ML20070K071
| ML20070K071 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 06/30/1994 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20070K060 | List: |
| References | |
| DPC-NE-3002, DPC-NE-3002-R01, DPC-NE-3002-R1, NUDOCS 9407260097 | |
| Download: ML20070K071 (69) | |
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DUKE POWER COMPANY MCGUIRE NUCLEAR STATION CATAWBA NUCLEAR STATION FSAR CHAPTER 15 SYSTEM TRANSIENT ANALYSIS METHODOLOGY DPC-NE-3002 Revision 1
/
June 1994 4
Safety Analysis Section Nuclear Engineering Division Nuclear Generation Department Duke Power Company I
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9407260097 94o73g PDR.
ADOCK 05000369 PDR
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FSAR CHAPTER 15. SYSTEM TRANSIENT ANALYSIS METHODOLOGY Table of Contents
1.0 INTRODUCTION
2.0 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 2.1 Feedwater System Malfunctions That Result In A Reduction In.
Feedwater Temnerature 2.2 Feedwater System Malfunction Causino an Increase in 3
Feedwater Flow 2.2.1 Nodalization 2.2.2 Initial Conditions 2.2.3 Boundary Conditions 2.2.4 Control, Protection, and Safeguards Systems Modeling 2.3 Fxcessive Tncrease in Secondary Steam Flow 2.3.1 Nodalization i
2.3.2 Initial Conditions 2.3.3 Boundary Conditions 2.3.4 Control, Protection, and Safeguards Systems Modeling 2.4 Inadvertent Oneninc of a Steam Generator Relief or Safety Valve 3.0 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 3.1 Turbino Trin 3.1.1 Peak RCS Pressure Analysis 3.1.1.1 Nodalization 3.1.1.2 Initial Conditions 3.1.1.3 Boundary Conditions 3.1.1.4 Control, Protection, and Safeguards System Modeling 3.1.2 Peak Main Steam System Pressure Analysis 3.1.2.1 Nodalization 3.1.2.2 Initial Conditions 3.1. 2. 3 Boundary conditions 3.1.2.4 Control, Protection, and Safeguards System Modeling 3.2 Loss of Non-Emeroency AC Power To The Station 3.2.1 Peak RCS Pressure Analysic 3.2.1.1 Nodalization 3.2.1.2 Initial Conditions 3.2.1.3 Boundary Conditions 3.2.1.4 Control, Protection, and 5sfeguards System Modeling 3.2.2 Peak Main Steam System Pressure Analysis 3.2.2.1 Nodal 1zation 3.2.2.2 Initial Conditions 3.2.2.3 Boundary Conditions 3.2.2.4 Control, Protection, and Safeguards System Modeling 3 2.3 Core cooling capability Analysis - short Tern l
3.2.3.1 Nodalization 3.2.3.2 Initial Conditions 3.2.3.3 Boundary Conditions i
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e-3.2.3.4 Control, Protection, and Safeguards System Modeling 3.2.4 Core Cooling capability Analysis - Long Term 3.2.4.1 Nodalization 3.2.4.2 Initial Conditions 3.2.4.3 Boundary Conditions 3.2.4.4 Control, Protection, and Safeguards System Modeling 3.3 Loss Of Normal Feedwater 3.3.1 Peak RCS Pressure Analysis 3.3.1.1 Nodalization 3.3.1.2 Initial Conditions 3.3.1.3 Boundary Conditions 3.3.1.4 Control, Protection, and Safeguards System Modeling 3.3.2 Peak Main Steam System Pressure Analysis.
3.3.2.1 Nodalization j
3.3.2.2 Initial Conditions 3.3.2.3 Boundary Conditions 3.3.2.4 Control, Protection, and Safeguards System Modeling 3.3.3 Core Cooling Capability Analysis 3.3.3.1 Nodalization 3.3.3.2 Initial Conditions 3.3.3.3 Boundary Conditions 3.3.3.4 Control, Protection, and Safeguards System Modeling 3.4 Feedwater system Pine Break 3.4.1 Short Term Core Cooling Capability 3.4.1.1 Nodalization 3.4.1.2 Initial Conditions 3.4.1.3 Boundary Conditions 3.4.1.4 Control, Protection, and Safeguards System Mode 3.4.2 Long Term Core Cooling Capability 3.4.2.1 Nodalization 3.4.2.2 Initial Conditions 3.4.2.3 Boundary Conditions 3.4.2.4 Control, Protection, and Safeguards System Modeling 4.0 DECREASE IN REACTOR COOLAtTP SYSTEM FLOW RATE 4.1 Partial Loss of Forced Reactor Coolant Flow 4.1.1 Nodalization 4.1.2 Initial Conditions 4.1.3 Boundary Conditions 4.1.4 control, Protbetion, and Safeguards System Modeling 4.2 comolete Loss of Forced Reactor coolant Flow 4.2.1 Nodalization t
4.2.2 Initial Conditions 4.2.3 Boundary Conditions 4.2.4 Control, Protection, and Safeguards System Modeling 1
4.3 Reactor Coolant PUC7 Locked Rotor 4.3.1 Peak RCS Pressure Analysis 4.3.1.1 Nodalization j
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4.3.1.2 Initial Conditions 1
4.3.1.3 Boundary Conditions 4.3.1.4 Control, Protection, and Safeguards System Modelina l-iii i
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6.0 INCREASE IN REACTOR COOLANT ItNENTORY 6.1 Inadvertent oneration of Eccs Durina power coeration 6.1.1 Core Cooling Capability Analysis 6.1.1.1 Nodalization 6.1.1.2 Initial Conditions 6.1.1.3 Boundary Conditions 6.1.1.4 Control, Protection, and Safeguards System Modeling 6.2.1 Pressurizer Overfill Analysis 6.2.1.1 Initial Conditions 6.2.1.2 Boundary conditions 6.2.1.3 Control, Protection, and Safeguards System Modeling 7.0 DECREASES IN REACTOR COOLANT ItWENTORY 7.1 Inadvertent Ooenina of a Pressurizer safety or Relief 7.1.1 Nodalization 7.1.2 Initial Conditions 7.1.3 Boundary Conditions 7.1.4 Control, Protection, and Safeguards Systems Modeling 7.2 steam cenerator Tube Ruoture 7.2.1 Core Cooling capability Analysis 7.2.1.1 Nodalization 7.2.1.2 Initial Conditions 7.2.1.3 Boundary Conditions 7.2.1.4 Control, Protection, and Safeguards System Modeling 7.2.2 Offsite' Dose Calculation Input Analysis 7.2.2.1 Nodalization 7.2.2.2 Initial Conditions 7.2.2.3 Boundary Conditions 7.2.2.4 Control, Protection, and Safeguards System Modeling 8.O
SUMMARY
9.0 REFERENCES
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e 2.0 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 2.1 Feodwater System Malfunctions That Result In A Reduction In Feedwater Temnerature A Feedwater System malfunction that results in a decrease in feedwater temperature will cause an increase in core power by decreasing reactor coolant temperature.
Physically, as the cooler feedwater reduces the reactor coolant temperature, positive reactivity will be inserted due to the effect of a negative moderator temperature coefficient.
Postulating that the Rod Control System is in automatic control, control rods would be withdrawn as RCS temperature decreased, inserting additional positive reactivity.
The net effect on the RCS due to a reduction in feedwater temperature would be similar to the effect of increasing feedwater flow or increasing. secondary steam flow; the reactor will reach a new equilibrium condition at a power level corresponding to the new steam generator AT.
A Feedwater System malfunction that results in a decrease in feedwater temperature can be initiated from the following types of events:
spurious actuation of a feedwater heater bypass valve, interruption of steam extraction flow to a feedwater heater (s), spurious startup of a single auxiliary feedwater pump, failure of a single feedwater heater drain pump or failure of all feedwater heater drain pumps. The above events are examined, with the most limiting determined to be a spurious 1
actuation of a feedwater heater bypass valve.
Howev'er, under the current Duke Power company method of analysis, this accident is bounded by quantitative analysis of the increase in feedwater flow event or the excessive increase in secondary steam flow event. These events bound tha reduction in feedwater temperature event by producing a greater RCS cooldown. The applicable acceptance criterion is that fuel cladding integrity shall be maintained by ensuring that.the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations.
i 2.2 Feedwater System Malfunction Causind an Increase in Feedwater Flow The malfunctions considered are 1) the full opening of a single main-feedwater control valve, 2) an increase in the speed of a single main feedwater pump, 3) the spurious startup of a single auxiliary feedwater pump, or 4) a malfunction which affects more than one loop.
The latter scenario has been identified saly recently and is currently'bsing evaluated to determinc applicability to ricCuirc and Catawba. The limiting scenario from among those listed above is evaluated to demonstrate that fuel cladding integrity is maintained by ensuring that' the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations using the Statistical Core Design Methodology.
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9 2.2.1 Nodalization of the events identified in the previous section, the latter, the multi-loop malfunction, is expected to be the most limiting, and is therefore the one that is discussed. This transient affects all loops equally.and wou4 dis therefore be-analyzed with a single-loop NSSS system model (Reference 2, Section 3.2).
If the asst limiting transicnt is not I
determined to bc cac which affects all locps cqually, a multiple icop modci wsuid be selected appr;priatcly.
The pressurizer modeling includes the use of the local conditions hca; transfer option-for the sessci conductors.
2.2.2 Initial Conditions Core Power Level High initial power level maximizes the primary system heat flux.
The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurizer Pressure The nominal pressure corresponding to full power operation is assumed, I
with the pressure initial condition uncertainty accounted for in the Statistical Core Design Methodology.
Pressurizer Level Since this accident involves a reduction in RCS volume due to coolant contraction, a positive level uncertainty is applied to the nominal programmed level to minimize the initial pressurizer steam bubble volume and therefore maximize the pressure decrease due to contraction.
Reactor Vessel Averace Tomoerature The nominal temperature corresponding *to full power operation i.s assumed, with the temperature initial condition uncertainty accounted for in the Statistical Core Design Methodology.
RCS Flcw The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Evoass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in.the Statistical Core Design Methodology.
Steam Generator Level A negative level uncertainty is assumed to maximize the margin to a high-high steam generator narrow range level reactor trip due to any temporary steam /feedwater flow mismatch.
This maximizes the duration of the overcooling before it is ended by feedwater isolation.
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Fuel Tecoerature A low initial temperature is assumed to maximize the gap conductivity calculated for steady-state conditions and used for the subsequent transient.
A high gap conductivity minimizes the fuel heatup and attendant negative reactivity insertion caused by the power increase.
This makes the power increase more severe and is therefore conservative for DNBR evaluation.
steam Generator Tube pluccina In order to maximize the effects of the increased secondary system heat removal, no tube plugging is assumed.
2.2.3 Boundary Conditions Main Feedwater Figg A conservatively large step change in main feedwater flow to all steam generators is assumed at the initiation of the transient. A step decrease in main feedwater temperature is assumed to account for.the increased main feedwater flow rate.
2.2.4 Control, Protection, and Safeguards Systems Modeling Reactor Trin The pertinent reactor trip functions are the low-low steam generator level, high flux and overpower AT.
The safety analysis setpoint or the initial condition for the monitored parameter contains an allowance for measurement instrumentation uncertainty and setpoint setting tolerance.
Pressurizer Level Control No credit is taken for pressurizer level control. system operation to compensate for the depressurization which accompanies RCS volume shrinkage.
Rod Control This accident will result in a decrease in RCS temperature. With the Rod Control System in manual control, tThe reduced temperature will cause a positive reactivity insertion through the negative moderator temperature coefficient. With the Rod Control System in automatic control, in whichthe control rods may insert due to the mismatch between NI power and turbine power and cause a negative reactivity insertion.
However, since the reactor vessel average temperature is maintained at a programmed value, the control rods may withdraw in an attempt to maintain this temperature andwt++ cause a positive reactivity insertion as they are withdrawn in an attcupt to maintaia this temperature.
Both eesesautomatic and manual control of the Rod Control System are analyzed in order to ensure that the worse cac worst case is determined.
Turbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
In this mode any decrease in steam pressure, due for example to a shift from latent to sensible heat 2-3
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- jv transfer because of the overfeed, would be compensated for by an opening-of the turbine control valves to maintain impulse chamber pressure at the programmed value.
?J3' flow is cscdited, after thc appropriate Tcchnical Specifica;ica respcase tiac delay,
-han the safety analysis value of the 1;w icw steam generator narrow rangc level setpoint is reached.
In c; der to miniai;e thc post trip steam generator heat rencval, the miniaua aux;1iary feedwater flow is assumed.
ArW flow would be credited, after the appropriate Technical Specification response tdme delay, when the safety analysis value of the low-low steam generator level setpoint is reached. However, the parameter of interest for this transient has reached its limiting value j
before the appropriate Technical Specification response time delay has 4
elapsed. Therefore, no ArW is actually delivered to the steen generators.
Turbine Trio Turbine trip is credited, after the appropriate Technical Specification response time delay, on high-high steam generator narrow range level or on reactor trip.
Feedwater Isolation Feedwater isolation is credited, after the appropriate Technical Specification response time delay, on high-high steam generator narrow range level.
i, 2.3 Excessive Increase in Secondary Steam Flow The accident analyzed is a step increase in secondary steam flow of a magnitude equal to that for which the Reactor Control System is designed, 10% of licensed core thermal power.
Increases of larger magnitude are discussed in Section 2.4 and in Chapter 5 of Reference 1.
The accident is analyzed to demonstrate that fuel cladding integrity is maintained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology.
t 2.3.1 Nodalization The accident analyzed is an excessive increase in secondary steam flow at power.
Flow increases from a zero power initial condition are evaluated in Section 2.4 and in chapter 5 of Reference 1.
Per Reference 3, Section 15.1.4, the power level analyzed for this accident should' be 102% of licensed core thermal power for the number of loops initially assumed to be operating. At power, the Technical Specifications require all four loops to be operating.
Therefore full power is assumed as the initial condition.
An increase in steam flow to the turbine would affect all loops equally, therefore, a single-loop NSSS system model (Reference 2, Section 3.2) is used. Thc pressuri;er asdeling ;n:1udes l
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2.3.2 Initial Conditions Core Power Level Per Reference 3, Section 15.1.4, the power level analyzed for this accident should be 102% of licensed core thermal power for the~ number of loops initially assumed to be operating.
At~ power, the Technical' specifications require all four loops to be operating. Therefore full power is assumed as the initial condition. The uncertainty in initial power level is accounted for in the Statistical Core Design Methodology.
Pressurizer Pressure The nominal pressure corresponding to full power operation is assumed, with the pressure initial condition uncertainty accounted for in the Statistical Core Design Methodology.
Pressurizer Level Since this accident involves, particularly for the manual Rod Control System operation scenario, a reduction in RCS volume due to coolant contraction, a positive level uncertainty is assumed to minimize the initial pressurizer steam bubble volume and therefore maximize the pressure decrease duo to contraction.
Reactor Vessel Averace Tornerature The nominal temperature corresponding to full
,wer operation is assumed, with the temperature initial conditin uncertainty accounted for in the Statistical Core Design Methodology.
RCS Flow The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Bvoass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
Steam Generator Level The results of this transient are not sensitive to the direction of steam generator level uncertainty as long as the transient level response is kept within the range that avoids protection or safeguards actuation.
Fuel Temoerature A low initial temparature is assumed to maximize the gap conductivity.
calculated for steady-state conditions and used for the subsequent transient.
A high gap conductivity minimizes the fuel heatup and attendant negative reactivity insertion caused by the power increase.
This makes the power increase more severe and is therefore conservative for DNB evaluation.
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m Steam Generator Tube Plucaina In order to maximize the effects of the increased secondary system heat removal, no tube plugging is assumed.
2.3.3 Boundary Conditions Main Steam Flow A step change in main steam flow to the turbine equal to 10% of full power flow is assumed at the initiation of the transient.
2.3.4 Control, Protection, and Safeguards Systems Modeling Reactor Trin How'ver, The reactor is not expected to trip for this transient.
e reactor trip is credited, after the appropriate Technical Specification response time delay, if the safety analysis setpoint is exceeded for any reactor trip function.
Pressurizer Level Control No credit is taken for pressurizer level control system operation to compensate for the depressurization which accompanies RCS volume shrinkage.
Steam Line PORVs and condenser St eam Dumn While the steam line PORVs and steam dump might be a source of the increased steam flow in this postulated accident, the case analyzed assumes the increased flow exits to the turbine.
Steam Generator Level Control The results of this transient are not sensitive to the mode ot steam generator level control as long as the level is kept within the range l
that avoids prneection or safeguards actuation.
MFW Pumn Sneed Cont M The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Rod Control This accident will result in a decrease in RCS temperature. With the Rod Control System in manual control, the reduced temperature will cause a positive reactivity insertion through the negative moderator r
temperature coefficient. With the Rod Control System in automatic control, in which the reactor vessel average temperature is maintained at a programmed value, the control rods will cause a positive reactivity insertion as they are withdrawn in an attempt to maintain this temperature.
Both cases are analyzed in order to ensure that the worse one is considered.
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3.0 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 3.1 Turbine Trin The turbine trip event causes a loss of heat sink to the primary system.
The mismatch between power generation in the primary system and heat removal by the secondary system causes temperature and pressure to increase in the primary and secondary until reactor trip and/or lift of the pressurizer safety valves and main steam safety valves. The transient is analyzed to ensure that both the peak Reactor Coolant System pressure and the peak Main Steam System pressure remain below the acceptance criterion of 1101 of design pressure.
Peak RCS pressure and peak Ma,in Steam System pressure are analyzed separately.due to the differences in assumptions required for a conservative analysis.
3.1.1 Peak RCS Pressure Analysis
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i 3.1.1.1 Nodalization Since the transient response of the turbine trip event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis.
The pressurizer 2.cdeling includes the usc of thc iccal ccacitions heat transfer cptica for the.cssel conductcra.
3.1.1.2 Initial Conditions Core Power Level High initial power level and a positive power uncertainty maximize the primary-to-secondary power mismatch upon turbine trip.
Pressurizer Pressure Positive uncertainty is applied to the initial pressurizer pressure.
.High initial pressure reduces the initial margin to the overpressure limit.
Pressurizer Level High initial level minimizes the initial volume of the pressurizer steam space, which maximizes the transient primary pressure response.
Reactor Vessel Averace Tomnerature High initial temperature maximizes the primary coolant stored energy, which maximizes the transient primary pressure response.
l RCS Flow Low initial flow minimizes the primary-to-secondary heat transfer.
Core Evoans Flow Core bypass flow is not an important parameter in this analysis.
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Steam Generator Level High initial level minimizes the initial volume of the steam generator steam space, which maximizes the transient secondary pressure response.
Maximum secondary pressurization causes maximum secondary temperature response, which minimizes primary-to-secor dary heat transfer.
Fuel Temnerature Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer'from the fuel to the coolant.
Steam Generator Tube Pluccina A bounding high tube plugging value degrades primary-to-secondary heat transfer.
3.1.1.3 Boundary Conditions Pressurizer Safety Valves The pressurizer safety valves are modeled witn lift, accumulation, and blowdown assumptions which maximize the pressurizer pressure.
Steam Line Safetv Vaives The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure and minimize transient primary-to-secondary heat transfer.
3.1.1.4 Control, Protection, and Safeguards System Modeling Reactor Trio The pertinent reactor trip functions are the overtemperature AT (OTAT),
overpower AT (OPAT), and pressurizer high pressure.
The response time of each of the two AT trip functions is the Technical Specification value.
The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2.4.2 of Referenco 2.
In addition, the AT coefficients used in the analysis account for instrument uncertainties.
The response time of the pressurizer high pressure trip function is the Technical Specification vaIue.
Since the pressure uncertainty is accounted for in the initial pressurizer pressure, the pressurizer high pressure reactor trip setpoint is the Technical Specification value.
Pressurizer Pressure Control Pressurizer pressure control is in manual with sprays and PORVs disabled in order to maximize primary pressure.
Pressurizer Level Control
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Pressurizer level control is in automatic manual with the pressurizer heaters locked on in order to elevate primary pressure.
l Charging / letdown has negligible impact.
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Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization
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and minimize transient primary-to-secondary heat transfer.
Steam Generator Level Control
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Feedwater is isolated upon turbine trip. The addition of subcooled feedwater would tend to subcool the water in the steam generator, and reduce secondary side pressure.
Rod Control No credit is taken for the operation of the Rod Control System.
Following turbine trip, the turbine impulse chamber pressure is rapidly reduced. The corresponding reduction in the Rod Control System reference temperature would lead to control rod insertion, which would lessen the severity of the transient.
Auxiliary Feedwater Auxiliary feedwater is disabled.
The addition of subcooled auxiliary feedwater would tend to subcool the water in the steam generator, and reduce secondary side pressure.
3.1.2 Peak Main Steam System Pressure Analysis 3.1.2.1 Nodalization Since the transient response of the turbine trip event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis. The pressurizar acdeling include; the use of thc local conditions heat tran;fer option.for the vcssci conductors.
3.1.2.2 Initial Conditions Core Power Level High initial power level and a positive power uncertainty maximize the primary-to-secondary power mismatch upon turbine trip.
Pressurizer Pressure Positive uncertainty is applied to the initial pressurizer pressure. As long as a high pressurizer pressure reactor trip is avoided, maximum primary system pressure is conservative in order to delay reactor trip on OTAT.
t Pressurizer Level High' initial level minimizes the initial volume of the pressurizer steam space, which maximizes the transient primary pressure response.
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4 Pressurizer Pressure Control Pressurizer pressure control is in automatic with sprays and PORVs enabled in order to prevent a high pressurizer pressure reactor trip actuation prior to OTAT trip actuation.
Pressurizer Level Control Pressurizer level control is in autor.atic sanual with the pressurizer heaters locked on in order to elevate primary pressure.
Charging / letdown has negligible impact.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and condenser steam dump is unavailable in order to maximize secondary side pressurization.
Steam Generator Level Control Feedwater is isolated upon turbine trip. The addition of subcooled feedwater would tend to subcool the water in the steam generator, and reduce' secondary side pressure.
Rod Control No credit is taken for the operation of the Rod Control System.
Following turbine trip, the turbine impulse chamber pressure is rapidly reduced. The corresponding reduction in the Rod Control System reference temperature would lead to control rod insertion, which would lessen the severity of the transient.
Auxiliarv Feedwater Auxiliary feedwater is disabled. The addition of subcooled auxiliary feedwater would tend to subcool the water in the steam generator, and
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reduce secondary side pressure.
3.2 Loss of Non-Emeroency AC Power To The Station Auxiliaries A loss of non-emergency AC power causes the power supply to all busses This leads 1
not powered by the emergency diesel generators to be lost.
to the trip of both the main feedwater pumps and the, reactor coolant A primary system heatup ensues, due to both the coastdown of the pumps.
reactor coolant pumps and the loss of main feedwater heat removal.
As a result of this heatup, the primary concerns for this event are short-term core cooling capability (DNBR), long-term core cooling capability (natural circulation), and primary and secondary system ~
overpressurization.
This transient differs from the complete loss of flow transient only in the timing of the insertion of the control rods. Both transients presume reactor coolant pump and feedwater pump trip as the initiating events.
In the l'oss of flow event, the reactor trips when the reactor coolant pump bus undervoltage setpoint is reached and the rods begin to fall into the core after an instrumentation delay.
In the loss of AC power the control rods begin to fall immediately due to the loss.of transient, gripper coil voltage. Therefore, the transient core power response and consequently the ONBRahort-term core cooling capability result (DNBR) is l 3-5
v bounded by the loss of flow event.
Long-term core cooling capability is shown by analyzing the transition from forced, flow to natural circulation following a loss of non-emergency AC power.
Similarly, the primary system temperature increase and, therefore, the peak primary system pressure is also bounded by the loss of flow event.
Secondary sido pressure does not rise significantly until the turbine trip occurs and steam flow is terminated. The magnitude of this pressure increase is largely determined by the amount of heat transferred from the primary system to the secondary once the pressure increase has begun.
For this event the reactor trip occurs prior to the turbine trip, such that the primary system heat generation is rapidly decreasing as secondary side pressure is increasing..Therefore, the peak secondary pressure result is bounded by the turbine trip event, in which the reactor trip occurs well after the turbine trip.
Based on the above qualitative evaluation, a quantitative analysis of this transient is not required except for the 1s_ -term core cooling capability analysis.
Should a reanalysis become necessary, either due to plant changes, modeling changes, or other changes which invalidate any of the above arguments, the analytical methodology employed would be as follows.
Peak RCS pressure, peak Main Steam System pressure and core cooling capability (short-term and long-term) are each analyzed separately due l
to the differences in assumptions required for a conservative analysis.
The short-term core cooling capability analysis demonstrates that fuel l
cladding integrity is maintained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations.
The minimum DNBR is determined using the Statistical Core Design Methodology.
The long-term core cooling capability analysis demonstrates that natural circulation is established.
3.2.1 Peak RCS Pressure Analysis 3.2.1.1 Nodalization Since the transient response of the loss of offsite power event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this annlysis.
The pressuri;er modeling I
includes the use of the local conditions heat ~ transfer option for the c:ssel conductors.
3.2.1.2 Initial Conditions Cpre Power Level High initial power level and a positive power uncertainty maximize the primary-to-secondary power mismatch.
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't Pressurizer Pressure Control Pressurizer pressure control is in manual with sprays and PORVs disabled in order to maximize primary pressure.
Pressurizer Level Control Pressurizer level control is in automatic in order to maximize primary pressure. Charging / letdown has negligible impact.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Auxiliary Feedwater Auxiliary feedwater. actuation occurs on the loss of of'fsite power after an appropriate Technical Specification response time delay.
If applicable, aA purgo volume of hot main feedwater is assumed to be dalivered prior.to the cold AFW water reaching the steam generators.
In order to minimize the post-trip steam generator heat removal, the minimum auxiliary feedwater flow is assumed.
Turbino Trio Turbine trip occurs on the loss of offsite power.
3.2.2 Peak Main Steam System Pressure Analysis 3.2.2.1 Nodalization Since the transient response of the loss of offsite power event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis.
The pressurizer modeling includes thc use of the local conditions heat transfer option for the vessel conductors.
3.2.2.2 Initial Conditions Core Power Level High initial power level and a positive power uncertainty maximize the primary-to-secondary heat transfer.
Pressurizer Pressure Pressurizer pressure is not an important parameter in this analysis.
Pressuriter Level since initial level primarily affects the transient primary pressure response, it is not an important parameter in this analysis.
Reactor Vessel Averace Tomnerature High initial temperature maximizes the initial Main Steam System pressure and the primary coolant stored energy.
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Auxiliary Feedwater Auxiliary feedwater actuation occurs on the loss of offsite power after the appropriate Technical Specification response time delay.
If applicable, aA purge volume of hot main feedwater is assumed to be delivered prior to the cold AFW water reaching the steam generators.
In order to minimize the post-trip steam generator heat removal, the minimum auxiliary f eedwater flow is assumed.
Turbine Trio Turbine trip occurs on the loss of offsite power.
3.2.3 Core Cooling Capability Analysis - short Term l
3.2.3.1 Nodalization 4
Since the transient response of the loss of offsite power event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis.
The pressuri;er acdeling includes the usc of the local conditions heat transfer option for the vessel conductors.
3.2.3.2 Initial Conditions Core Power Level High initial power level maximizes the primary system heat flux. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurizer Pressure Nominal full power pressurizer pressure is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurizer Level Low initial level increases the volume of the pressurizer steam space which minimizes the pressure increase resulting from the insurge.
Reactor Vessel Averace Tornerature j
Nominal full power vessel average temperature is assumed. The i
uncertainty in this parameter is accounted for in the Statistical Core i
Design Methodology.
RCS Flow j
Technical Specification minimum measured Reactor Coolant System flow is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Core Bvoass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical core Design Methodology.
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~-
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Auxiliary Feedwater Auxiliary feedwater actuation occurs on the loss of of fsite power af ter the appropriate Technical Specification response time delay.
If applicable, aA purge volume of hot main feedwater is assumed to be delivered prior to the cold AFW water reaching the steam generators.
In order to minimize the post-trip steam generator heat removal, the minimum auxiliary feedwater flow is assumed.
Turbine Trio Turbine trip occurs on the loss of offsite power.
3.2.4 Core Cooling capability Analysis - Long Term 3.2.4.1 Nodalization Since the transient response of the loss of offsite power event is the same for all. loops, the single loop model described in Section 3.2 of Reference 2 is utilized for this analysis.
i 3.2.4.2 Initial Conditions Core Power Level I
High initial power level and a positive power uncezt.ainty maximize the primary system heat source.
Pressurizer Pressure The nominal pressure corresponding to full power operation is assumed since the establishment of natural circulation is independent of initial pressurizer pressure.
Pressurizer Level The nominal level corresponding to full power operation is assumed since i
the establishment of natural circulation is independent of initial pressurizer level.
Reactor Vessel Averaae Temperature
.High initial temperature maximizes the amount of stored energy in the primary system that must be removed by the secondary system.
RCS Flow Technical Specification minimum measured Reactor Coolant System flow is assumed since initial RCS flow has little impact on the final natural circulation flow.
Core BvDass Flow Core bypass flow is not an important parameter in this analysis.
Steam Generator Level High initial steam generator level minimizes the initial volume of the steam generator steam space, which maximizes the transient secondary pressure response. Maximum secondary pressurization causes maximum 3-12
y secondary temperature response, which minimizes primary-to-secondary heat transfer.
Fuel Temoerature Initial fuel temperature is not an important parameter in this analysis.
Steam Generator Tube Pluccina A bounding high tube plugging value degrades primary-to-secondary heat transfer.
3.2.4.3 Boundary Conditions Reactor Coolant Pumns All reactor coolant pumps are assumed to trip on undervoltage at the initiation of the loss of offsite power.
D9sAY.JitA%:
End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is employed.
Steam Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.
3.2.4.4 Control, Protection, and safeguards System Modeling Reactor Trig The insertion of all control and shutdown banks occurs when the power is lost to the control rod drive mechanism.
Pressurizer Pressure Control Pressurizer sprays are lost when the reactor coolant pumps trip.
Pressurizer PORVs are lost when offsite power is lost.
Therefore, both are inoperable.
Pressurizer Level Cpntrol Pressurizer heaters are assumed to be inoperable since they are lost when offsite power is lost.
Charging / letdown has negligible impact.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable due to the loss of offsite power.
Auriliary Feedwater Auxiliary'feedwater actuation occurs on the loss of offsite power after f
the appropriate Technical Specification response time delay.
In order i
to minimize post-trip steam generator heat removal, the minimum auxiliary feedwater flow is assumed.
1 Turbine Trit 2 Turbine trip occurs on the loss of offsite power.
3-13 l
- v 3.3 Loss Of Normal Feedwater A loss of normal feedwater flow event could result due to the failure of both of the main feedwater pumps or a malfunction of the feedwater control valves. A primary system heatup ensues due to the degradation of the secondary heat sink. As a result of this heatup, the primary concerns for this event are DNB and primary and secondary system overpressurization.
The loss of normal feedwater transient is bounded by the turbine t' rip transient.
Both transients involve a mismatch between primary heat source and secondary heat sink, but the mismatch is greater for the turbine trip. This is mainly due to the reactor trip and turbine trip occurring simultaneously for the loss of feedwater event, whereas reactor trip lags the turbine trip during the turbine trip transient.
Based on the above qualitative evaluation, a quantitative analysis of this transient is not required.
Should a reanalysis become necessary, either due to plant changes, modeling changes, or other changes which invalidate any of the above arguments, the analytical ~ methodology
-cmployed would be as follows.
Peak RCS pressure, peak Main Steam ~ System pressure and core cooling capability are each analyzed separately due to the differences in assumptions required for a conservative analysis. The core cooling capability analysis demonstrates that fuel cladding integrity is main-tained by ensuring that the min,imum DNBR remains above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology.
3.3.1 Peak RCS Pressure Analysis 3.3.1.1 Nodalization Since the transient response of the loss of normal feedwater event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis.
The pressuri;ds n.cdcling imm mmme mumm icas hcat transfer optica for the eme mem vm mum.---.
vessel cc.-ductcrs.
i 3.3.1.2 Initial Conditions i
Core Power Level High initial power level and a positive power uncertainty maximize the primary-to-secondary power mismatch.
Pressurizer Pressurg Positive instrument uncertainty is applied to the initial pressurizer pressure.
High initial pressure reduces the initial margin to the overpressure limit.
3-14
. l m,..
Pressurizer Level High initial level minimizes the initial volume of the pressurizer steam space, which maximizes the transient primary pressure response.
Reactor Vessel Averace Tercerature High initial temperature maximizes the initial primary coolant stored energy, which maximizes the transient primary pressure response.
RCS Figg Low initial flow degrades the primary-to-secondary heat transfer.
Core Bvoass Flow Core bypass flow is not an important parameter in this analysis.
Steam Generator Level High initial icral in all stca; generators delays rcasear trip on low-low level and maximizes the haatup of the primary system. Low initial level is assumed'in order to minimize steam generator inventory at the time of reactor trip.
The low-low level trip setpoint is adjusted to account for the difference between actual level and indicated level.
Fuel Tomnerature Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer from the fuel to the coolant.
Steam Generator Tube Pluccina A bounding high tube plugging value degrades primary-to-secondary heat transfer.
3.3.1.3 Boundary Condition!
~
~
Pressurizer Safety Valves The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize the pressurizer pressure.
Steam Line Safetv valves The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure and minimize transient primary-to-secondary heat transfer.
Decav Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is employed.
3.3.1.4 control, Protection, and Safeguards System Modeling Reactor Trin Reactor trip occurs when the low-low level setpoint is reached in the steam generator.
3-15 l
~
~
~ _. -
y Proesurizer Level Pressurizer level is not an important parameter in this analysis.
Reactor Vessel Averade TemDerature High initial temperature maximizes the initial Main Steam System pres-sure and the primary coolant stored energy.
RCS Flow High initial flow maximizes the primary-to-secondary heat transfer.
1 I
Core Bvnass Flow Core bypass flow is not an important parameter in this analysis.
Steam Generator Level
!!igh initial levc1 in all steam generators delays readtcr trip on ic;.
low icvel.
l.l s o, high levc1 minimizes the initial volusc of the stcan generator stca; space, which maximizes the transient seccadary pressure respansc. Low initial level is assumed in order to minimize st;eam generator inventory at the time of reactor trip. The low-low level trip setpoint is adjusted to account for the difference between actual level and indicated level.
Fuel Tomnerature Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer from the fuel to the coolant.
Steam Generator Tube Pluccina Zero tube plugging is modeled to maximize primary-to-secondary heat transfer.
3.3.2.3 Boundary Conditions Pressurizer Safety valves The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize the pressurizer pressure.
Steam Line Safetv Valves The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure.
Decav Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is employed.
1 3.3.2.4 Control, Protection, and Safeguards System Modeling Reactor Trio Reactor trip occurs when the low-low level setpoint is reached in the steam generator.
1 i
1 j
3-17 i
i r-
Pressurizer Pressure Control The results of this transient are not sensitive to the operation of pressurizer pressure control as long as the pressure is controlled to within the range that avoids protection or safeguards actuation.
Pressurizer Level Control The results of this transient are not sensitive to the operation of pressurizer level control as long as the level is kept within the range that avoids protection or safeguards actuation.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and condenser steam dump is unavailable in order to maximize the transient secondary side pressurization.
Rod Control No credit is taken for the operation of the Rod Control System for this j
transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value.
Turbine Control The turbine is modeled in the load control mode, which'is described in Section 3.2.5.1 of Reference 2.
Auxiliary Feedwater Auxiliary feedwater actuation occurs on low-low steam generator level after the appropriate Technical Specification response time delay If applicable, aA purge volume of hot main feedwater is assumed to be delivered prior to the cold AFW Water reaching the steam generators.
In order to minimize the post-trip steam generator heat removal, the minimum auxiliary feedwater flow is assumed.
3.3.3 Core Cooling Capability Analysis 3.3.3.1 Nodalization
' Since the transient response of the loss of normal feedwater event is the same for all loops, the single-loop model described in section 3.2 of Reference 2 is utilized for this analysis.
The pressuri;er asdeling includcs the use of the local conditi;ns heat transfer cptica for the vessel conductors.
l 3.3.3.2 Initial Conditions Corg_ Power Level High initial power level maximizes the primary system heat flux. The l
uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
3-18
,A l
4
- Pressurizer Pressure Nominal full power pressurizer pressure is assumed.
The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurifer Level Low initial level increases the volume of the pressurizer steam space which minimizes the pressure. increase resulting from the insurge.
Reactor Vessel Averade Temnerature Nominal full power vessel average temperature is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
RCS Flow Minimum measured Reactor Coolant System flow is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Core Bvoass Flag The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
Steam Generator Tube Pluccina A bounding high tube plugging level impairs the ability of the secondary side to remove primary side heat.
Fuel Tomnerature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient.
A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNBR evaluation.
Steam Generator Level
- igh initial icval in all steam gencrators dclays reactor trip cn icw icu icycl and extend the primary system heatup. Low initial level is assumed in order to minimize steam generator inventory at the time of reactor trip.
The low-low level trip setpoint is adjusted to account for the difference between actual level and indicated level.
3.3.3.3 Boundary Conditions Steam Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.
Decav Heat End-of-cycle decay heat, based upon'the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is employed.
3-19
4 3.3.3.4 Control, Protection, and Safeguards System Modeling Reactor Trin Reactor trip occurs when the low-low level setpoint is reached in the steam generator.
Pressurizer Pressure Control Pressurizer sprays and PORVs are assumed to be operable in order to minimize the system pressure throughout the transient.
Pr.esntriter Level Control i;u credit is taken for pressurizer heater operation so that Reactor Coolant System pressure is minimized. Charging / letdown has negligible impact.
Steam Line FORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Rod control No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its non,inal value.
Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliary Feedwater Auxiliary feedwater actuation occurs on low-low steam generator level If after the appropriate Technical Specification response time delay.
applicable, aA purge volume of hot main feedwater is assumed to be delivered prior to the cold AFW water reaching the steam generators. In order to minimize the post-trip steam generator heat removal, the l
minimum auxiliary feedwater flow is assumed.
Turbine Trio Turbine trip occurs on reactor trip.
3.4 Feedwater System Pine Break The feedwater system pipe break event postulates a rupture of the Main Feedwater System piping just upstream of the steam generator (downstream of the final feedline check valve).
Following the blowdown of the faulted generator, there is a mismatch between the heat generation in the reactor and the secondary side heat removal rate.
Due to the mismatch, the primary concern for this transient is the capability to I
effectively cool the reactor core.
l l
1 3-20 l
~
(-
- D W
?
Steam Generator Level Low initial level in all steam generators decreases the long-term capability of the secondary system to remove primary system heat.
Fuel Tomoerature A conservatively high initial fuel temperature is assumed in order to maximize the amount of stored energy that must be removed.
Steam Generator Tube Pluccina A bounding high tube plugging level impair; the ability of thc se;cndary j
side tu removc primary sida heat. Tube plugging does not significantly
[
f affect the transient results so long as the minimum Technical Specification RCS flow rate is used.
3.4.2.3 Boundary Conditions Dreak Modelino The,feedline break is modeled as a double-ended rupture of the main feedwater line just upstream of the steam generator (downstream of the check valve). A bounding flow area of the break junction is assumed in order to maximize the break flowrate. The break flowrate is determined by the Henry (subcooled) and Moody (saturated) critical flow correlations.
f Reactor Coolant Pumns Thc timing ci-the opcrator action to trip the reactor.ccolant pump; is invc;tigated in a ;sa;itivity ;tudy. Cased en the result; cf thi; seasidivity study, an carly pump trip time, With the cerresponding natural circulation heatup, is conservative.
Thc nCr; are tripped at 15 seconds, which is assumcd to prcccdc the time at which the pump; usuid be manually tripped on high high c;ntainacnt pressure.The reactor coolant pumps are lost at the initiation of the loss of offsite power which occurs coincident with reactor trip.
Offsite Power Offsite power is assumed to be lost coincident with reactor trip to delay safety injection and accelerate the post-trip heatup due to the j
loss of the reactor coolant pumps.
)
Pressurizer Safety Valves accumulation, and The pressurizer safety valves are modeled with lift,
~
blowdown assumptions which minimize pressurizer pressure.
Steam Line Safetv Valves accumulation, The main steam code safety valves are modeled with' lift, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transf er.
Decav Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty,.is employed.
1 i
3-24
-~
y
.=
4 3.4.2.4 Control, Protection, and Safeguards System Modeling Reactor Trio The reactor is tripped 10 seconds into the transient. This is assumed to be after the occurrence of safety injection actuation on high containment pressure.
Pressuriter Pressure Control Since low Reactor Coolant System pressure is conservative and the blowdown pressure of a cycling safety valve is much lower than for a cycling PORV, the PORVs are assumed inoperable.
Pressurizer spray is assumed to be operable in order to minimize system pressure.
Pressuri2er_ Level Control Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System pressure is minimized. Charging / letdown has negligible impact, h&Gne POPNs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Rod Control No credit is taken for the operation of the Rod Control System for this transient, since the pre-trip RCS temperature change is insufficient to cause rod motion.which results in an increase in ROC temperature.
' ith thc ned Control Oystem in autcaatic, thc control rods would-cam;c ;
negati,c reactivity additiGL as they arc inserted in an attempt-CG maintain ECO tcapcrature at its nominal valuer Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Safety Iniection Safety injection actuation occurs at 10 seconds on high containment pressure.
Injection begins after the appropriate Technical Specification delay to allow for the startup of the diesel generators on
.the loss of offsite power. One-train minimum injection flow, as a function of RCS pressure, is assumed to minimize the delivery of cold SI j
water. Injection is stopped when the emergency procedure SI termination criteria are met.
i Auxiliarv Feedwater 1
Auxiliary feedwater actuation occurs on safety injection actuation af te-H If the appropriate Technical Specification response time delay.
applicable, aA purge volume of hot water is assumed to be delivered prior to the cold AFW water reaching the steam generators.
Operator action to isolate AFW flow to the faulted generator occurs with a conservative delay time to minimize the amount of cold AFW flow to the faulted generator.at 120 seconds as a result of a sensitivity study.
In order to minimize the post-trip steam generator heat removal, the minimum auxiliary f eedwater flow is assumed.
1 3-25
l a
)
~
)
MSIV closure Thc tiain; of thc closurc of thc nain steam isolation ialvcs.is the fcCus of a scasitivity study which shsws that cs.rly
'Z'.
closurc, which j
initiates the cycrhcating phese of the transicnt, is conservatisc.
The valvas are closed at 10 sacands, which is assuraad to presed; autcaatic closure on hign high contal.a.ont pressura.Early MSIV closure is+
conservative since it accelerates the heatup portion of the transient due to the faulted SG reaching dryout sooner following MSIV closure.
Main steam line isolation occurs on low steem line pressure or high-high containment pressure.
Since neither of these setpoints can be reached, before reactor trip, it is conservat$vely assumed that MSIV closure-occurs coincident with turbine trip.
O b
1 E
l l
3-26 l
!- ~ ~ ~
4.0 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE j
4.1 Partial Loss of Forced Reactor Coolant Flow i
A partial loss of forced reactor coolant flow can result from a mechanical or electrical failure in a reactor coolant pump, or from a fault in the power supply to the pump.
If the reactor is at power when a
such a fault occurs, this could result in DNB with subsequent fuel danage if the reactor is not tripped promptly. The necessary protection against a partial loss of coolant flow is provided by the low reactor coolant flow reactor trip signal.
The acceptance critoria for this analysis are to ensure that there is adequate core cooling capability and that the pressure in the Reactor Coolant System remains below 110% of design pressure.
The core cooling capability analysis demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology. The peak RCS pressure criterion is met through a comparison to the peak pressure results for the more limiting locked rotor transient.
In Section 4.3 of this report, the locked rotor event is shown to remain below 110% of the RCS design pressure.
4.1.1 Nodalization This non-symmetric transient is analyzed using a two-loop model, with a single loop for the tripped reactor coolant pump and an intact triple loop.
Thc prcssurizar acdoling includos the uso of the local conditions heat transfcr option for thc vc; sci conductors.
4.1.2 Initial Conditions Core Power Level High initial power level maximizes the primary system heat flux.
The uncertainty for this parameter is incorporated in the Statistical. core Design Methodology.
1 Erfanurizer Pressure The nominal pressure corresponding to full power operation is assumed, with the uncertainty for this parameter incorporated in the Statistical Core Design Methodology.
Pressurizer Level Low initial level increases the volume of the pressurizer steam space which minimizes the pressure increase resulting from the insurge.
l l
1 Reactor Vessel Averace Temoerature i
The nominal temperature corresponding to full power operation is assumed, with the uncertainty for this parameter incorporated in the i
Statistical Core Design Methodology.
4-1
+
Pressurizer Pressure Control Pressurizer sprays and PORVs are assumed to be operable in order to minimize the system pressure throughout the transient.
Pressurizer Level Control Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System pressure is minimized. Charging / letdown has negligible impact.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Steam Generator Level Control The results of this transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
MFW Pumn Soeed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Rod Control No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value.
Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliary Feedwater AFW flow would be credited when the safety analysis value of the low-low steam generator level setpoint is reached.
However, the parameter of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed.
Therefore, no AFW is actually delivered to the steam generators.
Turbine Trin The reactor trip leads to a subsequent turbine trip.
4.2 Comnlete Loss Of Forced Reactor Coolant Flow A complete loss of forced reactor coolant flow would occur if all four reactor coolant pumps tripped due to either a common mode failure or a simultaneous loss of power to the pump motors. The Reactor Protection System (RPS) senses an undervoltage condition at the pumps and initiates a teactor trip. The decrease in core flow which occurs prior to reactor trip causes a heatup of the Reactor Coolant System.
4-3
4 Fuel Temnerature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient.
A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNDR evaluation.
Steam Generator Tube Pluccina For transients of such short duration, steam generator tube plugging does not have an effect on the transient results.
4.2.3 Boundary Conditions RCP Oneration All four reactor coolant pumps are tripped at the initiation of the transient.
The pump model is adjusted such that the resulting coastdown
]
flow is conservative with respect to the flow coastdown test data.
Steam Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.
l 4.2.4 Control, Protection, and Safeguards System Modeling Reactor Trin Reactor trip occurs on reactor coolant pump undervoltage, after an appropriate instrumentation delay.
Pressurizer Pressure Control Pressurizer sprays and PORVs are assumed to be operable in order to minimize the system pressure throughout the transient.
Pressurizer Level control Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System pressure is minimized.
Charging /lctdown has negligible impact.
Steam Line PORVn and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Steam Generator Level Control The results of this transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
i 4-5
MFW Pumn Sreed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Red centrol No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in'an attempt to maintain RCS temperature at its nominal value.
Turbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliary Feedwater AFN flow would be credited when the safety analysis value of the low-low steam generator level setpoint is reached. However, the parameter of interest for this transient has reached its limiting value before the appropriate Tochnical Specification response time delay has elapsed.
Therefore, no AFW is actually delivered to the steam generators.
Turbine Trin The reactor trip leads to a subsequent turbine trip.
4.3 Reactor Coolant Pumn Locked Rotor The postulated accident involves the instantaneous seizure o% cne reactor coolant pump rotor.
Coolant flow in that loop is rapidly reduced, causing the Reactor Protection System (RPS) to inid ato a reactor trip on low RCS loop flow. The mismatch between power generar. ion and heat removal capacity due to the degraded flow condition causes a heatup of the primary system.
The acceptance criteria for this analysis are to ensure that'there is adequate core cooling capability and that the pressure in the Reactor Coolant System remains below 120% of design pressure.
Peak RCS pressure and core cooling capability are analyzed separately due to the differences in assumptions' required for a conservative analysis.
The core cooling capability analysis determines to what extent fuel cladding integrity is compromised by calculating the number of fuel rods that exceed the 95/95 DNBR limit based on acceptable correlations.
i 4.3.1 Peak RCS Pressure Analysis j
4.3.1.1 Ncdalization Due to the asymmetry of the transient, a two-loop model (Reference 2, Section 3.2), with a faulted single loop and ane intact triple loop, is utilized for this analysis.
The prcssurincr msdeling includca the use 4-6
of th: local conditicas heat transfcr option for th; vessci condu;;crs.
l l
4.3.1.2 Initial Conditions Core Power Level High initial power level and a positive power uncertainty maximize the primary system heat load.
Pressurizer Pressure High initial pressure yields a smaller margin to overpressurization.
Pressurizer Level High initial level decreases the volume of the pressurizer steam space I
which maximizes the pressure increa~seresulting from the insurge.
Reactor Vessel Averace Tomoerature High initial temperature maximizes the initial primary coolant stored energy, which maximizes the transient primary pressure response.
PCS Flow Low initial flow minimizes the primary-to-secondary heat transfer.
Core Dvoass Flow High core bypass flow minimizes coolant flow through the core and exacerbates heatup.
Steam-Generator Level Initial steam generator level is not an important parameter in this analysis.
Fuel Temoerature Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer from the fuel to the coolant.
Steam Generator Tube Pluccina For transients of such short duration, steam generator tube plugging does not have an effect on the transient results.
(,
1.3 Boundary Condi-tions Reactor Coolant Pumns The rotor of the reactor coolant pump in the faulted loop is assumed to seize at the initiation of the transient.
The remaining reactor coolant pumps trip on bus undervoltage following the loss of offsite power.
Offsite Power Offsite power is assumed to be lost coincident with the turbine trip.
Pressurizer Safetv Valves The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize pressurizer pressure.
4-7
l!e 4
Steam Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer..
4.3.1.4 Control, Protection, and Safeguards System Modeling Reactor Trin Reactor trip occurs on low Reactor Coolant System flow in the locked loop.
Pressurizer Pressure control In order to maximize primary system pressure, no credit is taken for pressurizer spray or PORV operation.
Pressurizer Level CQncrol Pressurizer heaters are assumed to be operable in order to maximize Reactor Coolant System pressure resulting from the insurge/ level increase. Charging / letdown has negligible impact.
Steam Line PORVs and Condenser Steam Dumo Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Steam Generator Level Control The results of this' transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
MFW Pumn SDeed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Rod control No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value.
Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliary Feedwater AFW flow would be credited when the safety analysis value of the low-low steam generator level setpoint is reached.
However, the parameter of 1
interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed.
Therefore, no AFW is actually delivered to the steam generators.
4-8
)
i
v e
Turbine Trio The reactor trip leads to a subsequent turbine tr'.p.
4.3.2 Core Cooling Capability Analysis 4.3.2.1 Modalization Due to the asymmetry of the transient, a two-loop model (Reference 2, Section 3.2), with a single (faulted) loop and a triple (intact) loop, is utilized for this analysis.
Tuc prcssuri;;; sodcling in;1udcs the usc of thc local conditions heat transfer option for the vessci ccnductors.
4.3.2.2 Initial Conditions Core Power Level High initial power icvci and a positiva pcwcr uncertainty saxini;c the primary systca heat load.High initial power level maximizes the primary system heat flux.
The uncertainty in this parzmeter is accounted for in the Statistical Core Design Methodology.
Pressurizer Pressure Low initial pressurc yicids a icwcr initial, and therefer; trac.sient, eNBR-The nominal pressure corresponding to full power operation is assumed, with the pressure initial condition uncertainty accounted for the in Statistical Core Design Methodology.
Eressurizer Level Low initial level increases the volume of the pressurizer steam space which minimizes the pressure increase resulting from the insurge.
Reactor Vessel Averace Temeerature Migt initial tempcraturc incrcases the stored cacrgy in thc primary systc which must bc rasovcd and miniai;es thc tran;ient DN0n.The nominal temperature corresponding to full power operation is assumed, wich the temperature initial condition uncertainty accounted for in the Statistical core Design Me,thodology.
RCS Flow Lcw 1sitial fi w degradcs the primary to secsadary heat transfer and mtniai;es trc transient ONOn.The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation.
The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Bvoass Flow High corc bypass f i c;. cxace&batcs heatup by ainiairing coolant ficW through thc core and miniai;cs thc transicnt DNER.The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
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4
-.-..m..
Steam Generator Level
. Initial steam generator level is not an important parameter in this
- analysis, i
Fuel Temocrature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient. A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNER evaluation.
Steam Generator Tube 91uccinc For transients of such short duration, steam generator tube plugging does not have an effect on the transient results.
4.3.2.3 Boundary Conditions Reactor Coolant Pumns The rotor of the reactor coolant pump in the faulted loop is assumed to seize at the initiation of the transient.
The remaining reactor coolant pumps trip on bus undervoltage following the loss of offsite power.
Offsite Power Of f sitc poucr is assuncd to bc ic;; coincident with the turbine tritr-cases with offsite power maintained as well as with offsite power lost coincident with the turbine trip are analyzed.
Pressurizer Safety Valves The pressurizer safety valves are not challenged by this transient.
Steam Line Safetv Valves The main steam code safsty valves are modeled with lift, accumulation, i
and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.
b 4.3.2.4 Control, Protection, and Safeguards System Modeling Reactor Trio Reactor trip occurs on low Reactor Coolant System flow in the loop with the locked rotor.
Pressurizer Pressure Control Credit is taken for both pressurizer spray and PORV operation in order to minimize primary system pressure.
Pressurizer Level control Pressurizer heaters are assumed to be inoperable so that Reactor Coolant 3
System pressure is minimized.
Charging / letdown has negligible impact.
I l
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k Steam Line PORVs and Condenser Steam Dumo Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Steam Generator Leve1 Control The results of this transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
MFW Pumo Soeed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Rod Control No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value.
7trbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliarv Feedwater AFW flow would be credited when the safety analysis value of the low-low steam generator level setpoint is reached.
However, the parameter of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed.
Therefore, no AFW is actually delivered to the steam generators.
Strbine Trio The reactor trip leads to a subsequent turbine trip.
l 4.3.2.5 Other Assumptions The peak clad temperature calculation employs the fuel conduction model as described in Section 4.2.2 of Reference 1.
1 1
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5.0 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 5.1 Uncontrolled Bank Withdrawal From a Suberitical or Low power Startuo condition A malfunction of the Rod Control System can result in an uncontrolled withdrawal of control rods.
Beginning from a low initial power typical of Modes 2 and 3, the resulting positive reactivity addition causes,a power excursion which is terminated by the high power range flux (low setpoint) or high pressurizer pressure RPS trip functions.
Since the l
initial condition requires as few as three reactor coolant pumps in operation, the minimum DNBR is of concern for peak transient power levels less than full power. The peak Reactor Coolant System pressure limit of 110% of design pressure is also of concern due to the mismatch between core power and the secondary heat sink during the power excursion. Peak RCS pressure and core cooling capability are analyzed separately due to the differences in assumptions required for a conservative analysis.
The core cooling capability analysis demonstrates that fuel cladding integrity is maintained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable
{
correlations.
The minimum DNBR is determined using the Statistical core i
Design Methodology.
5.1.1 Peak RCS Pressure Analysis 5.1.1.1 Nodalization The peak RCS pressure transient is analyzed with four reactor coolant pumps in operation.
Since all initial and boundary conditions are symmetric, a single-loop model or any multi-loop nodalization is appropriate.
The standard model (Reference 2, Section 3.2) is used with one significant exception. Since this transient initiates at zero power, and since the duration of the transient is very short, the steam generator secondary response is not important.
Rather than using the standard steam generator secondary nodalization, a single secondary volume is used.
The single volume uses the bubble rise option with the local-conditions heat transfer model applied to the steam generator tube conductors.
With this mod'eling approach the initial condition of zero power can be obtained, and the primary-to-secondary heat transfer that occurs following the power excursion can be simulated.
The prcssurizer nodcli-- includca the use of thc local conditicas hca; transfcr optica fcr tr
,cssel ccaductors.
5.1.1.2 Initial Conditions Core Power Level A minimum initial power level typical of a critical, zero power startup' condition maximizes the power excursion.
5-1
4 5.1.1.4 Control, Protection, and Safeguards dystem Modeling Reactor Trin The pertinent reactor trip functions are the high power range flux (low setpoint) and pressurizer high pressure.
The high power range flux (low setpoint) trip includes a conservative allowance to account for calibration error, and error due to rod withdrawal effects. The response time of the high flux trip function is the Technical Specification value.
The response time of the pressurizer high pressure trip function is the Technical specification value. Since the pressure uncertainty is accounted for in the initial pressurizer pressure, the pressurizer high press'ure reactor trip setpoint is the Technical Specification value.
Pressurizer Pressure Control Pressurizer spray _and PORVs are inoperable to maximize RCS pressure during the transient.
Pressurize Leggi control Due to the short duration of this transient, heaters, makeup and letdown are unimportant.
Steam Line PORVs and Condenser Steam Dumn Steam line PORVs and steam dump to condenser are unimportant for this transient and are inoperable.
5.1.2 Core Cooling Capability Analysis 5.1.2.1 Nodalization The core cooling capability analysis, which determines the minimum DNBR, is analyzed with three reactor coolant pumps in operation. A two-loop model with one single loop and one triple loop is utilized for this analysis. The standard model (Reference 2, Section 3.2) is used with one significant exception.
Since this transient initiates at zero power, and since the duration of the transient is very short, the steam generator secondary response is not important.
Rather than using the standard steam generator secondary nodalization, a single secondary volume is used.
The single volume uses the bubble rise option with the local-conditions heat transfer model applied to the steam generator tube conductors. With this modeling approach the initial condition of zero power can be obtained, and the primary-to-secondary heat transfer that i
occurs following the power excursion can be simulated No main or
]
auxiliary feedwater or initial steam flow is modeled.
The pressurizar modeling includcs the use of iccal conditions heat transfer optica for.
tt vcssci condu;;cr;.
5-3
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5.1.2.2 Initial conditions i
Core Power Level A minimum initial power level typical of a critical, zero power startup condition maximizes the power excursion.
Pressurizer Pressure Nominal pressure is assumed, with the pressure initial condition uncertainty accounted for in the Statistical Core Design Methodology.
Pressurizer Level Low initial pressurizer level minimizes the pressure increase following an insurge.
, 7 Reactor Vessel Averace Temoerature The nominal temperature corresponding to zero power operation is assumed, with the temperature initial condition uncertainty accounted for in the Statistical Core Design Methodology.
RCS Flow Nominal three pump flow is assumed since low flow is conservative for DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Bvoass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
Steam Generator Level Initial steam generator level is not an important parameter in this l
analysis.
Puol Temnerature Due to the initial zero power condition, the initial fuel temperature is 1
equal to T-ave.
The fuel-clad gap conductivity is set high to maximize heat transfer from the fuel.
Steam Generator Tube Pluccina No tube plugging is assumed to maximize the RCS volume and thereby minimize the insurge into the pressurizer.
i l
5.1.2.3 Boundary Conditions lion-Conductinc Heat Exchancers For initialization purposes, non-conducting heat exchangers are used to remove reactor coolant pump heat since the steam generators are passive at initialization. These are turned off prior to the start of the power excursion.
RCP Oceration Since low flow is conservative for DNBR, the minimum number of reactor coolant pumps (three) required for the modes for which this transient is r
i f
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I
3 applicable Modes 2 and 3) are assumed to be in operation.
Pressurizer Safetv Valves The pressurizer safety valves ara modeled with lift, accumulation, and blowdown assumptions to minimize RCS pressure during the transient.
Steam Line Safety Valves Although not importaat for this transient, steam line safety valves are modeled with lift, accumulation, and blowdown assumptions to maximize primary-to-secondary heat transfer.
5.1.2.4 Control, Protection, and Safeguards System Modeling Reactor Trin The pertinent reactor trip functions are the high power range flux (low setpoint) and pressurizer high pressure.
The high power range flux (low setpoint) trip includes a conservative allowance to account for calibration error, and error due to rod withdrawal effects. The response time of the high flux trip function is l
the Technical Specification value.
The response time of the pressurizer high pressure trip function is the Technical Specification value. The pressurizer high pressure reactor trip setpoint is the Technical specification value plus an allowance which bounds the instrument uncertainty.
Pressurizer Pressure Control Pressurizer spray and PORVs are operable to minimize RCS pressure during the transient. Heaters are not energized during the transient, i
Steam Line PORVs and Condenser Steam Dumn for this Steam line PORVs and steam dump to condenser are unimportant transient and are inoperable.
5.1.2.5 Other Assumptions Due to the potential for bottom-peaked power distributions during this and due to the'non-applicability of the Statistical Core transient, fuel Design Methodolo7y below the mixing vane grids in the current assembly designs, arceptable DNBRs are confirmed with the W-3S CHF correlation as neces9ary.
Explicit accounting for uncertainties (i.e.,
non-SCD) isare used with the W-3s correlation.
j 5.2 Uncontrolled Bank Withdrawal at Power The uncontrolled bank withdrawal at power accident is characterized by an increase in core power level that cannot be matched by the secondary heat sink. The resultant mismatch causes an increase in primary and The increases in power and secondary system temperatures and pressures.
along with a change in the core power distribution, present temperature, i
5-5
l a DNBR concern.
The primary and secondary overpresnure limits of 11'0%
of design pressure are also of concern.
t Peak RCS pressurc, peak : bin Otoaa Oystem pressure and core cooling capability are each-analyzed separately due to the differences in assumptions required for a conservative analysis.
The core cooling capability analysis demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology, i
5.2.1 Peak RCS Pressure Analysis i
5.2.1.1 Nodalization Since the transient response of..the uncontrolled bank withdrawal event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis'.
The pressuri cr acdcling includca the usc of the local conditions heat transfer option for the vessel conductors.
+
F 5.2.1.2 Initial Conditions Core Power Level Initial pressurizer pressure and, thus, initial margin to the Due l
overpressurization limit tsare independent of initial power level.
to the pressure overshoot during the reactor trip instrumentation delay, maximum pressure is achieved with the maximum pressurization rate.
The maximum pressurization rate is achieved with the maximum insertion of I
reactivity, provided that reactor trip on high f1'ux does not occur prior to significant system heatup.
Since the initial margin to the high flux reactor trip is greatest at a low power level, this power level yields the most rapid insertion of reactivity with significant system heatup, Pressurizer Pressure Initial pressurizer pressure is the nominal value, and the uncertainty in pressure is accounted for in the high pressure reactor trip setpoint.
Pressurizer Level High initial level minimizes the initial. volume of the pressurizer steam space, which maximizes the transient primary pressure response.
Reactor Vessel Averace Temnerature Initial temperature is not an important parameter in this analysis.
(
RCS Flow Initial RCS flowrate is not an important parameter in this analysis.
i Core Bvoass Flow Core bypass flow is not an important parameter in this analysis.
i 5-6
./
steam Generator Level
- igh initiai icvci nininitcs thc initial valuac of the ;tcam ; acrator s can ;pa;c,.chich maximirc; the transient secondary prc;surc rc;pon;c.
Maxncam sc ondary prc33uriza~ica causc; uaxus: ac;endary tc4esaturc response, which minimizca primary Oc seccadary haat transfer. Initial steam generator level is not an important parameter in this analysis.
Fuel Torneraturt Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer from the fuel to the coolant.
Steam Generator Tube Pluccina A bounding high tube plugging value degrades primary-to-secondary heat transfer.
5.2.1.3 Boundary conditions Pressurizer Safety Valves The pressurizer safety valves are modeled with lift, accumulation, and
. blowdown assumptions which maximize the pressurizer pressure.
Steam Line Safety Valves The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient recondary side pressure and minimize transient primary-to-secondary heat tran'sfer.
5.2.1.4 control, Protection, and Safeguards System Modeling Reactor Trio
.The pertinent reactor trip functions are the overtemperature AT (oTAT),
overpower AT (oPAT), pressurizer high pressure and power range high flux (high setpoint).
The response time of each of the two AT trip functions is the Technical Specification value.
The setpoint values.of the AT trip functions'are continuously computed from system parameters using the modeling described in Section 3.2 of Reference 2.
In addition, the AT coeffi'-
cients used in the analysis account for instrument uncertainties.
The response time of the pressurizer high pressure trip function is the Technical Specification value. The pressurizer high pressure reactor trip setpoint is the Technical Specification value plus an allowance which bounds the instrument uncertainty.
The response time of the power range high flux trip function is the Technical Specification value. The power range high flux trip high setpoint is the Technical Specification value plus an allowance which bounds the instrument uncertainty.
The high flux signal is adjusted to, account for the effects of-bank withdrawal.
5-7 5
~
v.
ik Pressurizer Pressure Control In order to maximize primary system pressure, no credit is taken for pressurizer spray or PORV operation.
Pressurizer Level Control Pressurizer level control system operation has negligible impact on the results of this analysis.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient. primary-to-secondary heat transfer.
Steam cenerator Level Control Feedwater control is in automatic to prevent steam generator low-low level reactor trip.
Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliary Feedwater Auxiliary feedwater is disabled. The addition of subcooled auxiliary feedwater would tend to subcool the water in the steam generator, and provide better heat removal capability.
Turbine Trio Turbine trip upon reactor trip is modeled in order to minimize the post-trip primary-to-secondary heat transfer.
5.2.2 Core Cooling Capability Analysis
.5.2.2.1 Nodalization Since the transient response of the uncontrolled bank withdrawal event is the same for all loops, the single-loop'model described in Section 3.2 of Reference 2 is utilized for this analysis.
The pressuri;er modeling includes the use of the local conditions heat transfer optica for the cassel conductors.
5.2.2.2 Initial Conditions Core Poger Level The uncontrolled bank withdrawal event is analyzed with a spectrum of-initial power levels which range from low power to full power.
Uncertainties in initial power level are accounted for in the Statistical Core Design Methodology.
Pressurizer Pressurg Initial pressurizer pressure is the nominal value, and the uncertainty in pressure is accounted for in the Statistical Core Design Methodology.
5-8 i
~
Pressuriter Level j
Initial pressurizer level is the nominal value which corresponds to the init:a1 power level, and uncertainties are accounted for in the initial
{
value.
Low initial level maximizes the initial volume of the pressurizer steam space, which minimizes the transient primary pressure retronse.
Reactor Vessel Averace Temoerature The nominal temperature corresponding to the initial power level is e assumed, with the temperature initial condition uncertainty accounted for in the Statistical Core Design Methodology.
ECS Flow The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Evoass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
Steam Generator Level Initial steam generator levci is the nominal value which corresponds to thc initial powcr level, and uncertainties are accounted for in the initial valuc.
- igh initial 1 vci minimizes the initial volusc of the steam gcncrator stca; spacc, which maximi;cs the transient scccadary prcssure responsc. l'aximum scccadary pressurization causes maxir-seccadary tempera;ure respense, which minimizes pri ::y to secondary haat transfer. Initial steam generator level is not an important parameter in this analysis.
Fuel Temoerature Initial fuel temperature is the value which corresponds to the initial power level.
Low fuel temperature maximizes the transient heat transfer from the fuel to the coolant.
Steam Generator Tube Pluccina A bounding high tube plugging value degrades primary to secondary heat transfer.The bounding tube plugging assumption (high or low) varies depending on other initial.and boundary conditions.
5.2.2.3 Boundary Conditions Pressurizer Safetv Valves The pressurizer _ safety valves are modeled with lift, accumulation, and blowdown assumptions which minimize the pressurizer pressure.
Steam Line Safetv Valves The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure and minimize transient primary-to-secondary heat transfer.
5-9 J
'l i
\\
e-i The acceptance criterion for this event is to ensure that there is adequate core cooling capability. The core cooling capability analysis determines to what extent fuel cladding integrity is compromised by calculating the number of fuel rods that exceed the 95/95 DNBR limit based on acceptable correlations.
\\
5.4.1 Nodalization Since the transient response of the single rod withdrawal event is the same for all loops, the single-loop model described in Section 3.2 of s
Reference 2 is utilized for this analysis.
Thc pressurizar asdeting includes thc use of th; local conditions hcat transfer optica fc; the vessc1 conductora.
5.4.2 Initial Conditions Core Power Level Initial power is the nominal full power value. Uncertainty in power l
level is accounted for in the Statistical Core Design Methodology.
~
Pressurizer Pressure Initial pressurizer pressure is the nominal value. Uncertainty in pressure is accounted for in the Statistical Core Design Methodology.
Pressurizer Level High initial level minimizes the initial volume of the pressurizer steam space, which maximizes the transient primary pressure response. Up to the limit of the ability of the pressurizer sprays to control pressure, maximum pressure is conservative in order to delay reactor trip on OTAT.
Reactor Vessel Averace Temoerature Initial temperature is the full power nominal value.
Uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
RCS Flow The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation.
The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Bvrass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
Steam Generator Level Higt initial icvel sinimizes the initial volumm of the steas generatcr etcam spacc, which maximizes the transient secondary pressure response.
Maximun seconds,ry p;cssurization causcs maximun sesondary ta.werature respansc, which miniai;e; primary to secondary hcat transfer. Initial I
5-12
steam generator level is not an important parameter in this analysis.
l Fuel Temnerature Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer from the fuel to the coolant.
Steam Generator Tube Pluccina A bounding high tube plugging alue degrades prir.ary to-secondary hcat transfer. steam generator tube plugging is not an important parameter in this analysis.
5.4.3 Boundary Conditions Pressurizer Saferv Valves The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which minimize the pressurizer pressure.
Steam Line Safetv Valves The steam line safety valves are modeled with lift, accumulation, an'd blowdown assumptions which maximize transient secondary side pressure and minimize transient primary-to-secondary heat transfer.
t 5.4.4 Control, Protection, and Safeguards System Modeling Reactor Trio The pertinent reactor trip functions are the overtemperature AT.(OTAT),
overpower AT (OPAT), pressurizer high pressure and power range high flux (high setpoint).
The response time of each of the two AT trip functions is the Technical Specification value. The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2 of Reference 2.
In addition, the AT coeffi-cients used in the analysis account for instrument uncertainties.
The response time of the pressurizer high pressure trip function is the Technical Specification value. The pressurizer high pressure reactor trip setpoint is the Techni, cal Specification value plus an allowance which bounds the instrument uncertainty.
The response time of the power range high flux trip function is the Technical Specification value. The power range high flux trip high setpoint is the Technical Specification value plus an allowance which bounds the instrument uncertainty.
The high flux signal is adjusted to account for the effects of rod withdrawal.
Pressurizer Pressure Control Pressurizer pressure control is in automatic with sprays enabled and 4
PORVs disabled in order to delay reactor trip on OTAT and delay reactor trip on high pressurizer pressure.
5-13
p,.
i 1
Precsurizer Level control Pressurizer level control is in manual with the pressurizer heaters disabled in order to delay reactor trip on high pressurizer pressure.
Charging / letdown has negligible impact.
Steam Line PORVs and Condenser steam Dumo Secondary steam relief via the steam line PORVs and the condenser steam l
dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transf 7r.
steam Generator Level Control Feedwater control is in automatic to prevent steam generator low-low level reactor trip.
Auxilia rv Feedwater Auxiliary feedwater is disabled. The addition of subcooled auxiliary feedwater would tend to subcool the water in the steam generator, and.
reduce secondary side pressure.
Turbine Centrol The turbine is modeled in the load' control mode, which is described in Section 3.2.5.1 of Reference 2.
Turbine Trio Turbine trip upon reactor trip is modeled in order to minimize the post-trip primary-to-secondary heat transfer.
5.5 startuo Of An inactive Reactor Coolant Pumn At An Incorrect Tamnerature The McGuire and Catawba plant Technical Specifications currently require that all four RCPs be running at power operation.
Furthermore, low flow in any RCS loop, coincident with reactor power above the P-8 interlock (currently at 48% of rated thermal power) will cause a reactor trip.
Therefore, the only situation in which the subject accident is possible is a trip of one RCP below P-8.
For this situation the operator might choose, during allowable at power outage time for.the fourth RCP, to attempt a restart of the tripped pump. The accident is analyzed from the most. conservative condition allowed by the Reactor Protection System, even though operator error is required for the analyzed scenario to occur. The acceptance criterion is that fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains the above the 95/95 DNER limit based on acceptable correlations.
?
5.5.1 Nodalization F
Because of the loop asymmetry between the inactive single loop and the three active loops, the double-loop RCS model described in Section 3.2 of Reference 2 is used. The pressuriner u.sdaling iaciudes the use of thc local conditicas heat transfar option for the ccasci conductors.
5-14 i
1 l
=
5.5.2 Initial conditions i
Core Power Since the positive reactivity insertion due to the colder moderator average temperature causes a power increase, the initial indicated power level must be sufficiently less than the P-8 setpoint such that, by the time the indicated power level reaches the P-8 setpoint, the indicated flow in the loop containing the restarted pump is greater than the low flow reactor trip setpoint. This delays or prevents reactor trip and is therefore conservative for DNBR evaluation.
Pressurizer Pressure A pressure initial condition uncertainty is applied to minimize pressure during the transient since this is conservative for DNBR evaluation.
Pressurizer Level The heatup of the colder water and the increase in core power will'cause an expansion of the reactor coolant and an increase in pressurizer level. A negative level uncertainty is used in order to maximize the size of the pressurizer steam bubble to be compressed, which minimizes the transient pressure response.
Reactor Vessel Averace Temnerature A positive temperature uncertainty is used to minimize the margin to DNB.
RCS Flow In order to minimize core flow, and therefore the margin to DNB, the three pump equivalent of the Technical specification minimum measured flow is adjusted by a negative flow uncertainty.
Core Bvnass Flow A positive flow uncertainty is used to minimize the margin to DNB.
Steam Generator Level The results of this transient are not sensitive to the direction of steam generator level uncertainty as long as the transient level l
response is kept within the range that avoids protection or safeguards actuation.
Fuel Temnerature A low initial temperature is assumed to maximize the gap conductivity calculated for steady-state conditions and used for the subsequent transient.
A bigh gap conductivity minimizes the fuel heatup and attendant negative reactivity insertion caused by the power increase.
This makes the power increase more severe and is therefore conservative for DNB evaluation.
Steam Generator Tube Pluccing Steam generator tube plugging is not an important parameter in this analysis.
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6.0 INCREASE IN REACTOR COOLANT INVENTORY 6.1 Inadvertent Ooeration Of ECCS Durina Power Ooeration The inadvertent operation of the Emergency Core cooling System could be caused by either operator error or a spurious electrical actuation signal. Upon receipt of the actuation signal, the centrifugal charging pumps begin delivering highly borated refueling water storage tank water to the Reactor Coolant System. The resultant negative reactivity insertion causes a decrease in core power and, consequently, a decrease in temperature.
Initially, Ecoolant shrinkage causes a reduction in both pressurizer water level and pressure. Core cooling capability (DNB) is the primary concern for this transient during this time period due to yhtsthe decrease in system pressure.
Following the initial depressurization, the increase in reactor coolant inventory causes pressurizer level to increase and pressurization to occur. Pressurizer level might increase sufficiently to overfill the pressurizer and.cause water relief through the pressurizer safety valves (PSVs). Water relief through the PSVs could degrade valve operability and lead to a Condition III event.
The magnitude of the pressure decrease for this transient is no more severe than that for the inadvertent opening of a pressurizer safety or.
relief valve transient, which also trips the reactor on low pressurizer pressure.
Furthermore, the opening of a safety valve does not introduce the core power and Reactor Coolant Sys' tem temperature decreases that are characteristic of the inadvertent ECCS' actuation. Neither event involves any reduction in the Reactor Coolant System flowrate, since the reactor coolant pumps are not tripped. Therefore, the DNB results of this transient are bounded by the inadvertent opening of a pressurizer safety or relief valve transient.
Based on the above qualitative evaluation, a quantitative core cooling capability analysis of this transient is not required.
Should a reanalysis become necessary, either due to plant changes, modeling changes, or other changes which invalidate any of the above arguments, the analytical methodology employed would be as follows.
The core cooling capability analysis demonstrates that fuel cladding integrity is maintained by. ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations.
The minimum DNBR is determined using the Statistical Core Design Methodology.
The concern in the pressurizer overfill analysis is that water relief through the PSVs will degrade valve operability and lead to a Condition III event.
However, even if water relief occurs, valve operability is not degraded provided that the temperature of the pressurizer water is
[
sufficiently high.
Therefore, the acceptance criterion for this analysis is the minimum water relief temperature to assure PSV operability.
6-1
6.1.1 core cooling capability Analysis l
6.1.1.1 Nodalization l
Since the inadvertent ECCS operation transient is symmetrical with respect to the four reactor coolant loops, a single-loop model (Reference 2, Section 3.2) is utilized for this analysis.
The pressurizar acdeling includcs the use of the local conditions heat transfer option for the vessel ccaductcrs.
4 6.1.1.2 Initial Conditions l
Core Power Level High initial power level maximizes the primary system heat flux. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurizer Pressure Nominal full power pressurizer pressure is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurizer Level High initial level minimizes the volume of the pressurizer steam space which maximizes the pressure decrease resulting from the outsurge.
Reactor Vessel Averace Temnerature Nominal full power vessel average temperature is assumed.
The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
RCS Flow The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.
Core Bvoass Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.
Steam Generator Level steam generator level is not an important parameter in this analysis.
Fuel Temnerature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient.
A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNBR evaluation.
6-2
v l
Steam cenerator Tube Pluacino Steam generator tube plugging is not an important parameter in this analysis.
6.1.1.3 Boundary Conditions l
ECCS Flow A maximum safety injection flowrate along with a conservatively high boron concentration yields the most limiting transient response.
In order to minimize the delay in the delivery of the borated injection water, no credit is taken for the purge volume of unborated water in the injection lines.
Steam Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.
6.1.1.4 Control, Protection, and Safeguards System Modeling l
Reactor Trio Reactor trip is assumed to occur on low pressurizer pressure, after an appropriate instrumentation delay.
Pressurizer Pressure Control Pressurizer sprays and PORVs are assumed to be operable in order to minimize the system pressure throughout the transient.
Pressurizer Level Control Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System pressure is minimized. Charging / letdown has negligible impact.
Steam Line PORVs and Condenser Steam Dumn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.
Steam Generator Level Cnntrol The results of this transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
MFW Pumn Speed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Rod Control i
No credit is taken for the operation of the Rod Control System for this transient, which results in aan decrease in RCS temperature. With the l
Rod Control System in automatic, the control rods would cause a positive reactivity addition as they are withdrawn in an attempt to maintain RCS i
6-3
- /.
l'
.e e
e temperature at its nominal value. The resultant power increase would retard the system depressurization.
Turbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
l Auxiliary Feedwater AFW flow would be credited when the safety analysis value of the low-low steam generator level setpoint is reached.
However, the parameter of interest for this transient has reached its limiting value before the appropriate Technical specification response time delay has elapsed.
j Therefore, no AFW is actually delivered to the steam generators.
Turbine Trio The reactor trip leads to a subsequent turbine trip.
6.1.2 Pressurizer overfill Analysis 6.1.2.1 Initial conditions Core Power Zero power is assumed in this analysis. Reference 3 states that the acceptable initial power for the analysis is the licensed core thermal
- power, i.e.,
full power. However, lower power is more limiting in order to minimize the initial NC system temperature.
If overfill occurs at lower initial power, then the water relief temperature is more likely to be less than the acceptance criterion.
Pressurizer Pressure Actual system response to an safety injection (sI; *ould be an initial pressure drop then subsequent pressurization hbove initial pressure.
During the depressurization phase, SI flow would increase above the initial flowrato, and during the pressurization phase, SI flow would decrease below initial flowrute. Initial pressure is assumed conservatively low to determine the SI flow during the event.
Enactor Vessel Averace Temeerature Low initial temperature is' conservative in order to minimize pressurizer water temperature.
Steam Generator Tube Pluacina High steam generator tube plugging is assumed in order to decrease the
)
volume of the initial RCS water, which will minimize the F.CS water j
temperature as it mixes with the cold SI water.
6.1.2.2 Boundary Conditions RCP Ooeration For Modes 1-3, the Technical Specificattons require at least one reactor coolant pump be operating.
6-4 l
l
,v 6.1.2.3 control, Protection, and Safeguards System Modeling Pressurizer Level Control The pressurizer heaters are assumed to be in manual and off since heater operation would increase the temperature of the pressurizer water.
Normal makeup is isolated upon SI, and credit is not taken for letdown.
ECCS Flow i
A r.aximum safety injection flowrate from both centrifugal charging pumps l
is assumed. RCS pressure remains above the shutoff head of the intermediate head and low head safety injection pumps for the duration of the event.
ECCS Temocrature Minimum injection temperature is conservative in order to minimize relief temperature.
6-5
=..-.- ~
s 7.0 DECREASES IN REACTOR COOLANT I!WENTORY 7.1 Inadvertent coenino of a pressurizer safety or Relief Valve The loss of inventory through the open valve causes a depressurization of the RCS.
Since the core power, flow, and temperature are relatively unaffected prior to reactor trip by this depressurization, the reduction in pressure causes a reduction in DNB margin. The applicable acceptance criterion is that fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains the above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology.
7.1.1 Nodalization Since the valve opening is in the pressurizer, it affects all RCS loops identically.
Therefore a single-loop RCS system model is used.--The-pressuri;er codeling includes the u;; of the 1 scal ccaditions heat transfer option for thc vcasal conductors.
7.1.2 Initial Conditions Power Level Full power is assumed in order to maximize the primary system heat flux.
The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.
Pressurizer Pressure Nominal pressure is assumed, with the pressure initial condition uncertainty accounted for in the Statistical Core Design Methodology.
Pressur'iler Level Since this accident involves a reduction in RCS volume due to inventory loss, a positive level uncertainty is assumed to minimize the initial pressurizer steam bubble volume and therefore maximize the pressure decrease due to inventory losr.
Egaetor Vessel Averace Temnerature The nominal temperature corresponding to full power operation is assumed, with the temperature initial condition uncertainty accounted for in the Statistical Core Design Methodology.
RCS Flow The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for'DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core.
Design Methodology.
Core Bvnass Flow 4
The nominal calculated flow is assumed, with the flow uncertainty
^
accounted for in the Statistical core Design Methodology.
7-1
,N' Steam Generator Level The results of this transient are not sensitive to the direction of steam generator level uncertainty as long as the transient level l
response is kept within the range that avoids protection or safeguards actuation.
Puel Temnerature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient. A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease due to moderator density. This makes the power decrease less severe and is therefore conservative for DNBR evaluation.
Steam Generator Tube Pluccina The results of this analysis are not sensitive to the amount of steam generator tube plugging.
7.1.3 Boundary Conditions Steam Line Safety Valves The steam line safety valves are modeled with setpoint drift, accumulation, and blowdown assumptions which maximize the transient secondary pressure and therefore minimize secondary side heat removal.
7.1.4 Control, Protection, and Safeguards Systems Mode]ing Reactor Trin Reactor trip is on either low pressurizer pressure or overtemperature AT.
The Technical Specification response times are used and the safety analysis setpoints include the effects of uncertainty in the monitored parameter and in the setpoint.
Pressuri7er Pressure Control No credit is taken for pressurizer heater operation to compensate for the decrease in pressurizer pressure which occurs due to the inventory loss.
This results in a lower post-trip pressurizer pressure, which is conservative for DNBR evaluation.
Steam Generator Level Control The results of this transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
MFW Pumn Sneed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
7-2 l
i --
)
1 s;
l l
Steam Gonerat or Level Control i
i The results of this transient are not sensitive to the mode of steam generator level control as long as'the level is kept within the range l
that avoids protection or safeguards actuation.
{
MFW Pumo SDeed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
Rod Control
'I A penalty is taken for automatic rod control to insert positive reactivity to increase power and reactor vessel average temperature, i
These parameters would otherwise decrease in response to the negative reactivity inserted by the moderator density reduction.
Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliarv Feedwater AFW flow would be. credited when the safety analysis value of the low-low steam generator level setpoint is reached. However, the parameter of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed.
Therefore, no AFW is actually delivered to the steam generators.
Turbino Trin The turbine is tripped on reactor trip. No titae delay is assumed since this assumption minimizes post-trip primary-to-secondary heat removal.
7.2 Steam Generator Tube Ruoture The steam generator tube rupture analyzed is a double ended guillotine break of a single tube.
This transient is evaluated in two parts; first to evaluate minimum DNBR, and secondly to provide offsite dose input data for a separate evaluation to determine whether the fission product release to the environment is within the established dose acceptance criteria.
The DNBR analysis for this transient is modeled as a complete loss of coolant flow event initiated from an off-normal condition, using the Statistical Core Design methodology. The loss of flow is assumed to occur subsequent to the OTAT reactor trip caused by the steam generator tube rupture depressurization.
The initiating event for the offsite dose input analysis is the double-ended guillotine break of a single steam generator tube.
This analysis generates the of f site steam release boundary condition f or the dose evaluation. The single failure identified for maximizing offsite dose is the failure of the PORV on the ruptured steam generator to close.
In this analysis, this valve remains open until operator action is taken to 7-3
>r
~
4 Fuel Te.moerature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient.
A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease.
This makes the power decrease less severe and is therefore conservative for DNBR evaluation.
Steam Generator Tube Pluccino For transients of such short duration, steam generator tube plugging does not have an effect on the transient results.
7.2.1.3 Boundary Conditions RCP Ooeration All four reactor coolant pumps are tripped on the loss of offsite power.
The pump model is adjusted such that the resulting coastdown flow is conservative with respect to the flow coastdown test data.
Steam iine Safetv Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.
Offsite Power offsite power is assumed to be lost coincident with t'urbine trip in order to minimize RCS flow following reactor trip. This isolates staan flow to the condenser, thcreby caximizing the atasspheric steam reicases.
7.2.1.4 Control, Protection, and Safeguards. System Modeling Reactor Trio Reactor trip is assumed to occur on overtemperature AT, after an appropriate instrumentation delay.
Pressuricer Pressure Control Pressurizar sprays and 70nVs are assumed to he cperabic in ordcr to minimize thc systcn pressure thrsughout the transient.
Following the tube rupture, RCS pressure continuously decreases through the time at which minimum DNBR occurs.
Thus, pressurizer sprays are not activated nor are the pressurizer PORVs challenged during the transient. '
Pressurizer Level Contr21 Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System pressure is minimized.
Charging /lctdcwn has negligible impact.
Charging and letdown are assumed to be balanced at all times during the j
event with no action taken to increase charging flow due to RCS pressure
]
and pressurizer level decreasing. This will maximize the RCS depressurization rate.
7-5
_m.._
m 9-i
.?
6 Steam Line PORVs and Condenser 9 team Dumn The main steam PORVs and condenser dumps valves are assumed to be _
unavailable during this transient. This maximizes the secondary side pressure and temperature and therefore reduces primary-to-secondary heat' transfer.
Steam Generator Level Control i
The results of this transient are not sensitive to the mode of steam generator level control as long as the level is kept within the range l
that avoids protection or safeguards actuation.
IfFW Pu*n SDeed Control The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.
I Rod Control No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted.in an attempt to maintain RCS temperature at its nominal value.
Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Auxiliarv Feedwater AFW flow would be credited when the safety analysis value of the low-low steam generator level setpoint is. reached.
However, the parameter of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed.
Therefore, no AFW is actually delivered to the steam generators.
1 Turbine Trin The reactor trip leads to a subsequent turbine trip.
7.2.2 offsite Dose Calculation Input Analysis 7.2.2.1 Nodalization i
Due to the asymmetry of this transient a twethree-loop model, with etwo single loops and a tripledouble loop, is utilized for this analysis.
The boundary conditions for the two intact steam generators with operable steam line PORVs are symmetric.synactric, enabling the use of the two loop ;; del.
The loop with the tube rupture requires separate modeling, as does the loop with the inoperable steam line PORV.
The pressurizar codcling includc; the usc of the local conditions heat trnasfer cption for the tc sel conductors.
7-6 1
1 l
l 7.2.2.2 Initial Conditions Core Power Level High ' initial core power and a positive uncertainty maximize the primary l
system heat load.
Pressurizer Pressure Highinitialpressureend-withapositiveuncertaintydelaysthetimeofl reactor trip.
This retards the primary system cooldown, extending primary-to-secondary leakage, and therefore maximizing the offsite dose.
Pressurizer Level High initial level,'ith a pocitive uncertainty maximizes primary-to-secondary leakage and maximizes pressurizer heater operation.
Reactor Vessel Averace Temnerature Nominal vessel average temperature with a negative uncertainty is used to minimize the M.itial steam generator steam pressure.
This maximizes the initial dirferential pressure across the steam generator tubes and therefore maximizes the initial primary-to-secondary leakage. A lower vessel average temperature also anximizes the initial mass flow rate through the ruptured tube.
RCS Flow Nominal primary system loop flow with a negative uncertainty is assumed.
Low forced circulation flow results in lower natural circulation flow during the post-trip cooldown.
This reduces primary-to-secondary heat transfer and extends plant cooldown.
Frictional and form losses will also be smaller throughout the RCS, resulting in a higher primary pressure at the break location. This maxLaizes primary to secondary leakage.
Core Bvoass Flow Core bypass flow is not an important parameter for this transient.
Steam Generator Level Minimum steam generator level reduces the initial secondary inventory available to mix with and dilute the primary-to-secondary leakage.
This also minimizes the secondary side static head at the break location, thus maximizing primary to secondary leakage.
Fuel Temoerature High initial fuel temperature maximizes the stored energy which must be removed during the post-tr.ip natural circulation cooldown.
Steam Generator Tube Pluccind A bounding high tubc plugging icycl niniaircs th$ initial steam gcnarator st:aa pressurc and thereforc saximi;cs the pressure diffc ential between the primary and accondary systcas. Steam generator tube plugging is not an important parameter in this analysis.
6 7-7
,+.r 7.2.2.3 Boundary conditions sinale Failure The single failure identified for maximizing offsite dose is the failure of the PORV on the ruptured steam generator to close.
In this analysis, this valve remains open until operator action is taken to isolate the PORV.
Per Reference 4, page 5-7, "The most limiting failure would be the loss of air supply or power which prevents actuation of the (PORVs) from the main control room. The valves could be operated (locally) by manual action to correct for this single failure." This failure is incorporated into the analysis as it prolongs the transient, maximizing the primary-to-secondary leakage.
Pressurizer Safety Valves The pressurizer code safety valves are not challenged during the course of this transient.
Steam Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which minimize secondery pressurc and maximizc atmospheric steam rolessesmaximize secondary pressure. This delays operator identification of the failed open steam line PORV.
Steam Line PORVs Only two of the three steam line PORVs on the intact steam generators are assumed to be operablo. This lengthens the cooldown time, thereby maximizing the atmospheric steem releases. A negative bias is applfed to the ruptured steam generator PORV control signal. This results in an earlier opening time which maximizes atmospheric releases and delays operator identification of the failed open steam line PORV.
A positive bias is applied to the intact SG PORV control signals to maximize secondary side post-trip' pressurization. This delays operator identification of the failed open steam line PORV.
Decay Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is employed.
Offsite Power offsite power is assumed to be lost coincident with turbine trip. This isolates steam flow to the condenser, thereby maximizing the atmospheric steam releases.
Break Model The break is assumed to be a double-ended guillotine break of a single steam generator tube at the tubesheet surface on the steam generator t
outlet plenum _. This location maximizes the mass flow through the break.
RCP Oneration The reactor coolant pumps are assumed to operate normally until offsite power is lost coincident with turbine trip.
7-8
9 Eccs Iniection SI actuation is assumed to occur on low pressurizer pressure at a setpoint with an applied positive uncertainty or on manual operator action. Maximum ECCS injection flow is assumed to maximize the primary-to-secondary leakage.
Main Feedwater Main feedwater. flow is assumed to terminate coincident with the loss of of fsite power to minimize the secondary inventory available to mix with and dilute primary-to-secondary leakage.
Charcinn Flow A conservatively high charging flow capacity is modeled to delay reactor trip and maximize total primary-to-secondary leakage.
Manual Actigns
- Immediate action to maximize charging flow (penalty).
- Immediate action to energize pressurizer heater banks (penalty).
operators identify the abnormal condition of the RCS at 20 minutes and manually trip the reactor if not already tripped.
- Identify and isolate ruptured steam generator consistent with assumptions in WCAP-10698 (Reference 5), 4915 minute minimum delay (credit).
- Isolate failed open steen line drains upstream of the main steam isolation valves. This action occurs 10 minutes after the ruptured steem generator is identified.
- Isolate the steam supply to the turbine-driven auxiliary feedwater pump from the ruptured steam generator after identification of the ruptured steam generator. An operator action delay time of 530 l
minutes is assumed (credit).
- Isolate f ailed open steam line PORV on the ruptured steam generator with an operator action delay time from when it should have closed normally. The delay times assumed are 510 minutes for control room and +530 minutes for local operation (credit).
- Manually control auxiliary feedwater to maintain zero power steam generator levels (nominal).
- Using the steam line PORVs, init. ate natural circulation cooldown of the primary system af ter identification of the ruptured steam generator. Operator action delay times of 515 minutes for control room action and +445 minutes for local action are assumed (credit).
- Initiate depressurization of the primary system using the pressurizer PORVs to terminate break flow 910 minutes after the primary system is 200F subcooled at the ruptured steam generator pressure, or 2 minutes after the cocidown has bcca ce;;pleted (credit).
7-9 1
~,. _-
?
Terminate safety inja:; ion ficu.:5 n pressurincr icv;; rcccycr; with a conscrvativc delay (pcnalty;.
Manually control charging fica after safety injection tcrmination to maintain the ncro scwc; pressurizer icvci (ncminal) 7.2.2.4 Control, Protection, and Safeguards System Modeling Reactor Trio A reactor trip occurs on eithGr low pressurizer pressure or manual operator action at 20 minutes. A negative uncertainty is applied to the low pressurizer pressure trip setpoint to delay reactor trip.
The overtemperature AT trip function is not credited. cvertu.perature AT.
Ths. Tc;hnical Opecification scspensc tiacs are used and thc safety analysis setpcints in:1ude thc cffcct-of uncertainty in the monitored parametcr and in the setpcint Pressurizer Pressure Control This control system is assumed to be in manual and therefore is not modeled. Operator action is assumed to energize the pressurizer heaters and control the PORVs.
Pressurizer spray is not available for the duration of this transient.
Pressuri7er Level Control This control system is assumed to be in manual and th'erefore is not modeled.
Operator action is assumed to maximize charging flow.
crer tir : ror'Is
=d condenser steam Dumn The acin ; teas PORVs are assuccd to be opcrable for thi-transient with a positive bias applied to the ccntrol signal.
This assumption minimi;e; sccondary pressure and saxisites atmospheric steam releases.
The condenser steam dump valves are not assumed to be operable.
Condenser steam dump would nonconservatively minimize offsite doses.
Steam Generator Level Control
.This control system-is assumed to operate to maintain the initial steam generator level prior to reactor trip.
Main Feedwater Pemn Soeed Control This control systen is assuned to cperatc to acintain the ini' t. E-eteem gersr;tc; icvel prics tc scactcr trip.The results of this tral.
nt are not sensitive to the mode of Mrw pump speed control as long as une steam generator level is kept within the range that avoid protection or safeguards actuation.
Rod Control No credit is taken for the operation of the Rod Contr :1 System for this transient, which results in a slight increase in RCS temperature. With the Rod Control System in automatic, the control rods would cause a to negative reactivity addition as they are inserted in an attempt maintain RCS temperature at its nominal value.
7-10
i s
Tarbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.
Turbine trip on reactor trip is delayed by 0.3 seconds to maximize primary to secondary leakage.
&;xiliarv Feedwater Auxiliary feedwater initiation occurs after the loss of offsite power with a delay, consistent with Technical Specifications.
If applicable, aA purge volume of hot water is assumed to be delivered before cold feedwater reaches the steam generators.
Minimum flow rates are assumed to minimize primary-to-secondary heat transfer.
M9IV Closure Automatic MSIV closure is assumed using both a dynamically compensated and a static steam line pressure signal.
Early closure maximizes the primary leakage released to the atmosphere through the failed open steam line PORV.
l l
l l
l l
1 i
7-11
)
1 1
?
Table 8-1 Accident Analysis Assumptions i
FSAR Section 15.1.2 15.1.3 15.2.3 15.2.3 15.2.6 15.2.6 15.2.6 15.2.6 15.2.7 l 15.2.7 Report Section 2.2 2.3 3.1.1 3.1.2 3.2.1 3.2.2 3.2.3 3.2.4 3.3.1 3.3.2 Nominal Nominal High High High High Nominal High High High High Nominal Power Pzr Pressure Nominal Nominal High High High High Low Pzr Level High High High High High RCS Temp Nominal Nominal High High High High Nominal Hich High High RCS Flow Nominal Nominal Low High Low High Nominal Low High Nominal Bypass Flow Nominal Nominal High High High High ilighLow llighLow SG Level Low Low Low Fuel Temp Low Low Low Low Low Low High SG Tube Plugging None None High None High None High High None Auto Off Off Pzr Spray Off Auto Off Off Off Auto Pzr Heaters Off
- Off Autcon Autcon Auto Auto closed Closed PZr PORVs Closed Auto Closed Closed Closed Closed Closed Closed Closed Closed Closed (D
Steam Dump Closed Closed Closed Closed Closed Closed Closed Closed SM PORVs k
SG Level 4
MFW Pump Speed Manual Manual Manual Manual Turbine Control Auto Auto Auto Auto Rod Control SI Signal SI Flow SI Delay LOSP LOSP LOSP LOSP SG Lvl SG Lvl AFW Signal se Min Min Min Min Min Min Lvl**
TS TS TS TS TS TS AFW Flow Min **
Turb Trip Signal SG Lvl LOSP LOSP LOSP LOSP Rx Trip Rx Trip AFW Delay 99**
Turb Trip Delay TS None None None None None None None None Stm Line Isol Signal Stm Line Isol MFW Isol Signal SG Lvl
~
~
Delay MFW Isol Delay TS Notes:Refer to the text discussion of this transient.
Results of_the transient are insensitive to the choice about this parameter.either because the tra Not applicable, malfunction of that system might be the cause of the transient.
9
-n-s-
n
N
- ---- -1
. -a
yn Table 8-1 (continued)
Accident Analysis Assumptions FSAR Section 15.2.7 15.2.8 15.2.8 15.3.1 15.3.2 15.3.3 15.3.3 15.4.1 15.4.1 15.4.2 Section 3.3.3 3.4.1 3.4.2 4.1 4.2 4.3.1 4.3.2 5.1.1 5.1.2 5.2.1 Power Nominal Nominal High Nominal Nominal High HtqhNominal 0
Nc.T. ir a 10
{
Pzr Pressure Nominal Nominal Low Nominal Nominal High bewNominal High Nominal High l
Pzr Level Low Low Low Low Low High Low High Low High RCS Temp Nominal Nomir:al High Nominal Nominal High HtqhNominal Nominal
}
l Nominal' RCS Flow Nominal Nominal Low Nominal Nominal Low bewNominal Nominal Nominal Nominal High HtehNominal Dyoass Flow Nominal Nominal SG Level HtghLow Low Mhth**
Fuel Temp High High High High High Low High Low Low Low SG Tube Plugging High Hteh**
High None High
[
Pzr Spray Auto Auto Auto Auto Auto Off Auto Off Auto Off Off Pzr Heaters Off
- Off Off Off Off Auto Off Pzr PORVs Auto Auto Closed Auto Auto Closed Auto Closed Auto Closed SM PORVs Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed 7
Steam Dump Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed W
SG Level Auto MF"4 Pump Speed Rod Control Manual Manual Manual Manual Manual Manual Manual Turbine Control Auto Auto Auto Auto Auto Auto Auto Auto SI Signal SI Flow Min SI Delay TS SI AFW Signal SG Lvl Min AFW Flow Min TS AFW Delay TS Turb Trip Signal Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip RX Trip Turb Trip. Delay None None None None None None None None Stm Line Isol
+
Signal Stm Line Isol Delay MFW Isol Signal MFW Isol Delay Notes: -Refer to the text discussion of this transient.
Results of the transient are insensitive to the choice about this parameter.
Not' applicable, either because the transient does not challenge that contol system or because the malfunction of that system might be the cause of the transient.
i-mw
--A, r-wy wm--
N y -',,
y w-
- -yaiN
- +
w-V-
^
m
a
}
Table 8-1 (continued)
Accident Analysis Assumptions 3
FSAR Section 15.4.2 15.4.3c 15.4.3d 15.4.4 15.4.6 15.4.7 15.5.1 15.5.1 15.6.1 15.6.3 15.6.3 Section 5.2.2 5.3 5.4 5.5 5.6 5.7 6.1.1 6.1.2 7.1 7.2.1 7.2.2 Nominal 0
Nominal Nominal High Power Nominal Nominal P2r Pressure Nominal Nominal Low Nominal Low Nominal Nominal High Pzr Level Low High Low High Low High Low High RCS Temp Nominal Nominal High Nominal Low Nominal Nominal Low RCS Flow Nominal Nominal Low Nominal Nominal Nominal Low Bypass Flos Nominal Nominal High Nominal Nominal Nominal Low SG Level
- igh* *
- igh**
Fuel Temp Low Low Low High High High High High
- igh *
- i SG Tube Plugging liich*
- igh**
-High P2r Spray Auto Auto Auto Auta -
Off P2r Heaters Off Off Off Off.
Off Off Off Manual Closed Auto Auto Auto -
Manual SM PORVs Closed Closed Closed Closed Closed P2r PORVs Steam Dump Closed Closed Closed Closed Closed Closed 7
SG Level Auto Auto Auto Auto Autc
- e Auto MFW Pump Speed Rod Control Auto Manual Auto Manual hManual Turbine Control Auto Auto Manual Auto Auto Auto Auto SI Signal LC-Ptt SI Flow Max Max Max SI Delay None Logp Min AFW Signal
=*
TS AFW Flow Turb Trip Signal Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip AFW Delay e
Turb Trip Delay None None None None-None Nenc Stm Line Isol
_a Sicnal utm Line Isol LOSP Delay None MFW Isol Signal
__MFW Isol Delay Notes: Refer to the text discussion of this transient.
Results of the transient are insensitive to the choice about this parameter.either because the tran Not applicable,
~
malfunction of that system might be the cause of the transient.
L
...,. - :r,
y.s
&wn.
ite%%&mg;mpmy y. g; _ gg u
I r
Table 8-1 (continued)
Accident Analysis Assumptions FFAR Section 15.4.2 15.4.3c 15.4.3d 15.4.4 15.4.6 15.4.7 15.5.1 15.5.1 15.6.1 15.6.3 15.6.3 Section 5.2.2 5.3 5.4 5.5 5.6 5.7 6.1.1 6.1.2 7.1 7.2.1 7.2.2 Nominal 0
Nominal Nominal High Power Nominal Nominal Pzr Pressure Nominal Nominal Low Nominal Low Nominal Nominal High Pzr Level Low High Low High Low High Low High RCS Temp Nominal Nominal High Nominal Low Nominal Nominal Low RCS Flow Nominal Nominal Low Nominal Nominal Nominal Low Bypass Flos Nominal Nominal High Nominal Nominal Nominal Low SG Level f&cyh**
H @ **
Fuel Temp Low Low Low High High High High High Migh**
SG Tube Plugging High*
Nigh**
-High Pzr Spray Auto Auto Auto Auto--
Off Pzr Heaters Off Off Off Off off Off Off Manual Closed Auto Auto Auto -
Manual Pzr PORVs SM PORVs Closed Closed Closed Closed Closed Steam Dump Closed Closed Closed Closed Closed Closed 7
SG Level Auto Auto Auto Auto Auto Aatc**
A MFW Pump Speed Rod Control Auto Manual Auto Manual
- Manual Turbine Control Auto Auto Manual Auto Auto Auto Auto SI Signal Lc;g ptr h*
SI Flow Max Max Max SI Delay None LoSp AFW Flow Min AFW Signal TS Turb Trip Signal Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip Rx Trip AFW Delay e
Turb Trip Delay None None None None None Nonc Stm Line Isol
~*
Stm Line Isol Signal Delay LOSP None MFW Isol Signal MFW Isol Delay Notes:
discussion of this transient.
Refer to the text this parameter.
Results of the transient are insensitive to the choice abouteither because the transient does not challenge th Not applicable, malfunction of that system might be the cause of the transient.
mmmmmm n _,..,,,
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