ML20247H443
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APPENDIX 15B METHODS FOR ANALYSIS OF THE LOSS OF FEEDWATER INVENTORY EVENTS O :
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Page Amendment i 8 ii 7 iii 7 l iv 8 v 8 Text Page g ndment 158-1 7 158-2 7 15B-3 7 l 15B-4 7 15B-5 7 15B-6 7 15B-7 10 15B-8 10 l (Ol 158-9 8 l
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CHAPTER 15 I APPENDIX 158 l Figures Figure No. Amendment 15B-18 7 !
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15B-22 7 l' 158-23 7 15B-24 7 15B-25 7 158-26 7 158-27 7 158-28 7 15B-29 7 158-30 7 158-31 8 15B-32 8 158-33 8 158-34 8 ,
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TABLE OF CONTENTS
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CHAPTER 15 APPENDIX 15B Section Sjject Page No.
15B.1 INTRODUCTION 15B-1 l 15B-2 DISCUSSION 158-1 l 158.3 METHOD OF ANALYSIS 15B-2 1 58.4 RESULTS 3B-7 1
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58.5 CONCLUSION
15B-9 15B.6 REANALYSIS OF SMALL BREAK LOSS OF FEEDWATER INVENTORY EVENTS WITH THE 8
LIMITING SINGLE FAILURE AND 0FFSITE 1 POWER AVAILABLE 15B-9 O
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i LIST OF FIGURES CHAPTER 15 l APPENDIX 158 i
Figure S_u_bj ec t 158-1 Loss of Feedwater Inventory Maximum RC System Pressure vs Break Area 158-2 Loss of Feedwater Inventory Maximum RC System Pressure j vs Initial RC System Pressure 15B-3 Loss of Feedwater Inventory Maximum RC System Pressure vs Initial Core Power 15B-4 Loss of Feedwater Inventory Maximum RC System Pressure vs Initial Reactor Vessel Flow 15B-5 Loss of Feedwater Inventory Maximum RC System Pressure vs Initial Pressurizer Water Volume 15B-6 Loss of Feedwater Inventory Maximum RC System Pressure vs Pressurizer Safety Valves Rated Flow 15B-7 Loss of Feedwater Inventory Maximum Reactor Coolant System l Pressure vs Doppler Multiplier
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i 15B-8 Loss of Feedwater Inventory Maximum Reactor Coolant System Pressure vs Core Life 15B-9 Loss of Feedwater Inventory Maximum RC System Pressure vs Fuel Gas Gap Heat Transfer Coefficient 15B-10 Loss of Feedwater Inventory Maximum RC System Pressure j vs Initial Steam Generator Inventory 158-11 Loss of Feedwater Inventory Maximum RC System Pressure l vs Feedwater Enthalpy j i
15B-12 Loss of Feedwater Inventory Maximum RC System Pressure l vs Initial Core Inlet Temperature 15B-13 Loss of Feedwater Inventory Limiting Case Core Power vs Time l 15B-14 Loss of Feedwater Inventory Limiting Case Core Average Heat Flux vs Time 15B-15 Loss of Feedwater Inventory Limiting Case Reactivity vs Time l II 9 i Amendment No. 7 March 31, 1982
1 C_./i CHAPTER 15 APPENDIX 15B l i
Figure Subject 15B-16 Loss of Feedwater Inventory Limiting Case Core Average Coolant Temperature vs Time 158-17 Loss of Feedwater Inventory Limiting Case Reactor Coolant Flows vs Time 158-18 Loss of Feedwater Inventory Limiting Case RCS and Pressurizer Pressure vs Time 15B-19 Loss of Feedwater Inventory Limiting Case RCS and Pressurizer Pressure vs Time ,
t 158-20 Loss of Feedwater Inventory Limiting Case Pressurizer Surge l Line Flow vs Time 15B-21 Loss of Feedwater Inventory Limiting Case Pressurizer Water Volume vs Time
/' 15B-22 Loss of Feedwater Inventory Limiting Case Pressurizer Safety Valve Flow vs Time l 1
158-23 Loss of Feedwater Inventory Limiting Case Stec Generator !
Pressures vs Time 15B-24 Loss of Feedwater Inventory Limiting Case Total Steam Flow vs Time 15B-25 Loss of Feedwater Inventory Limiting Case Total Steam Flow vs Time 15B-26 Loss of Feedwater It intory Limiting Case Break Discharge Flow vs Time 15B-27 Loss of Feedwater Inventory Limiting Case Break Discharge Enthalpy vs Tlme 158-28 Loss of Feedwater Inventory Limiting Case Steam Generator Liquid Mass vs Time i 15B-29 Loss of Feedwater Inventory Limiting Case Steam Generator l Water Level vs Time 15B-30 Minimum DNBR vs Time for the Loss of Feedwater Inventory i
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I LIST OF FIGURES (Cont'd) ]
CHAPTER 15 APPENDIX 15B
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Figure Subject l
15B-31 Reanalysis of Small Break Loss of Feedwater l Inventory Events, Maximum RCS Pressure vs Break Area 15B-32 Reanalysis of Small Break Loss of Feedwater Inventory Events - Limiting Case, Core Power vs Time 15B-33 Reanalysis of Small Break Loss of Feedwater f Inventory Events - Limiting Case, Core Heat 1 Flux vs Time l t
15B-34 Reanalysis of Small Break Loss of Feedwater Inventory Events - Limiting Case, Reactivities vs Time .
15B-35 Reanalysis of Small Break Loss of Feedwater Inventory Events - Limiting Case, Core Coolant Temperatures vs Time q 15B-36 Reanalysis of Small Break Loss of Feedwater Inventory Events - Limiting Case, Reactor Coolant Flow vs Time 15B-37 Reanalysis of Small Break Loss of Feedwater Inventory Events - Limiting Case, Primary ,
System Pressures vs Time {
15B-38 Reanalysis of Small Break Loss of Feedwater !
Inventory Events - Limiting Case, Steam l Generator Pressures vs Time !
I 158-39 Reanalysis of Small Break Loss of Feedwater l Inventory Events - Limiting Case, Steam Generator Liquid Mass vs Time i
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LIST OF TABLES 7
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) CHAPTER 15 APPENDIX 15B' Table Subject 158-1 Assumptions for the Limiting Case Loss of Feedwater Inventory Event 156-2 Sequence of Events for the Limiting Case '
Loss of Feedwater Inventory Event 158-3 Assumptions for the Reanalysis of the Small Break Loss of Feedwater Inventory Event 1 58-4 Sequence of Events for the Reanalysis of the Limiting Small Break Loss of Feedwater Inventory Event 1
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APPENDIX 15B MET 3005 FOR AW LYSIS OF THE LOSS OF FEEDWATER INVENTORY EVENTS 15B.1 INTRODUCTION ,
This appendix describes the methods utilized in the transient analysis of Section 15 to e_nvelope the overpressurization potential of the loss of feedwater inventory (LFI) event. The method involve simplifying, but conservative, modeling assumptions with respect to the feedwater line break mass and energy discharge rates and their affect on steam generator water level and heat transfer response. Using these assumptions sensitivity studies are performed to determine the most adverse set of initial plant operating conditions and transient parameters. The example used to demon-strate these methods corresponds to the limiting case feedwater line break defined in Reference 1 which includes consideration of the entire spectrum l of break sizes and locations with the most adverse set _of operating conditions, I a loss of normal on-site and off-site electrical power at the most adverse time, the most adverse active single failure, and the failure of the most reactive control element assembly to insert following a reactor trip signal. l l
158.2 DISCUSSION I The LFI event is initiated by a break in the main feedwater system (MFS) l piping. Depending on the break size and location and the response of the l MFS, the ef fects of a break can vary from a rapid heatup to a rapid cooldown l of the Nuclear Steam Supply System (NSSS). In order to discuss the possible effects breaks are categorized as small, if the associated discharge flow n
v is within the exce'ss capacity of the MFS, and as large, otherwise. Break locations are identified with respect to the feedwater line reverse flow check valves which are located between the steam generator feedwater nozzles and the containment penetrations. Closure of these valves to reverse flew from the nearest steam generator maintains the integrity of that generator in the presence of a break upstream of the valves.
i l Breaks upstream of the check valves can initiate one of the following tran-sients. If the MFS is unavailable following the pipe failure, a total loss l of normal feedwater flow (LOFW) results. With the MFS remaining in operation j no reduction in feedwater flow occurs for small breaks, while large breaks impose either a partial LOFW or a total LOFW, if the area is sufficient to discharge the entire feedwater pump flow capacity.
In addition to the possibility of partial or total LOFW events, breaks downstream of the check valves have the potential to establish reverse flow from the nearest steam generator (referred to as the " ruptured" generator) back to the break. Reverse flow occurs whenever the MFS is not operating subsequent to a pipe break or when the MFS is operating but without sufficient capacity to maintain pressure at the break above the steam generator pressure.
It is only these breaks which develop reverse flow that are of interest in this analysis.
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Depending on the enthalpy of the reverse flow and the ruptured stear, genera-tor's heat transfer characteristb, the reverse flow cy induce eit9r an RCS heatup or cooldown. However, excessive heat removal througn the break is not considrred in this analysis, because the cooldown potential is less than that of the loss of main steam inventory (LMSI) events. The maximum -
break size is smaller for the LFI events than for LMSI events. In addition, the LMSI breaks have a greater potential for discharging high enthalpy fluid due to the location of steam piping above feedwater piping within the steam generator. Furthermore, the LFI breaks cause an instant reduction in feedwater flow unlike LMSI breaks which results in a reduced heat removal capacity due to the lower liquid inventory. Since LFI breaks can cause a rapid depletion of ruptured steam ge.erator liquid mass, reducing the heat transfer capability and causing a rapid RCS heatup and pressurization, it is the heatup pot tial which is emphasized in this study.
A general description follows of the LFI event assuming a break downstream of the check valves, inoperability of the MFS, and low enthalpy break i discharge. The loss o' subcooled feedwater flow to both steam generators causes increasing steam generator temperatures and decreasing liquid inven-tories and water levels. The rising secondary temperatures reduce the primary-to-secnndary heat transfer and force a heatup and pressurization of the RCS. The heatup becomes more severe as the ruptured steam generator experiences a further reduction in its heat transfer capability due to insufficient liquid inventory as the break discharge coatinues. This initial sequence of events culminates with a reactor trip on high pressurizer pressure, icw steam generater water level or high containment pressure.
RCS heatup can continue after trip due to a total loss of heat transfer in the ruptured steam generator as it empties. Eventually the decreasing core power following reactor trip reduces the core heat rate to the heat removal capacity of the ir, tact steam generator.
The analysis methods address the influence of the four major controlling parameters; discharge enthalpy and flow, low water level trip condition in the ruptured steam generator, and the heat transfer characteristics of the l ruptured steam generator. ;
1 158.3 Method of Analysis Analysis of the LFI event is performed using the CESEC II computer program described in Section 15.0.3. along with several simplifying assumptions l which, with respect to RCS overpressurization, conservatively model the '
break discharge flow and enthalpy and the ruptured steam generator water level and heat transfer. In addition, sensitivity of the RCS overpres-surization to changes in various plant initial conditions is evaluated to determine the most adverse initial conditions for the LFI event.
Blowdown of the steam generator nearest the feedwater line break is modeled assuming frictionless critical flow as calculated by the Henry-Fauske correlation (Reference ?). Although the enthalpy of the blowdown physically depends upon the location of the breck relative to fluid conditions within the ruptured steam generator, it is assumed that saturated liquid is discharged until no liquid remains at which time saturated steam discharge is assumed.
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With respect to RCS overpressurization these assumptions result in.conserv-atively high mass flow and v>nservatively low energy flow from the steam generator to the break, thereby minimizing the rupturta generator hea' i removal capacity.
G' In lieu of detailed steam generator modelling to calculate the redistribution i of fluid under the influence of blowdown to the break, no credit is taken l for a low water level trip condition in the ruptured steam generator until l the generator is emptied of liquid. This conservatively delays the time of reactor trip, prolonging the RCS heatup and overpressurization. No credit is taken for the high containment pressure trip.
In order to determine the sensitivity of the RCS overpressurization to the ruptured steam generator heat transfer characteristics without implementing l a detailed steam generator model, the effective heat transfer area is ;
assumed to decrease linearly (from the design value to zero) as the steam l generator liquid mass decreases (from a selected value to zero). The mass l interval over which the rampdown is assumed to occur is referred to as i "M" . Therefore, decreasing values of M imply a more rapid loss of heat transfer in the ruptured steam generator.
Sensitivity studies are used to establish the most adverse set of initial operating and transient parameters with respect to RCS overpressurization.
These parameters include break size, M, initici core power, initial RCS pressure, initial reactor vessel ficw, initial pressurizer liquid volume, pressurizer safety valve rated flow, core physics conditions, fuel gas gap heat transfer coefficient, initial core inlet temperature, initial feedwater enthalpy and initial steam generator inventory.
The first parametric analysis includes various combinations of break sizes and steam generator heat transfer characteristics (M) with nominal full j power beginning-of-cycle plant operating conditions assumed. At the time j of turbine trip there is an assumed loss of both normal on-site and off- '
site electrica power. The break size is varied from 0.0 to the maximum areaof1.4ft}. The maximum area is restricted to the sum of the flow I distribution holes in the steam generator economizer section. The value of M is varied from 0 to 100,000 lbm to envelope all possible rates of decreasing the ruptured steam generator heat transfer. Results of this study are shown in Figure 158-i.
For each value of M the curve of peak RCS pressure versus break area is characterized by a relatively sharp rise in pressure with increasing break area followed by a gradual decline. Pressures for break area less than that corresponding to the inflection point are reduced due to a reactor trip on low water level in the intact steam generator before total heat transfer is lost in the ruptured steam generator. Larger breaks trip on high pressurizer pressure or low water level in the ruptured generator.
The relationship between pressure and break area after the inflection point is due to a combination of more rapid loss of heat transfe- with increasing area, off-set by greater steam relieving capacity of larger breaks once the ructured steam generator empties which is important in reducing the RCS p heatup following turbine trip.
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Except for the range of small breaks, larger values of AM result in lower RCS pressures. The more gradual decrease in heat transfer associated with a larger AM allows for more reactivity feedback from the moderator temperature which reduces core power prior to trip, and also allows for a greater shift of secondary hest transfer from the ruptured steam generator to the intact generator prior to trip. Both of these phenomena min hize the rate of RCS i '
heatup and pressurization after reactor trip. Howevei for the range of small breaks and small AM, no decrease in heat transter occurs before trip, but as AM increases, heat transfer reduction begins prior c.o and continues after reactor trip.
The study shows that a break area of 0.2 ft2 and AM equal to 0.0 is the most adverse combination. This combination is used as the base case for the rest of the sensitivity studies.
The sensitivity of peak RCS pressure to initial RCS pressure is shown in Figure 158-2. For pressures to the right of peak in the figure, reactor trip occurs on high pressurizer pressure prior to the ichs of heat transfer in the ruptured-steam generator. Lower initial pressures trip on low water level in the ruptured steam generator (once emptied). A.1d as the initial ;
pressure is lowered the RCS pressure at reactor trip decreases reducing the l peak pressure. The concavity of the curve up to 2200 psia is due to the limiting effects of the pressurizer safety valves.
Due to the sensitivity of the maximum RCS pressure to initial RCS pressure, each of the following parametric studied adjusts the initial RCS pressure within its full power operating band to provide a high pressurizer pressure trip signal coincicent with the first reactor trip signal, if possible. 1 This provides an equal basis for comparison.
The sensitivity of RCS pressure to initial core power is shown in Figure 158-3. Lowering the core power reduces the RCS heatup associated with losing one steam generator heat transfer capability.
The sensitivity of RCS overpressurization to initial reactor vessel flow is negligible as shown in Figure 158-4.
The initial pressurizer liquid volume shows no significant influence on the maximum RCS pressurization (~ Figure 15B-5). RCS pressurization prior to re-actor t. rip is more rapin for Targer initial liquid volumes, however, the l volume doec not affect the rate of RCS heatup nor the pressurizer safety valve opening ar i associated pressure required to accommodate the volumetric insurge flow due to the heatup. In the absence of significant sensitivity, the raximum pressurit er liquid volume is considered the most adverse due to the increased potential for completely filling the pressurizer with liquid during the course of the transient.
The rated flow of the four pressurizer safety valves is restricted to a minimum of 460,000 lbm/hr/ valve to a maximum of 575,000 lbm/hr/ valve. l Figure 158-6 shows that within this range the maximum RCS pressure decreases l only sligntly with increasing rated flow. This indicates that the maximum )
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volumetric insurge to the pressurizer during the maximum rate of RCS heatup is well within the relieving capacity of the safety valves (i.e., a valve with a icwer rated flow must open proportionately further to provide the
( required flow, but not beyond its iimit).
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' A multiplier applied to the Doppler reactivity feedback has negligible )
impact on the peak RCS pressure (see Figure 158-7). This result is due to compensating effects which the Doppler feedback has on the core power transient before reactor trip relative to after trip. Increasing the feedback (multiplier) slightly reduces the core power before, but reduces the rate of core power decay after the reactor trip.
Figure 15B-8 shows the response of the maximum RCS pr essure to relative changes in the core life. The decrease in RCS pressure is predominately l duetodecrgasingmoderatortemperaturecoefficientofreactivity(nominally
-1.13 x 10- ao/F at beginning-of-cycle to -2.46 x 10 -4 ap/F at end-of-1 l
cycle, assuming equilibrium xenon atd a core average temperature of 594 F).
The more negative coefficient preducts e greater reduction of core power prior to trip and thereby reduces the RCS heat up and pressurization following )
reactor trip. l The fuel gas gap heat transfer coefficient affects the initial fuel tempera-tures and the associated stored energy in the fuel. In the LFI event in-creased stored energy increases the core heat flux following reactor trip and hence the RCS heatup. Therefore, as shown in Figure 15B-9, the maximum l RCS pressure reaches a peak for the lowest value of fuel gas gap heat )
transfer coefficient (corresponding to cold clean fuel). )
n Because the core power has a significant impact on the peak RCS pressure (V) and can be influenced by moderator reactivity feedback prior to reactor I trip, sensitivity studies on parameters which strongly affect moderator temperature (i.e. , initial core inlet temperature, initial steam genera tor water mass and feedwater enthalpy) use the base case modified by implementing the most positive moderator reactivity versus temperature curve (beginning-of-cycle) including uncertainties.
The sensitivity of maximum RCS pressure to initial steam generator water mass shown in Figure 15B-10 has fcur characteristic segments. For, initial I masses between 90,000 lbm and 115,000 lbm, a reactor trip condition is first encountered on low water level in the intact steam generator event with an adjustment of the initial RCS pressure to the upper limit of operation (2400 psia). Within this range a decrease in initial mass forces an earlier reactor trip and lower RCS pressure at trip. Between 115,000 lbm and 135,000 lbm the initial RCS pressure can be adjusted to provide simultaneous reactor trip signals from high pressuri:er pressure and low water level in the intact steam generator and hence the plateau of maximum RCS pressure.
From 135,000 lbm to 155,000 lbm reactor trip still occurs on high pressurizer pressure and intact ste'im generator low water level, however heat transfer rampdown in the rupturec steam generator begins prior to reaching the maximum RCS pretsure. Abave 155,0C0 lbm the low water level trip condition is due to emptying cf the ruptured steam generator. Therefore, all heat transfer is lost in the ruptured steam generator prior to reactor trip f causing the most adverse RCS heatup initiated with the pressurizer pressure at the trip setpoint.
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Figure 158-11 shows the results of the parametric analysis on traximum RCS p, essure versus initid feedwater ent%1py. Raising the degree or feecuater ;
subcooling increases the rate of RCS heatup once main feedwater is terminated.
The RCS pressurization is greater, therefore, for decreasing feedwater enthalpy.
The final sensitivity study on initial core inlet temperature indicates only a small dependence of peak RCS pressure on temperature. There are several off-setting influences of temperature. Lowering the core inlet temperature increases the initial moderator temperature coefficient of reactivity. It decreases the secondary side tempera ~ture, thereby reducing the degree of feedwater subcooling. The corresponding decrease in the steam generator pressure reduces the break discharge ficw and also prevents opening of the main steam safety valves prior to reactor trip. The result of the interaction of these changes is shown in Figure 15B-12.
The results of these sensitivity studies provide a set of initial conditions and transient parameters which establish the limiting RCS overpressurization LFI event. In summary this set includes:
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- 1. 0.2 ft break area
- 2. Instantaneous loss of heat transfer in the ruptured steam generator (AM=0)
- 3. Initial RCS pressure which forces a high pressuri2er pressure trip co-incident with the first reactor trip signal
- 4. Nominal reactor vessel flow
- 5. Maximum initial core power
- 6. Maximum initial pressurizer liquid volume
- 7. Minimum pressurizer safety valve rated flow
- 8. Nominal !) oppler reactivity feedback
- 9. Most positive moderator temperature coefficient of reactivity
- 10. Minimum fuel gas gap heat transfer coefficient
- 11. Nominal initial steam generator water mass
- 12. Minimum initial feedwater enthalpy
- 13. Maximum initial core inlet temperature 1
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v 15B.4 RESULTS An example limiting analysis of the LFI transient suggested by Reference 1 was performed applying.the conservative methods with the most adverse set of initial plant conditions and transient parameters discussed above. Table ;
15B-1 lists the assumptions utilized in this worst case. The sequence of J events and the dynamic response of the important NSSS parameters are provided i in Table 150-2 and Figures 15B-13 through 15B-30, respectively.
2 A 0.2 ft crack in the main feedwater line is assumed to instantaneously ,
terminate feedwater flow to both steam generators and establish critical flow I
(-2000 lbm/sec of saturated liquid) from the generator nearest the break. The absence of subcooled water and pressurization of the steam generators during the first 33.82 seconds which reduces the primary-to-secondary heat transfer 10 rate. Rising reactor coolant temperatures and pressure' result. Due to temperature reactivity feedback during this period the core power decreases slightly from 102 percent to 98 percent of design full power. l At 33.82 seconds the ruptured steam generator is assumed to instantaneously l 10 lose all heat transfer capability due to total depletion of its liquid in-ventory by boil-off and the break discharge flow. This initiates a rapid heatup and pressurization of the reactor coolant system and depressurization of the steam generators. Once emptied, credit is taken for a low water level trip condition in the ruptured steam generator which' leads to a reactor trip signal at 34.82 seconds simultaneous with a high pressurizer pressure trip l 10 a signal. The rate of reactor coolant system pressurization is further l aggravated at 35.8 seconds. Closure of the turbine leaves the pipe break as the only steam relief path, thereby reducing the energy flow from the intact l 10 steam generator below that of the primary-to-secondary heat' transfer rate.
The resulting steam generator pressurization reduces the Primary-to-secondary temperature difference. In addition, the loss of reactor coolant flow following the loss of electrical power decreases the heat transfer coefficient of the coolant in the steam generator tubes. A significant heat transfer reduction occurs.
Compression of the pressurizer steam volume due to the high insurge flow raises the pressure to the safety valve setpoint at-34.6 seconds. Thereafter every increase in the surge flow causes a slight pressurization which opens the safety valves such t!at their volumetric discharge rate matches that of the insurge. The reactor coolant system pressure continues to increase to a maximum of 2843 psia at 38.2 seconds. At that time the increased pressure establishes a surge line pressure gradient which provides sufficient flow to allow the reactor coolant to expand under the existing heatup with no further pressurization. Pressurizer pressure and surge line flow are also at their maxima of 2587 psia and 2206 lbm/sec, respectively.
The rate of heatup decreases subsequent to core heat flux decay causing the primary pressures to drop. By 40.5 seconds the main steam safety valves open thus stabilizing the secondary side temperature and allowing the rising l
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primary coolant temperature to develop greater heat transfer to the intact i steam generator. The intact generator is forced to a maximum of 1318 psia i before the heat transfer begins to decrease. However, the core-to-steam !
generator heat rate mismatch is reduced sufficiently by 45.4 seconds to allow closure of the pressurizer safety valves and by 45.8 seconds the reactor coolant system enters a cooldown. Under the influence of steam blowdown through the ruptured steam generator to the break, the cooldown proceeds even af ter the steam generator safety valves close at 73.8 seconds.
A main steam isolation signal is generated at 166.0 seconds on low steam l10 l generator pressure which closes the main steam isolation valves decoupling the intact steam generator from the ruptured steam generator and the break. The intact steam generator repressurizes, thereby reducing its heat transfer and eventually causing a primary system heatup by 300 seconds. With the main l steam safety valves open by 314.2 seconds, the primary-to-secondary heat imbalance is eliminated by approximately 600 seconds. Thereafter the NSSS enters into a ouasi-steady state with a very gradual cooldown and depre';surization due to decreasing core decay heat and with emergency feedwater flow which was initiated at 90.0 seconds maintaining an adequate 10 liquid inventory within the intact steam generator for heat removal. By 1800 seconds the operator initiates a controlled cooldown to shutdown cooling utilizing the atmospheric dump valves.
i The minimum DNBR vs. Time as shown on Figure 158-30 remains above 1.19 throughout the transient.
During the first 30 minutes following the initiation of this LF1 event mass releases from the system amount to 2970 lbm of stean from the pressurizer safety valves to the reactor drain tank, 79,700 lbm of steam from the main steam safety valves to atmosphere, and 69,200 lbm of liquid and 34,200 lbm of steam from the feedwater line break to containment. Steam release to the reactor drain tank may burst the tank's rupture disc discharging its contents to containment.
During this event, two scurces of radioactivity contribute to the site boundary r dose, the initial activity in the steam generator inventory, and the activity 1 associated with primary to secondary leakaoe from the steam generator tubes which are assumed to be at the technical specification limits of 0.1 pCi/gm and 4.6 uCi/gm dose equivalent I-131 respectively. During the first two hours of this l event, the total activity from the steam generators includes 8.9 Ci from the j affected stean generator to the containment building including 1.6 Ci i associated with technical specification tube leakage (1 gpm) and 0.33 Ci total j activity released from unaffected steam generator to the containment and '
atmosphere. Assuming all the radioactivity is released to the atmosphere, the offsite dose due to feedwater line break with loss of offsite power results in no more than 9.5 rem two hour inhalation thyroid dose at exclusion area bounda ry.
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58.5 CONCLUSION
These conservative methods, even when applied to the limiting case of Reference 1, produce an NSSS transient with maximum pressures not greater than 7
2843 psia in the RCS and 1318 psia in the steam generators which is sufficiently low to ensure that feedwater line break with loss of offsite power produces a radiological dose which is well within 10CFR100 guidelines.
The minimum DNBR which remains above 1.19 indicates that no fuel cladding failure occurs.
158.6 REANALYSIS OF SMALL BREAK LOSS OF FEEDWATER INVENTORY EVENTS WITH THE LIMITING SINGLE FAILURE AND OFFSITE POWER AVAILABLE 15B.
6.1 INTRODUCTION
15B.6.1.1 Purpose The purpose of this reanalysis is to show that the results of the small break loss of feedwater inventory event with the limiting single failure and offsite power available produce maximum pressures less than 110% of design.
15B.6.1.2 Background l
The loss of feedwater inventory event presented in Section 15B.4 demonstrates that breaks of all sizes, when combined with the loss of offsite power, produce maximum pressures well below 120% of g
pi design. Based on the recurrence frequences provided in Reference 3, V the NRC has concluded that the 120% of design maximum pressure criterion is appropriate for large break loss of feedwater inventory events, and small break loss of feedwater inventory events combined with the loss of offsite power. However, as is stated in Reference 4, it must be shown that small break loss of feedwater inventory events with the limiting single failure and offsite power available meet the maximum pressure criterion of 110% of design.
In order to demonstrate compliance with this criW'on, a reanalysis of small breaks with a modified methodology was r voired. The methodology used in Section 158.4 is applicable the full spectrum of break sizes. However, it is extremely conservative when applied l to the smaller break sizes. As a result, a new method of analysis l which is still conservative was developed, and is discussed in the following section.
Since the recurrence frequencies presented in Reference 3 apply to pipsg"eaterthan6inchesindiameter,theregnalysisneedonly consider beaks less than approximately 0.20 ft . This is the same break size prcsented in Section 15B.4 as the limitinq break with the origind method, logy. Therefore, in the followin sections "small" breaks rer er tc those which are less than 0.20 ft l
15B-9 Amendment Number 8 May 10,1983
15B.6.2 METHOD OF ANALYSIS 15B.6.2.1 Mathematical Models
'b The methodology used in the reanalysis of small break loss of feed-water inventory events is the same as that applied in Section 15B.4 and described in Section 158.3 with the exception of the treatment of steam generator heat transfer and reactor trip on steam generator low water level . Predictions of steam generator heat transfer and level behavior are based on the model documented in References 5 through 8. As discussed below, this model is conservative when applied to the small break loss of feedwater inventory events.
Steam Generator Heat Transfer RCS pressurization is largely a function of the rate at which the ruptured steam generator's heat transke decreases as its inventory is depleted. (The " ruptured" generator refers to the steam generator nearest the pipe break). Section 15B.3 documents the sensitivity of RCS pressurization to steam generator heat transfer behavior. The study verified that RCS pressurization is maximized by under-estimating the affected steam generator liquid mass corresponding to the initiation of heat transfer degradation (i.e. ,
over-estimating the rate of heat transfer decrease). The original methodology took a simplistic and clearly conservative approach by assuming heat transfer degradation was instantaneous upon steam generator dryout. However, this approach is modified in order to more realistically predict the behavior.
A gradual heat transfer reduction is expected as the steam generator tubes are exposed to increasing void fractions which force the tubes from the normal nucleate boiling heat transfer regime into transition boiling and eventually into liquid deficient heat transfer. Transition boiling is anticipated when the local void fraction exceeds 0.9 (Reference 9). Liquid deficient heat transfer develops when local qualities approach 0.9. Under full power conditions and utilizing the steam generator model documented in References 5 through 8, the onset of these heat transfer regimes corresponds to steam generator liquid inventories of approximately 70,000 lbm and 35,000 lbm, respectively for the System 80 design.
However, the referenced model conservatively ignores the transition .
boiling regime, thereby delaying heat transfer degradation until j fluid conditions correspond to liquid deficient heat transfer. j Therefore, the modified treatment of steam generator heat transfer l behavior is conservative, since it ender-estimates the liquid mass associated with the initiation of heat transfer degradation.
Steam Generator Low Water Level Trip As discussed in Section 15B.3, the original loss of feedwater inventory event method credited low water > /el trip in the ruptured steam generator only after its liquid inventory had been depleted.
This assured conservative treatment of low level trip even if the loss of feedwater inventory event caused rapid steam generator depressurization (i.e., large breaks) and consequent swelli g of the i 15b10 Amendment Number 8 ,
May 10,1983
downcomer level due to flashing of the downcomer liquid. However,
(
(
for sufficiently small breaks the steam generator pressure remains constant or increases prior to reactor trip and no downcomer level swell will occur due to flashing. Therefore, in the reanalysis of small break loss of feedwater inventory events steam generator low water level trip is credited with a larger liquid inventory remaining. 1
('
For the System 80 design steam generators, the low level trip setpoint corresponds to a downcomer liquid level of approximately 24 feet above the tube sheet and a liquid inventory of over 70,000 lbm under full power conditions (based on the reference steam generator model). However, the reanalysis of small break loss of feedwater inventory events conservatively delays low level trip until heat transfer degradation begins with approximately 35,000 lbm of liquio remaining in the ruptured steam generator.
The NSSS response to the small break loss of feedwater inventory 4 event with the limiting single failure and offsite power available, '
was modeled using the CESEC computer program described in Section 15.0. In addition, the input to the CESEC code was modified to account for th9 steam generator low level trip and heat transfer degradation methodology described in the previous paragraphs.
158.6.2.2 Input parameters and initial conditions I
s The input parameters and initial conditions used to analyze the NSSS i response are discussed in Section 15.0. The initial conditions for i \ the principal process variables were varied within the range given in Table 15.0-5 to determine the set of initial conditions shown in Table 15B-3.
In addition to conservatively delaying steam generator low level trip coincident with the assumed heat transfer degradation, the initial primary system pressure was adjusted within the range specified in Table 15.0-5 to achieve, where possible, a coincident l reactor trip signal on high pressurizer pressure. This maximizes the primary pressurization potential of the small break loss of feedwater inventory event, by maximizing the primary system pressure at the time of the reactor trip.
To determine the limiting single failure of the loss of feedwater inventory event with offsite power available, Table 15.0-6 was used.
There are no single failures identified in this table which can adversely impact the consequences (i.e., pressurization) associated with the loss of feedwater inventory event. As a result of the evaluation method applied to the loss of feedwater inventory analysis, the only mechanisms for mitigation of the reactor coolant system (RCS) pressurization are the pressurizer safety valves, the reactcr coolant flow and the main steam safety valves. The last two influence the RCS-to-steam generator heat transfer rate.
15B-11 Amendment Number 8 May 10,1983
There are no credible failures which can degrade pressurizer safety valve or main steam safety valve capacity. Nor are there any crediblefgureswhichcanreducesteamflowtotherupturedsteam A decrease in RCS to steam generator heat transfer generator.
due to reactor coolant flow coastdown can only be caused by a failure to fast transfer to offsite power or a loss of offsite power following turbine trip (i.e., two or four pump coastdown, respectively). Because offsite power is assumed to be available for this analysis, the failure to fast transfer is assumed following the turbine trip. This results in the coastdown of two reactor coolant pumps in diagonally opposite loops.
2 A spectrum of small breaks, of size less than or equal to 0.20 ft ,
were analyzed using the methodology described in the preceeding paragraphs to determine the limiting break size. The results of this analysis are provided in Figure 158.31 which plots maximum primary pressure vs. breag size. As can been seen, the limiting break size is the 0.20 ft break.
The reason that the largest break produces the most adverse pressurization is due to the more rapid degradation of heat transfer in the ruptured steam generator. The rate of heat transfer degradation is a major factor that determines the primary coolant pressurization of the event (i.e. , the more rapid the reduction in steam generator heat transfer, the greater the primary pressurization). As was previously stated, heat transfer degradation is conservatively assumed to begin when the ruptured steam generator inventory decreased to 35,000 lbm. The larger break sizes require a shorter time interval to deplete this remaining inventory, resulting in a more rapid heat transfer degradation, and greater primary coolant pressurization.
Detailed results of this limiting break size are presented in the following section.
158.6.3 RESULTS The dynamic behavior of the important NSSS parameters following the small break loss of feedwater inventory event with the failure to fast transfer to offsite power following turbine trip is presented in Figures 158-32 to 39. The sequence of events provided in Table 15B-4 summarizes the important results of this event.
(1) It should be noted that the coincident occurrences (failua.s) considered in Chapter 15 do not include spurious independent failures, only consequential failures and pre-existing failures.
Accordingly, spurious closure of a main steam isolation valve is not considered credible during the loss of feedwater inventory event. .
15B-4 summarizes the important results of this event. (
O 158-12 Amendment Number 8 May 10,1983
f- A 0.20 ft 2rupture in the main feedwater line is assumed to instantaneously
( terminate feedwater flow to both steam generators and establish critical flow V) from the generator nearest the break an initial rate of 1979 lbm/sec. This causes a decrease in steam generator liquid mass as shown by Figure 15B-39. 8 The break discharge enthalpy is assumed to remain that of saturated liquid until the ruptured steam generator empties, at which time saturated vapor enthalpy is assumed.
The absence of subccoled feedwater flow causes a constant heatup and pressurization of the steam generators during the first 25.98 seconds which 10 reduces the primary-to-secondary heat transfer rate. Rising primary coolant temperatures and pressures result. Due to the temperature reactivity feedback 8 during this period core power is reduced from an initial value of 102% to 99.8% at 25.98 seconds. I 10 At 26.98 seconds the ruptured steam generator produces a low water level l 10 reactor trip signal. This reactor trip signal is coincident with a high 8 l
pressurizer pressure trip signal. At 25.98 seconds heat transfer in the 10 ruptured steam generator begins to degrade due to insufficient inventory. 8 This degradation initiates a rapid heat up and pressurization of the reactor coolant system. At 27.13 seconds the reactor trip breakers open followed by 10 an assumed instantaneous turbine trip at 27.97 seconds. Immediately following turbine trip, the failure to fast transfer to offsite power occurs, resulting in the coastdown of two reactor coolant pumps. These occurrences further aggravate the primary pressurization.
Closure of the turbine leaves the pipe break as the only steam relief path, thereby reducing the energy flow from the intact steam generator below that of the primary-to-secondary heat transfer rate. The resulting steam generator pressurization reduces the primary-to-secondary temperature difference. In addition, the loss of reactor coolant flow following the loss of electrical power to two pumps decreases the heat transfer coefficient of the coolant in the steam generator tubes. A significant heat transfer reduction occurs. 8 Compression of the pressurizer steam volume due to the high insurge flow raises the pressure to the safety valve setpoint at 28.3 seconds. Thereafter, every increase in the surge flow causes a slight pressurization which opens the safety valves such that their volumetric discharge rate matches that of '
the insurge. At 30.2 second.s, the surge line flow reaches its maximum value of 1458 lbm/sec.
At this point in time, the reactor coolant system pressure is at a maximum of 2712 psia. Also, the increased pressure establishes a surge line pressure gradient which provides sufficient flow to allow the reactor coolant to expand O
158-13 Amendment No. 10 l June 28, 1985 j l
a
the reactor coolant to expand under the existing heatup with no ,
further pressurization. The rate of heatup decreases subsequent to core heat flux decay, causing primary pressures to drop.
At 30.0 seconds the main steam safety valves opened stablizing the secondary side temperature and allowing the rising primary coolant temperature to develop greater heat transfer to the intact s?am genera ter. The intact generator is forced to a maximum of 1342 psia at 33.8 seconds before the heat transfer begins to decrease. The core-to-steam generator heat mismatch is reduced sufficiently by 37.4 seconds to allow closure of the pressurizer safety valves, and the reactor coolant system enters a cooldown. Under the influence of steam blowdown through the ruptured steam generator to the break, the cooldown proceeds even after the steam generator safety valves close.
Subsequently, a main steam isolation signal is generated on low steam generator pressure which closes the mains team isolation valves, decoupling the intact steam generator from the ruptured steam generator and the break. The intact steam generator repressurizes, thereby reducing its heat transfer and eventually causing a primary system heatup. With the main steam safety valves re-opening, the primary-to-secondary heat imbalance is eliminated shortly thereafter. The NSSS enters into a quasi-steady state with a very gradual cooldown and depressurization due to decreasing core decay heat and with emergency feedwater flow maintaining an adequate liquid inventory within the intact steam generator for heat removal.
By 1800 seconds the operator initiates a controlled cooldown to shutdown cooling utilizing the atmospheric dump valves.
158.
6.4 CONCLUSION
This evaluation shows that the plant response to the limiting small feedwater line break event with the most adverse single failure with offsite power available produces a maximum RCS pressure which is within 110% of design (2750 psia).
i O1 1 15B-14 Amendment Number 8 j May 10,1983 >
L References for Appendix 15B
- 1. "USNRC Standard Review Plan, Section 15.8.8, Feedwater System Pipe Breaks Inside and Outside Containment (PWR)", NUREG-75/087, November 24, 1975.
i
- 2. R.E. Henry, 4.K. Fauske, "The Two Phase Critical Flow of One-Component 4 Mixtures in Nozzles, Orifices, and Short Tubes", Journal of Heat Transfer, Transactions of the ASME, May,1971.
- 3. " Response to NRC Round One Question 440.42 on the CESSAR-FSAR", enclosure to Letter LD-81-069, A. E. Scherer to D. G. Eisenhut, dated October 8, .
I 1981.
- 4. " Safety Evaluation Report Related to the Final Design Approval of the Combustion Engineering Standard Nuclear Steam Supply System (CESSAR)"
NUREG-0852 (Section 15.3.2).
- 5. CENPD-107 Supplement 1, "ATWS Model Modification to CESEC," September 1974. (Section 3.0).
- 6. CENPD-107 Supalement 1, Amendment 1-P, "ATWS Model Modification to CESEC," Novemaer 1975. (Section 3.3).
- 7. CENPD-107 Supplement 3 "ATWS Model Modification to CESEC," August-1975.
TSections 240.8, 240.11 and 240.9).
O 8. CENPD-107 Supplement 4, "ATWS Model Modification to CESEC," December j Q 1975. (Section 1.6, 1.8 and 4.2).
- 9. Forced Convection Boiling Studies, Final Report on Forced Convection Va Jorization Project V. E. Schrock and L.M. Grossman, TID-14632 (1959).
I l
O 158-15 Amendment Number 8 '
May 10,1983
O THIS PAGE INTENTIONALLY BLANK.
O l
l 9
TABLE 158-1
/ ASSUMPTIONS FOR THE LIMITING CASE
( )) LOSS OF FEEDWATER INVENTORY EVENT Nominal Assumed Parameter Value Value Initial Core Power, MWt 3800 3876 Initial Core Inlet Temperature, F 565 560 Initial Reactor Vessel Flow, GPM 446000 446000 Initial Pressurizer Pressure, psia 2250 1920 Fuel Gas gap Heat Transfer Coefficient >540 540 BTU /HR-ft -F Doppler Coefficient Multiplier 1.0 1.0 Pressurizer Safety Valves Rated Flow, lbm/hr >460000 460000 3
Initial Pressurizer Liquid Volume, feet 930 1120 Initial Steam Generator Inventory, lbm 173000 173000 g3 Initial Feedwater Enthalpy, BTV/lbm 430 376 Steam Bypass Control System Automatic Manual ]
Normal On-Site or Off-Site Electrical Avai? Sie Unavailable Power After Turbine Trip !
2 Feedwater Pipe Dreak Area, feet --
0.2 CEA Worth at Trip, 10-2 4 -14.8 -10.0 l 4
I f~h, b
Amendment Number 8 May 10,1983
TABLE 15B-2 8 (Sheet 1 of 3)
SEQUENCE OF EVENTS FOR THE LIMITING CASE LOSS !
l 0F FEEDWATER INVENTORY EVENT 8 l Time Setpoint (Sec) Event or Value 2 10 0.0 Break in the liain Feedwater Line, ft 0.2 i
0.0 Instantaneous Loss of All Feedwater Flow to Both Steam Generators 8
0.0 Instantaneous Development of Critical --
Flow from the Ruptured Steam Generator to the Break ,
)
33.82 Instantaneous Loss of All Heat Transfer --
to the Ruptured Steam Generator 33.82 Steam Generator Water Leve' Aeaches Empty Reactor Trip Analysis Setpoi.it in the Ruptured Generator 1
33.82 Steam Generator Water Level Reaches Empty '
Emergency Feedwater Actuation Signal Analysis Setpoint in the Ruptured 10 Generator 33.82 Pressurizer Pressure Reaches Reactor 2475 Trip Analysis Setpoint, psia 34.6 Pressurizer Safety Valves Open, psia 2525 34.82 Low Water Level Trip Signal Generated --
34.82 Emergency Feedwater Actuation Signal --
Generated 3t.82 High Pressurizer Pressure Trip Signal --
Generated 34.97 Trip Breakers Open --
35.8 instantaneous Closure of the Turbine --
Stop Valves 8 O
Amendment No. 10 June 28, 1985
v I
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TABLE 15B-2 (Cont'd.) (Sheet 2 of 3) 10 SEQUENCE OF EVENTS FOR THE' LIMITING CASE LOSS OF FEEDWATER INVENTORY EVENT ,
Setpoint 8 Time 4 (Sec) Event or Value 35.8 Loss of Normal On-Site and 0ff-Site --
Electrical Power 36.4 Sicam Generator Water Level Reaches 35 i
Reactor Trip Analysis Setpoint in the Intact-Generator, percent oF wide range Maximum Reactor Coolant Pressure, psia 2843 I l
38.2 l
Maximum Pressurizer Pressure, psia 2587 Maximum Pressurizer Surge Line Flow, 2206 lbm/sec 40.5 Main Steam Safety Valves Open 1282 44.0 Steam Generator Water Level Reaches 10 Emergency Feedwater Actuation Signal Analysis Setpoint in the Intact 10 Generator, percent of wide range 45.0 Emergency Feedwater Actuation Signal Generated 44.8 Maximum Steam Generator Pressure, psia 1318 45.4 Pressurizer Safety Valves Close, psia 2525 45.8 Minimum Pressuriz'er Steam Volume, ft 3 138 73.8 Main Steam Safety Valves Close, psia 1218 90.0 Emergency Feedwater Flow Initiated 875 to the Intact Steam Generator, gpm 165.0 Steam Generator Pressure Reaches 810 Main Steam Isolation Signal Analysis Setpoint, psia l
Amendment No.10 June 28, 1985
l l
TABLE 15B-2 (Cont'd.) (Sheet 3 of 3) l 10 .
t i
SEQUENCE OF EVENTS FOR THE LIMITING CASE LOSS !
0F FEE 0 WATER INVENTORY EVENT 8
Time Setpoint (Sec) Event or Value 166.0 Main Steam Isolation Signal Generated 170.6 Minimum Intact Steam Generator 8100 Liquid Mass, lbm 10 170.6 Main Steam Isolation Valves Closed 314.2 Main Steam Safety Valves Open, psia 1282 1800.0 Operator Opens the Atmospheric Stream Dump Valves to Begin Plant Cooldown to Shutdown Cooling O
I i
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O Amendment No.10 June 28, 1985
TABLE 15B-3 ASSUMPTIONS FOR THE REANALYSIS OF THE LIMITING SMALL. BREAK LOSS OF FEEDWATER INVENTORY EVENT Assumed Parameter Value Initial Core Power, MWt 3893 l Core Inlet Temperature, F 560 0
Core Mass Flowrate, 10 lbm/hr 164.9 i
Reactor Coolant System P sure, psia 2115 Steam Generator Pressure, psia 1026 CEA Worth for Trip,10-2 ap -10.0 Pressurizer Safety Valves Rated Flow, lbm/hr 460,000 Initial Pressurizer Liquid Volume, tc 3 1120 Initial Steam Generator Inventory, lbm 173,000 Feedwater Pipe Break Area, ft 2 0.20 Steam Bypass Control System Manual Pressurizer Pressure Control System Manual Pressurizer Level Control System-- Manual l
D U
Amendment Number 8 May 10,1983
TABLE 15B-4 SEQUENCE OF EVENTS FOR THE REANALYSIS OF THE LIMITING SMALL BREAK I
LOSS OF FEEDWATER INVENTORY EVENT Time Setpoint 8 (sec) Event or Value '
l 2
0.0 Rupture in the Main Feedwater Line, ft 0.20 0.0 Complete Loss of Feedwater to Both Steam Generators ----
0.0 Initial Steam Generator Break Flow, lbm/sec 1079 25.98 Pressurizer Pressure Reaches Reactor Trip Analysis Setpoint, psia 2475 25.98 Steam Generator Water !r.ei Reaches Reactor 35000 Trip Analysis Mtgoint in the Ruptured Genera +m , ib 1iquid remaining m
25.98 Heat Transfer Degradation in Ruptured SG Begins ----
26.98 High Pressurizer Pressure Trip Signal Generated ----
10 26.98 Low Water Level Trip Signal Generated ----
]
27.13 Reactor Trip Breakers Open ----
27.97 Turbine Trip on Reactor Trip ----
27.97 Failure to Fast Transfer - Two Reactor Coolant Pumps Coast Down ---- l 3
28.3 Pressurizer Safety Valves, psia 2525 30.0 Main Steam Safety Valves Open 1282 30.2 Maximum Surge Line Flcw, lbm/sec 1458 8
30.2 Maximum RCS Pressure, psia 2712 33.8 Maximum Steam Generator Pressure, psia 1342 36.8 Ruptured SG Dries Out ----
37.a Primary Safety Valves Close, psia 2523 O
Amendment No.10 June 28,1985
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i, RUPTURED
~
J s
M \
i 250 \
l
\
\
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g b____4_____i______4_____L__-__
0 300 600 900 1200 1500 1800 TIME, SECONDS 1
Amendment No. 7 March 31, 1982 C-E LOSS OF FEEDWATER INVENTORY Rgure E LIMITING CASE ISB-23 STEAM GENERATOR PRESSURES vs TIME
'1 O 4500 i. i i i i l
3250 -- -
8 : i y2000 -- -
1
$ i i i l b 750 INTACT STEAM GENERATOR _
E
$ 4 g _ _ _ _ _
-500 --
R RUPTURED STEAM GENERATOR
-1750 -- -
I I I I I
-3000 0 300 600 900 1200 1500 1800 TIME, SECONDS Amendment No. 7 '
March 31, 1982 C-E L0SS OF FEEDWATER INVENTORY Figure LIMITING CASE- ISB-24 TOTAL STEAM FLOW vs TIME l
l l
4500 i i i i 1
3250 g -
0 2000 - -
2
$ 1 B INTACT STEAM GENERATOR
$ 750 - -
1 5 _ _ _ _ _ _ ____________________
o
-500 -- -
" ] RUPTURED 1 STEAM GENERATOR I
i -1750 - -
I I I I
-3000 0 40 80 120 160 200 TIME, SECONDS Amendment No. 7 March 31, 1982 C-E LOSS OF FEEDWATER INVENTORY Revre LIMITING CASE 15B-25 TOTAL STEAM FLOWS vs TIME
O 2400 i i i ' '
2000 sw iiE - -
cg 1600 ]
3:
b d 1200 - -
E 5
m O 5 6
800 - -
l s
400 -
1 1 I I 0
0 300 600 900 1200 1500 1800 TIME, SECONDS 11 l-Amendment No. 7 March 31, 1982-c-E LOSS OF FEEDWATER INVENTORY Figure g LIMITING CASE BREAK DISCHARGE FLOW vs TIME 15B-26 l
l 1
O
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1150
- E B
1000 g
b I
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5 5
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$5 s
l 550 4 -
I I I I I 400
! O 300 600 900 1200 1500 1800 TIME, SECONDS l-l j
Amendment No. 7 March 31, 1982 C-E LOSS OF FEEDWATER INVENTORY Figure LIMITING CASE 15B-27 BREAK DISCHARGE ENTHALPY vs TIME
--___________._-____m___
180 i i i i i E
$ 150 -
vi
$120 2 '
ca l 'D l 90 8
ti
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E 05 b;
30 INTACT D , , , ,
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_ STEAM GENER ATOR WATER LEVEL vs TIME
i
(
4.0 3,0- -
2.0 -
a
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=
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0 O 10 20 30 40 50 TIW,SEC000S Amendment No. 7 March 31,1982 C-E Tir [.]glg.3.j g.qq y3 7p.r t0R 'Or LOSS OF F8,f1't ItfEHORY II9 "
2@8855 6[/ APKt0lX15B 15E-30
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p -
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$ 2650 BREAK MAXIMUM EE AREA PRESSURE m (SQ. FT.) (PSIA) o lO 0.01- 2684-5
- ~
@ 2600 0.05 2686 E
g 0.10 2697 s 0.20 2712 2550
- INCLUDES ELEVATION AND REACTOR COOLANT PUMP HEADS 2500 0 0.05 0.10 0.15 0.20 BREAK SIZE, SQ. FT.
Amendment Number 8
- May 10,1983 C-E REANALYSIS OF SMALL LOSS OF pS"
/ FEEDW ATER INVENTORY EVENTS 88f9f8//// MAXIMUM RCS PRESSURE
- vs BREAK AREA 15B-31 ,
J
>OV 120 i i i 1 l 100 f5 3
2 80 - -
d i
2 8
g 60 - -
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a.
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0 10 20 30 40 50 TIME, SECONDS Amendment Number 8 May 10,1983 REANALYSIS OF SMALL LOSS OF C-E pS**
FEEDWATER INVENTORY EVENTS - LIMITING CASE 55 / - l CORE POWER vs TIME 15B -32 I
I i
120 i i i i l
l 100 -
8 CL 1
La i l
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M 60 - -
E '
D i m
g 40 - -
s E
8 20 - -
' ' ' I 0
0 10 20 30 40 50 TIME, SECONDS I
Amendment Number 8 May 10,1983 C-E REANALYSIS OF SMALL LOSS OF p' 8 "
FEEDWATER INVENTORY EVENTS - LIMITING CASE S9E7 CORE HEAT FLUX vs TIME 15B-33
O PLER
\ MODERATOR e.
$ -3 _ _
s M
e l p
N 0
l 6x L ,
EAs d
-11 0 10 20 30 ;0 50 TIME, SECONDS Amendment Number 8
_ May 10,1983 REANALYSIS OF SMALL LOSS OF C-E ' p' S *
- FEEDWATER-INVENTORY EVENTS - LIMITING CASE SEf8. / REACTIVITIES vs TIME 15B-34
)
)
l l
/m U
670 i i i 650 - -
u_
. OUTLET O 630 - -
E
!E Ei a.
- s 610 -
AVERAGE l $ %
z A $
U o o
" 590 m INLET 8
o l
570 - -
550 0 10 20 30 40 50 TIME, SECONDS 1
Amendment Number 8 May 10, 1983 C^E REANALYSIS OF SMALL LOSS OF
/
// FEEDWATER INVENTOR" EVENTS - LIMITING CASE p'S$*
15B 35 S2I@ CORE COOLANT TEMPER ATURES vs TIME
l 24000 ; i i i 20000 LOOP WITH -
INTACT STEAM 1 GENERATOR i S
w l
$16000 a
Ef LOOP WITHA-S RU PTURED z 12000 -
STEAM-GENERATOR 5
l 8O e
8000 se 4000 -
0 0 10 20 30 40 50 TIME, SECONDS Amendment Number 8 May 10,1983 C-E #
REANALYSIS OF SMALL LOSS OF s 8 y,,
FEEDWATER INVENTORY EVENTS - LIMITING CASE f SEE8E.l// t REACTOR COOLANT FLOW vs TIME 15B-36
____._____._..____m._______________________m_____m _ _ _ _ _ _ _ _ _ . . _ _ _ ___ - _ _ _ ...__.__..___.______m--.._...m
u i
O
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l 2700 REACTOR /
COOLANT SYSTEM
- E '7 -
g 2500 -
PRESSURIZER - -
E a ,
C EE2300 - -
s b
O =
g2100 2
EG c_
1900 - -
- DOES NOT INCLUDE ELEVATION OR <
REACTOR COOLANT PUMP HEADS 1700 0 10 20 30 40 50 TIME, SECONDS Amendment Number 8 May 10,193
.i REANALYSIS OF SMALL LOSS OF c~E geut, FEEDWATER INVENTORY EVENTS - LIMITING CASE Sf25 PRIMARY SYSTEM PRESSURES vs TIME ISB _37-
l' l
- O 1400 i i i i l
INTACT j STEAM GENERATOR 1300 - -
RU PTURED
< STEAM G GENERATOR
' 1200 - -
d E
0; g 1100 - -
c O
1 <C 51000 - -
( &
5 w
900 - -
l 800 l 0 10 20 30 40 50 TIME, SECONDS l
Amendment Number 8 May 10,1983 I
c-E REANALYSIS OF SMALL LOS'S OF p' S "
FEEDWATER INVENTORY EVENTS - LIMITING CASE Sff5 STEAM GENERATOR PRESSURES vs TIME ISB-38 ,
E_ _ _-.
180000 i i i i 150000 - -
3
. INTACT STEAM
$120000 .
GENERATOR -
2 a
E
$90000 - -
h RUPTURED STEAM 23 GENERATOR h @60000 5
w w
30000 - -
0 0 10 20 30 40 50 TIME, SECONDS Amendment Number 8 May 10,1983
~
C-E / REANALYSIS OF SMALL LOSS OF r;gur, y FEEDWATER INVENTORY EVENTS - LIMITING CASE 1888d'./ STEAM GENERATOR LIQUID MASS vs TIME 15B-39 l
__ ___ _ _ - - _ _ _ - _ _ _ _ -