ML20195E224
| ML20195E224 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/20/1988 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20195E221 | List: |
| References | |
| NUDOCS 8806230328 | |
| Download: ML20195E224 (60) | |
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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE i SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 , (TVA-SQN-TS-88-21) e LIST OF AFFECTED PAGES h Unit 1 3/4 7-14 . ,
- B 3/4 7-4 Unit 2
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3/4 7-14 3 3/4 7-4 s i u 9 b
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1 1 I g a PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7. The ultimate heat sink shall be OPERABLE with the average temper ce of wate t the ERCW system suction of less than or equal to 83*F. ADDLICABILITY: ES 1, 2, 3 and 4 - ACTION: With the average temperature of t water the ERCW system suction greater than 83'F be in at least HOT STANDBY - in 6 hours and in COLD SHUTDOWN within the following 30 hours. R12 SU.VE!LLANCE RE0VIRMENT . 4.7.5 Th ltimate heat sink shall be determined OPERABLE at lea once per 24 ne by verifying the average temperature to be within its limits. bekb en be psai/$a 4n esNel ref /ace w;f')( pey;,,) ytcikca-lo n 3l9' ?, 5 l I
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I 1 A;;ndscnt S- i SEQUOYAH - UNIT 1 3/47-1.4 7 45/S { 1 I l l l l l
kwhd J oro'8dl>., f J/% 7I* t, . PLANT SYSTEMS l 3/4.7.5 ULT! MATE HEAT SINK e LIMITING CCND! TION FOR OPERATION 3.7.5 The ultimate heat sink shall be OPERABLE with:
- a. A minimum water level at or above elevation 6 70 Hean Sea Level,
' USGS catu:n, and
- b. An averagea water temcerature of less than or ecual to 9%I'F.
ERcW supp/g Ar*4-APoLICABILIT'f: MODES 1, 2, 3, and 4 ACTION: Wita One recuirements of the acove scecification not satisfiec, De in at least HOT STANCSY witnin 6 hours and in COLD SHUT 00'nN within ne following 30 nours. SURVEILLANCE REOUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours by verifying the averagegater temperature and water level to be within their limits. jed - E/c W iy,dy 4
PLANT SYSTEMS BASES 3/4.7.5 ULTIMATE HEAT SINK UH.S leve/ a d The limitations on' temperature ensure that sufficient cooling capacity is l R12 available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits. The limitations on the maximum temperature are ' based on providing a 30 day cooling water supply to safety related equipment without exceeding their design l R12 basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants", March 1974. ZNJERT Mfsde/ /%yyA 3/4.7.6 FLOOD PROTECTION The requirements for flood protection ensures that facility protective actions will be taken and operation will be terminated in the event of flood conditions. A Stage 1 flood warning is issued when the water in the forebay is predicted to exceed 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 Feet Mean Sea Level USGS datum during April 15 through September 30. A Stage 11 flood warning is issued when the water in the forebay is predicted to exceed 703 feet Mean Sea Level USGS datum. A maximum allowed water level of 703 Mean Sea Level USGS datum provides sufficient margin to ensure waves due to high winds cannot disrupt the flood mode preparation. A Stage I or Stage II flood warning requires the imple-mentation of procedures which include plant shutdown. Further, in the event of a loss of communications simultaneous with a critical combination flood, heacwaters, and/or seismically induced dam failure the plant will be sr.utdown and flood protection measures implemented. 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel nccupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50. ANSI N510-1975 will oe used as a procedural guide for surveillance testing. SEQUOYAH - UNIT 1 B 3/4 7-4 -Amendment No. 8
Insert for Bases Section 3/4.1.5 l The limitations on minimum water level are based on providing sufficient flow to the ERCW serviced heat loads after a postulated event assumlns a time-dependent drawdown of reservoir level. Flow to the major transient heat loads (CCS and CC heat exchangers) is balanced assuming a reservoir level of elevation 670 The time-independent heat loads (ESF room coolers, etc.) are balanced assuming a reservoir level of ulevation 636. 1
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PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 e ultimate heat s:nk shall be OPERABLE with the average temper ure of water the ERCW system suction of less than or equal to 83*F. APPLICABI,LITY: MOD , 2, 3 and 4. ACTION: With the average temperature of the wa at the ERCW system suction greater than 83*F be in at least HOT STANDBY ithi 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIRM. TS 4.7.5 e ultimate heat sink shall be determined OPERABLE at least once r 24 urs by verifying the average temperature to be within its limits. deb enb y<,,A4- ,, A , y 6 4 ,, M ' ,,,;, ,/ ypaa.edL, d>- ap4<z.r . l 1
- I SEQUOYAH - UNIT 2 3/4 7-14
kvhecl-pSid!a-/l6., .3/V. 7 T PLANT SYSTEMS 3/4.7.5 ULT! MATE HEAT SINK t LIMITING COND! TION FOR OPERATION , 3.7.5 The ultimate heat sink shall be OPERABLE with:
- a. A minimum water level at or above elevation 670 Mean Sea Level, '
'USGS catum, and
- b. An average 3 water temcerature of less than or ecual to 848'F.
ERed <~y/s had< APDLICAE! LIT!: MCOES 1, 2, 3, and 4 ACT:CN: Wita the recuirements of the acove scecification not satisfiec, be in at least HOT STANDSY witnin 6 hours and in COLD SHUTDOWN witnin :ne following 30 hours. SURVE:LUNCE RECUIREMENTS 1 a.7.5 The ultimate heat sink shall be determinec OPERABLE at least once per 24 hours by verifying the averagegater temperature and water level to be within :neir limits. E,9cw .typ/y b M W*" I l . I I i 1 1 l l 1 4 ,i h
, PLANT SYSTEMS
- BASES 3/4.7.5 ULTIMATE HEAT SINK
/wI ed The limitations on the ultimate heat sink 4 temperature ensures that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.
The limitation on maximum temperature is based on providing a 30 day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants", March 1974. 3MERT AMd /myyA 3/4.7.6 FLC00 PROTECTION The requirements for flood protection ensures that facility protective
. actions will be taken and operation will be terminated in the event of flood conditions. A Stage 1 flood warning is issued when the water in the forebay is predicted to exceed 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 Feet Mean Sea Level USGS datum during April 15 through Sectember 30. A Stage 11 flood warning is issued when the water in the forebay is predicted to exceed 703 feet Mean Sea Level USGS datum. A maximum allowed water level of 703 Mean Sea Level USGS datum provides sufficient margin to ensure waves due to high winds cannot di.r upt the flood mode preparation. A Stage I or Stage 11 flood warning requires the implementation of procedures wnich include plant shutdown. Further, in the event of a loss of communications simultaneous with a critical combination flood, headwaters, and/or seismically inouced dam failure the plant will be shutdown and flood protection measures implemented.
3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of
- this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the recuirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.
ANSI N510-1975 will be used as a procedural guidc for surveillance testing. SEQUOYAH - UNIT 2 8 3/4 7-4
Insert for Bases section 3/4.7.5 The limitations on minimum water level are based on providing sufficient flow to the ERCW serviced heat loads after a postulated event assuming a time dependent drawdown of reservoir level. Flow to the major transient heat loads (CCS and CS heat exchangers) is balanced assuming a reservoir level of el. 670. The time independent heat loads (ESF room coolers, etc.) are balanced assuming a reservoir level of el. 636.
ENCLOSURE 2 PROPOSED TECHNICAL SPcCIFICATION CHANGE
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I SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (IVA-SQN-TS-88-21) r DESCRIPTION AND JUSTIFICATION FOR AMENDING SPECIFICATION 3/4.7.5 "ULTIMATE HEAT SINK" a
O 5 ' ENCLOSURE 2 1 1.0 Description of Change Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN) units 1 and 2 technical specifications to revise limiting condition for operation (LCO) 3.7.5 to increase the maximum allowable ultimate heat sink (UHS) temperature from 83 degrees Fahrenheit (F) to 84.5 degrees F. A minimum water-level requirement is added to LCO 3.7.5 by the proposed change. The wording of LCO 3.7.5 and surveillance requirement (SR) 4.7.5 is modified to clearly specify that the UHS temperature limit applies to the essential raw cooling water (ERCW) supply water temperature. The action statement and surveillance requirements are modified to be consistent with the addition of an LCO for reservoir level. The bases for technical specification 3.7.5 are modified to reflect all these changes. 2.0 Reason for Change The Tennessee River watershed is currently experiencing an extended drought. Figure i shows that, over the last four years, total rainfall is over 60 inches below normal in the vicinity of SQN. The lack of rainfall has significantly decreased the upstream cold water reservoir volumes. This results in decreased river flowrates and increased river temperatures at the site. Current projections indicate that a maximum river temperature of ; 84.4 degrees F may be reached this summer at the site. Final Safety ' Analysis Report (FSAR) section 9.2.5 identifies the Tennessee River as SQN's UHS. The Chickamauga Dam, located downstream of SQN, forms a reservoir at the site. Therefore, if the temperature of the Tennessee River exceeds 83 degrees F. LCO 3.7.5 would require a forced unit shutdown. TVA has evaluated the effects of incicasing the maximum UHS temperature above 83 degrees F. This evaluation has determined that increasing the maximum UHS temperature to 84.5 degrees F is acceptable. The 84.5-degree-F temperature limit should minimize the potential of a forced shutdown as a result of exceeding the UHS limit. l The minimum water-level requirement is added to the technical specification in order to support flow balance test changes that take credit for the dynamics of reservoir drawdown af ter an assumed loss of downstream dam coincident with a loss of coolant accident (LOCA). These changes have been made to provide additional margin for available ERCW pressure and flows. The action statements and surveillance requirements were modified to be consistent with the standard technical specifications and the revised limiting conditions for operability. Provisions were added to address remedial actions and surveillance requirements for the reservoir water level.
3.0 Background Information The UHS is the combination of water sources and associated structures used to remove waste heat frem the plant. The UHS reservoir for SQN is the Tennessee River. The UHS has been previously evaluated by TVA and NRC. The design basis of the UHS is described in FSAR section 9.2.5. The effects of potential low flow and low water-level conditiona have also been evaluated and described in FSAR section 2.4.11. NRC review of this information was documented in the safety evaluation report for SQN (NUREG-0011, March 1979) in sections 2.4.3 and 9.2.3. The 83-degree-F temperature limit was developed during the initial design stage of SQN. It was based on the maximum temperature found during a review of 12 years of river temperature data that spanned a 25-year period. All other annual maximum temperatures were below 80 degrees F. TVA used the 83-degree-F temperature as a design input. Consistent with this assumption, the 83-degree-F temperature also became the technical specification limit. Another initial design assumption was the instantaneous loss of downstream dam coincident with a LOCA. In effect, the design of the ERCW system was based on an assumed reservoir level of elevation 636 feet mean sea level (MSL). This elevation is associated with minimum river flows in the original Tennessee riverbed. As such, no technical specification was required for reservoir level. The SQN design utilizes the UHS as the water supply for the ERCW system. ERCW is described in FSAR section 9.2.2. The ERCW system is the qualified raw cooling water system for SQN. The system services essential heat loads that are necessary for both normal operation and accident mitigation. These heat loads include the containment spray (CS) heat exchangers, component cooling system (CCS) heat exchangers, and various attendant equipment, such as the diesel generator engine coolers and the engineered safety feature (ESP) space coolers. An increase in the UHS temperature will potentially affect the performance of these heat exchangers. FSAR figure 9.2.2-21 indicates that the major accident mitigation heat loads serviced by ERCW are time dependent. As seen in the figure, the peak heat rejection rates are of relatively short duration, and the average rates are significantly less than the peaks. The time-dependent decay cf the heat rejection rates allows for ERCW flow changes with time without the loss of capability of adequate heat rejections. The heat loads rejected to the UHS under postulated accident conditiona are bounding for a normal plant cooldown. All of the stored and decay energy is released to the containment and ultimately rejected to the UHS for the worst-case ac cident scenario. Auxiliary feedwater, however, is used to remove a s gnificant portion of the stored and decay energy for a normal plant cooldown. The heat load rejected to the UHS is not significantly increased until the residual heat removal (RHR) system is placed in service.
4.0 Justification for UHS Temperature Change New Analytical Assumption The proposed change to the UHS specification was evaluated, and the following changes were made to the FSAR analyses:
- 1. The UHS temperature was assumed to be at least 84.5 degrees F for the duration of the event. This is 1.5 degrees F above the value currently assumed in FSAR analyses.
- 2. The ERCW flow rates and overall heat coefficient for the CCS and CS heat exchangers have been revised to reflect actual as-built plant conditions. The new values used in the analyses are listed in table 1.
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These new assumptions ensure that plant performtnce during the postulated events will be bounded by the revised analyses. New Test Method The ERCW flow balance test method will be changed. The flow rates to the CS and CCS heat exchangers will be established assuming a reservoir level of elevation 670. The remaining heat load flows will still be established assumlng a reservoir level of elevation 636. This allows the flow rates to the major time-dependent accident I mitigation heat loads to take advantage of the additional reservoir l 1evels present early in the assumed scenario. l Analysis Methods ! The change in the UHS temperature potentially affects the heat removal rate from many areas of the plant. In order to demonstrate the acceptability of the proposed change. TVA has performed an evaluation of the effect of the increased temperature on the following key plant analyses: Emergency core cooling system (ECCS) Other FSAR chapter 15 accident analyses Containment subcompartment pressure analysis Peak containment temperature Peak containment pressure Long-term containment cooling Long-term cooling for pipe breaks outside containment Equipment qualification (EQ) temperature profiles ; ECCS Analysis The primary function of the ECCS is to cool the reactor core by removing stored and fission product decay heat from the reactor core so that fuel rod damage remains within prescribed limits. The requirements for ECCS evaluation models are described in 10 CFR 50 I Appendix K in subsections (1) Sources of Heat During a LOCA; J 4 _ , , - - _ _ - - - - -, ,
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(2) Swelling and Rupture of the Cladding and Fuel Rod Thermal Parameters; (3) Blowdown Phenomena; and (4) Post-Blowdown Phenomena, Heat Removed by the ECCS. The FSAR analyses, which demonstrate compliance to the requirements of 10 CFR 50.46, show the peak clad temperatures and core reflood/ quenching occur many minutes before any heat removal from the core to the UHS begins. The peak cladding temperature occurs at approximately 180 seconds into the ECCS event (for a discharge coefficient CD = 0.4), and core reflood is completed around 500 seconds. Heat removal to the UHS does not occur until switchover of the RHR to the emergency sump at approximately 1600 seconds. Therefore, because the parameters that demonstrate compliance to 10 CFR 50.46 are not affected by an increase in the UHS temperature, the FSAR ECCS analyses (FSAR sections 15.3.1 and 15.4.1) will not be changed. Other FSAR Chapter 15 Analyses The remaining FSAR analyses for Condition III and IV faults address transients and accidents that may cause core overcooling or overheating from reductions in shutdown margin, excessive or insufficient heat removal, or loss of or change in forced reactor coolant system (RCS) flow. In addition, only radiological i consequence events are addressed. (Note: Condition I and II events i are not addressed because these conditions represent either normal operation or operational transients or faults of moderate frequency that at worst result in reactor shutdown with the plant being capable of returning to operation.) The other events addressed in the FSAR are:
- 1. Major or minor secondary system ruptures (FSAR sections 15.3.2 and 15.4.2) l 2. Complete loss of forced RCS flow or single reactor locked rotor (FSAR sections 15.3.4 and 15.4.4)
- 3. Rod cluster withdrawal at full power (FSAR section 15.3.6) l
- 4. Rod cluster control assembly ejection (FSAR section 15.4.6) '
- 5. Steam generator tube rupture (FSAR section 15.4.3)
- 6. Fuel handling accident (FSAR section 15.4.5)
- 7. Waste gas decay tank rupture (FSAR section 15.3.5)
- 8. Inadvertent loading of a fuel assembly into an improper location (FSAR section 15.3.3)
The first four events listed above do not depend upon heat removal to the UHS for mitigation of the consequences that occur early in the event. Therefore, FSAR analyses for these events will not be altered by the proposed change. In a similar manner, the consequences associated with a steam generator tube rupture (SGTR) will not be altered by the proposed change. However, the last mitigative action item listed for the operator in the FSAR analysis for an SGTR is initiation of RHR for cooldown. The RHR heat exchanger does transfer its heat load to the UHS via the CCS. Therefore, cooldown of the RCS may be slightly extended. The extended cooldown does not represent 4
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l l r any unacceptable consequences. The consequences of the waste gas decay and fuel handling accident are radiological and cannot be affected by the proposed change. The inadvertent loading of the fuel assembly into an improper location does not impact heat transfer to the UHS. Subcompartment Pressure Analysis The peak subcompartment pressures given in the FSAR will not change because of an increase in UHS temperature. In order to maximize pressure, the subcompartment pressure analyses in FSAR section 6.2 assume an instantaneous, double-ended guillotine rupture of the largest pipe within a given subcompartment. The resulting flow because of the rapid depressurization of the pipe or system produces the peak subcompartment pressure in a matter of seconds. No heat removal to the UHS is assumed in the FSAR analyses. Therefore, they are unaffected by changes in UHS temperature. Peak Containment Temperatures The peak containment temperature results from a main steam line break (MSLB) and occurs very early in the transient during blowdown from the faulted steam generator. During this period, increases in containment temperature and pressure are mitigated by the ice condenser, the CS, and passive heat sinks. The CS system is supplied with constant temperature water from the refueling water storage tank (RWST), which is assumed to be at the maximum technical specification temperature, without any heat removal by the CS heat exchanger. The mass and energy releases from the faulted steam generator to the containment are terminated by steam generator dryout within 30 minutes (even for small breaks). The ice bed does not melt out until many hours after an MSLB and continues to remove energy from the containment. By the time switchover of the CS system to the emergency sump occurs and heat removal to the UHS begins, temperatures in containment have been decreased substantially because of heat removal from flow through the ice condenser caused by the air return fans. Thus, peak containment temperatures will not be affected by the proposed change because heat is not transferred to the UHS during the time of peak containment temperature. Peak Containment Pressure The peak containment pressure is a result of a large-break LOCA. During a large-break LOCA, heat transfer from containment to the UHS begins at approximately 1600 seconds via the coupled RHR and CCS heat 4
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j i exchangers*. Thus, the containment temperatures and pressures predicted in the FSAR analysis for a design basis LOCA before 1600 seconds will not change regardless of the UHS temperature. This was confirmed by TVA analyses using the MONSTER containment analysis code. (Documentation of the MONSTER code was submitted to the NRC as : a topical report by letter from J. A. Domer to Dr. C. O. Thomas dated July 13, 1985.) l After RHR switchover to the emergency sump at 1600 seconds, the temperature of the injection flow to the core is affected by the proposed change. The increased core inlet temperature would result in a slight increase in mass release to containment from core ! boiloff. However, containment conditions are still controlled by the l CS system, air return fan, and ice condenser; and the design basis ' containment analysis presented in the FSAR is not significantly impacted. The CS heat exchanger system begins to transfer its heat to the UHS at approximately 2800 seconds following switchover from the RWST to the emergency sump. But until ice bed meltout at approximately 3000 seconds, switchover of the containment spray does I not appreciably change the FSAR analysis. Following ice bed meltout, l the containment pressure and temperature in ths containment begin to I increase noticeably. At 3600 seconds, the RHR spray is initiated to increase the total containment heat removal capability. The containment pressure continues to increase until the heat removal to the UHS via the RHR and CS heat exchangers and through the containment shell exceeds heat addition to the containment j atmosphere. In order to quantify the effect on containment peak pressure because of a change in the UHS temperature, TVA has l performed a series of containment analyses using a MONSTER model l benchmarked against the FSAR design basis analysis. l The benchmark analysis was performed using the FSAR chapter 6 model and associated data (i.e., volumes, flow paths, ice weight, and heat ' sinks), blowdowns, pump flows, and UHS temperature of 83 degrees F. The model is documented in TVA calculation TI-ANL-104. The heat exchanger parameters from FSAR section 9.2.2.2 were used in this analysis for the CS, CCS, and RHR heat exchangers. It should be noted that the FSAR analysis modeled the CS heat exchanger as a U-tube, one-shell pass and two-tube, pass-type heat exchanger with an ) overall heat transfer coefficient of 3.125x106 (based on Figure l 10-8 of Heat Transfer by J. P. Holman, Fourth Edition). The CS heat I exchangers currently in use at SQN are a U-tube, two-shell pass and two-tube, pass-type heat exchangers that are strictly countercurrent. l The overall heat transfer coef ficients for a heat exchanger with the design used in the FSAR analysis would have the same performance as the heat exchanger design currently in use at SQN with an overall heat transfer coefficient of 2.44x106. Figures 2 and 3 show the upper and lower compartment pressure profiles from the benchmark analysis. A peak pressure of 11.03 pounds per sqaure inch gauge (1b/in 2g) was obtained for the lower compartment. The FSAR peak pressure is 11.09 lb/in g2 calculated with the LOTIC containment code. The difference between the two codes is less than 1 percent, which is an excellent comparison. Using 85 degrees F as
-7 t s the UHS temperature increased the peak lower compartment pressure by approximately 0.13 lb/in g.2 Figures 4 and 5 show the upper and lower pressure profiles for this study.
The distribution of flows in the ERCW has been revised since the FSAR analysis wae performed. A new containment analysis was performed using the FSAR model but with measured flow rates from surveillance instruction (SI) 566 and revised heat exchanger coefficients for CCS, , CS, and RHR heat exchangers. The containment analysis shows that the pressure profile predicted by the analysis (peak pressure 10.91 lb/in g) 2 was bounded by the design basis FSAR analysis. Figures 6 and 7 show the upper and lower compartment pressure profiles for this analysis. In order to ensure that the heat removal from containment is conservatively modeled, the ERCW flow rates to the heat exchangers were reduced by 10 percent. The reduced flow rates for the RHR, CS, and CCS are provided on Table 1. A containment LOCA analysis using the revised ERCW flows (with a 10-percent margin reduction), 83-degree F UHS temperature, and corresponding duties was performed. The peak pressure for the lower compartment in this analysis was 11.36 lb/in28 Figures 8 and 9 show, respectively, the upper and lower pressure profiles for this + analysis. A parametric study of this model, using an 85-degree-F UHS temperature, increased the peak lower compartment pressure by approximately 0.14 lb/in2 8 to 11.50 lb/in Zg. Figures 10 and 11 show, respectively, the upper and lower pressure profiles for this study. The results of there analyses, along with the results of the independent verification analyses performed by Westinghouse Electric Corporation with the LOTIC code, are tabulated in Table 2. The data in Table 2 shows that the design basis LOCA analysis with an 85-degree-F ultimate sink temperature does not exceed the containment design pressure of 12 lb/in 28 Thus, the peak containment pressure will not be unacceptably increased by the proposed change. Long-Term Containment Cooling Long-term cooldown involved in recovery from a postulated accident , scenario or normal cooldovn with the RHR system will involve heat transfer to the UHS. Any increase in the UHS temperature will decrease the rate of cooldown. Because heat is transferred to the UHS via heat exchangers with given duties based on specifled temperature differentials, heat removal would not be changed if the source temperature (i.e., CS and RHR and ultimately the containment atmosphere) increased by the same amount as the sink temperature (assuming that the heat exchangers are 100-percent efficient). Therefore, for an increase of 1.5 degrees F in the UHS temperature, an increase of 1.5 degrees F can be expected (in the limit) in the source temperature profile. Therefore, using this rationale, the lower compartment coolers that are initiated within four hours after a postulated MSLB would remove loss energy from containment than predicted by the current analysis yielding a containment temperature increase of 1.5 degrees F. (The lower compartment coolers are more efficient than assumed in design calculations, and the long-term
_e_ temperature should remain unchanged for this event. The cooler performance is verified in TVA calculation SQNAPS4-008.) Also, during a design basis LOCA, the CS and RHR heat exchangers would not remove the energy predicted in the FSAR analysis unless the containment tempcrature eventually increased in the limit by 1.5 degrees F. Therefore, the long-term containment temperature profile can be expected to increase by a maximum of 1.5 degrees F as a result of the proposed change. The increased temperature will extend the predicted amount of time required to reach normal temperature provided that the elevated temperatures exist for the entire cooldown period. Historical data on river water temperature at the ERCW pump intake indicates that this is extremely unlikely (see figure 12). The long-term containment cooldown will also be affected by the postulated loss of downstream dam assumed concurrent with the design basis LOCA. The postulated dam failure will result in a reduction in the total flow capacity of the ERCW system. The total ERCW flow capacity at normal reservoir levels (see FSAR figure 2.4.1-3, sheet 1 of 14) is greater than design flows used in the containment analyses and will remain so for approximately 10 hours after the postulated LOCA and dam failure. After approximately 10 hours, the reservoir level will drop below the 670-foot level used to determine ERCW design flow rates for the various analyses. The reservoir level will stabilize at minimum level in approximately two days, causing a 7-percent reduction in the total ERCW flow rate. The 7-percent flow reduction would cause a decrease in the heat removal capability of . the ERCW system. The decreased heat removal would result in increases on the order of 3 degrees F in the long-term containment temperature after two days. Therefore, in conjunction with the increased river water temperature, the long-term temperature inside containment would increase by no more than 4.5 degrees F starting at two days and continuing through the remaining duration of the accident. The increased long-term containment temperature will affect the qualified postaccident degradation equivalency calculations for 10 CFR 50.49 equipment, This effect is addressed in the section on EQ Temperature Profiles. No other parameters are affected by the increased long-term containment temperatures. Long-Term Cooling for Pipe Breaks Outside Containment long-term cooling for pipe breaks outside containment is affected by an i.' creased UHS temperature because the UHS serves as the cooling water au,a,-1 'or ESF room coolers. Cooling water temperature dominates the performance of the room coolers. The performance of the ESF room coolers was modeled assuming a maximum UHS temperature of 84.5 degrees F. The evaluation was to determine if the coolers would maintain theit itspective areas at or below the 100-day postaccident average EQ temperature. Figure 13 shows a typical profile of the time-varying river and room temperatures. This profile is based on a 10-year average profile normalized to the worst-case UHS temperature of 84.5 degrees F. A typical room temperature profile is also shown against the 100-day EQ temperature,
, 9_
which is assumed to remain constant. Evaluations of the profiles indicate that the 100-day average temperature profiles are not exceeded. Additional degradation to equipment resulting from the deviations from the average EQ temperature profile will be reflected in the EQ binders on a case-by-case basis. In a few cases, small increases in ERCW flow (approximately 3 gallons per minute) were required to Jatisfy the average temperature limit. ' EQ Temperature Profile The EQ temperature profile conservatively represents the worst possible containment accident temperatures versus time. The EQ profile is a composite profile constructed from a short-term MSLB profile and a long-term LOCA profile. As discussed previously (see peak containment temperature), the short-term MSLB temperatures are unaffected by the proposed increase in the UHS temperature. Therefore, the short-term EQ profile is still valid and bounding. The proposed increase will result in a corresponding temperature increase (1.5 degrees F in the limit) in the long-term LOCA/EQ profile provided that the river water temperature remained at 84.E degrees F for the entire cooldown period. Historical data of ri; I water intake temperature at SQN shows temperatures within 2 degrees F or closer of the present technical specification limit occur very rarely and generally only last no more than a week (see figure 12). Therefore, based upon this historical data, the long-term UHS temperature will not remain at 84.5 degrees F during a l 30-day plant cooldown period following an accident. The most likely scenario is that the average long-term containment temperature during the 30 days following an accident will be less than currently assumed. As a result, no impact on established operability of the plant's safety-related equipment contained in the qualification documentation should occur. Therefore, the plant's ability to safely cool down to normal temperatures within 30 days should not be impacted. The ef fects of a postulated loss of downstream dam were not included in the preceding discussion. As indicated previously in the section on long-term containment cooling, the increast in containment temperature could be as much as 4.5 degrees F two days after a LOCA from the combined effects of the proposed 1.5-degree-F increase in ERCW temperature and the loss of downstream dam. TVA has performed an evaluation of the impact of P.his increased temperature on postaccident degradation equivalency calculations for the plant's 10 CFR 50.49 equipment. The evaluation determined that no significant impact on the ability of the safety-related equipment to operatc as required during cooldown will occur as a result of the increased long-term containment temperature. In summary, the evaluation performed by TVA indicates that increasing ' the maximum UHS temperature does impact the peak containment pressure analysis and long-term cooling for pipe breaks outside containment. These impacts, however, are neither significant nor detrimental to the plant for the proposed increase in the UHS temperature limit from 83 degrees F to 84.5 degrees F.
5.0 Justification for Reservoir Elevation In accoYd ance with Regulatory Guide (RG) 1.27, "Ultimate Heat Sink for Nuclear ?cwer Plants," a loss of the downstream dam is postulated to occur simultaneously with a LOCA. TVA previously used a simple but very conservative assumption of an instantaneous drop in reservoir level coincident with the loss of downstream dam. It is easily realized that ~.he decrease in reservoir level is a time-dependent function. The use of a time-dependent model of a loss of dam is also allowed by RG 1.27. TVA has chosen to model the time-dependent drawdown of the reservoir. The analytical method for determining the time-dependent drawdown of the reservoir is included as attachment 1. The analysis is based on an assumed 1000-foot breach of the Chickamauga Dam. The initial reservoir level is assumed to be the minimum normal reservoir level expected during the hot summer months as shown in FSAR figure 2.4.1-3 (sheet 1 of 14). The calculated time-varying reservoir level is shown on figure 14. It should be noted that the minimum possible reservoir level of elevation 636 is only approached after several days and then only if river flow is at an extreme minimum. The heat load that is ultimately removed by the ERCW is composed of both time-dependent and time-independent loads. As shown on FSAR figure 9.2.2-21, the major time-dependent loads are the CS and CCS heat exchangers. The time-dependent loads associated with the CS and CCS are based on decay heat and are included in the design basis FSAR - analyses and in the analyses previously described in the discussion of peak containment pressure (see section 4). The time-independent loads consist of ESF room coolers and other station auxiliaries. The diesel generator loads are also modeled as time-independent because their load profile depends on operator action to remove loads that are no longer required and are, therefore, not as predictable. The design basis LOCA analysis and the analyses described in section 4.0 assume a constant flow rate to the CS and CCS heat exchangers to simplify the calculations. The heat exchanger flow rates used in these analyses are based upon system flow balancing for a reservoir level of 670 feet. As the reservoir level drops below this elevation, a reduction in t'ae heat removal capacity of the ERCW system will occur for the CS and CCS loads. The ERCW flow rates to the time-independent loads are based on a reservoir level of elevation 636, the minimum river level after a loss of downstream dam. Therefore, the ERCW flow rates for heat removal of time-independent loads will not be affected. As shown on figure 14, reservoir drawdown to elevation 670 will not occur earlier than 10 hours after the initiation of the postulated event. The peak time-dependent heat loads occur approximately tnree hours into the scenario, as shown on FSAR figure 9.2.2-21. A margin of 10 feet on reservoir elevation exists between the normal minimum elevation (680 feet) and the elevation used for flow balancing
i (670 feet). Hence, there are several hours between the peak heat , loads and the drawdown to the flow balance elevation. Therefore, the ; heat removal assumed in the design basis analysis is not affected by l the postulated dam failure during this period. i The minimum reservoir elevation 636 feet is reached no earlier than two days after the postulated event (see figure 14). The change in total system flow because of this decrease in elevation is on the l order of 7 percent. The decreased heat capacity of the ERCW because , of loss of downstream dam would result in temperature increases on i the order of 3 degrees F in the long-term containment temperature. Therefore, in conjunction with the river water temperature increase, ! the long-term temperature inside containment would increase by no ! more than 4.5 degrees F starting at two days after the postulated j accident. The increased containment temperature will affect the i qualified postaccident degradation equivalency calculations for 10 CFR 50.49 equipment. This effect was found to have no significant impact on the ability of the plant to safely cool down (see , section 4.0, "EQ Temperature Profiles"). l l As previously discussed, the ERCW can remove both time-dependent and time-independent heat loads without impacting the ability of the ' plant to effect a safe, long-term cooldown following a design basis event. The adoption of an LCO for a reservoir level of elevation 670 MSL is consistent with this conclusion. In addition, a review of FSAR section 2.4.11.3 and FSAR figure 2.4.1-3 (sheet 1 of 14) indicates that elevation 670 is a reservoir level not expected to occur during any expected mode of river operation. In fact, the 4 reservoir level has never dropped below elevation 673 since Chickamauga Dam was closed. A reservoir level of elevation 670 is indicative of an extremely unusual situation (i.e., loss of downstream dam). The incorporation of an LCO for a reservoir level of elevation 670 is appropriate because it would require a plant shutdown in the event that the reservoir is affected by a loss of downstream dam. This action ensures that the analytical basis for the ERCW heat loads is preserved. 6.0 Justification for Other Administrative Changes i The LCO and the SR are modified to clearly indicate that the temperature measurement is made of the ERCW supply header water. The temperature measurement is currently made with instrumentation installed in the ERCW intake station. The temperature measurement is appropriate because it represents the temperature of the water going to the critical heat exchangers.
]
j The action statement and SR are modified to include provisions for reservoir levels. These changes are consistent with the addition of j an LCO for reservoir level. The changes are also consistent with the
; h7C standard technical specification for the UHS.
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0 . TABLE 1 Revised Heat Exchanger Flow Rates and Overall Heat Transfer Coefficient HEAT EXCHANGER L:,0W RATES HEAT XCHR UA (Gallons Per Minute) x100 RHR 1 & 2 A&B Tube Side 3074 1.539 Shell Side 5000 CS 1A ' Tube Side 4650 3.092 Shell Side 3708 CS 1B Tube Side 4750 1x112 Shell Side 3776 CS 2A Tube Side 4750 3.123 Shell Side 3744 CS 2B Tube Side 4750 3.078 Shell Side M j CCS A Tube Side 1224 2.845 ' Shell Side 5000 ) CCS B Tube Side 4284 1322] Shell Side 5000 CCS C Tube Side 4284 2.802 Shell Side 5000 1 NOTE: Underlined values were used in the revised containment analysis. l
e e TABLE 2 Results of Containment Analysis Peak Pressure FSAR NEW PRESSURE PRESSURE (psig) (psig) M 83*F ERCW 11.03 11.36 85'F ERCW 11.16 11.50 WESTINGHOUSE 83*F ERCW 11.09 11.19 85'F ERCW 11.25 11.31
. TUB 40103 9 49) (J 82-Al Embn -t h6&J1 UNITED STATES GOVERN 5 TENT Memorandum TENNESSEE VALLEY AUTHORITY 0 95 TO . Mark Burzgnski, Supervisor, Regulatory Licensing,jSequoyah Nuclear Plant FR05! - Donald W. Newton, Supervisor, Hydrology Section, Flood Protection Branch, 200 LB-K DATE : June 6, 1988
SUBJECT:
SEQUOYAH NUCLEAR PLANT (SNP) - RESERVOIR RECESSION CURVES - 20STULATED ' FAILURE OF CHICKAMAUGA DAM This provides the reservo'ir recession curves for your use in requesting a revision to the Technical Specification Limit for ERCW Reservoir Tempera-ture and the supporting data requested by Rex Wescott of NRC. Attachment 1 is the recession curve for Chickamauga Reservoir at the Sequoyah plant intake (Tennessee River mile 484.7) assuming a nonflood failure of Chickamauga Dam. The assumed initial pool level was eleva-tion 681, the level expected to be maintained this year, and the normal August-oaptember level as shown on Attachment 2, the seasonal guide curve for the Chickamauga project. Inflows were varied between 6,000 and 30,000 cfs. Those flows cover the range of inflows from the minimum required to meet TVA commitments to maintain low flow at Chattanooga up to the maximum normally used in generating power. Attachment 3 provides the general plan, elevation, and sections for Chickamauga. For our ai.alysis we assumed a breach size based on the relationships developed by Thomas C. MacDonald and Jennifer Langridge-Monopolis and I published in ASCE's Journal of Hydraulic Engineering, Volume 110, No. 5, ) May 1984 The size of breach was estimated to be 1,000-feet wide with vertical side slopes and bottom at elevation 630. The discharge rating curve shown on Attachment 4 was developed using the open channel i contraction discharge equation Q = CA 2g( oh+ag v[-h ) with C assumed to l f 2 be 0.7 and the term a}V
-h g toke 0. Attachment 5 W
provides the tailwater rating curve and Attachment 6 provides the area-volume relationships for Chickamauga. The surface elevation of tho, Tennessee River at the Sequoyah site following dam failure was computed using TVA's Simulated Open Channel Hydraulics computer model. The control section is a high point at elevation 632 in the river channel at Tennessee River mile 480, 4.7 miles downstream of Sequoyah Nuclear Plant and 9 miles upstream of Chickamauga Dam. Attachment 7 shows the streambed profile for the . Tennessee River, upstream of Chic;kamausgg,g JUN 71988 LREGULATORY UCENSING ir
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2 Mark Burzynski June 6, 1988 SEQUOYAH NUCLEAR PLANT (SNP) - RESERVOIR RECESSION CURVES - POSTULATED FAILURE OF CHICKAMAUGA DAM Attachment 8 provides the results of our analysis to identify sediment concentrations and the potential for blockage of the ERCW pump intakes following a loss of Chickamauga Dam. b h
" 'Donal'd W. Newton DWN:RGL:CH Attachments Prepared by Don Newton and Ramon Lee 1447K l
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.5 ;:
5 d 630 630 620 620
} 610 0 10 0 200 300 400 610 500 600 700 600 900 1000 Vot,UME IN THOUSA'#05 CF ACRE-FEET tity ' ""'
gy AAt h 19 80 it4 7 1954 1956 1961 AC A C. F T AC rf AC rf AC.F T AC Ff 490 d e,8 00 918.000 939,000 950,000 945,000 944,000 045.44 40,800 747,000 735,000 745,000 740,000 139,000 stS 39,900 727,000 715,000 723,000 729,000 719,000 ese 32,200 347,000 $ 3 5,0 00 144,000 $ 3 9,0 0 0 $ 3 7,000 SIS 23,600 402,000 300,000 198.000 394.000 392,000 Int 18,500 260,000 250,000 263,000 283,000 241,000 tot 12,400 140,000 153,000 150,000 115,000 133,000 slo 10,$ 00 137,000 131,000 133,000 132.000 130,000 850 S,720 37,700 33,700 54,700 $ 3,6 0 0 $ 2.20 0 840 2,020 20,000 10,000 18,600 18,600 18,000 830 1,140 4,140 4,020 3,970 3,950 3,860 820 10 4 135 to e6 47 at
$17 0 0 0 0 0 0 NOTES-Aeservetr areas at etevat,en 660 and be;oe .ere n'easured on a composite rnap prepared by the Hydraulic Data Branch eith centeurs dra.n al 10' intervais. The mag was prepared from Tennessee River Survey Maps made by the U. $ Army Engineers,
. th contevra at 620, 630, 640,650 and 660. Centeurs sere made to conform !a eleveteens en TVA sedirment range cross sections located at one to five mite intervals.
Areas above elevatron 660 ere ineasured en TVA navigation traps at eleva4e 662,669,675,680,685 and 690 from TVA navigation maps adjusted to agree with the area at eievation 683 from itA land maps. Dam clesure Ja nu a ry 15,1940. The 1940 volure was comovted by the conteur seethed. Volumes of sediment en succseding dates sere cei*puted by the constant factor nethod, Elevations are referred to the USC & GS 1936 Supplerrentary Adjustment. TENNESS EE RIVER-MILE 4 71. O Area of ernqinal rner eithin reservoir.9,500 stres. crainne area at dam.20,n0 neare miles. RESERVOIR ARE AS r AND VOLUMES l','l we l we i e a i u, i =w C H IC K A M AUG A PROJECT A AUT TE,N,N..C.S
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1 CHI KAMAUGA 10iAL LIUEL STORACE IN 1000 DAY-SICOND-TEIT
==============s==2:==========,..=ss========,s:ss.:. =.z=======za. .. l' ACE 1 0F 2 ILEU. 0.00 .01
!...c=== =.....== 3============............... === ...
.02 .03 .04 .05 .06 .07 =x.........ss=====....... ass............ ass .== ..a 12"SEE .08 .09 ILEU. 0.00 .01 .02 .03 .04 .05 .06 670.0 3:33311:12E3:232=32E2==35E33 288332333=C:E232==33E3332=23E223:322EEJ: 323333332233 2333E3 .07 .08 . 0<
.1 141.20 141.30 141.40 141.49 141.59 141.69 141.79 141.89 141.99 142.09 675.0 2E133:3232323:33333:23333333333sEEEaza233EE23233332EEE2 3323333322333222333523.
.2 142.19 142.29 142.38 142.48 142.58 142.68 142.78 142.88 142.98 143.08 5.1 197.63 197.76 197.89 198.02 198.15 198.28 198.41 198.53 198.66 198.:
.3 143.18 143.28 143.38 143.48 143.58 143.68 143.78 143.88 143.98 144.08 5.2 198.92 199.05 199.18 199.31 199.44 199.57 199.70 199.83 199.96 200.4
.4 144.18 144.28 144.38 144.48 144.58 144.68 144.78 144.88 144.98 145.08 5.3 200.22 200.35 200.48 20G.61 200.74 200.87 201.01 201.13 201.27 201.
T55 145.18 145 28 145.38 145 48 145.58 145.69 145.79 145.89 145.99 146.09 201.53 201.66 201.79 201.92 202.05 202.19 202.32 202.45 202.58 202.;
.6 14EIFI4E2F1403n4GF14GFI405 I4Gn4GM4E00 14DO- T53 54 202.84 202.97 203.11 203.24 203.37 203.50 20? 63 203 76 203 90 204 I
.7 147.20 147.31 147.41 147.51 147.61 147.71 147.82 147.92 14S.02 148.125.6 20Ef6 25Gn0Gn005n0G9 20E8F20E9F20DF20DF205?.
.8 148.22 148.33 148.43 148.53 148.63 148.74 148.84 148.94 149.04 149.14 5.7 205.49 205.62 205.76 205.89 206.02 206.16 206.29 206.42 206.56 20
.9 149.25 149.35 149.45 149.56 149.66 149.76 149.87 149.97 150.07 150.18 5.8 206.82 206.96 207.09 201.23 207.36 201.49 207.63 207.76 207.90 208.i TE0 150 28 150.38 150.49 150 59 150 69 150.80 150.90 151 00 151.11 151.21 5.9 208.17 208.30 208.44 208.57 208.71 208.84 208.98 209.11 209.24 209..
209.51 209.65 209.79 209.92 210.06 210.19 210.33 210 46 210.60 210., 1.1 1517F15I32 15f 52 I5GFISOFI5CGn50F15E0n5E15 15E25' TIO 1.2 152.35 152A6 152.56 152.67 152.77 152.88 152.98 153.09 153.19 153.30 6.1 ' 2iO7 2fE0nlIJF2108 2fGi 2fE55 2iDF2ID2 2iOF2IEl 1.3 153.40 153 51 153.61 153.72 153.82 153.93 154.03 154.14 154.24 154.35 6.2 212.23 212.37 212.51 212.64 212.78 212.92 213.05 213.19 213.33 213. 14 154.45 154.56 154.66 154.77 154.88 154.98 155.09 155.19 155.30 155.40 6.3 213.60 213.74 213.88 214.01 214.15 214.29 214.43 214.56 214.70 214 T E5 155 51 155 62 155 72 155 83 155 94 1",6.04 156.15 156 25 156.36 156 47 64 214.98 215.12 215.26 215.39 215.53 215.67 215.81 215.95 216.08 216 1.6 15E5F15G5 1509 1509 15D5 15Df lSDFI5E32-'15D3 15EST TO 216 36' 216 50 216.64 216.78 216.92 217 06 217.20 217 33 217.47 2 1.7 157.64 157.75 157.86 157.96 158.07 158.18 158.29 158.39 158.50 158.61 6.6 21D5 21D9 2IDF219:1F21Oi 2fG5 2IB 59 21Dn1DF2197 f.8 158.72 158.82 158.93 159.04 159.15 159.26 159.36 159.47 159.58 159.69 6.7 219.15 219.29 219.43 219.57 219.71 219.85 219.99 220.13 220.27 220 1.9 159.80 159.91 160.01 160.12 160.23 160.34 160.45 160.56 160.66 160.77 6.8 220.55 220.69 220.84 220.98 221.12 221.26 22I.40 221.54 221.68 221 160.88 160 99 161 to 161.21 161.32 161.43 161.54 161 65 161.75 161 86 221.97 222.11 222.25 222.39 222.53 ??2.67 222.82 222.96 223.10 223 372TO 6.9 2.1 16D7 16DFIIE19 3DD 16DI 16E52 15DFIED4 IED5 IEDFTED 223.38 223.53 223 67 223.81 223 95 224 10 224 24 224.38 224 52 224 2.2" 163.07 163.18 163.29 163.40 163.51 163.62 163.73 163.84 163.95 164.05 7.1 22OI 2205 22D9 225:2F225'38 225 52'22DF225 8I 225.W220 2.! 164.17 164.29 164.40 164.51 164.62 164.73 164.81 164.95 165.06 165.17 7.2 226.24 226.38 226.53 226.67 226.82 226.96 227.10 227.25 227.39 227 2.4 165.28 165.40 165.51 165.62 165.73 165.84 165.95 166.06 166.18 166.29 7.3 727.68 227.82 227.97 228.11 228.26 228.40 228.55 228.69 228.83 22 166.40 166.51 166.6? 116 74 166.85 166 96 167.07 167.18 167.30 147.41 7.4 229.13 229.27 229.42 229.56 229.71 229.85 230.00 230.14 230.29 23 T72'5 2.6 167:52~ 16DI 16D5 16DFl6D7 !EDFlEDG 1EDF16D2 1 EOF ~377:5 230 58 230.72 230.87 231.02 231.16 231.31 231.45 231.60 231.74 231.8' 2.7 168.65 168.76 168.87 168.99 169.10 169.21 169.33 169.44 169.55 169.67 7.6 23DF23D8 23DF23278 232'62 23D7 23D2 23DF23DF23D 2.8 169.78 169.90 170.01 170.12 170.24 170.35 170.46 170.58 170.69 170.81 7.7 233.50 233.65 233.80 233.95 234.09 234.24 234.39 234.53 234.68 234 29 170.92 171.04 171.15 171.27 171.38 171.49 171.61 171.72 171.84 171.957.8 234.98 235.12 235.27 235.42 235.57 235.72 235.86 236.01 236.16 236 T r0 172 07 17? 18 172.30 172.41 172 53 172 64 172.76 172.87 172.99 173 to 7.9 236.46 236.61 236.75 236.90 237.05 237.20 237.35 237.50 237.65 237 3.1 17E22 17E3F17D5 173 57 17D9 17D5 17DI I703 17EfFl7DFTD 237.94 238.09 238.24 238.39 238.54 238.69 238 84 238.99 439.14 239. 3.2 174.38 121.49 174.61 174.73 174.84 174.96 175.08 175.19 175.31 175.428.1 ~23GF2309 23DF23OF24GF2409 24DF24G9 24GF240! 3.3 175.54 175.66 175.78 175.89 176.01 176.13 176.24 176.36 176.48 176.59 8.2 240.94 241.09 241.24 241.39 241.54 241.69 241.84 241.99 242.14 24 34 176.71 176.83 176.95 177.07 177.18 177.30 177.42 177.54 177.65 177.778.3 242.t3 242.60 242.75 242.90 243.05 243.20 243.35 243.50 243.66 24 TIS 177.89 178 01 178.13 178 25 178 36 178.48 178.60 178.72 178.84 17E.96 8.4 243.96 244.1I 244.26 244.42 244.57 244.72 244.87 245.02 245.18 245.3 3.6 7707 17EIFI7DFI703 1705FI7G7 17DFl?OI 12GI IBO4~~ DB:5 245.48 245.63 245.79 245.94 246.09 246 25 U6.40 246.55 246.70 246 8 3.7 180.26 180.38 180.50 180.62 180.74 180.86 180.98 181.10 181.22 181.34 8.6 24 EOF 24D6 24D2'247 4F24D2 24DF24D3 24GF24DF24Q! 3.8 181.46 181.58 181.70 181.82 181.94 182.06 182.18 182.30 182.42 182.54 8.7 248.54 248.70 248.85 249.01 249.16 249.31 249.47 249.62 249.77 2 39 182.67 182.79 182.91 183.03 183.15 183.27 183.39 183.51 183.63 183.75 8.8 250.08 250.24 250.39 250.55 250.70 250.86 251.01 251.17 251.32 2 T7EO 183 68 184 CG 184.12 184.24 184.36 184.48 184 60 184 73 184.85 184.618.9 251.63 251.79 251.94 252.10 252.26 252.41 252.57 252.72 252.88 253.0; l 4.1 165:0Fl85:2F18DFIEDFIB5:5FlBD5 155 BFIDOS ISO 7 IBO9- 253.19 253.34 253.50 253.66 253.81 253.97 254.13 254.28 254 44 254. 4.2 186.31 186.44 186.56 186.68 186.81 186.93 187.05 187.17 107.30 187.42 T70l9.l 25D5 250f 25D6 25D2 25D8 255 53 25GF25D5 25GF25G4 4.3 187.54 187.67 187.79 187.91 188.04 188.16 188.29 188.41 188.53 188.66 9.2 256.32 256.47 256.63 256.79 256.95 257.10 257.26 257.42 257.58 2 44 188.78 188.91 189.03 189.15 189.28 189.40 169.53 189.65 189.78 189.909.3 t 257.89 258.05 258.21 258.37 258.52 258.68 258.84 259.00 259.16 259
~67ES 190.02 190 15 190 27 190.40 190 32 190.65 190.77 190.90 191.02 191.159 '. , 259.47 259.63 259.79 259.95 260.11 260.27 260.43 260.59 260.74 260.90 4.6 19f 2FI9DFI9E5F19GF19DS 19DG 19DI 19E15 19DB 19Df- ~6705 261.06 261.22 261.38 261.54 261.70 261.86 262.02 262.18 262.34 26 4.7 192.53 192.66 192.78 192.91 193.04 193.16 193.29 193.42 193.54 193.67 9.6 26 D F264.42 264.26 262 264.58 02 26264.74 D B 26 rfr26DF26DF26DF26DF20:9F26EID 4.8 193.00 193.92 194.05 194.18 194.30 194.43 194.56 194.68 194.81 194.949.7 264.90 265.06 265.22 265.38 265.54 265.71 49 195.07 195.20 195.32 195.45 195.58 195.71 195.83 195.96 196.09 196.22 9.8 265.87 266.03 266.19 266.35 266.51 266.67 266.84 267.00 267.16 267 99 267.48 267.65 267.81 267.97 268.13 268.29 268.46 268.62 268.78 268 196 35 196 47 196.65 196.73 196 86 196.99 197.12 197.24 269.11_ 269.27 275~73 197.37 269.43 197.50 269.59 269.76 269.92 ~B50270.08 270.24 270 41 1986 ED1110WR08 UOLUnfS BASED CH 1961 SURUIV.
III8DUA11R 670- 681
~ _ ___
~
Q y s . CHI KAMAUGA - 1 TOTAL LIUIL S10RAff IN 1000 DAY-SICOND-IIET 2
............. P6CEk0F IIIU. 0.00. ......................................................
.01 .02 .03 .04 .05
~...................... ...................................................................................
. 2..3 .06 .07 .08 .09 ELIU. 0.00
. . . . = 212. 2 2 3. . Z . 2. . . 3 . F . . . . Z . . . . . 3 . . z . 3 3. . E Z 3. . . = 3 . 3.. . = E Z . = .12 & . . . . . . . . . . Z . = 2..... .01 .02 .03 .04 .05 .06 .07 .08 EE...&...E.. 2=... 23 2..Z.... 32..E2. 3..E..E.
.09 2......z.32. 23 2...z. 2... 3E.a 680.0.1 270.73 270.90 271.06 271.22 271.39 271.55 271.71 271.88 272.04 272.20
.2 272.37 272.53 272.70 272.86 273.02 273.19 273.35 273.52 273.68 273.85 685 5.1 0 362.49 362.70 362.91 363.11 363.32 363.53 363.74 363.95 364.15 364.36
.3 274.01 274.17 274.34 274.50 274.67 274.83 275.00 275.16 275.33 275.49 5.2 364.57 364.78 364.99 365.20 365.41 365.62 365.83 366.03 366.24 366.45
.4 275.66 275.83 275.99 276.16 276.32 276.49 276.65 276.82 276.98 277.15 5.3 366.66 366.87 367.08 367.29 367.50 367.71 367.92 368.13 368.34 368.53 277 32 277.48 277.65 277.81 277 98 178.15 278.31 218 48 278 64 278 81 54 368.77 363.98 369.19 369.40 369.61 369.82 370.04 370.24 370.46 370.67 370.88 371.09 371.30 371.52 371.73 371.94 372.15 372.36 372.58 372.79 755~5 .6
.7 280.65 2783F279TF27DI 280.81 280.98 281.15 281.32 27DF2793F2791F27933 281.49 281.65 281.82 281.99 282.16 2853n001 5.6 20570 TD5 5 3?DFJ7D2 37DI 1733F37 375.14 375.35 375.57 375.78 375.99 376.21 376.42 376.64 376.85 377.0)
.8 282.32 282.49 282.66 282.83 283.00 283.17 283.33 283.50 283.67 283.84 5.7
.9 284 01 284.18 284.35 284.52 284.69 284.85 285.02 285.19 285.36 285.51 5.8 377.28 377.50 377.71 377.93 378.14 378.36 378.57 378.79 379.00 379.22 75E0 285 70 285 87 286.04 286 21 286.38 286.55 286 72 286 89 287.06 287.23 5.9 379.44 379.65 379.87 380.08 380.30 380.52 380.73 380.95 381.16 381.3E 381.60 381.81 382.03 382.25 382.47 382.68 382.90 383.12 383.33 383.53 1.1 1.2 2BDF25E5F2BD4 2873f 22DE 233'2F2002 283~5F28DF2BB37BCD 289.10 289.28 289.45 289.62 289.79 289.96 290.13 290.30 290.47 290.65 6.1 3BDMBD9 3BOF3802 3BOMBO 1.3 290.82 290.99 291.16 291.33 291.51 291.68 291.85 292.02 292.19 292.37 6.2 385.95 3G6.17 386.39 386.61 386.83 387.04 387.26 387.48 387.70 387.92 1.4 292.54 292.71 292.89 293.06 293.23 293.41 293.58 293.75 293.92 294.10 6.3 388.14 388.36 388.58 388.80 389.02 389.24 389.46 389.68 389.90 390.12 T E5 294 27 294 44 294.62 294 79 294 97 295 14 295 31 295.49 295.66 295.83 6.4 390.34 390.56 390.78 391.00 391.22 391.44 391.66 391.88 392.10 392.33 7901 29CF290F29C51 290F290F29D6 29DF2985 29EST TC5 39?.55 392.77 39?.99 393.21 393.43 393.65 393.88 394 10 394,32 394 54 1.6 1.7 297.75 297.93 298.10 298.28 298.45 298.63 298.81 298.98 299.16 299.33 6.6 39Cn909 19DI 395 41 19535 39D7 3900 39CF390M906 1.8 299.51 299.68 299.86 300.03 300.21 300.39 300.56 300.74 300.91 301.09 6.7 396.98 397.21 397.43 397.65 397.88 398.10 398.32 398.54 398.77 398.91 19 301.27 301.45 301.62 301.80 301.98 302.16 302.33 302.51 302.68 302.86 6.8 399.22 399.44 399.66 399.89 400.11 400.34 400.56 400.78 401.01 401.23 7320 303.04 303.22 303.40 303 57 303.75 303.93 304.11 304 28 304.46 104.64 6.9 401.46 401.68 401.91 402.13 402.36 402.58 402.80 403.03 403.25 403.48 2.1 3002 30DM053F30D5'~~305~5n0Dn0D9'30Gn0D5 3002~ TBD 401.70 401 93 404.15 404 38 404.60 404 83 405 06 405 28 405 50 '405 73 2.2 306.60 306.78 306.96 307.14 307.32 307.50 307.68 307.86 308.04 308.22 7.1 40B6 4008 40Gl'~400F4DO6 40Gn0T31-~4DD4 40D6 40D9 2.3 3C8.40 308.58 308.76 308.94 309.12 309.30 309.48 309.66 309.84 310.02 7.2 408.22 408.44 408.67 408.90 409.12 409.35 409.58 409.80 410.03 410.26 24 310.20 310.39 310.57 310.75 310.93 311.11 311.29 311.47 311.65 311.83 7.3 410.48 410.11 410.94 411.17 411.39 411.62 411.85 412.07 412.30 412.53 TBE5 312.02 312.20 312.38 312.56 312.75 312.93 313.11 313 29 313 47 113.66 7.4 412.76 412.99 413.22 413.44 413.67 413.90 414.13 414.35 414.58 414.81 2.6 31ICnlO2~ 31DG 310F3105F3105 31Gnl5~I2 3!DF315 4r TBT5 415.04 415.27 415.50 415.73 415.95 416.18 416.41 416.64 416.87 417.10 2.7 315.67 315.85 316.04 316.22 316.40 316.59 316.77 316.95 317.14 317.327.6 4fDMID5-~4IDn!OI 4fOniunioniB3nIOniD9 419.62 419.85 420.08 420.31 420.54 420.77 421.00 421.23 421.46 421.69 2.8 317.51 317.69 317.88 318.06 318.25 318.43 318.62 318.80 318.98 319.17 7.7 29 319.36 319.54 319.73 319.91 320.10 320.29 320.47 320.65 320.84 321.03 7.8 421.92 422.15 422.38 422.61 422.84 423.07 423.30 423.53 423.76 423.99 TBr0 321 21 321.40 321 59 321.77 321.96 322 15 322.33 322 52 3??.71 322 89 79 424.22 424.45 124.68 424.92 425.15 425.38 425.61 425.84 426.07 426.30 3.1 323 03 32DF323 45723~6M23:8F32On2Of 32On2058 3207~ 7B8 0 426.53 426.76 427.00 427.23 427.46 427 69 427.92 428.15 428.38 428 61 3.2 324.96 325.14 325.33 325.52 325.71 325.90 326.09 326.27 326.46 326.55 8.1 4205 4208 420F429:54 42OM30tDI 4104 41Dn1Dn303 3.3 326.84 327.03 327.22 327.41 327.60 327.79 327.98 328.16 328.35 328.54 8.2 431.17 431.40 431.63 431.86 432.10 432.33 432.56 432.79 433.02 433.26 34 328.74 328.93 329.12 329.31 329.50 329.69 329.88 330.07 330.26 330.45 8.3 433.49 433.72 433.96 434.19 434.42 434.66 434.89 435.12 435.35 435.59
- TT5 330 64 330 81 331.02 331.21 331.41 331.60 331.79 331.98 33217 332.36 8.4 435 82 436.06 436.29 436.52 436.76 436.99 437.22 437.45 437.69 437.92 3.6 438.15 438.39 "8.62 438.86 439.09 439 32 419 56 439.79 440 02 440 26 3.7 332~55 33DS 33DM3DM3D2 33E52 33DF33GF3309 334.48 334.67 334.86 335.06 335.25 335.45 335.64 335.83 336.02 336.22 8.6 33Or7893 ~440M4DM4Dn4D0 44OF44176 3.8 336.41 336.61 336.80 336.99 337.19 337.38 337.58 337.77 337.97 338.16 8.7 412.83 443.07 443.30 443.54 443.77 444.01 444.24 444.47 444.71 444.94 3.9 333.36 338.55 338.75 338.94 339.14 339.33 339.53 339.72 339.92 340.11 8.8 445.18 445.41 445.65 445.88 446.12 446.35 446.59 446.82 447.06 447.29 340.31 340.51 340.70 340.90 341.10 341.29 341.49 341.68 341.88 342.08 8.9 447.53 447.77 448.00 448.24 448.47 448.71 448.94 449.18 449.41 449.65 754.10 ~34DF342.4n4DF34E87 34D6 34DO4DO4335 34TB5 1405'3 00 1 449.88 450.12 450 35 450.59 450.83 451.06 451.30 451.53 451 77 452 00 4.2 344.25 344.45 344.64 344.84 345.04 345.24 345.44 345.63 345.83 346.03 9.1 45E24 45E4n5DMS2'.95 45D8 45D2 45DS 45DF45C12 4506 4.3 346.23 346.43 346.f3 346.83 347.03 347.23 347.43 341.63 347.83 348.03 ' 9.2 454.60 454.83 455.07 455.30 455.54 455.78 456.01 456.25 456.48 456.72 4.4 348.23 348.43 348.63 348.83 349.03 349.23 349.43 349.63 349.83 350.03 9.3 456.16 457.19 457.43 457.67 457.90 458.14 458.38 458.61 458.85 459.08 7 05 350 23 350 43 350 64 350.84 351 04 351.24 351.44 351.64 351.84 35E 05 94 459.32 459.56 459.E0 460.03 460.27 460.51 460.74 460.98 461.22 461.45 4.6 35D5-~35D5 35D5 35EBn5Dn5Dn5Dn5DF35Dn5GF TDE5 461.69 461.93 462.16 462 40 462.64 462.87 463.11 463 35 463 58 463.82 4.7 354.28 354.48 354.68 354.89 355.09 355.29 355.50 355.70 355.90 356.11 9.6 4EO6 4609 46C5F46Dn65~0n4DM65 48 46Dn6535 4609 l 4.8 356.31 356.52 356.72 356.93 357.13 357.34 357.54 357.14 357.95 358.15 9.7 466.43 466.67 466.90 467.14 467.38 467.61 467.85 468.09 163.32 468.56
[ 4.9 358.36 358.57 358.77 35C.90 359.19 359.39 359.60 359.80 360.01 360.21 9.8 468.80 469.04 469.27 469.51 469.75 469.99 470.23 470.46 470.70 470.94 360.4? 360.63 360 83 361.04 361.?S 361.46 361.66 361.87 362.03 362.20 9.9 471.19 471.41 471.65 471.89 472.13 472.36 472.60 472.84 473.07 473.31 T90 473.55 425T93 473.79 474.01 474.26 474.50 474.74 474.58 475.21 175.45 475.69 19% IDITI0tUH08 UOLbnIS BASED Off 1961 SURUIV. l t HEADUAlfR 680- 690
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3 ... R&ckk 8 SEQUOYAH NUCLEAR PLANT Sedimentation Effects The following provides an identification of sediment concentrations and potential blockage of the intake channel at Sequoyah Nuclear plant resulting from flooding or loss of Chickamauga Dam downstream. Reservoir Characteristics Following Loss of Downstream Dam An analysis was performed to estimate how quickly the water level at the Sequoyah sita falls following failure of Chickamauga Dam. The assumptions used for the loss of downstream dam analysis were as follows:
- 1. Chickamauga Dam was postulated to fail totally and instantaneously.
- 2. Initial reservoir elevation at the tine of postulated failure was assumed to be the April 15-June 30 full pool level of 682.5.
- 3. Discharge from Chickamauga Dam before failure was assumed te be 32,000 cf s, an average daily flow, based on 5 years of record for the April 15-June 30 period.
- 4. Channel roughness was estimated from analysis of the March 1973 flood.
The surface elevation of the Tennessee River at the Sequoyah site following failuro of Chickamauga Dam was computed using TVA's Simulated Open Channel Hydraulics (SOCH) computer program (Reference 1). This unsteady flow model was previously calibrated to reproduce the observed March 1973 flood water surface profiles throughout Chickamauga Reservoir. Our dam failure analysis showed the water surface at the plantsite would drop to elevation 668 about twelve hours following failuro. It was assumod that the CCW pumps would operate throughout this period. The water surface would drop to elevation 665 about fifteen hours following failure. Figure 1 is a plot of water surface elevation at the Sequoyah site following Chickamauga Dam failure. Sediment Concentrations at ERCW and CCW pump Intakes Sedimont concentrations were estimated at the CRCW pump intakes, located at the streambed of the Tennosseo River, and at the CCW pump intakes, located in the intake channel 30 foot above tho main river streambed. The following assumptions woro mado in the sediment calculations:
- 1. Based on samples from Chickamauga Lake during 1982, about 40 percont of the bottom sediment is silt and about 40 percent is sand.
- 2. Clay-sizsd sadim2nt was excludsd from tha analysis bscause it will neithsr block ths intake ch:nns1 nor clog pumps.
- 3. Silt-sized sediment is adequately represented by medium silt (0.016-0.031 mm).
- 4. Very fine sand (0.062-0.125 mm) adequately represents the sand-sized sediment.
- 5. All sediment particles are spherical grains with a specific gravity of 2.65.
- 6. Vertical turbulence and sediment particle interaction does not affect settling velocity.
- 7. Sediment concentration is uniform throughout the vertical profile.
- 8. pumping rates produce uniform velocitios throughout the effective flow area of the intake channel between the skimmer wall and the CCW pumps.
l Sediment concentrations for the ERCW oump intakes were estimated using the Laursen equation (Reference 2) and a modification factor for large streams (Reference 3). 1 The Laursen equation relates sediment transport to the ratio of actual I tractive force (To ) to critical tractive force (Tce) at any time. If I To is less than Ter no transport occurs. The actual tractive force ) is a function of average channel velocity and stream depth (both obtained ) from the unsteady flow analysis). Critical tractivo force is a function ' of grain shear velocity, which was computed using average stream velocity. Silt and sand transport were computed for every hour for the first 4 hours j following failure, then for 6, 8, 10, 12, 14, and 16 hours, i I The maximum sediment concentrations for the 16 hours following dam failuro would be less than 2,000 parts per million (ppm) at the ERCW pump intakes. Concentrations after 16 hours would be less than 1,700 ppm. Sediment l concentrations at these levels are not expected to cause fouling in areas { of low flow in various p.lant systems; however, operator action can be I relied upon to open valvos as needed to flush out debris; therefore, this is not a problem. The estimated suspended silt and sand concentration in the main channel at the ERCW pump intakes versus time following failure is shown in Figure 2. Even though sediment concentrations are not extremoly high, there is potential for significant deposition of silt in the excavated sump or forebay area at the front of the ERCW pumping station. A detailed analysis of this situation will not be needed because even if the areas immediately adjacent to the ERCW pump suction elevation woro heavily silted, the operating ERCW pumps would continue to operato forcing flow channels through the silt. The worst situation that could happen would be that the ERCW pumps initially on standby could not be restarted if silt was up to the elevation of the pump suction. However, the required number of pumps would be started or continuo operating immediatoly after dam failure and before dotrimontal silt buildup could occur (Reference AOI-22). Therefore, lack of operability of the additional ERCW pumps would pose no safety concern.
For a Tennssseo River flood, stdim3nt concsntration during extrama Tennessee River flooding would be similar to those produced by failure of Chickamauga Dam. However, because channel velocities would be less than for the dam failure case, the potential for transporting enough sediment into the ERCW pump intake area to cause blockage is low. Only a fraction of the bottom sediment suspended following dam failure is elevated to the level of the intake channel 30 feet above the streambed of the Tennessee River. The Rouse suspended load distribution equation (Reference 2) provides such estimates and depends on sediment size, water depth, sediment settling velocity, and sediment concentrations near the streambed. It is estimated based on this equation that the silt conce1-tration at the intake channel entrance is about 75 percent of the concentration near the streambed, while the sand concentration is only about 2 percent. The estimate of sediment concentration at the CCW pumps was determined using settling basin calculations. The first step was to determine resident time in the channel. Residence time is the time it takes sediment-laden water to travel from the channel entrance to the pumps and is a function of flow-through velocity. Flow-through velocity was estimated by dividing the pumping rate by the effective flow area of the channel. Flow area, which changed with time, was determined using the elevation hydrograph of Figure 1 and a relationship developed for depth of flow in the channel and average effective flow area. Secondly, the settling velocity of medium silt and very fine sand was computed using a standard text equation (Reference 4) and the assumptions stated previously. The third stop was to determine what percent of incoming sediment settled out before reaching the pumps. With the residence time and settling velocity determined, the depth that medium silt and very fine sand settled while traveling to the pumps was computed. Since it was assumed that sediment cor.contration was uniform throughout the vertical profile, the settling depth divided by the total depth represents the portion of sediment that settlos out before reaching the pumps. The sediment that does not settle out reaches the CCW pumps. The residence time associated witn the CCW pumping rate is such that about 90 percent of all silt entering the intake channel reaches the pumps without settling out. The maximum silt concentration at the CCW pump intakes for the 11 hours the pumps operate following dam failure is 1,100 ppm. The estimated silt concentration at the CCW pumpt versus time is shown in Figure 2. Essentially all sand entering the intake channel settles out before reaching the CCW pumps. The potential for blocking the CCW pump intakes is low because of the low concentrations of inflowing silt, the volumesof sand entoring the channel is low, and because the channel is isolated from the Tennesseo River 15 hours af ter dam failure. Also because the intake channel is located 30 feet about the Tennesseo River streambed, we see no chance for intake channel blockago.
REFERENCES o .
- 1. Garrison, J. M., Gran.'e, J. P., and Price, J. T., "Unsteady Flow Simu-lation in Rivers and Reservoirs," Journal of the Hydraulics Division, ASCE, Volume 95, No. HYS, Proceedings Paper 6771, September 1969, pages 1559-1576.
- 2. Graf, Walter Hans, 1971. Hydraulics of Sediment Transport, Second Edition. New York, New York: McGraw-Hill Book Company
- 3. U.S. Army Corps of Engineers, March 1977. HEC-6 Scour and Deposition in Rivers and Reservoirs, Users Manual.
- 4. Haan, C. T., and B. J. Barfield, 1978. Hvorology and Sedimentology of Surface-Mined Lands. Office of Continuing Education and Extension, College of Engineering, University of Kentucky, Lexington, Kentucky.
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e- . ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-21) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS 5 k ii.i iii i i.
ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the proposed technical specification change and has determined that it does not represent a significant hazards consideration based on criteria cctablished in 10 CFR 50.92(c). Operation of SQN in accordance with the proposed amendment will not: 1 (1) involve a significant increase in the probability or consequences of an accident previous [y evaluated. l The function of ERCW as described in FSAR section 9.2.2 and the function of the UHS as described in FSAR section 9.2.5 remain unchanged. The UHS temperature is not assumed to be an initiating cause for FSAR evaluation events. As such, increasing the UHS temperature will not affect the probability of any previously evaluated accidents. The increase in the maximum UHS temperature was found to be acceptable by evaluating the impacts on various systems, components, and analyses. The following were evaluated: ECCS Other FSAR chapter 15 accident analyses Containment subcompartment pressure analysis (FSAR section 6.2) Peak containment temperature Peak centsinment pressure Long-term containment cooling Long-term cooling for pipe breaks outside containment EQ temperature profiles The increase in UHS temperature was found to impact the peak containment pressure analysis and long-term cooling for pipe breaks inside and outside of containment. These impacts, however, are neither significant nor detrimental to the plant for the proposed increase in the UHS temperature limit from 83 degrees F to 84.5 degrees F. Therefore, the consequences of a previously evaluated accident are not significantly increased. The addition of a minimum UHS res~;veir level requirement ensures that plant operation is bounded by the analysis that established ERCW flow rates to the major transient heat load components. The establishment of flows based on time-dependent heat loads merely provides more operational margin without being overly conservative. As such, the proposed specification addition does not increase the probability or consequences of a previously evaluated accident. The wording changes to specification 3/4.7.5 are administrative in nature and are made for consistency and clarification. The revisions to the action statement and the SR are made for consistency with the new requirement for UHS reservoir level. The additional wording change is made to cler-ly specify that the URS temperature limit applies to the ERCW supply water temperature. The change does not affect the probability or consequences of a previously evaluated accident.
c . (2) create the possibility of a new or different kind of accident from any previously analyzed. The proposed increase in the maximum allowable UHS temperature is made to accommodate elevated river temperatures at the SQN site. These elevated river temperatures are the result of an extended drought in the Tennessee River watershed. The performance of the UHS and ERCW has been analyzed and found to be acceptable at the increased maximum temperature. Because the UHS and ERCW still perform their intended function and these functions remain unchanged, the probability of a new or different kind of accident from any previously analyzed is not created. The addition of a minimum UHS reservoir level ensures that plant operation is bounded by analyses performed to establish ERCW flow rates to transient heat load equipment. Adequate performance of the heat exchangers is ensured. The addition of the specification will not create the possibility of a new or different accident. The wording changes to specification 3/4.7.5 are administrative in nature and are made for consistency and clarification. The revisions ; to ehe action statement and the SR are made for consistency with the new requirement for UHS reservoir level. The additional wording change is made to clearly specify that the LHS temperature limit applies to the ERCW supply water temperature. As such, the proposed change will not create the possibility of a new or different accident. (3) involve a significant reduction in a margin of safety. The proposed increase in the UHS maximum temperature has only minor impacts on the peak containment pressure analysis and long-term cooling for pipe breaks outside containment. The peak containment pressure increases by approximately 0.14 lb/in g2 and remains below the containment design pressure of 12 lb/in 2g. The ESF room cooler performance remains acceptable and maintains the 100-day average room temperature below EQ limits. As such, the increase in the UHS temperature limit does not significantly reduce the margin of safety. The addition of a minimum UHS reservoir level ensures that the ERCW flow rates to the CCS and CS heat exchangers meet or exceed the values assumed in the peak containment pressure analysis as performed for this change. This establishes a level of operational margin for l performance of the heat exchangers. The margin of safety is not I significantly reduced by the proposed change. The wording changes to specification 3/4.7.5 are administrative in nature and are made for consistency and clarification. The revisions to the action statement and the SR are made for consistency with the new requirement for UHS reservoir level. The additional wording change is made to clearly specify that the UHS temperature limit applies to the ERCW supply water temperature. The proposed change does not reduce the margin of safety.
TABLE 1 Revised Heat Exchanger Flow Rates and Overall Heat Transfer Coefficient HEAT EXCHANGER FLOW RATES HEAT XCHR UA (Gallons Per Minute) x106 RHR 1 & 2 A&B Tube Side 3074 1.539 Shell Side 5000 CS 1A Tube side 4650 3.092 Shell side 3708 CS 1B Tube side 4750 26502 Shell Side 3776 CS 2A Tube Side 4750 3.123 Shell Side 3744 CS 2B Tube Side 4750 3.078 Shell Side aggg CCS A Tube Side aggg 2.845 Shell Side 5000 CCS B Tube Side 4284 2.111 Shell Side 5000 CCS C Tube Side 4284 2.802 Shell Side 5000 NOTE: Underlined values were used in the revised containment analysis.
e .. TABLE 2 Results of Containment Analysis Peak Pressure FSAR NEW PRESSURE PRESSURE (psig) (psig) TVA 83*F ERCW 11.03 11.36 85*F ERCW 11.16 11.50 WESTINGHOUSE 83*1 ERCW 11.09 11.19 85*F ERCW 11.25 11.31 i t 1 l l l i if I 1 1 1 I
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