ML20210H063
| ML20210H063 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/28/1999 |
| From: | Campbell G CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20210H053 | List: |
| References | |
| NUDOCS 9908030267 | |
| Download: ML20210H063 (20) | |
Text
Docket Number 50-346 -
License Number NPF-3 Serial Number 2397 lPage1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3. Also included is the Safety Assessment and Significant Hazards Consideration.
- The proposed changes (subm tied under cover letter Serial Number 2397) concern
Appendix A, Technical Specifications (TS):
3/4.7.5.1 - Plant Systems - Ultimate Heat Sink I, Guy G. Campbell, state that (1) I am Vice President - Nuclear of the FirstEnergy Nuclear Operating Company, (2) I am duly authorized to execute and file this certification on behalf of the Toledo Edison Company and The Cleveland Electric Illuminating Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.
By:
hw Guy G. Ca@ bell, ViceMsident - Nhclear Affirmed and subscribed before me this 28th day of July,1999.
h
. Notary Public, State of Ohio - Nora L. Flood My_ commission expires September 4, 2002, l
\\
M[M~
v e
m
Docket Number 50-326 License Number NPF-3 Serial Number 2397 Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specification (TS) 3/4.7.5.1, Plant Systems -
A.
Time Required to implement: The License Amendment associated with this license amendment application is to be implemented within 120 days after NRC issuance.
B.
Reason for Change (License Amendment Request Niunber 96-0008):
The proposed changes would increase the allowable water temperature for continued plant operation, as specified in Technical Specification Limiting Condition for Operation 3.7.5.1.b, from 5 85 F to 5 90 *F.
C.
Safety Assessment and Significant Hazards Consideration:
See Attachment 1.
D.
Drawings: provides drawings showing the configuration of the intake canal and conduit.
1
Docket Number 50-346 License Number NPF-3 Serial Number 2397 Attachment i SAFETY ASSESSMENT AND SIGNIFICANT IIAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 96-0008 (19 pages follow) i
LAR 96-0008 Page1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 96-0008 TITLE:
Proposed Modification to the Davis-Besse Nuclear Power Station Unit Number 1, Facility Operating License NPF-3, Appendix A, Technical Specifications, to Increase the Water Temperature Limit in Technical Specification (TS) 3/4.7.5.1, Plant Systems-Ultimate Heat Sink.
DESCRIFI' ION:
The purpose of this proposed change is to modify the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A, Technical Specifications (TS). The proposed change would increase the allowable Ultimate Heat Sink (UHS) average water temperature, as specified in Technical Specification Limiting Condition for Operation 3.7.5.1.b, from s 85 *F to s 90 F.
Although having the UHS temperature reach 85 F at the DBNPS is a rare phenomenon, the UHS temperature did approach 85 F in August 1995 due to an extended period of unusually hot weather. At that time, the DBNPS submitted a Request for Enforcement Discretion to allow continued plant operation,in the event the UHS temperature continued to increase and the TS limit was exceeded. A follow-up emergency License Amendment Request (LAR) was submitted to change the UHS TS temperature limit to s.90 F on a temporary basis. However, the UHS temperature did not exceed 85 F and the LAR was withdrawn by DBNPS. Nuclear Regulatory Commission (NRC) approval of this request to permanently change the UHS temperature to 90 F would eliminate the potential for a future UHS TS-related request for enforcement discretion and an emergency LAR.
In addition to the above benefit, this LAR addresses an issue previously posed by the NRC regarding the Service Water (SW) System's maximum water temperature used in the existing DBNPS Updated Safety Analysis Report (USAR) containment response analysis for a design basis accident. The existing containment response analysis used a constant SW System temperature, assuming its source, the UHS, remained connected to l
Lake Erie. The containment response analysis performed to support this LAR does not I
assume the UHS remains connected to Lake Erie following a design basis accident.
Therefore, the SW System temperature was determined as a function of time following the accident.
The proposed change is shown on the attached, marked-up TS page.
1 LAR 96-0008 Pcge 2 SYSTEMS, COMPONENTS, ANL) ACTIVITIES AFFECTED:
The Service Water System, Component Cooling Water System and components cooled i
by these systems are affected by the proposed TS change. Technical Specification Limiting Condition for Operation (LCO) 3.7.5.1.b, which presently requires an UHS average water temperature of g 85 F, is affected by the proposed TS change.
FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES:
As described in USAR Section 9.2.5, Ultimate Heat Sink, the UHS for the DBNPS is Lake Erie, which is the source of cooling water for the SW System. Lake Erie water flows through a buried intake conduit,96 inches in diameter, to the intake canal. The intake canal flows to the intake structure, where the service water pumps are located.
There is an open forebay area ahead of the intake stmeture.
The SW System is described in USAR Section 9.2.1, Service Water System. During normal operation, the SW System supplies cooling water to the Containment Air Coolers (CACs) and the Component Cooling Water (CCW) heat exchangers. The CACs are described in USAR Section 6.2, Containment Systems, and the CCW System is described in USAR Section 9.2.2, Component Cooling Water System. The SW System also provides flow to the Emergency Core Cooling System (ECCS) room coolers during normal plant operation to aid in preventive maintenance and to provide supplemental room cooling, as required. The ECCS room coolers are described in USAR Section 9.2, Service Water. The SW System also performs numerous other functions which support power operation, but which are not important to safety, such as supplying cooling water to the Turbine Plant Cooling Water (TPCW) heat exchangers and providing makeup to the Circulating Water (CW) System.
During a Design Basis Accident (DBA), the SW System supplies cooling water from the UHS intake forebay to the CACs, the CCW heat exchangers, the ECCS room coolers, and the Control Room Emergency Ventilation System (CREVS) (USAR Section 9.4.1, Control Room). The SW System also provides seal water to the Hydrogen Dilution System blowers (USAR Section 6.2.5, Combustible Gas Control in Containment Vessel) and provides a safety grade backup source of water for the Auxiliary Feedwater (AFW)
)
System (US AR Section 9.2.7, Auxiliary Feedwater System) and the Motor-Driven Feedwater Pump (MDFP) (USAR Section 9.2.8, Motor Driven Feedwater Pump) in the j
event the normal Condensate Storage Tanks' water supply is unavailable. The SW j
System is also a backup source of makeup water for the CCW System if no other CCW makeup water sources are available following a Safe Shutdown Earthquake (SSE).
The CACs provide containment heat removal during normal plant operation and following a DBA.
F LAR 96-0008 '
Page 3 '
The CCW heat exchangers supply cooling to various components during normal operation and to the following essential components during a DBA:
High Pressure Iniection (HPD oumns 1 and 2 bearine oil coolers The HPI System (USAR Section 6.3, Emergency Core Cooling System) provides emergency core cooling for Small Break Loss-of-Coolant Accidents (SBLOCAs) and makeup to the Reactor Coolant System (RCS) to compensate for contraction during overcooling of the RCS. The HPI System also provides a source of borated water from the Borated Water Storage Tank (BWST) for Shutdown Margin (SDM) requirements.
Decay Heat / low Pressure Iniection (DH/LPD pumps 1 and 2 bearing housine coolers The LPI System (US AR Section 6.3, Emergency Core Cooling System) provides emergency core cooling and refill of the reactor vessel following a Large Break Loss-of-Coolant Accident (LOCA), post-LOCA containment sump recirculation capabilities, long term decay heat removal, and post-LOCA boron dilution. The DH Removal System (USAR Section 9.3.5, Decay Heat Removal System) removes decay heat from the core and sensible heat from the reactor coolant system during the later stages of normal plant cooldown.
Decay Heat (DH) Coolers 1 and 2 The DH coolers (USAR Section 9.3.5, Decay Heat Removal System) provide post LOCA Containment (CTMT) sump cooling for emergency shutdown cooling requin:ments. During the later stages of normal plant cooldown, the DH coolers remove decay heat from the core and sensible heat from the reactor coolant system.' The DH coolers can also be used to provide cooling for the Spent Fuel Pool.
Containment (CTMT) Gas Analyzer heat exchaneers 1 and 2 The CTMT gas analyzers (USAR Section 6.2.5, Combustible Gas Control in Containment Vessel) provide the capability of analyzing the CTMT atmosphere to detect the build up of hydrogen in CTMT following a LOCA. This function is required to monitor hydrogen levels such that they can be maintained within acceptable limits.
- Emernency Diesel Generator (EDG) iacket cooline water heat exchancers 1 and 2 The EDG jacket cooling water heat exchangers (US AR Section 8.3.1.1.4, Diesel Generators) maintain the required temperature of the lube oil and diesel engine to ensure that the EDGs will perform their intended safety function during and after an accident.
LAR 96-0008 Page 4 The ECCS room coolers, supplied by SW, maintain a suitable temperature in the ECCS rooms during a LOCA. The ECCS room coolers are also available to provide supplemental cooling during normai plant operation, as required.
The CREVS is not required to operate during normal plant operations. When required to operate, following a loss of non-essential power, a station vent monitor high radiation signal, or a Safety Features Actuation System (SFAS) signal, the CREVS maintains Control Room (CTRM) temperature s95 *F and Cabinet Room temperature $110 F for habitability considerations. The CREVS also maintains the entire control room complex at a positive pressure.
The Hydrogen Dilution System blowers provide aur.iliary building air to the containment vessel for the dilution of the post-LOCA hydrogen concentration.
EFFECTS ON SAFETY:
The discussion below addresses the effect of an increase of 5 F in the UHS temperature and the SW supply temperature during both normal plant operating and emergency operating conditions. The einergency operating conditions assessment includes consideration of a loss of the connection with Lake Erie and the effect of an SSE on the intake canal.
A.
Normal Operation Containment AirCoolers(CACs)
The SW System provides cooling water from the UHS to the CACs. Under normal operating conditions the CACs maintain the containment average air temperature, which is a design input to the LOCA and EQ analyses, below the current Technical Specification LCO 3.6.1.5 limit of 120 *F. A change to the current LCO 3.6.1.5 CTMT air temperature limit is not being proposed in this submittal. Extrapolation of normal operating condi; ions to 90 F UHS temperature indicate that the CACs can maintain the containment average air temperature below 120 F. Since LCO 3.6.i.5 will continue to limit the containment air temperature during normal operating conditions, there will be no adverse effect on plant safety by an increase to 90 F of the UHS temperature.
Component Cooling Water (CCW)
During normal plant operating conditions, there is sufficient cooling capacity available in the CCW System to accommodate an increase in the SW supply (UHS) temperature to 90 F. The automatic controls on the CCW System are set procedurally to maintain the CCW heat exchanger outlet temperature between 80 *F and 120 *F, with the outlet temperature normally maintained between 85 *F and 95 F. An evaluation of the CCW heat exchangers has determined that
LAR 96-0008 Page 5 an outlet temperature of approximately 97 *F can be maintained with a 90 *F SW inlet temperature.
The DH Coolers and Spent Fuel Pool (SFP) Coolers (cooled by CCW) provide cooling for the SFP. It may be noted that a CCW outlet temperature of 95 F, rather than the approximately 97 *F discussed above, was assumed in the SFP thermal-hydraulic analysis in support of another DBNPS LAR recently submitted to the NRC, LAR 98-0007. LAR 98-0007 was submitted to the NRC on May 21,1999 (FirstEnergy Letter, Serial Number 2550) (Reference 15) requesting NRC approval to use additional spent fuel racks in the cask pit area.
As stated in LAR 98-0007, American Concrete Institute (ACI) Code t
Requirements for Nuclear Safety Related Concrete Structures, ACI-349 (Reference 9), permits long-term concrete temperatures of up to 150 F and short term excursions in localized areas up to 350 F. The results of the analysis performed in support of LAR 98-0007 indicate that, under the worst-case transient conditions, the bulk SFP water temperature will be above 150 F for less than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />, and should not exceed 151.5 F, and the bulk Cask Pit water temperature will be above 150 *F for approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, and should not exceed 154.5 F. An increase in CCW temperature of approximately 5 F(to
)
approximately 100 F) would increase these times above 150 F to approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for SFP water and 170 hours0.00197 days <br />0.0472 hours <br />2.810847e-4 weeks <br />6.4685e-5 months <br /> for Cas'. t water. The maximum expected bulk temperature for the SFP would bc less e an 157 F and less than 160 F for the Cask Pit. These temperatures and durations are not significant when considering the thermal inertia of the concrete walls and floor.
The concrete temperature will lag the water temperature such that the bulk of the concrete mass cross-section will remain well below the 150 *F range. In fact, a very small depth of the concrete will actually experience temperatures in excess of 150 *F. Accordingly, the SFP and Cask Pit wall and floor concrete temperatures in excess of 150 F, under the worst case conditions, are acceptable. since they will exist for a short duration. This evaluation is based on information obtained from Reference 8.
Consequently, all loads served by the CCW heat exchangers during normal power operation will continue to be operated in their normal ranges, and there is no adverse effect on plant safety.
ECCS Room Coolers The SW System supplies flow to the ECCS Room coolers to provide supplemental cooling. An increase in the SW supply temperature of 5 *F during i
normal plant operation would have no adverse effect on this function.
LAR 96-0008 Page 6 b
Other Components and Systems The other functions of the SW System are not impodant to plant safety during.
I normal plant operation, therefore, an increase in the SW supply (UHS) temperature would have no adverse effect on plant safety. However, an increase in the SW supply temperature could affect TPCW-cooled components and, hence, limit plant capacity. The TPCW System temperature is normally maintained at 85 F at the outlet of the heat exchangers. A 90 *F SW supply temperature would not permit this TPCW System temperature to be maintained.
The thermal limitations for the balance-of-plant equipment have been assessed.
This assessment determined that power operation may continue with SW System temperatures of 90 *F. However, at elevated SW temperatures TPCW-cooled components would be closely monitored. This monitoring would be no different than that which would be performed if the UHS temperature approached the current TS LCO 3.7.5.1.b limit of 85 F. Appropriate actions, including possible turbine load reductions, would be taken to ensure that acceptable equipment operating conditions are maintained.
B.
Emergency Operation Desien Basis Accident Seauence An increase in SW supply temperature would affect the temperature of the CCW System, the containment building, and ECCS rooms following a DBA. A design basis LOCA imposes the greatest performance requirement on the SW System.
In this postulated accident, it is assumed that a loss of offsite power occurs, with the concurrent failure of an Emergency Diesel Generator. Thus, only one train of SW and one train of ECCS pumps are assumed available. Davis-Besse does not postulate a concurrent LOCA and seismic event. However, only safety grade (seismic) equipment will be credited to mitigate the LOCA. The connection to Lake Erie, although rugged, is not Seismic Class I. Therefore, the connection between Lake Erie and the forebay is assumed to be lost in support of this LAR.
The containment performance analysis for a design basis LOCA is presented m USAR Section 62, Containment Systems.
In a design basis LOCA, the critical period for refilling and cooling the reactor core occurs within the first few minutes of the accident. Following refill of the reactor vessel, the fuel will be adequately cooled. Service Water temperature does not directly impact core cooling during this portion of the event because the reactor is refilled via the Core Flood Tanks (CFF) and the Low Pressure Injection (LPI) pump using water from the Borated Water Storage Tank (BWST). In addition, the available Containment Spray (CS) pump injects water directly into containment from the BWST without any cooling supplied by the SW System. Until the BWST is exhausted, cooling is not required for the Decay Heat Coolers. The initial blowdown is sufficiently rapid that heat removal via
Pag 7 the CACs is not effective in reducing the initial temperature / pressure spike in containment.
Following depletion of the BWST inventory, containment cooling for preservation of containment equipment qualification must be maintained. At this point, more than 30 minutes into the transient (depending on pump combindions in service), the suctions of both the LPI pump and the CS pump are transferred to the containment emergency sump, where effluent from the break and CS has accumulated. A CCW heat exchanger, cooled by SW, is required to' provide cooling to the Decay Heat Removal (DHR) heat exchanger at this point.
' Upon making the transfer to the emergency sump (which is at a higher temperature than the BWST), both the LPI temperature and the CS injection temperature increase, giving rise to a second, temporary increase in containment temperature and pressure. This increase peaks at approximately 12,000 seconds and is lower than the initial pressure / temperature peak. The long term
- containment temperature following a LOCA is controlled by heat removal through the DHR System and the CACs.
Containment Response Re-Analysis The post-LOCA containment response analysis was re-performed (Reference 3) assuming that the CACs receive SW for cooling at a conservative flowrate of 1150 gpm and assuming that the connection to Lake Erie does not exist. With no flow from or to Lake Erie, the forebay temperature increase was analyzed. The results of this analysis were used as the SW temperature in the containment response re-analysis. Additional discussions regarding the loss of the connection to Lake Erie are provided later. The CAC heat removal duty was developed using the COOLNUC program developed by the CAC manuf acturer (Reference 6). The benchmarking runs for the COOLNUC program show that the heat removal duties predicted by this program are conservative when compared to the CAC heat removal duties used in the previous USAR analyses. A fouling factor of 0.003 was used in this analysis, which is consistent with the use of lake water.
Figures 1 and 2 show a comparison of calculated containment temperature and pressure profiles for: (1) the current USAR Section 6.2 analysis with a constant SW temperature of 85 'F, assuming that the connection to Lake Erie is.
i maintained (Reference 7), and (2) a variable SW/ forebay temperature with an initial temperature of 90 F, assuming there is no connection to Lake Erie (Reference 3). The peak containment temperature occurs during the initial reactor coolant system blowdown phase, and is therefore not affected by the initial SW temperature, as discussed above. The peak containment pressure occurs at approximately 15 seconds after the CACs are started. During this short time, the difference in the effectiveness of the CACs with 85 F or 90 F SW is very small and has an insignificant effect on the containment response.
However, modeling changes made between the time the existing analysis was
r LAR 96-0008, l-Page 8 performed and now have resulted in very slight increases in the peak temperature and pressure. The maximum peak temperature and pressure remain well below the maximum design conditions for the containment vessel. As the transient progresses, the containment temperature and pressure responses are higher than previously analyzed. These increases are primarily due to the increased SW temperature. The second temperature and pressure peaks remain bounded by the initial peaks.
The new containment temperature profile was evaluated for its impact on Environmental Qualification (EQ) of equipment located in containment (References 3 and 4). This evaluation concluded that the equipment qualification tests demonstrate that the instrumentation required to monitor the course of an accident have a qualified life of approximately 1 year or greater, which is consistent with the Toledo Edison Design Criteria Manual Section 8.3.3, Environmental Conditions (Reference 13). The time dependent temperature profile for the ECCS Pump rooms was also analyzed (Reference 5).
The resulting temperature profile is conservatively bounded by the temperature profile used for the present environmental qualifications for the equipment within these rooms (References 10 and 11).
The Main Steam Line Break (MSLB) containment response analysis was also re-performed (Reference 3). The peak containment pressure is essentially the same as the previous analysis. The peak containment temperature is slightly higher (i.e., approximately 1.5 *F), than the previous analysis. As with the current analysis, the peak temperature spike is of such duration that equipment temperatures remain below those encountered during a LBLOCA. Therefore, the new analysis does not impact the environmental qualifications of any equipment and is bounded by the LBLOCA analysis.
With an isolated forebay and an increased initial SW temperature, CCW temperatures are also predicted to be higher than in the previous analysis. The peak CCW temperature predicted by the LOCA containment response analysis
- (Reference 3) is slightly less than 120 F at approximately 40000 seconds following the accident, which is within the temperature requirements of the essential components served by CCW, as discussed below.
The maximum allowable bearing temperature for the HPI pumps is 165 *F.
Plant surveillance testing data show that the maximum bearing temperature is normally below 115 F, and well below 165 F. Therefore, the HPI bearing i
temperature will remain well below 165 *F even with the increased initial CCW temperature. Time dependent SW temperature increases beyond the initial l
increase to 90 *F can be accommodated by the HPI bearing margin.
The DH/LPI pumps, Emergency Diesel Generators and CTMT Gas Analyzens l
are designed to operate with a cooling water (CCW) temperature of 120 F or l
less, which is above the predicted maximum CCW temperature. Therefore,
LAR %-0008 Page 9 essential CCW loads will be adequately served following a DBA even during the time of peak CCW thermal loading. Non-essential CCW loads are automatically
' isolated by SFAS.
Two ECCS room coolers, of 50 percent capacity each, are provided in each of the two ECCS rooms and one cooler is in the DH cooler room. The containment response analysis does not take credit for the cooler in the DH cooler room.
Based on an evaluation (Reference 5), two ECCS room coolers are adequate, assuming the increased post-LOCA SW temperature profile and an initial temperature of 90 F, even with flow rates substantially degraded from normally accepted values. During periods of reduced SW temperature, one ECCS Room cooler may be removed from service, provided the remaining cooler can provide 100 percent of the required cooling capacity as determined by engineering analysis. Removal of one cooler from service is procedurally controlled.
The CREVS condenser cooling units receive cooling water directly from the SW System and are required to operate during a LOCA. These units are also provided with a safety grade, air-cooled condensing coil. The air-cooled condensing unit is automatically selected if refrigerant pressure increases due to inadequate water-cooled condenser cooling. This is expected to occur at a SW temperature of approximately 110 *F. Since the maximum SW temperature is predicted remain below 110 *F (as shown in Figure 3), an increase in the initial SW temperature does not impact the availability of CREVS.
The Containment Hydrogen Dilution Blowers are "Nash" pumps that utilize less than 10 gpm of SW to provide seal water. The inlet temperature of SW for this use is not critical.
The SW makeup connection to the CCW System will be unaffected since the inlet temperature of SW would be lower than the CCW temperature.
The emergency suctions for the AFW System and the MDFP are supplied by the SW System. A conservatively assumed increase in SW temperature to 110 *F represents a very small increase in initial liquid enthalpy when compared to the large increase in feedwater enthalpy as it is boiled in the steam generators. Due to the ample flow capacity of the AFW pumps and the MDFP, the increase in SW inlet enthalpy will have a negligible effect on system response. A 90 *F SW temperature will provide adequate pump bearing cooling when the pumps are taking suction from the SW System._ Time dependent SW increases beyond the initial increase to 90 *F are not necessary to consider because the SW heatup is based on a LBLOCA and the SW feed to AFW or the MDFP is not required following a LBLOCA.
The piping and supports for the SW and TPCW Systems have been reviewed for the proposed normal and emergency operating temperatures. The existing piping stress analysis temperatures bound the proposed increases for both systems.
l
LAR 96-0008 Page 10 In summary, the existing margin in equipment performance has been reviewed and determined to maintain the containment response within design parameters following a LOCA. In addition, all equipment will operate as designed for all transients. Hence, there is no adverse effect on plant safety.
C.
Losgor Connection Between Lake Erie and the Intake Canal The int tke canal is connected to Lake Erie by a 96" diameter buried intake conduit. Lake water enters the intake conduit via an intake crib located in the lake. 'I he level of the intake crib is slightly below 562' International Great Lakes l
Daium :lGLD). Thus, the connection between Lake Erie and the intake canal could b : lost if the Lake Erie water level were to decrease below approximately 562' IGLD.
The low water datum of Lake Erie is 568.6'IGLD. The maximum variations in the mean monthly level are 4.2' above and 1.2' below the datum for the 110-year period that the data has been collecte1 Therefore, during normal operation, the lake water level will be significantly higher than 562'IGLD. The Lake Erie water level could be expected to decrease below 562'IGLD if a maximum probable veteorological event, consisting of a sustained WSW wind of 70 mph for a six hour duration, were to occur. The low level condition resulting from such an event would last for a period of approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Technical Specification Limiting Condition for Operation 3.7.5.1 requires that a plant shutdown be commenced if the forebay level decreases to below 562.0'IGLD. Since the low water condition due to a meteorological event is of limited duration, and would result in the commencement of a plant shutdown, such a circumstance has no j
impact on the proposed change to increase the water temperature limit to s 90 F.
The connection with Lake Erie could also be lost if an earthquake was assumed to collapse the non-seismic Class Iintake conduit. DBNPS had an evaluation performed on the intake conduit from Lake Erie, associated non-seismic Class I components, the intake canal dikes, and the Lake Erie crib stmeture to determine their ability to withstand the effects of a Safe Shutdown Eanhquake (SSE). The evaluation, performed by EQE International, considered ground wave motion induced strains and displacements, inertial loads, and potential soils-related failures. Although the entire intake system was not designed as a seismic Class I system, the evaluation concluded that the above-mentioned structures and components were capable of resisting the effects of an SSE with no loss of function. A detailed discussion of this evaluation was provided to the NRC by DBNPS letter dated January 31,1996 (Serial No. 2347) (Reference 14). The evaluation concluded that a loss of connection with Lake Erie due to a seismic event is not credible. However, even though the postulation of an earthquake combined with a LOCA is outside the licensing basis of the DBNPS, the containment response to a LOCA was reanalyzed, as previously discussed, assuming that the connection with Lake Erie was lost.
LAR 96-0008 Page11 If a seismic event were to result in a loss of the connection between the intake canal and Lake Erie, a Plant Emergency Procedure provides direction to establish temporary pumping to the intake forebay. It is expected that the connection between Lake Erie and the intake canal could be re-established well within a 30 day period, s currently stated in USAR Section 9.2.5.2.
In summary, the proposed change to increase the allowable water temperature, as specified in TS LCO 3.7.5.1.b, from 5 85 F to g 90 F, will not adversely affect plant safety.
)
i
LAR 96-0008 Page 12 SIGNIFICANT HAZARDS CONSIDERATION:
The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance wit.. the proposed changes would:
(1) Not involve a significant increase in the probability or consequences of an accident previously evaluated;(2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. The Davis-Besse Nuclear Power Station has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit No.1, in accordance with these changes would:
l a.
Not involve a significant increase in the probability of an accident previously evaluated because no accident initiators, conditions, or assumptions are significantly affected by the proposed change. The proposed change would increase the allowable Ultimate Heat Sink (UHS) water temperature, as specified in TS LCO 3.7.5.1.b, from s 85 F to s 90 F. This water is used by the Service Water System to provide cooling to equipment that is used to mitigate accidents such as a Large Break Loss of Coolant Accident. This increase in Service Water temperature has been evaluated and the proposed change does not result in the operation of equipment important to safety outside their acceptable operating ranges.
I b.
Not involve a significant increase in the consequences of an accident previously evaluated because the proposed change does not change the source term, containment isolation, or allowable releases. The proposed increase in the Service Water System temperature has been evaluated with respect to the containment and equipment used to mitigate the consequences of accidents previously evaluated. These evaluations have determined that there are no significant increases in consequences.
2.
Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assumptions are introduced by the proposed 5 F increase in UHS temperature. The proposed change does not result in installed equipment being operated outside their design operating ranges. No new or different e.quipment failure modes or mechanisms are introduced by the proposed change.
3.
Not involve a significant reduction in a margin of safety because the proposed 5 F increase in UHS temperature does not result in significant changes to the initial conditions contributing to accident severity or consequences.
LAR 96-0008 Page 13 CONCLUSION:
On the basis of the above, the Davis-Besse Nuclear Power Station has determined that this License Amendment Request does not involve a significant hazards consideration.
As this License Amendment Request concerns a proposed change to the Technical Specifications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.
ATTACHMENT:
Attached are the proposed marked-up changes to the Operating License.
REFERENCES:
1.
DBNPS Operating License NPF-3, Appendix A Technical Specifications through Amendment 232.
' 2.
DBNPS Updated Safety Analysis Report through Revision 21.
Section 2.4.11, Imw Water Consideration Section 6.2, Containment Systems Section 6.2.5, Combustible Gas Control in Containment Vessel Section 6.3, Emergency Core Cooling Section 8.3.1.1.4, Diesel Generators Section 9.2.1, Service Water System Section 9.2.2, Component Cooling Water System Section 9.2.5, Ultimate Heat Sink Section 9.2.7, Auxiliary Feedwater System Section 9.2.8, Motor Driven Feedwater Pump
)
Section 9.3.5, Decay Heat Removal System
)
Section 9.4.1, Control Room 3.
DBNPS Calculation C-NSA-060.05-008, rev. O, Containment Post LOCA Response with Variable SW Temperature.
4.
DBNPS Calculation C-ECS-201.10-001, rev.1, Post-Accident Operating Time Test Profile Extrapolation.
5.
Bechtel Calculation 12501-M-003 rev. O, ECCS Room Temperature with Initial 90 F Forebay.
6.
DBNPS Calculation C-NSA-060.05-007, rev. O, CAC Heat Duty at Elevated SW j
Temperatures.
)
LAR 96-0008 Page 14 7.
Bechtel Calculation 512-80-12501, Rev 0, LOCA Analysis with Degraded ESW Flow of 1150 GPM to Containment Air Coolers.
8.
Holtec Report HI-981864 Rev 5, DB Vendor Calc # C-63Q-1.
9.
American C nerete Institute (ACI) 349-85 and 349R-85," Code Requirements for Nuclear Safety Rt ated Concrete Structures and Commentary".
l
- 10. DBNPS Calculation C-ECS-099.16-146, rev 00, Thermal Aging Effects of ECCS Room Post LOCA Temperature.
I1. EQ Package DB1-036C, Rev 4, Electrical Equipment Environmental Qualification Package for Westinghouse HPI Pump Motor.
- 12. DBNPS System Descriptions:
- a. SD-003, Rev. 3, Emergency Diesel Generators
- b. SD-016, Rev. 3, Component Cooling Water c.
S D-018, Rev. 2, Service Water
- d. SD-022B, Rev. 2, Containment Air Coolers
- e. SD-023, Rev. 3, Hydrogen Control f.
SD-029B, Rev. 2, Control Room Emergency Ventilation
- g. SD-038, Rev. 2, High Pressure Injection
- h. SD-042, Rev. 2, Decay Heat Removal
- 13. Toledo Edison Design Criteia Manual, Rev 3.
- 14. Toledo Edison Letter to the h RC dated January 31,1996 (Serial Number 2347),
"Ubimate Heat Sink / Service Water Temperature."
- 15. pirstEnergy Ietter to the NRC dated May 21,1999 (Serial Number 2550), License Amendment Request 98-0007.
1
0 0
0
- h l
.I 0
F 3
!1i 0
s llwi I
I1i' 0
4 x
ll-1 6
D>lL 2
>li If f
x o
A,ll>l' e
ur t
x 0
ar 0
e N'lilo'L
.ll 0
p ll ie ll fIi' 0
m l,
l l;Ii 0
e r
llI1;'
iL ili
!l E
ll
!1' 1
t N
e n
k g
x l'
a is s
L e
n x
~
e o d
o N
i t
r t
t E n n
s
\\
1 x
o L
\\
ei e
0 e
i m
,lll_' A k
t s
r c
a 0
n a iillI e
n p
Ap L
0 i
o n
o n 0
a t
Cm t
o 1
n s
,.z,, /
it u
e n C o
R o
c Oo g
t c
R e
o L
a r
e u
C A
r se
,!N"s, h
n t
t n i S
a t
w o
U C.
lli i'
)c ep os W
e n
h m
P n X
it S
0 s a
,1l N
wF 0
(
s 1
e o
i1; 0 e u
T t
e np
, g 1
s 1l' We m
r t
r u
n e
Il' g
e e s,,1l S D l
i]i:
iT v
F m
e,tlJ.~
F 0 ll' f'
F i
mR g
9 n
o ar ;ilW-ia n
ll e
l 8
a e
D it t
3 i
n i
6 5
I 2
n 8
C t u nt s
A
,l 0
e
~
p p 0
i C
oa
,l'
~
mm
,l!
I' 1
r r
,I' e e
,l t
O Ce Iin u
- i T T L
a p
,i T T r
e t
MM p
so m
,I'-
T T m
P e
,I' CC e
t T
IA Il'
,l,+
0 n
t ne m
1Il'
- l i;'
1 i
- I' 1I;!,
a L;
,illI!l,-
tn
,:llI/li, il l
1 c
o k
,lll1ll,,
!,l a
e e
e' 1'
i' d
p
'e e
77 ta l
1 u
c l
5 0
5 0
5 0
5 0
5 0
5 a
7 5
2 0
7 5
2 0
7 5
2 c
80 2
2 2
2 1
1 1
1 e
0 h
0 c eo B 2 $ t a E oF ceE j e,g O T
n e
65 e
91 to Re N
g Aa LP
.g is 0
p 0
0 0
Mn': 0 0
0 4
i' 0
f o
N u
1
=
er 2
Ni sse r
N p
0 n
N1l7 0
g i
- 0 ilt s
n' 0
e s
0 d
e liili Il,
m i
x i;Itli 1
r E
s ill's u
\\
e m
a k
i
~
L
\\
x e
1 a
n io
\\
m r
t i
o E n
\\
t o
0 n
s e
it n
0 e
k e
i a c 0
m L e
/
V sn Ar a
n i
/
0 n
o C
o n
/
1 i
p p
o
/
ll a
t t
s Om n C I?
n o
/n
=
e e
o R
it t
u
\\-
c L
c o
\\
e o
e h
\\
R r
t n
sC
\\u n
it u
A s
o w
\\'.
)
W c
S s
o e C e
e U
S
\\a P s h
r 0 s n
P 2
it
.I!l.:
0
(
N
.ii 1
t t n w
a ii; li:l 0 e s
e n
,F r
e no w
g m
u ug m
ep S
e s
i n
D r
T e
i ms F
F C
v i
0 a
n n
e l
N~
e g 9 t
g R D a 5 i i
i o
s C
a it p
A nr In t e 8
7 6
C ou s s s s 0
3 O
L Cs e e 0
s i
r r
s P P 1
e ts e T T ru o
r M M s
P P
T T se C C J.
rp t
n e
=
m n
i 0
a
/
1 t
/
n
/,,
li' co i
k
/ /
a fi e
p
/
d r.
e
/
t a
f l
u I
c 1
la O
5 0
5 0
5 0
5 0
5 0
c 80
.o 4
4 3
3 2
2 1
1 e
h 0
T 0
3~g e=aoE EoEE 5oo 66 e
t 91 o
Re N
g Aa LP
0000003 0000052 e
irE e
iI' ka 0
e 0
s L
0 n
0 o
o 0
p t
0 s
2 e
n R
o i
e et ru r c ue tar t n
)
e a
0 c p
r n 0 e m
eo 0 s 3
(
e 0
r e
pC 0 e u
T mt 5 m g
y 1
i F
a e u T
i b
To e
h ro St i
F H W A
U 0
0 C
F 0
O
~
0 L
g 00 e
t 1
so D
s-P 0
9 la i
0 w
t 0
in 00 N
i 05 l'
O 0
5 0
5 0
5 0
5 0
2 1
1 0
0 9
9 8
8 1
1 1
1 1
8 Cc$o I3Ee$ $ g$.}a3 n
000-67 91 Re g Aa LP
-