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Response of gas-cooled fast breeder reactors to depressurization accidents
NUCLEAR ENGINEERING AND DESIGN 26 (1974) 195-200. © NORTH-HOLLAND PUBLISHING COMPANY
RESPONSE
OF GAS-COOLED
FAST BREEDER
TO DEPRESSURIZATION
REACTORS
ACCIDENTS
David R. BUTTEMER and James A. LARRIMORE Gulf General Atomic Company, San Diego, California 92138, USA Received 16 July 1973
The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV). The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient response following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor. Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.
1. Introduction
2. Descriptive background
Depressurization accidents play a central role in gas-cooled reactor safety studies since the gaseous coolant is pressurized so as to improve the heat transfer and pumping characteristics. The prediction of detailed reactor response during a depressurization accident is of more importance in a fast gas-cooled, reactor with low core heat capacity and high power density than in a thermal gas-cooled reactor with large core heat capacity and low power density. Therefore, considerable effort has been given over the past four years to the accurate analysis of depressurization accidents as part o f the safety program on the gas-cooled fast breeder reactor (GCFR) at Gulf General Atomic (GGA) described in a companion paper [1]. Results o f other studies of depressurization accidents in gas-cooled fast breeder reactors have also been reported recently [2, 3 ].
To discuss the depressurization accident in a gascooled breeder reactor it is necessary to first describe the relevant design features of the reactor and cooling systems. In this paper, we shall present details for the helium-cooled 300 MW(e) G C F R demonstration plant designed at GGA [1] but the characteristics of the system response apply generally to gas-cooled breeder reactors in which the nuclear steam supply system is contained within a prestressed concrete reactor vessel (PCRV). The main nuclear steam supply components are the reactor core and its supporting structure, the three main cooling loops, and the three auxiliary cooling loops. Each main cooling loop contains a steam generator and a steam-turbine-driven helium circulator and each auxiliary-cooling loop contains a helium-to-pressurized water heat exchanger and an electrically-driven helium circulator. The auxiliary
196
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
loops are designed to remove the reactor decay heat and provide means of core cooling independent of the main loops which continue to provide core cooling after reactor trip during normal and accident conditions (see fig. 3 of ref. [1 ] ). The reactor core is located within the central PCRV cavity and is supported at its upper end by a thick grid plate. An 11.5 ft diameter concrete closure plug is located above the reactor cavity to facilitate installation of reactor components during plant construction. The core is made up of stainless steel clad rods containing mixed oxide fuel pellets which are arranged within hexagonal shaped fuel elements. The fuel rod cladding surface is roughened over part of the length to enhance heat transfer in the hotter part of the core at the expense of a slightly higher system pressure drop. The helium coolant flows downward through the core, entering at 593°F and exiting into the large reactor outlet plenum at 1007°F during full power operation. Helium then enters the three steam generators through the lower cross-over ducts, flows up through the central duct within each steam generator, reverses direction and flows down over the helically wound tube bundle. Uphill boiling occurs in the steam generator and all sections are cross-counterflow. The helium leaving the lower end of the steam generator reverses direction and flows up through an annular space surrounding the steam generator bundle into the circulator inlet plenum. The helium enters the circulator and then passes into the reactor inlet plenum through the upper cross-over duct which contains the helium loop isolation valves. At the top of each steam generator cavity there is a large concrete closure plug for steam generator installation. In addition to the karge PCRV closures over the reactor and steam generator cavities, there are smaller PCRV penetrations with closures for such things as control rod drives, fuel handling apparatus, and reactor instrumentation. The secondary coolant system and its control during an accident situation, and particularly the operation of the steam-turbine-driven helium circulator, influence the consequences of a depressurization accident. (A flow diagram of the secondary coolant system is shown in fig. 4 of ref. [1 ] .) During normal plant operation, feedwater is provided to each of the three steam generators by two extraction steam-
driven feed pumps. Flow to each steam generator is controlled by feedwater control valves. Superheated steam leaving each steam generator passes through two parallel circulator turbine steam control valves and then into the circulator turbine. The large valve is controlled so as to maintain the correct helium-tosteam flow ratio over the operational load range. The smaller of the two valves is used for control of the circulator turbine after reactor shutdown when the larger valve is closed. Steam leaving the circulator turbine is resuperheated, combines with the other loops and enters the main turbine. The remainder of the secondary coolant system is similar to that of a conventional non-reheat turbine system. Upon generation of a reactor trip signal, the main turbine is tripped, the main boiler feed pumps are tripped, the large circulator turbine control valves are closed over a few second period and control is assumed by the small control valves. Resuperheat bypass valves open which divert the steam leaving the circulator turbines directly through desuperheaters to the condenser. Electrically-driven shutdown boiler feed pumps located in each loop start within one minute after reactor trip and provide a constant 2% of normal feedwater flow to each loop. The same control actions occur during all reactor trips regardless of what initiated the trip. Reactor shutdown cooling is provided by continued operation of the main cooling loops using the steam stored in the steam generators and produced from residual heat to drive the turbocirculators. Closing the large circulator turbine control valve rapidly after the trip slows the circulators down which prevents core overcooling and reduces the subsequent thermal transients on the coolant system components. The small circulator turbine control valves are controlled so as to keep the reactor outlet temperature constant. Resuperheat bypass valves are controlled using a helium pressure signal to maintain the circulator turbine exhaust pressure proportional to helium pressure. During a depressurization accident, this control reduces the circulator turbine outlet pressure causing more energy to be extracted per pound of steam passing through the circulator turbine and consequently delivering the increased circulator power required to obtain adequate helium flow at the lower pressure condition. The depressurization accident is assumed to be initiated by a failure in the primary hold-down system
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
of the large PCRV reactor cavity closure which, in turn, is assumed to cause the primary seal to fail. The closure would be retained by the structurally independent secondary hold-down system. The depressurization rate would be controlled by flow restrictors incorporated into the closure design structurally independent from the hold-down systems. The secondary containment building in which the PCRV is located limits the minimum helium pressure in the PCRV at the end of the depressurization. The preliminary sizing of the secondary containment building for the GCFR demonstration plant would result in a minimum pressure of about 2 atm absolute. This back pressure, above atmospheric, aids core cooling following depressurization since less pumping power is required for a given mass flow with higher density. Core cooling continues to be provided by the main cooling loops with the auxiliary cooling loops capable of independently providing core cooling if necessary. Studies have shown that a failure of the reactor cavity closure seal results in higher core temperatures than if the same failure occurred in one of the steam generator cavity closures so the results presented below are for leakages from the reactor cavity closure. The potential leak areas around the reactor cavity or steam generator cavity closures are essentially the same; however, the leak rate from the reactor cavity closure is somewhat higher because cooler, and therefore more dense helium, is leaking.
3. Analysis method To accurately analyze depressurization accidents, an elaborate computer program has been developed at GGA. This program subdivides the primary coolant system into several volumes which are interconnected with detailed mathematical models representing either a reactor core, a steam generator, a helium circulator, or an adiabatic connection simulating the leak. Compressible flow equations are used only in evaluating the flow through the leak; all other connections are treated as incompressible flow which is sufficiently accurate for the leak rates studied. About 700 system state variables are evaluated each calculational time step and computational time is typically three to four times real time. A schematic diagram of the model used to analyze
197
"P~'q
Q
HELIUM VOLUME LOOP ISOLATION VALVE
C~HELIUM CIRCULATOR
~
STEAMGENERATOR
I I
FLOWRESTRC I TOR
Fig. 1. Schematic diagram of analysis model for the depressurization accident. Legend: R.I.P. = reactor inlet plenum, R.O.P. = reactor outlet plenum, C.I.P. = circulator inlet plenum, and C.B. = containment building.
depressurization accidents is shown in fig. 1. This case represents a leak in the reactor inlet plenum and continued cooling with all three main loops. Six volumes are modeled: the containment building, the reactor inlet and outlet plenums, and the three circulator inlet plenums. The program has been written so as to allow various arrangements of volumes and interconnections. The thermal response of the core is represented by an average powered fuel rod modeled with eight radial and 18 axial nodes. A similar parallel calculation of the maximum powered fuel rod with statistically evaluated hot channel and hot spot factors is also performed in order to monitor maximum cladding temperature during the depressurization accident. Variations of both the heat transfer and friction factor multipliers with Reynolds number in the roughened portions of the core are accounted for. An accurate dynamic representation of the steam generator has been developed which accounts for movements of both the liquid-to-mixture and the mixture-to-vapor transition zones and changes in the water/steam inventory are evaluated. Tube side heat transfer is evaluated as a function of flow rate and
198
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
scram delay is assumed. Reactor trip results in a main turbine trip, the main feedwater pumps being shut down, the resuperheaters being bypassed, and the closure in a few seconds of the large circulator turbine throttle valves in the three main loops. The circulator speed increases somewhat before reactor trip because of the decreasing pressure head and helium density;it then decreases rapidly as the large circulator turbine control valves close to avoid excessively overcooling the core and then gradually increases as the system depressurizes. The steam flow to the circulators is thereafter controlled by the small control valves. The cladding hot spot temperature increases slightly prior to reactor trip, falls initially after trip then increases somewhat while fuel temperatures rapidly decrease as the stored heat within the fuel rods is removed. As depressurization continues, cladding temperature decreases and then slowly increases to a final peak which occurs shortly after depressurization is complete. The maximum cladding temperature remains well below the damage level. During the transient, the steam generators provide steam to drive the helium circulators and the steam generator pressure slowly decreases due to reduction in water inventory and in decay heat input. The steam generator pressure decreases from 2900 to 1000 psi 30 min after the accident, and the steam generator inventory reduces to a minimum of 60% of its initial value. Continued long-term core cooling will be maintained by driving the circulators with steam produced in
phase and the tube bundle heat capacity is accounted for. Steam and feedwater valve controllers are included and the resulting valve actuation is calculated. The performance of the circulator and its steam turbine drive are included over the entire operating range by a set of performance tables which are built into the code. Inertial effects of the turbocirculator unit are used in determining the speed change and the circulator water bearing frictional losses, which are quite a significant factor following a depressurization accident, are accounted for. The connection used to model the leak solves the adiabatic compressible flow equation treating both the choked and subsonic flow conditions.
4. Results of depressurization analyses The system response during the design basis depressurization accident is shown in fig. 2. This case assumes a failure in the reactor cavity closure seal with the helium leaking from the reactor inlet plenum into the containment building through the flow restrictor which limits the leak area to 25 in 2. Depressurization from 85 arm to 2 atm is complete in 4 min, which corresponds to an exponential pressure decay time constant of about 60 sec. Reactor trip is initiated by a low primary coolant system pressure signal which occurs 4 sec after the failure and an additional 0.5 sec detection-actuation 1600 1400 TEMPERATURE 1200 (°p) 1000
125 100
80O
STEAM GENERATOR INVENTORY (%)
5O
TIME SCALE CHANGE 15,000
10,000 CIRCULATOR SPEED (RPM)
1600 1000 PRESSURE (PSi) 500
5,000
6 2
6
10
14
10
22
26
30
TIME (MIN)
Fig. 2. System response during the design basis depressurization accident.
199
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
auxiliary boilers which are started up after a reactor trip signal. It should be emphasized that the core temperature response in a depressurization accident in a GCFR is primarily determined by the mass flow through the core. The core heat capacity is relatively low so that the stored heat in the fuel is removed in about the first 10 sec after reactor trip (this time depends on the gap conductance between fuel and cladding). The steam generators have a relatively large heat capacity and thus serve both as good primary heat sinks and as sources of energy, in the form of steam, to drive the helium circulators. The core flow rate can be affacted through control of the circulator turbine steam flow and pressure level. The present method of control, which maintains the circulator turbine exhaust pressure proportional to the helium pressure level by means of the resuperheated bypass pressure regulating valve and which adjusts the circulator turbine steam flow rate by the small throttle valve by sensing the reactor outlet temperature, appears to function adequately over a wide range of operational and accident conditions which have been investigated. The maximum leak area can be limited by the PCRV penetration closure and flow restrictor design, which is a considerable advantage in the use of the PCRV to enclose the entire reactor coolant system. To determine safety margins in the design, the effects of leak areas considerably larger than the design basis value of 25 in 2. have been investigated. The principal effect of a large leak area is that depressurization is completed more rapidly so that core heat removal reaches its minimum capability at an earlier time when the decay heat level is higher. Heat removal capability during the period when the stored heat leaves the fuel is also somewhat reduced. The result is that both the initial and final cladding temperature peaks tend to increase with increasing leak area. The final cladding temperature peak remained the highest in the cases studied. Maximum cladding hot spot temperature does not increase very rapidly with increasing leak area, as shown in fig. 3. For example, a 200 in 2. leak area, one eight times that limited by the flow restrictor, would result in a maximum cladding hot spot temperature of 1800°F and a maximum cladding temperature for the average fuel rod of 1450°F. With the leak occur-
i i
DESIGN BASIS LEAK AREA
o v
1800
~
-
16oo WITH HOT-SPOT FACTORS IkO0
1200
--
IOO0
25
1
I
1
50
100
15o
200
LEAKAREA(IN. 2) Fig. 3. Maximumcladdingtemperatures as a function ofleak area in depressurization accidents. ring in the reactor inlet plenum it is possible to have a temporary flow reversal within the core during the depressurization if the flow out of the leak exceeds the flow supplied by the circulators. Momentary core flow reversal would occur, at the time the circulator speed reaches its minimum value, for a leak area of 180 in 2. It would last for such a short time that the effect on fuel rod temperature would be small. The secondary containment atmosphere pressure and temperature responses resulting from the design basis depressurization accident are shown in fig. 4. The calculated helium temperature and leak rate into
24
600
20
5OO 400
~..
12
=..
8
*
300 200 I---
TEMPERATURE I O0
I
I
I
I
I
5
I0
15
20
25
0
30
TIME (MIN)
Fig. 4. Containment building atmosphere pressure and temperature transients during the design basis depressurization accident. Containment volume = 1.18 × 106 ft 3, helium leaked = 8.200 lb.
200
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
the containment were used as input for a modified version of the CONTEMPT [4] computer code. This code performs detailed mass and energy balances within the containment and accounts for energy removed from the containment atmosphere by heat transfer to internal structures. The maximum temperature is 280 ° F and the maximum pressure is 16 psig, which occurs near the time when helium depressurization is complete. If the entire PCRV helium inventory were to be instantaneously mixed with the air within the secondary containment building the peak containment pressure would be only 20 psig, so the peak containment pressure is also not strongly dependent upon the depressurization leak rate.
5. Conclusion In conclusion, substantial progress has been made in understanding the response of the gas-cooled fast
breeder reactor to depressurization accidents. Detailed analyses for the 300 MW(e) demonstration plant have shown that this class of accident can be safely handled with a considerable safety margin.
References [ 1] S.J. Milioti and J.A. Larrimore, Status and safety aspects of the 300 MW(e) GCFR demonstration plant, Nucl. Eng. Des. 26 (1) (1973). [2] J.A. Dupont, G. Sarlos and A. Tiberini, The rapid loss of pressure in a GCFR plant using a direct cycle, Trans. Amer. Nucl. Soc. 15, November (1972) 849. [3] F. Decamps, J.R. Stanbridge and E. Turrini, Summary of the safety work on some gas breeder reference designs, J. Brit. Nucl. Energ. Soc. II (4) October (1972) 345. [4] L.C. Richardson et al., CONTEMPT - a computer program for predicting the containment pressure-temperature response to a loss-of-coolant accident, USAEC Report, IDO-17220, Phillips Petroleum Company, June (1967).
RESPONSE
OF GAS-COOLED
FAST BREEDER
TO DEPRESSURIZATION
REACTORS
ACCIDENTS
David R. BUTTEMER and James A. LARRIMORE Gulf General Atomic Company, San Diego, California 92138, USA Received 16 July 1973
The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV). The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient response following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor. Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.
1. Introduction
2. Descriptive background
Depressurization accidents play a central role in gas-cooled reactor safety studies since the gaseous coolant is pressurized so as to improve the heat transfer and pumping characteristics. The prediction of detailed reactor response during a depressurization accident is of more importance in a fast gas-cooled, reactor with low core heat capacity and high power density than in a thermal gas-cooled reactor with large core heat capacity and low power density. Therefore, considerable effort has been given over the past four years to the accurate analysis of depressurization accidents as part o f the safety program on the gas-cooled fast breeder reactor (GCFR) at Gulf General Atomic (GGA) described in a companion paper [1]. Results o f other studies of depressurization accidents in gas-cooled fast breeder reactors have also been reported recently [2, 3 ].
To discuss the depressurization accident in a gascooled breeder reactor it is necessary to first describe the relevant design features of the reactor and cooling systems. In this paper, we shall present details for the helium-cooled 300 MW(e) G C F R demonstration plant designed at GGA [1] but the characteristics of the system response apply generally to gas-cooled breeder reactors in which the nuclear steam supply system is contained within a prestressed concrete reactor vessel (PCRV). The main nuclear steam supply components are the reactor core and its supporting structure, the three main cooling loops, and the three auxiliary cooling loops. Each main cooling loop contains a steam generator and a steam-turbine-driven helium circulator and each auxiliary-cooling loop contains a helium-to-pressurized water heat exchanger and an electrically-driven helium circulator. The auxiliary
196
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
loops are designed to remove the reactor decay heat and provide means of core cooling independent of the main loops which continue to provide core cooling after reactor trip during normal and accident conditions (see fig. 3 of ref. [1 ] ). The reactor core is located within the central PCRV cavity and is supported at its upper end by a thick grid plate. An 11.5 ft diameter concrete closure plug is located above the reactor cavity to facilitate installation of reactor components during plant construction. The core is made up of stainless steel clad rods containing mixed oxide fuel pellets which are arranged within hexagonal shaped fuel elements. The fuel rod cladding surface is roughened over part of the length to enhance heat transfer in the hotter part of the core at the expense of a slightly higher system pressure drop. The helium coolant flows downward through the core, entering at 593°F and exiting into the large reactor outlet plenum at 1007°F during full power operation. Helium then enters the three steam generators through the lower cross-over ducts, flows up through the central duct within each steam generator, reverses direction and flows down over the helically wound tube bundle. Uphill boiling occurs in the steam generator and all sections are cross-counterflow. The helium leaving the lower end of the steam generator reverses direction and flows up through an annular space surrounding the steam generator bundle into the circulator inlet plenum. The helium enters the circulator and then passes into the reactor inlet plenum through the upper cross-over duct which contains the helium loop isolation valves. At the top of each steam generator cavity there is a large concrete closure plug for steam generator installation. In addition to the karge PCRV closures over the reactor and steam generator cavities, there are smaller PCRV penetrations with closures for such things as control rod drives, fuel handling apparatus, and reactor instrumentation. The secondary coolant system and its control during an accident situation, and particularly the operation of the steam-turbine-driven helium circulator, influence the consequences of a depressurization accident. (A flow diagram of the secondary coolant system is shown in fig. 4 of ref. [1 ] .) During normal plant operation, feedwater is provided to each of the three steam generators by two extraction steam-
driven feed pumps. Flow to each steam generator is controlled by feedwater control valves. Superheated steam leaving each steam generator passes through two parallel circulator turbine steam control valves and then into the circulator turbine. The large valve is controlled so as to maintain the correct helium-tosteam flow ratio over the operational load range. The smaller of the two valves is used for control of the circulator turbine after reactor shutdown when the larger valve is closed. Steam leaving the circulator turbine is resuperheated, combines with the other loops and enters the main turbine. The remainder of the secondary coolant system is similar to that of a conventional non-reheat turbine system. Upon generation of a reactor trip signal, the main turbine is tripped, the main boiler feed pumps are tripped, the large circulator turbine control valves are closed over a few second period and control is assumed by the small control valves. Resuperheat bypass valves open which divert the steam leaving the circulator turbines directly through desuperheaters to the condenser. Electrically-driven shutdown boiler feed pumps located in each loop start within one minute after reactor trip and provide a constant 2% of normal feedwater flow to each loop. The same control actions occur during all reactor trips regardless of what initiated the trip. Reactor shutdown cooling is provided by continued operation of the main cooling loops using the steam stored in the steam generators and produced from residual heat to drive the turbocirculators. Closing the large circulator turbine control valve rapidly after the trip slows the circulators down which prevents core overcooling and reduces the subsequent thermal transients on the coolant system components. The small circulator turbine control valves are controlled so as to keep the reactor outlet temperature constant. Resuperheat bypass valves are controlled using a helium pressure signal to maintain the circulator turbine exhaust pressure proportional to helium pressure. During a depressurization accident, this control reduces the circulator turbine outlet pressure causing more energy to be extracted per pound of steam passing through the circulator turbine and consequently delivering the increased circulator power required to obtain adequate helium flow at the lower pressure condition. The depressurization accident is assumed to be initiated by a failure in the primary hold-down system
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
of the large PCRV reactor cavity closure which, in turn, is assumed to cause the primary seal to fail. The closure would be retained by the structurally independent secondary hold-down system. The depressurization rate would be controlled by flow restrictors incorporated into the closure design structurally independent from the hold-down systems. The secondary containment building in which the PCRV is located limits the minimum helium pressure in the PCRV at the end of the depressurization. The preliminary sizing of the secondary containment building for the GCFR demonstration plant would result in a minimum pressure of about 2 atm absolute. This back pressure, above atmospheric, aids core cooling following depressurization since less pumping power is required for a given mass flow with higher density. Core cooling continues to be provided by the main cooling loops with the auxiliary cooling loops capable of independently providing core cooling if necessary. Studies have shown that a failure of the reactor cavity closure seal results in higher core temperatures than if the same failure occurred in one of the steam generator cavity closures so the results presented below are for leakages from the reactor cavity closure. The potential leak areas around the reactor cavity or steam generator cavity closures are essentially the same; however, the leak rate from the reactor cavity closure is somewhat higher because cooler, and therefore more dense helium, is leaking.
3. Analysis method To accurately analyze depressurization accidents, an elaborate computer program has been developed at GGA. This program subdivides the primary coolant system into several volumes which are interconnected with detailed mathematical models representing either a reactor core, a steam generator, a helium circulator, or an adiabatic connection simulating the leak. Compressible flow equations are used only in evaluating the flow through the leak; all other connections are treated as incompressible flow which is sufficiently accurate for the leak rates studied. About 700 system state variables are evaluated each calculational time step and computational time is typically three to four times real time. A schematic diagram of the model used to analyze
197
"P~'q
Q
HELIUM VOLUME LOOP ISOLATION VALVE
C~HELIUM CIRCULATOR
~
STEAMGENERATOR
I I
FLOWRESTRC I TOR
Fig. 1. Schematic diagram of analysis model for the depressurization accident. Legend: R.I.P. = reactor inlet plenum, R.O.P. = reactor outlet plenum, C.I.P. = circulator inlet plenum, and C.B. = containment building.
depressurization accidents is shown in fig. 1. This case represents a leak in the reactor inlet plenum and continued cooling with all three main loops. Six volumes are modeled: the containment building, the reactor inlet and outlet plenums, and the three circulator inlet plenums. The program has been written so as to allow various arrangements of volumes and interconnections. The thermal response of the core is represented by an average powered fuel rod modeled with eight radial and 18 axial nodes. A similar parallel calculation of the maximum powered fuel rod with statistically evaluated hot channel and hot spot factors is also performed in order to monitor maximum cladding temperature during the depressurization accident. Variations of both the heat transfer and friction factor multipliers with Reynolds number in the roughened portions of the core are accounted for. An accurate dynamic representation of the steam generator has been developed which accounts for movements of both the liquid-to-mixture and the mixture-to-vapor transition zones and changes in the water/steam inventory are evaluated. Tube side heat transfer is evaluated as a function of flow rate and
198
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
scram delay is assumed. Reactor trip results in a main turbine trip, the main feedwater pumps being shut down, the resuperheaters being bypassed, and the closure in a few seconds of the large circulator turbine throttle valves in the three main loops. The circulator speed increases somewhat before reactor trip because of the decreasing pressure head and helium density;it then decreases rapidly as the large circulator turbine control valves close to avoid excessively overcooling the core and then gradually increases as the system depressurizes. The steam flow to the circulators is thereafter controlled by the small control valves. The cladding hot spot temperature increases slightly prior to reactor trip, falls initially after trip then increases somewhat while fuel temperatures rapidly decrease as the stored heat within the fuel rods is removed. As depressurization continues, cladding temperature decreases and then slowly increases to a final peak which occurs shortly after depressurization is complete. The maximum cladding temperature remains well below the damage level. During the transient, the steam generators provide steam to drive the helium circulators and the steam generator pressure slowly decreases due to reduction in water inventory and in decay heat input. The steam generator pressure decreases from 2900 to 1000 psi 30 min after the accident, and the steam generator inventory reduces to a minimum of 60% of its initial value. Continued long-term core cooling will be maintained by driving the circulators with steam produced in
phase and the tube bundle heat capacity is accounted for. Steam and feedwater valve controllers are included and the resulting valve actuation is calculated. The performance of the circulator and its steam turbine drive are included over the entire operating range by a set of performance tables which are built into the code. Inertial effects of the turbocirculator unit are used in determining the speed change and the circulator water bearing frictional losses, which are quite a significant factor following a depressurization accident, are accounted for. The connection used to model the leak solves the adiabatic compressible flow equation treating both the choked and subsonic flow conditions.
4. Results of depressurization analyses The system response during the design basis depressurization accident is shown in fig. 2. This case assumes a failure in the reactor cavity closure seal with the helium leaking from the reactor inlet plenum into the containment building through the flow restrictor which limits the leak area to 25 in 2. Depressurization from 85 arm to 2 atm is complete in 4 min, which corresponds to an exponential pressure decay time constant of about 60 sec. Reactor trip is initiated by a low primary coolant system pressure signal which occurs 4 sec after the failure and an additional 0.5 sec detection-actuation 1600 1400 TEMPERATURE 1200 (°p) 1000
125 100
80O
STEAM GENERATOR INVENTORY (%)
5O
TIME SCALE CHANGE 15,000
10,000 CIRCULATOR SPEED (RPM)
1600 1000 PRESSURE (PSi) 500
5,000
6 2
6
10
14
10
22
26
30
TIME (MIN)
Fig. 2. System response during the design basis depressurization accident.
199
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
auxiliary boilers which are started up after a reactor trip signal. It should be emphasized that the core temperature response in a depressurization accident in a GCFR is primarily determined by the mass flow through the core. The core heat capacity is relatively low so that the stored heat in the fuel is removed in about the first 10 sec after reactor trip (this time depends on the gap conductance between fuel and cladding). The steam generators have a relatively large heat capacity and thus serve both as good primary heat sinks and as sources of energy, in the form of steam, to drive the helium circulators. The core flow rate can be affacted through control of the circulator turbine steam flow and pressure level. The present method of control, which maintains the circulator turbine exhaust pressure proportional to the helium pressure level by means of the resuperheated bypass pressure regulating valve and which adjusts the circulator turbine steam flow rate by the small throttle valve by sensing the reactor outlet temperature, appears to function adequately over a wide range of operational and accident conditions which have been investigated. The maximum leak area can be limited by the PCRV penetration closure and flow restrictor design, which is a considerable advantage in the use of the PCRV to enclose the entire reactor coolant system. To determine safety margins in the design, the effects of leak areas considerably larger than the design basis value of 25 in 2. have been investigated. The principal effect of a large leak area is that depressurization is completed more rapidly so that core heat removal reaches its minimum capability at an earlier time when the decay heat level is higher. Heat removal capability during the period when the stored heat leaves the fuel is also somewhat reduced. The result is that both the initial and final cladding temperature peaks tend to increase with increasing leak area. The final cladding temperature peak remained the highest in the cases studied. Maximum cladding hot spot temperature does not increase very rapidly with increasing leak area, as shown in fig. 3. For example, a 200 in 2. leak area, one eight times that limited by the flow restrictor, would result in a maximum cladding hot spot temperature of 1800°F and a maximum cladding temperature for the average fuel rod of 1450°F. With the leak occur-
i i
DESIGN BASIS LEAK AREA
o v
1800
~
-
16oo WITH HOT-SPOT FACTORS IkO0
1200
--
IOO0
25
1
I
1
50
100
15o
200
LEAKAREA(IN. 2) Fig. 3. Maximumcladdingtemperatures as a function ofleak area in depressurization accidents. ring in the reactor inlet plenum it is possible to have a temporary flow reversal within the core during the depressurization if the flow out of the leak exceeds the flow supplied by the circulators. Momentary core flow reversal would occur, at the time the circulator speed reaches its minimum value, for a leak area of 180 in 2. It would last for such a short time that the effect on fuel rod temperature would be small. The secondary containment atmosphere pressure and temperature responses resulting from the design basis depressurization accident are shown in fig. 4. The calculated helium temperature and leak rate into
24
600
20
5OO 400
~..
12
=..
8
*
300 200 I---
TEMPERATURE I O0
I
I
I
I
I
5
I0
15
20
25
0
30
TIME (MIN)
Fig. 4. Containment building atmosphere pressure and temperature transients during the design basis depressurization accident. Containment volume = 1.18 × 106 ft 3, helium leaked = 8.200 lb.
200
D.R. Buttemer, J.A. Larrimore, GCFR response to depressurization accidents
the containment were used as input for a modified version of the CONTEMPT [4] computer code. This code performs detailed mass and energy balances within the containment and accounts for energy removed from the containment atmosphere by heat transfer to internal structures. The maximum temperature is 280 ° F and the maximum pressure is 16 psig, which occurs near the time when helium depressurization is complete. If the entire PCRV helium inventory were to be instantaneously mixed with the air within the secondary containment building the peak containment pressure would be only 20 psig, so the peak containment pressure is also not strongly dependent upon the depressurization leak rate.
5. Conclusion In conclusion, substantial progress has been made in understanding the response of the gas-cooled fast
breeder reactor to depressurization accidents. Detailed analyses for the 300 MW(e) demonstration plant have shown that this class of accident can be safely handled with a considerable safety margin.
References [ 1] S.J. Milioti and J.A. Larrimore, Status and safety aspects of the 300 MW(e) GCFR demonstration plant, Nucl. Eng. Des. 26 (1) (1973). [2] J.A. Dupont, G. Sarlos and A. Tiberini, The rapid loss of pressure in a GCFR plant using a direct cycle, Trans. Amer. Nucl. Soc. 15, November (1972) 849. [3] F. Decamps, J.R. Stanbridge and E. Turrini, Summary of the safety work on some gas breeder reference designs, J. Brit. Nucl. Energ. Soc. II (4) October (1972) 345. [4] L.C. Richardson et al., CONTEMPT - a computer program for predicting the containment pressure-temperature response to a loss-of-coolant accident, USAEC Report, IDO-17220, Phillips Petroleum Company, June (1967).