Thermal Response of a High Temperature Reactor during Passive ...
2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY
#Paper F02
Beijing, CHINA, September 22-24, 2004
Thermal Response of a High Temperature Reactor during Passive Cooldown under Pressurized and Depressurized Conditions
Hamidul Haque, Wolfgang Feltes and Gerd Brinkmann Framatome ANP, GmbH Freyesleben Strasse 1, 91058 Erlangen, Germany Tel.: +49-9131-1897529, Fax: +49-9131-189-137529 Email: [email protected]
Abstract: HTR design is characterized by its inherent safety features with respect to passive decay heat removal through conduction, radiation and natural convection. This passive concept was first introduced in the German HTR-Module (pebble fuel) design and subsequently extended to other modular HTR design in recent years e.g. MHTGR (prismatic fuel), PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). In the past numerical simulations of the HTR-Module have been performed using the thermalhydraulic code THERMIX which was developed for the analysis of HTRs with pebble fuels and verified by experiments. Following modifications in the code it has been applied recently for the calculation of GT-HTR with prismatic fuels. These calculations have been checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. Based on the promising results of the benchmark calculation, the code has now been extended to the thermal analysis of the V/HTR which is conceived primarily for the production of hydrogen requiring very high temperatures (> 850 °C) for the electrolysis or thermo-chemical processes. In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e. g. 490/850, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters i.e. decay heat, graphite conductivity, surface emissivity etc on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 °C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures etc.
Key Words: VHTR, THERMIX, CRP-3, Prismatic fuel, Thermal-hydraulic, HTR Safety, gascooled, modular, nuclear reactor, passive, inherent, pressurized and depressurized cooldown
1
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
1
#F02
Introduction
HTR design is characterized by its inherent safety features with respect to passive decay heat removal through conduction, radiation and natural convection. This passive concept was first introduced in the German HTR-Module (pebble fuel) design and subsequently extended to other modular HTR design in recent years e.g. MHTGR (prismatic fuel), PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). In the past numerical simulations of the HTR-Module have been performed using the thermalhydraulic code THERMIX which was developed for the analysis of HTRs with pebble fuels and verified by experiments. Following modifications in the code it has been applied recently for the calculation of GT-HTR with prismatic fuels. These calculations have been checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. Based on the promising results of the benchmark calculation, the code has now been extended to the thermal analysis of the V/HTR which is conceived primarily for the production of hydrogen requiring very high temperatures (> 850 °C) for the electrolysis or thermo-chemical processes. In this paper the results of the following analyses have been presented: • • • •
2
Simulation of the IAEA CRP-3 benchmark problem to compare the results with those performed by various codes of other countries (China, France, Netherlands, Russia and USA) who participated in the CRP-3 program. Thermal hydraulic calculations of the VHTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions. Investigations to determine the influence of various inlet/outlet operating temperatures (e. g. 490/850, 490/1000, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature. Sensitivity analyses to evaluate the effect of varying the parameters i.e. decay heat, graphite conductivity, surface emissivity, reactor cavity cooling system temperature on the maximum fuel and pressure vessel temperature.
Short description of the reactor
The data of the reactor chosen for the investigations are taken from the CRP-3 benchmark problem for the GT-MHR Plutonium Burner /Ref. 1/. The reference plant is a 600 MWt (286MWe) heliumcooled, graphite-moderated reactor. The major design parameters are listed below. The core of this reactor consists of 1020 hexagonal prismatic fuel elements, stacked in an annular array of 102 columns of ten elements each (Fig. 1-2). The fuel elements contain cylindrical compacts formed from TRISO-coated fuel particles mixed with graphite material. The core is enclosed in a steel pressure vessel, which is connected to an adjacent pressure vessel containing the Power Conversion System (PCS) by a short cross-vessel. During normal operation helium coolant exits the reactor core at 850 °C and 7.01 MPa, flows through the center hot duct within the cross vessel, and is expanded through the turbine in the power conversion vessel. The turbine directly drives the electric generator and the high and low pressure compressors. Helium exits the turbine at 510 °C and 2.64 MPa and flows through the highly effective recuperator to return as much energy as possible to the cycle, and then through the precooler to reject heat to the ultimate heat sink. Relatively cold helium at 26 °C passes through the recuperator. Helium
2
HTR 2004
Beijing, CHINA, 2004.9
at 490 °C and 7.07 MPa flows from the recuperator exit, through the outer annulus within the cross vessel, and back to the core inlet and downward through the core to complete the loop. During post shutdown cooling a Reactor Cavity Cooling System (RCCS) is used to passively remove the core decay heat. This heat is transferred from the core to the RCCS, which surrounds the reactor vessel, by natural convection, conduction and radiation inside the vessel and by radiation and natural convection outside of the vessel. It is then removed by the RCCS via a cooling water system.
Fig. 1: Axial section of the reactor
Fig. 2: Radial section of the reactor
Major design parameters of the reference plant: Core thermal power Average power density Helium pressure at pressure vessel inlet Helium flow rate through the core Helium temperature at reactor inlet/outlet Core inner/outer average diameter Core height Vessel outer diameter (m) Average temperature of RCCS surface Coolant channel diameter Fuel compact diameter Gap between blocks
600 MW 6.5 MW/m3 7.07 MPa 320 kg/s 490/850 °C 2.96/4.84 m 8m 7.7 m 65 °C 16 mm 12.5 mm 2 mm
3
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
3
#F02
Code and numerical model
Computation Code For the analyses the thermal hydraulic code THERMIX was applied. This code was originally developed at KFA (now FZJ) Jülich and subsequently used for HTR-Module analysis in Siemens following various modifications in the program. The code has been verified through various experiments conducted at KFA and benchmark calculations. The code has been successfully applied for thermal hydraulic analyses of the PBMR (spherical fuel) and GT-MHR (block fuel) with annular core configuration. THERMIX is a transient two-dimensional code for thermal hydraulic analysis of high temperature gas cooled pebble bed nuclear reactors. The basic code structure consists of a general purpose steady-state or transient heat conduction module (THERMIX) and a quasi steady-state convection module (KONVEK). Both modules of the code use the finite-difference method with the successive pointwise over-relaxation solution technique. The heat conduction module solves the time dependent general heat conduction equation with temperature-dependent material properties. The convection module called by the conduction module solves the steady-state continuity, momentum and energy equations for the core and the adjacent flow regions within the core cavity for a given time-dependent temperature profile of the solid structures. It iterates on the pressure, flow and gas temperature fields until the desired convergence is achieved. Based upon the latest values of the gas temperatures, the code calculates the heat sources or sinks which are then transmitted to the conduction module for subsequent calculation of solid temperatures. This procedure is repeated for all the time steps. Numerical Model The calculation models for the THERMIX/KONVEK modules use two-dimensional rotation symmetric models in (r, z) –geometry. The THERMIX model covers the whole reactor in its cavity including the reactor cavity cooling system. It consists of 33 radial, 46 axial mesh points and 38 compositions which are used to describe core, reflector, ceramic and metallic components. The KONVEK model describes only the regions of the reactor with fluid flow. It consists of 27 radial, 33 axial mesh points and 16 compositions.
4
Input data and assumptions
The physical properties of the materials used for the calculation are: Helium: Density, specific heat and thermal conductivity according to German nuclear standard KTA 3102.1 Air:Density, specific heat, thermal conductivity and dynamic viscosity Graphite, Reactor pressure vessel, Core barrel and shell
4
HTR 2004
Beijing, CHINA, 2004.9
Graphite: Density = 1740 kg/m3 , Specific heat = 1840 J/kg/K , Emissivity = 0.9 , Thermal conductivity = f (T) using different functions for reflector graphite type H451 and fuel blocks. Reactor pressure vessel: Steel 10Cr9MoVNb Density = 7800 kg/m3, Specific heat = 520 J/kg/K, Thermal conductivity = 33 W/m/K, Emissivity = 0.8 Core barrel and shell: Steel 03Cr21Ni32Mo3Nb Density = 7800 kg/m3, Specific heat = 520 J/kg/K, Thermal conductivity = 24 W/m/K, Emissivity = 0.8 Power density distribution Power density distribution at normal operation and during accident conditions were generated for the THERMIX mesh points using the axial power profiles of the three active core layers and the decay heat evolution as shown in Fig. 3. 6
Deacy heat (% Power)
5
4
3
2
1
0 0
10
20
30
40
50
60
70
80
90
100
Time (hour)
Fig. 3: Decay heat curve used for the calculation
Boundary conditions and assumptions The upper and lower ends of the model are assumed to be adiabatic. The outer radial boundary which represents the reactor cavity cooling system shell is prescribed with a constant average temperature of 65 °C. Depressurized conduction cooldown: This case is characterized by the reactor scram from normal operation (100% power) following loss of forced cooling with rapid depressurization of the primary system (assumption for the calculation: p=1 bar immediately after scram). In this case the heat is transferred from the core to the RCCS, which surrounds the reactor vessel, by conduction and radiation inside the vessel and by radiation and natural convection outside of the vessel.
5
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
#F02
Pressurized cooldown: This case is characterized by the scram of the reactor from normal operation following loss of forced cooling but the primary system remains at full pressure (assumption for the calculation: p=55 bar immediately after scram). In this case the heat is transferred from the core to the RCCS, which surrounds the reactor vessel, by natural convection, conduction and radiation inside the vessel and by radiation and natural convection outside of the vessel.
Analyses and Results
5 5.1
CRP-3 benchmark problem (GT-MHR Pu-Burner)
Depressurized cooldown The maximum fuel and reactor pressure vessel temperatures obtained from this analysis for the depressurized condition are compared with those of the analyses performed by various other codes of the countries who participated in the CRP-3 program (Figures 4-5). The differences in the peak temperatures indicated in these Figures are probably attributable to various computation methods, modeling, material properties used by the various codes. In general the THERMIX code produced satisfactory results.
1700
1600
1500 Germany (FANP GmbH)
Temperature (°C)
1400
USA 1300
Russia France (CEA)
1200
France (FANP SAS) 1100
China Netherlands
1000
900
800 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.4: CRP-3 benchmark, Comparison of the code results of various countries for the peak fuel temperature during depressurized cooldown
6
HTR 2004
Beijing, CHINA, 2004.9
650
600
550 Germany (FANP GmbH) Temperature (° C)
500
USA Russia
450
France (FANP SAS) 400
France (CEA) China
350
Netherlands 300
250
200 0
10
20
30
40
50
60
70
80
Time (hour)
90
100
Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.5: CRP-3 benchmark, Comparison of the code results of various countries for the maximum reactor pressure vessel temperature during depressurized cooldown Pressurized cooldown The maximum fuel and reactor pressure vessel temperatures obtained from this analysis for pressurized condition are also compared with those of the analyses performed by various other codes of the countries who participated in the CRP-3 program (Figures 6-7). The differences in the peak temperatures indicated in these figures are probably attributable to various computation methods, modeling and material properties used by the various codes. In this case the results are not satisfactory. This leads to the necessity of defining a precise benchmark problem for pressurized cooldown.
1500
1400
1300 Germany (FANP GmbH) Temperature (° C)
1200 USA 1100 Russia 1000
China Netherlands
900
800
700
600 0
10
20
30
40
50 Time (hour)
60
70
80
90
100
Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.6: CRP-3 benchmark, Comparison of the code results of various countries for the peak fuel temperature during pressurized cooldown
7
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
#F02
600
550
500
Temperature (° C)
Germany (FANP GmbH) 450
USA Russia
400
China 350 Netherlands 300
250
200 0
10
20
30
40
50 Time (hour)
60
70
80
90
100
Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.7: CRP-3 benchmark, Comparison of the code results of various countries for the maximum reactor pressure vessel temperature during pressurized cooldown
5.2
VHTR
Within the framework of Gen IV (Generation IV) concept renewed interests have been focused on the VHTR (Very High Temperature Reactor) which is conceived primarily for the production of hydrogen requiring very high temperatures (> 850 °C) for the electrolysis or thermo-chemical processes. Possible layout may be a direct cycle plant with helium turbine or an indirect cycle plant using intermediate helium heat exchanger as an interface between primary and secondary systems. In the current design concept the inlet and outlet helium temperatures (operating temperatures) of this reactor with a thermal power of 600 MW are 490 °C and 1000 °C respectively. The other important data are identical to those of GT-MHR as shown in section 2. For this reactor concept the passive cooldown (natural convection, conduction and radiation) of the reactor under the postulated pressurized and depressurized accident conditions has been analyzed.
5.2.1
Pressurized Cooldown
This case is characterized by the scram of the reactor from normal operation following loss of forced cooling but the primary system remains at full pressure (assumption for the calculation: p=55 bar immediately after scram). Furthermore, the RCCS remains in operation and the RCCS cooling tubes have an average temperature of 65 °C. At high system pressure the gas density is high and as such high buoyancy effect leads to natural circulation of helium in the core as well as in the top and bottom reflectors. As a result, the hot gas from the bottom region of the core rises upwards in the inner region of the fuel blocks, transfers energy to the upper regions of the core and to the top reflectors and then flows downwards along the core boundary transferring energy further to the side structures and outer regions of the core. Fig. 8 shows the core internal natural convection induced velocity vectors in the core at times t=48 h and
8
HTR 2004
Beijing, CHINA, 2004.9
100 hours following the accident. The velocity vectors at t=0 hour indicate the conditions at normal operation.
t= 0 hour
t=48 hours
t=100 hours
Fig. 8: VHTR, Velocity vectors in the core for pressurized cooldown (attn.:different scales)
As a consequence of natural circulation strong temperature redistribution in the core and its adjacent structures takes place whereby relative cold regions, e.g. top reflector, heat up and the lower hot region of the core, cool down. In addition to the natural convection, heat transport mechanism also takes place by conduction and radiation to the adjacent structures. In the helium and air gaps the heat transfer takes place additionally by natural convection. Accordingly the heat is transferred from the core to the adjacent structures and the pressure vessel and from there finally to the reactor cavity cooling system. Fig. 9 shows the temperature distribution in the reactor at times 0 h, 48 h and 100 hours. Fig. 10 shows the temperature transients in the central graphite, fuel, side reflector, core barrel and reactor pressure vessel of the VHTR at the position where the maximum temperature occurs following the accident. The core is heated up to a maximum temperature of 1214 °C in 48 hours and then slowly cools down to about 1120 °C in 100 hours. Due to natural convection in the core the maximum fuel temperature in this case remains well below that of the depressurized case (sec. 5.2.2). The maximum pressure vessel temperature remains below 400 °C. Fig. 11 shows the radial temperature distributions in the VHTR at the axial position where the maximum temperature occurs following the accident.
9
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
t= 0 hour
t=48 hours
#F02
t=100 hours
Fig. 9: VHTR, Temperature distribution for pressurized cooldown
1400
1200
Temperature (°C)
1000
Central graphite Fuel Side reflector Core barrel Reactor pressure vessel
800
600
400
200
0 0
10
20
30
40
50
60
Time (hour)
70
80
90
100
FANP GmbH/Dr. Haque/HTR2004
Fig. 10: VHTR, Temperature transients in the central graphite, fuel, side reflector, core barrel and reactor pressure vessel for pressurized cooldown
10
HTR 2004
Beijing, CHINA, 2004.9
1400
1200
Temperature (°C)
1000
0h 10 h 48 h 74 h 100 h
800
600
400
200
0 0
100
200
300
Radius (cm)
400
500
FANP GmbH/Dr. Haque/HTR2004
Fig. 11: VHTR, Radial temperature distribution at the axial position where maximum temperature occurs for pressurized cooldown
5.2.2
Depressurized cooldown
This case is characterized by the reactor scram from normal operation following loss of forced cooling with rapid depressurization of the primary system (assumption for the calculation: p=1 bar immediately after scram). Furthermore, the RCCS remains in operation in this case and the RCCS cooling tubes have an average temperature of 65 °C. At low pressure (p= 1 bar) the heat transport by natural convection in the core is negligible due to low gas density and as such low buoyancy effect leads to negligible natural circulation in the core. The heat transport mechanism takes place primarily by conduction and radiation in the core and its adjacent structures. In the helium and air gaps the heat transfer takes place additionally by natural convection. Accordingly the heat is transferred from the core to the adjacent structures and the pressure vessel and from there finally to the reactor cavity cooling system. Fig. 12 shows the temperature distribution in the reactor at times t=0 h, 70 h and 100 hours following depressurization accident. Fig. 13 shows the temperature transients in the central graphite, fuel, side reflector, core shell and reactor pressure vessel of the VHTR at the position where the maximum temperature occurs following depressurization accident. The core is heated up in the beginning, attains its maximum temperature in 70 hours and then cools down slowly. The nominal peak fuel temperature amounts to 1587 °C which is below the limiting temperature of 1600 °C with respect to excessive release of fission products. The maximum reactor pressure vessel temperature amounts to 476 °C in this case. The radial distributions in the VHTR following depressurization accident are shown in the Fig. 14.
11
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
t= 0 hour
t=70 hours
t=100 hours
Fig. 12: VHTR, Temperature distribution in the reactor for depressurized cooldown
. 1800
1600
1400
Temperature (°C)
1200 Central graphite Fuel Side reflector Core barrel Reactor pressure vessel
1000
800
600
400
200
0 0
10
20
30
40
50
60
Time (hour)
70
80
90
100
FANP GmbH/Dr. Haque/HTR2004
Fig. 13: VHTR, Temperature transients in the central graphite, fuel, side reflector, core barrel and reactor pressure vessel for depressurized cooldown
12
#F02
HTR 2004
Beijing, CHINA, 2004.9
1800
1600
1400
Temperature (°C)
1200 0h 10 h 70 h 85 h 100h
1000
800
600
400
200
0 0
100
200
300
Radius (cm)
400
500
600
FANP GmbH/Dr. Haque/HTR2004
Fig. 14: VHTR, Radial temperature distribution at the axial position where the maximum temperature occurs for depressurized cooldown.
5.2.3
Influence of operating temperatures on the peak temperatures under accident conditions
The analyses to determine the influence of operating temperatures on the peak fuel and pressure vessel temperatures of VHTR were concentrated only on the depressurized cooldown case. In this section the results of the analyses are presented for the VHTR (Thermal power 600 MWth) with the following inlet/outlet temperatures: 490/850, 490/900, 490/950, 490/1000, 450/850, 400/850, 400/850, 350/850 and 350/1000 °C. For each of these cases calculations were first performed for steady state conditions which provide the starting temperatures for the final transient calculations for depressurized conduction cooldown. Fig. 15 shows the influence of inlet/outlet temperature on peak fuel temperature of VHTR for depressurized conduction cooldown. Fig. 16 shows the influence of inlet/outlet temperature on maximum reactor pressure vessel temperature of VHTR for depressurized conduction cooldown.
13
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
#F02
1600
Peak fuel temperature (° C)
1500
1400
Inlet / Outlet temp. (° C)
1300 490/850 490/900 490/950 490/1000 450/850 400/850 350/850 350/1000
1200
1100
1000
900 0
10
20
30
40
50
60
70
Time (hour)
Framatome ANP GmbH
80
90
100
Dr. Haque / NGPS7/HTR2004
Fig. 15: Influence of inlet/outlet temperature on the peak fuel temperature of VHTR for depressurized conduction cooldown.
500
450 Inlet/Outlet temp. (° C)
Temperature (°C)
400
350 490/850 490/900 490/950 490/1000 450/850 400/850 350/850 350/1000
300
250
200 0
10
Framatome ANP GmbH
20
30
40
50
60
Time (hour)
70
80
90
100
Dr. Haque / NGPS7/HTR2004
Fig. 16: Influence of inlet/outlet temperature on maximum reactor pressure vessel temperature of VHTR for depressurized conduction cooldown.
14
HTR 2004
Beijing, CHINA, 2004.9
For radiological analysis (Fission products release etc.) it is necessary to have information about the volume fractions of fuel elements above a certain temperature as a function of time. Fig. 17 provides this information for the VHTR with operating temperatures (inlet/outlet) 490/850 °C and 490/1000 °C respectively. Less than 5% of the fuel elements have temperatures above 1550°C for the case 490/1000°C. 60 >1200 °C VHTR with Inlet/Outlet temperature (°C)
50
490/850 490/1000 490/850 490/1000 490/850 490/1000 490/850 490/1000 490/850 490/1000
Volume fraction (%)
40 >1300 °C
30 >1400 °C
20
10 >1500 °C >1550 °C 0 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR2004
Fig. 17: VHTR: Time dependent volume fractions of fuel elements above a certain temperature in the VHTR (inlet/outlet temperatures 490/850 °C and 490/1000 °C) for depressurized cooldown.
5.2.4
Sensitivity analysis
Sensitivity analyses have been carried out for the V/HTR (inlet/outlet temperature 490/850 °C) for the depressurized cooldown to evaluate the effect of the following input data variation on the fuel and reactor pressure vessel temperatures: • Increase of decay heat by 10 % • Decrease of metallic surface emissivity from 0.8 to 0.6 • Decrease of graphite thermal conductivity of core and reflector blocks by 25 % • Increase of average temperature of RCCS cooling tubes (boiling off mode) to 140 °C Figures 18 and 19 show the influence of variation of the above parameters on the evolution of maximum fuel and reactor pressure vessel temperatures. These influences are expected to be of the same order of magnitude for the VHTR plant with operating temperatures 490/1000 °C.
15
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
1800
1700
1600
Temperature (° C)
1500
1400
Ref. case Decay heat (+10%) Met. emissivity 0.6 Gr. conductivity (-25%) RCCS Temp. 140 °C
1300
1200
1100
1000
900
800 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR2004
Fig. 18: Sensitivity analysis for V/HTR (inlet/outlet temperature 490/850°C) Peak fuel temperature vs. Time for depressurized cooldown
550
500
Temperature (° C)
450
Ref. case Decay heat (+10%) Met. emissivity 0.6 Gr. conductivity (-25%) RCCS Temp. 140 °C
400
350
300
250 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR2004
Fig. 19: Sensitivity analysis for V/HTR (inlet/outlet temperature 490/850°C) Maximum pressure vessel temperature vs. Time for depressurized cooldown
16
#F02
HTR 2004
Beijing, CHINA, 2004.9
6 Conclusion The results of the analyses lead to the following conclusions: • • • •
For the CRP-3 benchmark problem the code THERMIX produced satisfactory results in comparison with those of the other codes of various countries who participated in the CRP-3 program. For the VHTR the results indicate that the nominal peak fuel temperatures remain below 1600 °C which is the limiting temperature relating to radioactivity release from the fuel. The results provide some fundamental information for the design optimization of VHTR with respect to its maximum thermal power, operating temperatures etc. The code THERMIX can be applied for the thermal-hydraulic calculations of HTRs with annular core configuration having prismatic fuels.
It may be pointed out that all the calculated temperatures shown above are nominal temperatures. An additional temperature increment must be added to these peak temperatures in order to consider the effects of unfavorable initial state (e. g. >100% power) prior to scram and uncertainties of input data (e. g. decay heat, power density distribution, core thermal conductivities, specific heats, metallic emissivities, computation model etc.). To determine these systematical and statistical effects an uncertainty analysis has to be carried out. On the basis of past experience (licensing procedure) an uncertainty effect of approximately 150 K is to be expected. Furthermore, the volume fraction (Fig. 17) of fuel above 1550 °C (nominal) are relatively high for VHTR with operating temperatures 490/1000 °C. If the results of the sensitivity analysis (sec. 5.2.3) or the proposed uncertainty analysis are taken into consideration, the volume fraction of fuel above 1550 °C would be even higher. These facts should be taken into consideration for the final design (i.e. maximum thermal power, operating temperatures) of VHTR so that the peak fuel temperature (nominal plus uncertainty) remains below 1600 °C.
References /1/
IAEA-TECDOC-1163: Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions, Chapter 3.4 GT-MHR Plutonium Burner
Author introduction Hamidul Haque, born in 1943, is now a senior project engineer of Framatome ANP GmbH/Germany, a subsidiary of AREVA/France and SIEMENS/Germany and currently involved in the VHTR and OL3/Finnland projects. He obtained his Masters and Ph.D degrees in Nuclear Engineering from the University of Birmingham, U.K. in 1968 and 1970 respectively. Since 1975 he is working in the Power Generation Division of SIEMENS/Germany and responsible for the thermalhydraulic and accident analyses of HTRs (HTR-Module, HTR-10 etc.) and is also involved in various PWR activities, i.e. Functional requirements on reactor protection system, Technical specifications, Safety report, Accident instrumentation etc. for various PWR plants.
17
#Paper F02
Beijing, CHINA, September 22-24, 2004
Thermal Response of a High Temperature Reactor during Passive Cooldown under Pressurized and Depressurized Conditions
Hamidul Haque, Wolfgang Feltes and Gerd Brinkmann Framatome ANP, GmbH Freyesleben Strasse 1, 91058 Erlangen, Germany Tel.: +49-9131-1897529, Fax: +49-9131-189-137529 Email: [email protected]
Abstract: HTR design is characterized by its inherent safety features with respect to passive decay heat removal through conduction, radiation and natural convection. This passive concept was first introduced in the German HTR-Module (pebble fuel) design and subsequently extended to other modular HTR design in recent years e.g. MHTGR (prismatic fuel), PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). In the past numerical simulations of the HTR-Module have been performed using the thermalhydraulic code THERMIX which was developed for the analysis of HTRs with pebble fuels and verified by experiments. Following modifications in the code it has been applied recently for the calculation of GT-HTR with prismatic fuels. These calculations have been checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. Based on the promising results of the benchmark calculation, the code has now been extended to the thermal analysis of the V/HTR which is conceived primarily for the production of hydrogen requiring very high temperatures (> 850 °C) for the electrolysis or thermo-chemical processes. In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e. g. 490/850, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters i.e. decay heat, graphite conductivity, surface emissivity etc on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 °C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures etc.
Key Words: VHTR, THERMIX, CRP-3, Prismatic fuel, Thermal-hydraulic, HTR Safety, gascooled, modular, nuclear reactor, passive, inherent, pressurized and depressurized cooldown
1
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
1
#F02
Introduction
HTR design is characterized by its inherent safety features with respect to passive decay heat removal through conduction, radiation and natural convection. This passive concept was first introduced in the German HTR-Module (pebble fuel) design and subsequently extended to other modular HTR design in recent years e.g. MHTGR (prismatic fuel), PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). In the past numerical simulations of the HTR-Module have been performed using the thermalhydraulic code THERMIX which was developed for the analysis of HTRs with pebble fuels and verified by experiments. Following modifications in the code it has been applied recently for the calculation of GT-HTR with prismatic fuels. These calculations have been checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. Based on the promising results of the benchmark calculation, the code has now been extended to the thermal analysis of the V/HTR which is conceived primarily for the production of hydrogen requiring very high temperatures (> 850 °C) for the electrolysis or thermo-chemical processes. In this paper the results of the following analyses have been presented: • • • •
2
Simulation of the IAEA CRP-3 benchmark problem to compare the results with those performed by various codes of other countries (China, France, Netherlands, Russia and USA) who participated in the CRP-3 program. Thermal hydraulic calculations of the VHTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions. Investigations to determine the influence of various inlet/outlet operating temperatures (e. g. 490/850, 490/1000, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature. Sensitivity analyses to evaluate the effect of varying the parameters i.e. decay heat, graphite conductivity, surface emissivity, reactor cavity cooling system temperature on the maximum fuel and pressure vessel temperature.
Short description of the reactor
The data of the reactor chosen for the investigations are taken from the CRP-3 benchmark problem for the GT-MHR Plutonium Burner /Ref. 1/. The reference plant is a 600 MWt (286MWe) heliumcooled, graphite-moderated reactor. The major design parameters are listed below. The core of this reactor consists of 1020 hexagonal prismatic fuel elements, stacked in an annular array of 102 columns of ten elements each (Fig. 1-2). The fuel elements contain cylindrical compacts formed from TRISO-coated fuel particles mixed with graphite material. The core is enclosed in a steel pressure vessel, which is connected to an adjacent pressure vessel containing the Power Conversion System (PCS) by a short cross-vessel. During normal operation helium coolant exits the reactor core at 850 °C and 7.01 MPa, flows through the center hot duct within the cross vessel, and is expanded through the turbine in the power conversion vessel. The turbine directly drives the electric generator and the high and low pressure compressors. Helium exits the turbine at 510 °C and 2.64 MPa and flows through the highly effective recuperator to return as much energy as possible to the cycle, and then through the precooler to reject heat to the ultimate heat sink. Relatively cold helium at 26 °C passes through the recuperator. Helium
2
HTR 2004
Beijing, CHINA, 2004.9
at 490 °C and 7.07 MPa flows from the recuperator exit, through the outer annulus within the cross vessel, and back to the core inlet and downward through the core to complete the loop. During post shutdown cooling a Reactor Cavity Cooling System (RCCS) is used to passively remove the core decay heat. This heat is transferred from the core to the RCCS, which surrounds the reactor vessel, by natural convection, conduction and radiation inside the vessel and by radiation and natural convection outside of the vessel. It is then removed by the RCCS via a cooling water system.
Fig. 1: Axial section of the reactor
Fig. 2: Radial section of the reactor
Major design parameters of the reference plant: Core thermal power Average power density Helium pressure at pressure vessel inlet Helium flow rate through the core Helium temperature at reactor inlet/outlet Core inner/outer average diameter Core height Vessel outer diameter (m) Average temperature of RCCS surface Coolant channel diameter Fuel compact diameter Gap between blocks
600 MW 6.5 MW/m3 7.07 MPa 320 kg/s 490/850 °C 2.96/4.84 m 8m 7.7 m 65 °C 16 mm 12.5 mm 2 mm
3
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
3
#F02
Code and numerical model
Computation Code For the analyses the thermal hydraulic code THERMIX was applied. This code was originally developed at KFA (now FZJ) Jülich and subsequently used for HTR-Module analysis in Siemens following various modifications in the program. The code has been verified through various experiments conducted at KFA and benchmark calculations. The code has been successfully applied for thermal hydraulic analyses of the PBMR (spherical fuel) and GT-MHR (block fuel) with annular core configuration. THERMIX is a transient two-dimensional code for thermal hydraulic analysis of high temperature gas cooled pebble bed nuclear reactors. The basic code structure consists of a general purpose steady-state or transient heat conduction module (THERMIX) and a quasi steady-state convection module (KONVEK). Both modules of the code use the finite-difference method with the successive pointwise over-relaxation solution technique. The heat conduction module solves the time dependent general heat conduction equation with temperature-dependent material properties. The convection module called by the conduction module solves the steady-state continuity, momentum and energy equations for the core and the adjacent flow regions within the core cavity for a given time-dependent temperature profile of the solid structures. It iterates on the pressure, flow and gas temperature fields until the desired convergence is achieved. Based upon the latest values of the gas temperatures, the code calculates the heat sources or sinks which are then transmitted to the conduction module for subsequent calculation of solid temperatures. This procedure is repeated for all the time steps. Numerical Model The calculation models for the THERMIX/KONVEK modules use two-dimensional rotation symmetric models in (r, z) –geometry. The THERMIX model covers the whole reactor in its cavity including the reactor cavity cooling system. It consists of 33 radial, 46 axial mesh points and 38 compositions which are used to describe core, reflector, ceramic and metallic components. The KONVEK model describes only the regions of the reactor with fluid flow. It consists of 27 radial, 33 axial mesh points and 16 compositions.
4
Input data and assumptions
The physical properties of the materials used for the calculation are: Helium: Density, specific heat and thermal conductivity according to German nuclear standard KTA 3102.1 Air:Density, specific heat, thermal conductivity and dynamic viscosity Graphite, Reactor pressure vessel, Core barrel and shell
4
HTR 2004
Beijing, CHINA, 2004.9
Graphite: Density = 1740 kg/m3 , Specific heat = 1840 J/kg/K , Emissivity = 0.9 , Thermal conductivity = f (T) using different functions for reflector graphite type H451 and fuel blocks. Reactor pressure vessel: Steel 10Cr9MoVNb Density = 7800 kg/m3, Specific heat = 520 J/kg/K, Thermal conductivity = 33 W/m/K, Emissivity = 0.8 Core barrel and shell: Steel 03Cr21Ni32Mo3Nb Density = 7800 kg/m3, Specific heat = 520 J/kg/K, Thermal conductivity = 24 W/m/K, Emissivity = 0.8 Power density distribution Power density distribution at normal operation and during accident conditions were generated for the THERMIX mesh points using the axial power profiles of the three active core layers and the decay heat evolution as shown in Fig. 3. 6
Deacy heat (% Power)
5
4
3
2
1
0 0
10
20
30
40
50
60
70
80
90
100
Time (hour)
Fig. 3: Decay heat curve used for the calculation
Boundary conditions and assumptions The upper and lower ends of the model are assumed to be adiabatic. The outer radial boundary which represents the reactor cavity cooling system shell is prescribed with a constant average temperature of 65 °C. Depressurized conduction cooldown: This case is characterized by the reactor scram from normal operation (100% power) following loss of forced cooling with rapid depressurization of the primary system (assumption for the calculation: p=1 bar immediately after scram). In this case the heat is transferred from the core to the RCCS, which surrounds the reactor vessel, by conduction and radiation inside the vessel and by radiation and natural convection outside of the vessel.
5
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
#F02
Pressurized cooldown: This case is characterized by the scram of the reactor from normal operation following loss of forced cooling but the primary system remains at full pressure (assumption for the calculation: p=55 bar immediately after scram). In this case the heat is transferred from the core to the RCCS, which surrounds the reactor vessel, by natural convection, conduction and radiation inside the vessel and by radiation and natural convection outside of the vessel.
Analyses and Results
5 5.1
CRP-3 benchmark problem (GT-MHR Pu-Burner)
Depressurized cooldown The maximum fuel and reactor pressure vessel temperatures obtained from this analysis for the depressurized condition are compared with those of the analyses performed by various other codes of the countries who participated in the CRP-3 program (Figures 4-5). The differences in the peak temperatures indicated in these Figures are probably attributable to various computation methods, modeling, material properties used by the various codes. In general the THERMIX code produced satisfactory results.
1700
1600
1500 Germany (FANP GmbH)
Temperature (°C)
1400
USA 1300
Russia France (CEA)
1200
France (FANP SAS) 1100
China Netherlands
1000
900
800 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.4: CRP-3 benchmark, Comparison of the code results of various countries for the peak fuel temperature during depressurized cooldown
6
HTR 2004
Beijing, CHINA, 2004.9
650
600
550 Germany (FANP GmbH) Temperature (° C)
500
USA Russia
450
France (FANP SAS) 400
France (CEA) China
350
Netherlands 300
250
200 0
10
20
30
40
50
60
70
80
Time (hour)
90
100
Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.5: CRP-3 benchmark, Comparison of the code results of various countries for the maximum reactor pressure vessel temperature during depressurized cooldown Pressurized cooldown The maximum fuel and reactor pressure vessel temperatures obtained from this analysis for pressurized condition are also compared with those of the analyses performed by various other codes of the countries who participated in the CRP-3 program (Figures 6-7). The differences in the peak temperatures indicated in these figures are probably attributable to various computation methods, modeling and material properties used by the various codes. In this case the results are not satisfactory. This leads to the necessity of defining a precise benchmark problem for pressurized cooldown.
1500
1400
1300 Germany (FANP GmbH) Temperature (° C)
1200 USA 1100 Russia 1000
China Netherlands
900
800
700
600 0
10
20
30
40
50 Time (hour)
60
70
80
90
100
Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.6: CRP-3 benchmark, Comparison of the code results of various countries for the peak fuel temperature during pressurized cooldown
7
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
#F02
600
550
500
Temperature (° C)
Germany (FANP GmbH) 450
USA Russia
400
China 350 Netherlands 300
250
200 0
10
20
30
40
50 Time (hour)
60
70
80
90
100
Framatome ANP GmbH / Dr. Haque / NGPS7/HTR/CRP3
Fig.7: CRP-3 benchmark, Comparison of the code results of various countries for the maximum reactor pressure vessel temperature during pressurized cooldown
5.2
VHTR
Within the framework of Gen IV (Generation IV) concept renewed interests have been focused on the VHTR (Very High Temperature Reactor) which is conceived primarily for the production of hydrogen requiring very high temperatures (> 850 °C) for the electrolysis or thermo-chemical processes. Possible layout may be a direct cycle plant with helium turbine or an indirect cycle plant using intermediate helium heat exchanger as an interface between primary and secondary systems. In the current design concept the inlet and outlet helium temperatures (operating temperatures) of this reactor with a thermal power of 600 MW are 490 °C and 1000 °C respectively. The other important data are identical to those of GT-MHR as shown in section 2. For this reactor concept the passive cooldown (natural convection, conduction and radiation) of the reactor under the postulated pressurized and depressurized accident conditions has been analyzed.
5.2.1
Pressurized Cooldown
This case is characterized by the scram of the reactor from normal operation following loss of forced cooling but the primary system remains at full pressure (assumption for the calculation: p=55 bar immediately after scram). Furthermore, the RCCS remains in operation and the RCCS cooling tubes have an average temperature of 65 °C. At high system pressure the gas density is high and as such high buoyancy effect leads to natural circulation of helium in the core as well as in the top and bottom reflectors. As a result, the hot gas from the bottom region of the core rises upwards in the inner region of the fuel blocks, transfers energy to the upper regions of the core and to the top reflectors and then flows downwards along the core boundary transferring energy further to the side structures and outer regions of the core. Fig. 8 shows the core internal natural convection induced velocity vectors in the core at times t=48 h and
8
HTR 2004
Beijing, CHINA, 2004.9
100 hours following the accident. The velocity vectors at t=0 hour indicate the conditions at normal operation.
t= 0 hour
t=48 hours
t=100 hours
Fig. 8: VHTR, Velocity vectors in the core for pressurized cooldown (attn.:different scales)
As a consequence of natural circulation strong temperature redistribution in the core and its adjacent structures takes place whereby relative cold regions, e.g. top reflector, heat up and the lower hot region of the core, cool down. In addition to the natural convection, heat transport mechanism also takes place by conduction and radiation to the adjacent structures. In the helium and air gaps the heat transfer takes place additionally by natural convection. Accordingly the heat is transferred from the core to the adjacent structures and the pressure vessel and from there finally to the reactor cavity cooling system. Fig. 9 shows the temperature distribution in the reactor at times 0 h, 48 h and 100 hours. Fig. 10 shows the temperature transients in the central graphite, fuel, side reflector, core barrel and reactor pressure vessel of the VHTR at the position where the maximum temperature occurs following the accident. The core is heated up to a maximum temperature of 1214 °C in 48 hours and then slowly cools down to about 1120 °C in 100 hours. Due to natural convection in the core the maximum fuel temperature in this case remains well below that of the depressurized case (sec. 5.2.2). The maximum pressure vessel temperature remains below 400 °C. Fig. 11 shows the radial temperature distributions in the VHTR at the axial position where the maximum temperature occurs following the accident.
9
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
t= 0 hour
t=48 hours
#F02
t=100 hours
Fig. 9: VHTR, Temperature distribution for pressurized cooldown
1400
1200
Temperature (°C)
1000
Central graphite Fuel Side reflector Core barrel Reactor pressure vessel
800
600
400
200
0 0
10
20
30
40
50
60
Time (hour)
70
80
90
100
FANP GmbH/Dr. Haque/HTR2004
Fig. 10: VHTR, Temperature transients in the central graphite, fuel, side reflector, core barrel and reactor pressure vessel for pressurized cooldown
10
HTR 2004
Beijing, CHINA, 2004.9
1400
1200
Temperature (°C)
1000
0h 10 h 48 h 74 h 100 h
800
600
400
200
0 0
100
200
300
Radius (cm)
400
500
FANP GmbH/Dr. Haque/HTR2004
Fig. 11: VHTR, Radial temperature distribution at the axial position where maximum temperature occurs for pressurized cooldown
5.2.2
Depressurized cooldown
This case is characterized by the reactor scram from normal operation following loss of forced cooling with rapid depressurization of the primary system (assumption for the calculation: p=1 bar immediately after scram). Furthermore, the RCCS remains in operation in this case and the RCCS cooling tubes have an average temperature of 65 °C. At low pressure (p= 1 bar) the heat transport by natural convection in the core is negligible due to low gas density and as such low buoyancy effect leads to negligible natural circulation in the core. The heat transport mechanism takes place primarily by conduction and radiation in the core and its adjacent structures. In the helium and air gaps the heat transfer takes place additionally by natural convection. Accordingly the heat is transferred from the core to the adjacent structures and the pressure vessel and from there finally to the reactor cavity cooling system. Fig. 12 shows the temperature distribution in the reactor at times t=0 h, 70 h and 100 hours following depressurization accident. Fig. 13 shows the temperature transients in the central graphite, fuel, side reflector, core shell and reactor pressure vessel of the VHTR at the position where the maximum temperature occurs following depressurization accident. The core is heated up in the beginning, attains its maximum temperature in 70 hours and then cools down slowly. The nominal peak fuel temperature amounts to 1587 °C which is below the limiting temperature of 1600 °C with respect to excessive release of fission products. The maximum reactor pressure vessel temperature amounts to 476 °C in this case. The radial distributions in the VHTR following depressurization accident are shown in the Fig. 14.
11
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
t= 0 hour
t=70 hours
t=100 hours
Fig. 12: VHTR, Temperature distribution in the reactor for depressurized cooldown
. 1800
1600
1400
Temperature (°C)
1200 Central graphite Fuel Side reflector Core barrel Reactor pressure vessel
1000
800
600
400
200
0 0
10
20
30
40
50
60
Time (hour)
70
80
90
100
FANP GmbH/Dr. Haque/HTR2004
Fig. 13: VHTR, Temperature transients in the central graphite, fuel, side reflector, core barrel and reactor pressure vessel for depressurized cooldown
12
#F02
HTR 2004
Beijing, CHINA, 2004.9
1800
1600
1400
Temperature (°C)
1200 0h 10 h 70 h 85 h 100h
1000
800
600
400
200
0 0
100
200
300
Radius (cm)
400
500
600
FANP GmbH/Dr. Haque/HTR2004
Fig. 14: VHTR, Radial temperature distribution at the axial position where the maximum temperature occurs for depressurized cooldown.
5.2.3
Influence of operating temperatures on the peak temperatures under accident conditions
The analyses to determine the influence of operating temperatures on the peak fuel and pressure vessel temperatures of VHTR were concentrated only on the depressurized cooldown case. In this section the results of the analyses are presented for the VHTR (Thermal power 600 MWth) with the following inlet/outlet temperatures: 490/850, 490/900, 490/950, 490/1000, 450/850, 400/850, 400/850, 350/850 and 350/1000 °C. For each of these cases calculations were first performed for steady state conditions which provide the starting temperatures for the final transient calculations for depressurized conduction cooldown. Fig. 15 shows the influence of inlet/outlet temperature on peak fuel temperature of VHTR for depressurized conduction cooldown. Fig. 16 shows the influence of inlet/outlet temperature on maximum reactor pressure vessel temperature of VHTR for depressurized conduction cooldown.
13
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
#F02
1600
Peak fuel temperature (° C)
1500
1400
Inlet / Outlet temp. (° C)
1300 490/850 490/900 490/950 490/1000 450/850 400/850 350/850 350/1000
1200
1100
1000
900 0
10
20
30
40
50
60
70
Time (hour)
Framatome ANP GmbH
80
90
100
Dr. Haque / NGPS7/HTR2004
Fig. 15: Influence of inlet/outlet temperature on the peak fuel temperature of VHTR for depressurized conduction cooldown.
500
450 Inlet/Outlet temp. (° C)
Temperature (°C)
400
350 490/850 490/900 490/950 490/1000 450/850 400/850 350/850 350/1000
300
250
200 0
10
Framatome ANP GmbH
20
30
40
50
60
Time (hour)
70
80
90
100
Dr. Haque / NGPS7/HTR2004
Fig. 16: Influence of inlet/outlet temperature on maximum reactor pressure vessel temperature of VHTR for depressurized conduction cooldown.
14
HTR 2004
Beijing, CHINA, 2004.9
For radiological analysis (Fission products release etc.) it is necessary to have information about the volume fractions of fuel elements above a certain temperature as a function of time. Fig. 17 provides this information for the VHTR with operating temperatures (inlet/outlet) 490/850 °C and 490/1000 °C respectively. Less than 5% of the fuel elements have temperatures above 1550°C for the case 490/1000°C. 60 >1200 °C VHTR with Inlet/Outlet temperature (°C)
50
490/850 490/1000 490/850 490/1000 490/850 490/1000 490/850 490/1000 490/850 490/1000
Volume fraction (%)
40 >1300 °C
30 >1400 °C
20
10 >1500 °C >1550 °C 0 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR2004
Fig. 17: VHTR: Time dependent volume fractions of fuel elements above a certain temperature in the VHTR (inlet/outlet temperatures 490/850 °C and 490/1000 °C) for depressurized cooldown.
5.2.4
Sensitivity analysis
Sensitivity analyses have been carried out for the V/HTR (inlet/outlet temperature 490/850 °C) for the depressurized cooldown to evaluate the effect of the following input data variation on the fuel and reactor pressure vessel temperatures: • Increase of decay heat by 10 % • Decrease of metallic surface emissivity from 0.8 to 0.6 • Decrease of graphite thermal conductivity of core and reflector blocks by 25 % • Increase of average temperature of RCCS cooling tubes (boiling off mode) to 140 °C Figures 18 and 19 show the influence of variation of the above parameters on the evolution of maximum fuel and reactor pressure vessel temperatures. These influences are expected to be of the same order of magnitude for the VHTR plant with operating temperatures 490/1000 °C.
15
THERMAL RESPONSE OF A HIGH TEMPERATURE REACTOR DURING PASSIVE COOLDOWN UNDER PRESSURIZED AND DEPRESSURIZED CONDITIONS
1800
1700
1600
Temperature (° C)
1500
1400
Ref. case Decay heat (+10%) Met. emissivity 0.6 Gr. conductivity (-25%) RCCS Temp. 140 °C
1300
1200
1100
1000
900
800 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR2004
Fig. 18: Sensitivity analysis for V/HTR (inlet/outlet temperature 490/850°C) Peak fuel temperature vs. Time for depressurized cooldown
550
500
Temperature (° C)
450
Ref. case Decay heat (+10%) Met. emissivity 0.6 Gr. conductivity (-25%) RCCS Temp. 140 °C
400
350
300
250 0
10
20
30
40
50
60
70
80
90
100
Time (hour) Framatome ANP GmbH / Dr. Haque / NGPS7/HTR2004
Fig. 19: Sensitivity analysis for V/HTR (inlet/outlet temperature 490/850°C) Maximum pressure vessel temperature vs. Time for depressurized cooldown
16
#F02
HTR 2004
Beijing, CHINA, 2004.9
6 Conclusion The results of the analyses lead to the following conclusions: • • • •
For the CRP-3 benchmark problem the code THERMIX produced satisfactory results in comparison with those of the other codes of various countries who participated in the CRP-3 program. For the VHTR the results indicate that the nominal peak fuel temperatures remain below 1600 °C which is the limiting temperature relating to radioactivity release from the fuel. The results provide some fundamental information for the design optimization of VHTR with respect to its maximum thermal power, operating temperatures etc. The code THERMIX can be applied for the thermal-hydraulic calculations of HTRs with annular core configuration having prismatic fuels.
It may be pointed out that all the calculated temperatures shown above are nominal temperatures. An additional temperature increment must be added to these peak temperatures in order to consider the effects of unfavorable initial state (e. g. >100% power) prior to scram and uncertainties of input data (e. g. decay heat, power density distribution, core thermal conductivities, specific heats, metallic emissivities, computation model etc.). To determine these systematical and statistical effects an uncertainty analysis has to be carried out. On the basis of past experience (licensing procedure) an uncertainty effect of approximately 150 K is to be expected. Furthermore, the volume fraction (Fig. 17) of fuel above 1550 °C (nominal) are relatively high for VHTR with operating temperatures 490/1000 °C. If the results of the sensitivity analysis (sec. 5.2.3) or the proposed uncertainty analysis are taken into consideration, the volume fraction of fuel above 1550 °C would be even higher. These facts should be taken into consideration for the final design (i.e. maximum thermal power, operating temperatures) of VHTR so that the peak fuel temperature (nominal plus uncertainty) remains below 1600 °C.
References /1/
IAEA-TECDOC-1163: Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions, Chapter 3.4 GT-MHR Plutonium Burner
Author introduction Hamidul Haque, born in 1943, is now a senior project engineer of Framatome ANP GmbH/Germany, a subsidiary of AREVA/France and SIEMENS/Germany and currently involved in the VHTR and OL3/Finnland projects. He obtained his Masters and Ph.D degrees in Nuclear Engineering from the University of Birmingham, U.K. in 1968 and 1970 respectively. Since 1975 he is working in the Power Generation Division of SIEMENS/Germany and responsible for the thermalhydraulic and accident analyses of HTRs (HTR-Module, HTR-10 etc.) and is also involved in various PWR activities, i.e. Functional requirements on reactor protection system, Technical specifications, Safety report, Accident instrumentation etc. for various PWR plants.
17
