
~
~
I9 l
I x
SAFETY EVALUATION
SUMMARY
FOR UCN 94-082 (SE 94-0043) l TITLE:
Grout Penetration Seal Inspections CHANGE:
Support s'one time change'to the Operating Specification surveillance requirements found in the Fire Hazard Analysis Report (FHAR) for penetration seal inspections.
REASON FOR CHANCE:
It is proposed that the 10% seal inspection of grout seals (GFS-1).be halted with the 4th inspection group.
The only failures in the 3rd inspection group-were in barrier 304E/310W.
It was decided that if the only failures in the 4th inspection group were in the same barrier, a 100% inspection of'the barrier ~
would be performed.
The 4th inspection group had failures which were limited to the same barrier. A 100% barrier inspection was performed and-additional failures were found. Work orders have been written to fix all-identified deficiencies.
This evaluation is to stop the 10% sample continuation requirement of Operating Specifications 8.1.4 and 8.2.5, Surveillance Requirement C and supports the one time change to the FEAR to note that.this has occurred.
.j SAFETY EVALUATION SUHHARY:
t The intent of the 10% sample groups is to minimize the amount of inspection time required for each seal type but at the same time provide reasonable i
assurance of the status of the specific seal type (e.g., no generic degradation or other generic failure mechanism that is rendering the specific seal type inoperable).
If a failure is found in the first 10% sample group, additional 10% sample groups are inspected until either 100% of the specific seal types have been inspected or there is a sample group with no failures.- In this specific case, it was documented in PCAQs that after several sample groups the failures were limited to barrier 304E/310W indicating a problem with this barrier and not with the GFS seals in general.
Thus, the correct action is to perform a 100% barrier inspection of the identified barrier and terminate the increased inspections of other GFS-1 seals.
l i
i DBP 5302FFF/1 l
..,, ~,..-_
SAFETY EVALUATION SUMHARY l
FOR UCN 94-088 (SE 94-0036)
. TITLE:
~
Implementation of License Amendment 186-i CHANGE:
' Allows the use of containment purge and exhaust system noble gas radiation monitor RE5052C to mitigate the consequences of a fuel handling accident inside the containment. As this is to be done in lieu of requiring.SFAS to be OPERABLE in MODE 6, an evaluation was performed to ensure that other components which would have previously received SFAS level 1 actuation signals would be in a safe state given a postulated fuel handling accident.
REASON'FOR CHANGE:
License Amendment No. 186 permits the use of the containment purge and exhaust system noble gas monitor to terminate the release of radioactive gases during a postulated fuel handling accident inside the containment.
When using this noble gas monitor, SFAS is not required to be OPERABLE in MODE 6.-
Since the NRC license amendment safety evaluation did not specifically address other components potentially actuated by SFAS in MODE 6 (level 1 - high containment radiation), this safety evaluation is prepared to demonstrate that the required changes to the USAR and plant operating procedures to implement this amendment will not result in an unreviewed safety question.
SAFETY EVALUATION
SUMMARY
i Upon detection of high radiation, the RE5052C signal will initiate a plant computer alarm in the control room, trip the purge system supply and exhaust fans, open the bypass duct to EVS damper, and close various inlet and outlet dampers in the containment purge and exhaust system.
This effectively isolates the potential radioactivity leakage path.
In lieu of automatic containment, isolation via SFAS, operators would then manually isolate the containment purge and exhaust system containment isolation valves from the control room.
.i Other containment isolation valves that receive isolation signals from SFAS incident level 1 include containment air sample system valves CV5010A, B.-C, D,
E and CV5011A, B, C, D, E.
The penetrations associated with these valves do not meet the Technical Specification criteria of providing direct access from l
the containment atmosphere to the outside, as the containment air sample system is a closed system which returns sampled air directly back to the containment.
Administrative requirements exist to verify no work is being done on this.or
.other systems which would present a leakage path from the containment atmosphere during CORE ALTERATIONS or movement of irradiated fuel.
- Thus, i
isolation of these valves by SFAS level 1 in MODE 6 is not required.
Per Technical Specification OPERABILITY of EVS and shield building integrity is j
not required in MODES 5 and 6.
Further the radiation dose calculations given 1
in USAR for a fuel handling accident inside containment did not take credit for containment isolation or filtration by charcoal filters.
Thus isolation of DBP 5302FFF/47 l
-l
HVAC dampers in the Auxiliary Building (CV5009. CV5016, HA5439, HA5440 HA5716A
& B, CV5024, CV5004, CV5021. HA5441, HA5442, HA5715A & B, CV5025) and automatic initiation of station EVS by SFAS level 1 actuation in MODE 6 is not necessary.
Any radioactivity trapped between the purge system and the EVS can be removed by subsequent operation of the EVS.
Similarly the isolation of control room normal HVAC via SFAS level 1 is not necessary because there is no potential for the leakage of radioactive gases once the release is effectively terminated by the containment purge and exhaust noble gas monitor.
Since the containment purge exhaust is released to the environment via the station vent, any radioactive releases from the system will be detected by the station vent radiation monitors.
Upon detection of bigh radiation, the station vent monitors will automatically isolate the control room normal HVAC.
If both station vent radiation monitors are inoperable, procedural guidance exists to place the control room HVAC in the recirculation mode. Additional procedural guidance directs operators to align and start both trains of the control room emergency ventilation system upon high alarm of RE5052C.
Thus, automatic actuation of control room HVAC dampers HA5301A-H and HA5311A-H in MODE 6 is not necessary.
USAR section 6.2.3.2 contains a discussion of the capability of utilizing the EVS to aid in containment cleanup after a postulated fuel handling accident. A change to the original FSAR description of this capability was made in a previous revision to the USAR.
As part of the overall implementation of this License Amendment, the USAR section 6.2.3.2 discussion is being clarified and expanded.
Therefore the implementation of Amendment No. 186 has no negative safety implications.
DBP 5302FFF/48
1 l
SAFETY EVALUATION
SUMMARY
FOR UCN 94-100 (SE 94-0049)
TITLE:
Hiscellaneous FHAR Changes CHANGE:
The areas being changed in the FHAR are related to deleting duplicate information, making two minor editorial corrections, deleting specific manufacturer part/model numbers and to add a fire test standard for penetration seals.
REASON FOR CHANGE:
Deleted the discussions in section 4.6.FF.3 of the FHAR that involve internal separation of the circuits in the control room cabinets.
This discussion duplicates one that is found in Chapter 8 of the USAR.
Since the discussion deals with electrical separation and not 10CFR50 Appendix R type fire area separation, the appropriate place for this is Chapter 8 of the USAR.
Deleted the specific part/model numbers discussed in Appendix D of the FHAR for the ceiling tiles used in the Control Room complex and the for the covering used in the new fuel storage area.
The intent of both of these statements was to show that the products used were fire resistant.
By using the specific part/model number, ongoing replacement of these materials is not possible with newer fire resistant components, even if the requirements are met, due to the specific part/model number in the FHAR.
Added a reference to fire test standard used in qualifying the fire barrier penetration seals at Davis-Besse. This test has been part of the documentation of the seal program reviewed with the NRC, yet the FHAR was never revised.
Revised the discussion of fire and smoke propagation for shops and storage areas in Appendix D of the FRAR so that it only addresses the NRC's requirement.
The current test discusses specific room name and contents.
This type of information is more than required for the response.
Clarified the words on the use of polyvinyl chloride (PVC) in one section of Appendix D of the FHAR to be consistent with the wording on cable construction in another section of Appendix D.
The intent is to avoid the use of PVC to the extent practical and to not permit it in cable trays in the plant.
SAFETY EVALUATION
SUMMARY
The proposed changes do not impact safety because: the deletion of duplicate information and obsolete part/model numbers are clarifications in the text; the deletion of shop and storage room names is a simplification of the text; and the addition of a test reference is an enhancement.
In summary, the proposed changes are safe, claritive in nature and do not change the conclusions of the fire analysis.
DBP 5302FFF/2
SAFETY EVALUATION
SUMMARY
FOR UCN 94-122 (SE 94-0059)
TITLE:
Water Treatment Building Fire Detectors.
- CilANGE :
Revised the FHAR to clarify that there are actually five smoke detectors and one heat detector installed in the Water Treatment Building rather than six smoke detectors as listed in Table 8-3.
No physical change in the plant is being made, this is only to correct the table to reflect the plant design.
REASON FOR CHANGE:
This change will correct an error in Table 8-3 of the FHAR.
SAFETY EVALUATION
SUMMARY
The proposed change does not impact safety because the number and type of fire detectors is not changed from that originally designed for that area.
The heat detector is used in a small room and is well within the maximum allowed spacing for the device.
In summary, the proposed change is safe, claritive in nature and does not change the conclusions of the fire analysis.
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l SAFETY EVALUATION
SUMMARY
FOR
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UCN 94-155 (SE 95-0033)
TITLE:
Revision of USAR Chapters 11 and 12
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CHANGE:
Information regarding radioactive concentrations, process parameters, and other information which was utilized to design appropriate supporting SSCs to achieve.
the goal of. keeping offsite radioactive releases as lov as reasonable achievable vill be noted as historical, and not routinely updated.in USAR Chapters 11 and 12.
Information regarding the overall manner in which a process is performed, such as flow paths, equipment utilized, etc. vill be maintained.
REASON FOR CHANGE:
To more accurately reflect current plant operating practices and to denote material that is considered historical design information.
SAFETY EVALUATION
SUMMARY
i USAR Chapter 11 (Radioactive Vaste Management) contains descriptions of the design and operation of various plant SSCs which were intended to ensure that radioactive releases from the plant vould be as lov as reasonably achievable.
Chapter 12 (Radiation Protection) contains information'related to radiation protection design and operational considerations.
Of the information contained in these chapters, included were estimates of various activity levels and process parameters such as isotope concentrations, process liquid volumes and flow rates, cycle lengths, etc.
As summarized in USAR Section 11.0, plant radvaste systems are designed such that "All operations associated with radioactive materials are based on criteria and procedures ensuring that the operations can be performed in accordance with 10CFR20 and the "as lov as reasonable achievable" standard set forth in 10CFR50." In practice, the plant operates within the guidance of the site Of fsite Dose Calculation Manual (ODCH). The ODCM is consistent with the USAR intent.
As the plant has operated, various changes in these original process estimates such as the operating cycle length have occurred, and some of the process equipment such as the Degasifier is no longer utilized.
This is acceptable, since much of the assumed design amounts of radioactivity are conservative with respect to actual operational inventories, and the overall level of radioactive releases from the plant is monitored as specified in the ODCH.
Information presented in USAR sections for activities in various components for shielding design was based on continuous operation of the plant with 1% failed fuel. The Davis-Besse Technical Specifications and administrative limits, however, limit the RCS activity to well below the level corresponding to 1%
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failed fuel for.any significant duration. _Further, to minimize the radiation releases from. the plant to meet the numerical objectives of 10 CFR 50 Appendix I, the RCS activity is maintained to levels much lower than Technical Specification limits. Demonstrating compliance with 10 CFR 50 Appendix I and-10 CFR 20 using ODCM guidance ensures that there is no effect on safety when some of the plant radioactive vaste process equipment in not used.
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i SAFETY EVALUATION
SUMMARY
l FOR UCN 95-006 (SE 95-0017)
UCN 95-007 (SE 95-0016)
TITLE:
Changes to the DBNPS Nuclear Quality Assurance Program Site Organization' CHANGE:
The proposed change is to restructure the Management organization from four directors reporting to the Vice President to three directors reporting to the Vice President.
With this restructing comes the realignment of responsibilities.
REASON FOR CHANGE:
These changes are being proposed under Centerior's overall effort to continue to streamline the organization, realign groups into a more effective organization to improve communications, work flow and effectiveness.
SAFETY EVALUATION
SUMMARY
The proposed changes do not affect the safety function of any structures, systems and components and are considered to be safe. The proposed changes are only administrative in nature and maintain programmatic controls to assure that j
pertinent requirements in the Nuclear QA Program-(NOAP) are completely addressed.
j The changes proposed by USAR Change Notice (UCN)95-006 vere approved by the NRC in Log 1-3573 dated April 20, 1995.
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m SAFETY EVALUATION
SUMMARY
POR HUCN 95-018 (SE 95-0018)
TITLE:
I
- Revision of USAR Discussion of the Battery Room Ventilation-1 CHANGE:
This UCN clarifies USAR Section 9.4.2.1.3.1 and Table 9.4-7 on the functionality of the ventilation system and the inter-dependence and delineates.
required precautions when the battery room' ventilation system is not functioning as designed.
REASON FOR CHANGE:
Due to an identified concern,-the battery room ventilation systems were re-evaluated. The USAR implied that the operability of the station batteries depended upon the operability of the battery room ventilation system. The station batteries present a negligible heat load and hydrogen generation rate
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compared to the duration of time that the batteries would be required to function.
SAFETY EVALUATION
SUMMARY
Hydrogen Production and Concentration:
On float charge at 2.2 VPC, 0.0022 cfm of hydrogen would normally be produced
' f in each room.
If both batteries in a room vere placed on equalize charge at 2.33 VPC, 0.012 cfm hydrogen would be produced. The forced ventilation systems each exchange orders of magnitude more air than is' required to dilute the hydrogen if it is uniformly mixed with air.
Based on this information, the quantity of hydrogen production currently provided in the USAR will be revised. The purpose of the ventilation, not the flow rate, will be specified in the USAR.
t The forced ventilation system vill provide abundant dilution, even if an adverse thermal layer should form. The USAR currently states that the exhaust openings are located close to the ceiling which aids in exhausting any light gases collected there.
Due to the lov hydrogen generation rates I
expected and natural mixing mechanisms available, the ventilation openings need not be located near the roof. Therefore, the ventilation opening location information is being deleted from the USAR.
Without forced ventilation, the vorst case scenario would involve battery room A.- At the maximum equalize charge rate (2.39 VPC) with a hydrogen production rate of 0.026 cfm, the concentration limit of 2 percent in the battery room could be reached within approximately 2 days.
Opening the door to the battery room would limit maximum hydrogen concentration to less than 2 percent for over 3 days, while still crediting only molecular diffusion within the battery room.
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Due to the much larger hydrogen generation rate, continuous forced ventilation is prudent to maintain, especially when equalize charging is in progress.
USAR section 9.4.2.1.3.1 currently indicates that a 4 percent (combustible) concentration would be reached in 10-15 days.
While 4 percent by volume is the actual combustible concentration limit, 2 percent is the FHAR administrative limit. This information is being revised in the USAR to reflect the 2 percent administrative limit and the time duration based on the 135 VDC float charge voltage.
Conditions During and Following Loss of AC pover Station batteries are sized to supply the station DC and Instrument AC loads for a period of one hour without AC power supply availability.
During this time period, due to loss of AC power, no ventilation systems would be operating. However, during battery discharge, no significant quantity of hydrogen vould be produced.
Some internal heat vould be generated by the cells due to internal resistance and the rapid rate of discharge.
- However, this heat generation would not be adverse if the cells and the battery room were at an acceptable temperature prior to the onset of discharging.
Following discharge of the batteries, even without forced ventilation, the batteries could be recharged without any substantial increase in the rate of hydrogen buildup.
After a full charge were to be reached by the majority of cells, a relatively high hydrogen evolution rate could occur.
However, the batteries vould be voll attended following a loss of AC power event, such that ventilation vould be assured within the required time.
Battery Room Conditions Following Loss of Ventilation, Prior to Compensatory Action:
The two battery room ventilation systems provide an annunciator alarm in the control room when lov air flow occurs in either energized fan or when neither fan is energized.
Up to the point when constant ventilation is lost, temperature and hydrogen concentration are controlled.
Following loss of coastant ventilation, room temperature vould change very slowly over the course of many hours.
As stated above, even with no ventilation and charging at the normal " equalize" rate, an operator would have at least four days to respond to the alarm for hydrogen control.
If the safety grade system should fail, the non-safety grade ventilation vill not automatically start, but the control room annunciator vill actuate. The IISAR allows a room temperature (electrolyte temperature) from 60 to 104'F, while the automatic fan starts at approximately 77'F and stops at approximately 70'F.
Based on the overlap between fan actuation temperatures and the USAR allowable temperatures and based on the large thermal capacity of the room, an operator vould have many hours to respond to the loss of ventilation annunciator alarm.
Opening the battery room doors vill encourage sufficient free air exchange to eliminate hydrogen concentration buildup and vill moderate air temperature changes.
Therefore, action to open a doorway will be credited in USAR section 9.4.2.1.3.1 and " equalize" charging vill be restricted when no continuous forced ventilation is provided.
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i It is noted that the safety grade train of ventilation ensures that full ventilating capability is not lost even under design conditions.
However, the battery room ventilation is not immediately required.
With the procedural controls in place to open the battery room door when ventilation fans are not available, combustible concentrations vill not be reached and room temperature vill normally be adequate to maintain battery temperature.
Based on the above, operability of the station batteries is not directly linked to immediate operability of the battery room ventilation system.
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SAFETY EVALUATION
SUMMARY
FOR UCN 95-025 (SE 95-0023)
TITLE:
Update Description of the Solid Radvaste Handling System CHANGE:
Deleted the description of solid radvaste handling equipment that is no longer used. This change also updated the location where solid vaste handling occLrs to the Low Level Radvaste Storage Facility.
REASON FOR CHANGE:
This UCN provided a updated description of the Solid Radvaste Handling System.
SAFETY EVALUATION
SUMMARY
The proposed changes to the USAR describe improvements in facilities and equipment for the safe handling of solid radioactive vaste. The safety of the Low Level Radvaste Storage Facility (LLRVSF) was previously analyzed and found to be adequate. At that time, the handling of solid vaste was relegated to that facility. This change to the USAR provides a more accurate description of the process.es nov in place end deletes a process description that is no longer pertinent. There is no effect on safsjty from this change.
There are no effects on safety due to the proposed changes.