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Helium-cooled test blanket module box behaviour under accidental pressurisation
Fusion Engineering and Design 83 (2008) 1738–1741
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Helium-cooled test blanket module box behaviour under accidental pressurisation Christian Girard a,∗ , Nicolas Schmidt a , Jean-Franc¸ois Salavy b , Gilles Rampal b a b
CEA, Cadarache, 13108 St Paul lez Durance, France CEA, Saclay, 91191 Gif sur Yvette, France
a r t i c l e
i n f o
Article history: Available online 21 September 2008 Keywords: Fusion Safety Test blanket ITER
a b s t r a c t The helium-cooled lithium–lead (HCLL) breeder blanket concept is one of the two breeder blanket lines presently developed by the EU for DEMO reactor. In the short-term so-called DEMO relevant test blanket modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. For the purpose of licensing such test module in the ITER facility, a safety assessment of the different possible accidental conditions has to be performed. This paper presents the results of the thermo-mechanical calculations that have been obtained in case of internal module leak occurring between the 8.0 MPa pressurised helium circuit and the lithium–lead circuit. In a conservative manner, it is also assumed that helium cooling of the TBM box is stopped at the accident initiation. In such situation the TBM box will be entirely pressurised at 8.0 MPa and the different challenges are related to the thermo-mechanical behaviour, the plasma power hold up and the heat removal capacity of the module. An assessment to estimate the time when the TBM box can withstand the different loads induced by the situation is presented. Particularly the available time span to trigger an emergency plasma shutdown is estimated in regards of the maximal allowable stress for the EUROFER which is the test blanket module material. A 3D finite element model has been developed with CAST3M computer code taking into account the beryllium layer, the decay heat after shutdown and the thermal radiation phenomena. After a description of the model, the paper presents and explains the results and particularly the methodology followed to determine the maximal allowable stress location which must combine the thermal transient calculations and the pure mechanical calculation under the 8.0 MPa loading. The results show that the hottest spot of the first wall (FW) was the most challenged location and that the structure can withstand such accidental conditions without plasma shutdown up to 15 s. Afterwards, the structural design criteria defined by ITER for in-vessel components (SDC-IC) is not any more fulfilled. This does not necessary mean that a lithium–lead leakage can occur in the vacuum vessel. More research and development are needed to have a clear understanding of crack propagation and break size in a material like EUROFER where few data are available, particularly at the temperature reached in this accident. Finally as a conclusion of the studies presented in this paper, it can be stated that, although very conservative assumptions were taken, the time span is large enough to trigger a plasma shutdown in order to avoid more severe operational and safety consequences of this accidental situation. © 2008 Elsevier B.V. All rights reserved.
1. Introduction The helium-cooled lithium–lead (HCLL) breeder blanket concept is one of the two breeder blanket lines presently developed by the EU for DEMO reactor. In the short-term so-called DEMO relevant test blanket modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed,
∗ Corresponding author. Tel.: +33 442254956. E-mail address: [email protected] (C. Girard). 0920-3796/$ – see front matter © 2008 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2008.07.042
commissioned and operated in ITER for first tests in a fusion environment. For the purpose of licensing such test module in the ITER facility, a safety assessment of the different possible accidental conditions has to be performed. This paper presents the results of the thermo-mechanical calculations obtained in case of internal module leak occurring between the 8.0 MPa pressurised helium circuit and the lithium–lead circuit (in-TBM LOCA: loss of coolant accident inside the TBM). In a conservative manner, it is also assumed that the helium cooling of the TBM box is stopped at the accident initiation.
C. Girard et al. / Fusion Engineering and Design 83 (2008) 1738–1741
In such situation the TBM box will be entirely pressurised at 8.0 MPa and the different challenges are related to the thermomechanical behaviour, the plasma power hold up and the heat removal capacity of the module. The time span when the TBM box can withstand the different loads induced by the situation without an emergency plasma shutdown is estimated in regards of the maximal allowable stress for the EUROFER which is the test blanket module material. 2. Process followed The goal of the study is to determine the HCLL-TBM temperature field during the accidental transient and to compute the mechanical stress caused by the helium pressurisation of the HCLL-TBM box. To this end, a finite element 3D model was developed using the CAST3M code [1]. This model allows the thermal and the stress fields to be computed within the HCLL-TBM during the transient. The HCLL-TBM mechanical integrity is assessed in three steps: • The first stage is the achievement of thermal computation on the model. This calculation refers to permanent conditions as well as the accidental transient simulation and the plasma shutdown effect. • The second stage is the mechanical calculation under accidental pressure conditions. • The last stage consists in verifying the HCLL-TBM box behaviour according to the stress level and the temperature reached during the transient by determining the allowable stress intensity. The mechanical analysis is performed using the structural design criteria for in-vessel components (SDC-IC) defined by the ITER project [2–5]. 3. TBM modelling
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it describes the precise geometry of the junction between the first wall (FW) and the side wall (SW). For simplicity, the model does not account for the horizontal stiffening plates nor the back plates, which are replaced at this step by mechanical boundary conditions. The thermal models used are • Thermal conduction in all the metallic structures and in the lithium–lead coolant, • Thermal exchange between structures and helium by forced convection, • Thermal radiation between the beryllium layer and the vacuum vessel (thermal radiation between the side wall and the frame is not considered at this step). For the mechanical computations, an elastic model is used. The physical properties of helium and lithium–lead come from literature, the thermo-mechanical characteristics of the steel structures, EUROFER, are found in Ref. [6]. The thermo-mechanical characteristics of beryllium layer are from Ref. [7]. 3.2. Thermal and mechanical loadings The thermal loading is due to the neutrons and the ␥ flux (neutron wall loading) produced by the plasma, which induces heating of the structures and of the lithium–lead. In addition, a heat flux is applied on the external face of the FW. The nominal thermal loadings considered in this study are: • Power density distribution related to a neutron wall loading (NWL) of 0.78 MW m−2 , • Heat flux (HF) on the beryllium surface, plasma side, of 0.5 MW m−2 .
3.1. Meshing, thermal and mechanical models For geometrical simplification the model represents one and a half Breeder Unit in width (see Fig. 1). The model is innovative as
The mechanical loading is due to the internal pressure of 8.0 MPa. This pressure is applied to the helium channels and to the surfaces in contact with lithium–lead to simulate an in-TBM helium
Fig. 1. Exploded view of the 3D model.
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leak pressurisation. An equivalent stress representative of the end effect, is applied to the upper surface of the model. 4. Thermo-mechanical assessment of the HCLL TBM behaviour 4.1. Simulation assumptions The simulation establishes, first, the steady state thermal field under normal operating conditions, and then considers the loss of helium cooling thermal transient by assuming an instantaneous and total loss of heat exchange with helium. In addition, it is assumed that the accident remains undetected and the plasma continues to burn producing a surface heat flux of 0.5 MW m−2 and a nominal neutron wall loading of 0.78 MW m−2 . It is obvious that the TBM cannot withstand such a thermal loading if any plasma shutdown does not happen; the purpose of the calculation is to determine the time interval t0 available during which the TBM can withstand the thermal loading (due to the loss of helium cooling capacity) combined with the internal pressure loading (due to the in-TBM LOCA). 4.2. Determination of the time interval The time interval t0 during which the HCLL-TBM can comply with the ITER safety criteria after an in-TBM LOCA without plasma shutdown, is determined by using the following methodology: • A thermal transient computation is performed up to a FW temperature of 1300 ◦ C (assuming plasma heating and no cooling from helium), • A pure mechanical computation, simulating the 8.0 MPa pressurisation of the HCLL-TBM box is performed without accounting for the thermal stresses (since the LOCA is classified as a level D event, the analysis deals only with the primary stresses), • The analysis is then performed at the location of the maximal stress obtained in the mechanical computation and at the location of the maximal temperature obtained during the thermal transient. • The maximal mechanical stress is then related to a maximal allowed temperature (from the table of maximal allowable stress as a function of temperature). It is then investigated into the results from the thermal transient when, at the same location, this maximal temperature is reached, this gives the time t1 , • The analysis is also performed for the mechanical stress obtained at the maximal temperature location, this stress is related to a maximal allowable temperature and from the thermal transient to a time t2 when this temperature is reached. • The minimum of t1 and t2 gives the time t0 = min(t1 , t2 ) when the HCLL-TBM box will start not to fulfill the SDC-IC criteria (primary stress upper than the maximal allowable stress).
Fig. 2. Temperature field in the beryllium layer under nominal conditions.
The loss of helium cooling transient is then simulated for 100 s. The thermal field is calculated as a function of time. As soon as the helium flow is stopped, the FW temperature tends toward an homogeneous thermal field. The temperature exceeds 700 ◦ C, 15 s after the transient initiation then reaches some 1000 ◦ C at 50 s. At the end of simulation (100 s) the FW temperature is about 1300 ◦ C (Fig. 3). In the frame of this study, the consequences of a plasma shutdown with disruption and the potential to remove decay heat by thermal radiation were also assessed. The assumption is that the plasma shutdown creates a disruption with a peak heat flux (5.5 MW m−2 during 100 ms) adding an additional heat loading that must be taken into account in the analysis. As an example, a plasma shutdown has been simulated when the FW temperature reaches 1115 ◦ C (approximately melting temperature of the beryllium layer), i.e. at 65 s after the transient initiation. The thermal computation shows that the increment of temperature caused by the disruption is weak (only 8 ◦ C). After the plasma shutdown, the thermal radiation allows the temperature to be decreased rapidly (Fig. 4). For the mechanical assessment, since the LOCA is classified as a level D event, the mechanical analysis deals only with the primary stresses. Consequently, the stress field does not depend on the thermal field. The equivalent primary stresses (Von Misès) are displayed in the meshing (Fig. 5). The calculations lead to a maximal stress intensity of 324 MPa.
5. Thermal and mechanical assessment First, the permanent thermal field before the accidental transient is calculated (Fig. 2). At this stage, the temperature fields under nominal working conditions are as follows: • • • • •
361 ◦ C ≤ Tberyllium layer ≤ 572 ◦ C (see Fig. 2), 310 ◦ C ≤ Tfirst wall–side wall ≤ 563 ◦ C, 429 ◦ C ≤ Tstiffening plate vertical ≤ 486 ◦ C, 390 ◦ C ≤ Tcooling plate ≤ 509 ◦ C, 369 ◦ C ≤ Tlithium–lead ≤ 534 ◦ C.
Fig. 3. FW temperature evolution following a loss of helium cooling.
C. Girard et al. / Fusion Engineering and Design 83 (2008) 1738–1741
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pressurisation, is the location where structural damages will appear first. 6. Conclusion The thermo-mechanical study presented in this paper deals with the HCLL-TBM behaviour under an in-TBM LOCA event without active plasma shutdown leading to: • Total loss of helium cooling, • TBM box pressurisation at 8.0 MPa. The thermal simulations performed have shown that:
Fig. 4. Decay heat removal by thermal radiation after plasma shutdown.
• The FW temperature increases rapidly up to 1300 ◦ C after 100 s, not very far from the EUROFER melting temperature which is around 1500 ◦ C, • Plasma shutdown simulation indicated that the disruption effect is not significant causing an additional heating of only 8 ◦ C and that the thermal radiation phenomena after shutdown is able to remove the decay heat. The mechanical analysis has been particularly performed at two places: • Over-stress place at the junction FW–SW, • Hot-spot place on the FW.
Fig. 5. Equivalent primary stress under TBM pressurisation at 8.0 MPa.
As seen before (Section 4.2) two locations have been analysed: • The location of the maximal stress, • The location of the hottest spot where the temperature is maximal. At the location of the maximal stress, the analysis shows that the ITER safety criteria are fulfilled as long as the temperature stays below 634 ◦ C. From the thermal transient calculations, this time is reached in a time t1 equals to 70 s. At the location of the maximal temperature the stress analysis shows that the criteria are fulfilled when the temperature is lower than 681 ◦ C. The time to reach this temperature is derived from the thermal transient calculations and it is found that this time, t2 , equals 15 s. Since this time is lower than the time obtained at the maximal stress location, it is concluded that the hottest spot in case of a combined loss of cooling and in-TBM LOCA producing a 8.0 MPa
The results show that the hottest spot of the first wall was the most challenged location and that the structure can withstand such accidental conditions without plasma shutdown up to 15 s. Afterwards, the structural design criteria defined by ITER for invessel components is not any more fulfilled. This does not necessary mean that a lithium–lead leakage can occur in the vacuum vessel. More research and development are needed to have a clear understanding of crack propagation and break size in a material like EUROFER where few data are available, particularly at the temperature reached in this accident. Finally as a conclusion of the studies presented in this paper, it can be stated that, although very conservative assumptions were taken (particularly the instantaneous total loss of heat removal), the time span is large enough to trigger a plasma shutdown in order to avoid more severe operational and safety consequences of this accidental situation. References [1] Cast3M FE code, CEA/DEN/DM2S/SEMT, Industrial Version 2000 for Windows
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Helium-cooled test blanket module box behaviour under accidental pressurisation Christian Girard a,∗ , Nicolas Schmidt a , Jean-Franc¸ois Salavy b , Gilles Rampal b a b
CEA, Cadarache, 13108 St Paul lez Durance, France CEA, Saclay, 91191 Gif sur Yvette, France
a r t i c l e
i n f o
Article history: Available online 21 September 2008 Keywords: Fusion Safety Test blanket ITER
a b s t r a c t The helium-cooled lithium–lead (HCLL) breeder blanket concept is one of the two breeder blanket lines presently developed by the EU for DEMO reactor. In the short-term so-called DEMO relevant test blanket modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. For the purpose of licensing such test module in the ITER facility, a safety assessment of the different possible accidental conditions has to be performed. This paper presents the results of the thermo-mechanical calculations that have been obtained in case of internal module leak occurring between the 8.0 MPa pressurised helium circuit and the lithium–lead circuit. In a conservative manner, it is also assumed that helium cooling of the TBM box is stopped at the accident initiation. In such situation the TBM box will be entirely pressurised at 8.0 MPa and the different challenges are related to the thermo-mechanical behaviour, the plasma power hold up and the heat removal capacity of the module. An assessment to estimate the time when the TBM box can withstand the different loads induced by the situation is presented. Particularly the available time span to trigger an emergency plasma shutdown is estimated in regards of the maximal allowable stress for the EUROFER which is the test blanket module material. A 3D finite element model has been developed with CAST3M computer code taking into account the beryllium layer, the decay heat after shutdown and the thermal radiation phenomena. After a description of the model, the paper presents and explains the results and particularly the methodology followed to determine the maximal allowable stress location which must combine the thermal transient calculations and the pure mechanical calculation under the 8.0 MPa loading. The results show that the hottest spot of the first wall (FW) was the most challenged location and that the structure can withstand such accidental conditions without plasma shutdown up to 15 s. Afterwards, the structural design criteria defined by ITER for in-vessel components (SDC-IC) is not any more fulfilled. This does not necessary mean that a lithium–lead leakage can occur in the vacuum vessel. More research and development are needed to have a clear understanding of crack propagation and break size in a material like EUROFER where few data are available, particularly at the temperature reached in this accident. Finally as a conclusion of the studies presented in this paper, it can be stated that, although very conservative assumptions were taken, the time span is large enough to trigger a plasma shutdown in order to avoid more severe operational and safety consequences of this accidental situation. © 2008 Elsevier B.V. All rights reserved.
1. Introduction The helium-cooled lithium–lead (HCLL) breeder blanket concept is one of the two breeder blanket lines presently developed by the EU for DEMO reactor. In the short-term so-called DEMO relevant test blanket modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed,
∗ Corresponding author. Tel.: +33 442254956. E-mail address: [email protected] (C. Girard). 0920-3796/$ – see front matter © 2008 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2008.07.042
commissioned and operated in ITER for first tests in a fusion environment. For the purpose of licensing such test module in the ITER facility, a safety assessment of the different possible accidental conditions has to be performed. This paper presents the results of the thermo-mechanical calculations obtained in case of internal module leak occurring between the 8.0 MPa pressurised helium circuit and the lithium–lead circuit (in-TBM LOCA: loss of coolant accident inside the TBM). In a conservative manner, it is also assumed that the helium cooling of the TBM box is stopped at the accident initiation.
C. Girard et al. / Fusion Engineering and Design 83 (2008) 1738–1741
In such situation the TBM box will be entirely pressurised at 8.0 MPa and the different challenges are related to the thermomechanical behaviour, the plasma power hold up and the heat removal capacity of the module. The time span when the TBM box can withstand the different loads induced by the situation without an emergency plasma shutdown is estimated in regards of the maximal allowable stress for the EUROFER which is the test blanket module material. 2. Process followed The goal of the study is to determine the HCLL-TBM temperature field during the accidental transient and to compute the mechanical stress caused by the helium pressurisation of the HCLL-TBM box. To this end, a finite element 3D model was developed using the CAST3M code [1]. This model allows the thermal and the stress fields to be computed within the HCLL-TBM during the transient. The HCLL-TBM mechanical integrity is assessed in three steps: • The first stage is the achievement of thermal computation on the model. This calculation refers to permanent conditions as well as the accidental transient simulation and the plasma shutdown effect. • The second stage is the mechanical calculation under accidental pressure conditions. • The last stage consists in verifying the HCLL-TBM box behaviour according to the stress level and the temperature reached during the transient by determining the allowable stress intensity. The mechanical analysis is performed using the structural design criteria for in-vessel components (SDC-IC) defined by the ITER project [2–5]. 3. TBM modelling
1739
it describes the precise geometry of the junction between the first wall (FW) and the side wall (SW). For simplicity, the model does not account for the horizontal stiffening plates nor the back plates, which are replaced at this step by mechanical boundary conditions. The thermal models used are • Thermal conduction in all the metallic structures and in the lithium–lead coolant, • Thermal exchange between structures and helium by forced convection, • Thermal radiation between the beryllium layer and the vacuum vessel (thermal radiation between the side wall and the frame is not considered at this step). For the mechanical computations, an elastic model is used. The physical properties of helium and lithium–lead come from literature, the thermo-mechanical characteristics of the steel structures, EUROFER, are found in Ref. [6]. The thermo-mechanical characteristics of beryllium layer are from Ref. [7]. 3.2. Thermal and mechanical loadings The thermal loading is due to the neutrons and the ␥ flux (neutron wall loading) produced by the plasma, which induces heating of the structures and of the lithium–lead. In addition, a heat flux is applied on the external face of the FW. The nominal thermal loadings considered in this study are: • Power density distribution related to a neutron wall loading (NWL) of 0.78 MW m−2 , • Heat flux (HF) on the beryllium surface, plasma side, of 0.5 MW m−2 .
3.1. Meshing, thermal and mechanical models For geometrical simplification the model represents one and a half Breeder Unit in width (see Fig. 1). The model is innovative as
The mechanical loading is due to the internal pressure of 8.0 MPa. This pressure is applied to the helium channels and to the surfaces in contact with lithium–lead to simulate an in-TBM helium
Fig. 1. Exploded view of the 3D model.
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C. Girard et al. / Fusion Engineering and Design 83 (2008) 1738–1741
leak pressurisation. An equivalent stress representative of the end effect, is applied to the upper surface of the model. 4. Thermo-mechanical assessment of the HCLL TBM behaviour 4.1. Simulation assumptions The simulation establishes, first, the steady state thermal field under normal operating conditions, and then considers the loss of helium cooling thermal transient by assuming an instantaneous and total loss of heat exchange with helium. In addition, it is assumed that the accident remains undetected and the plasma continues to burn producing a surface heat flux of 0.5 MW m−2 and a nominal neutron wall loading of 0.78 MW m−2 . It is obvious that the TBM cannot withstand such a thermal loading if any plasma shutdown does not happen; the purpose of the calculation is to determine the time interval t0 available during which the TBM can withstand the thermal loading (due to the loss of helium cooling capacity) combined with the internal pressure loading (due to the in-TBM LOCA). 4.2. Determination of the time interval The time interval t0 during which the HCLL-TBM can comply with the ITER safety criteria after an in-TBM LOCA without plasma shutdown, is determined by using the following methodology: • A thermal transient computation is performed up to a FW temperature of 1300 ◦ C (assuming plasma heating and no cooling from helium), • A pure mechanical computation, simulating the 8.0 MPa pressurisation of the HCLL-TBM box is performed without accounting for the thermal stresses (since the LOCA is classified as a level D event, the analysis deals only with the primary stresses), • The analysis is then performed at the location of the maximal stress obtained in the mechanical computation and at the location of the maximal temperature obtained during the thermal transient. • The maximal mechanical stress is then related to a maximal allowed temperature (from the table of maximal allowable stress as a function of temperature). It is then investigated into the results from the thermal transient when, at the same location, this maximal temperature is reached, this gives the time t1 , • The analysis is also performed for the mechanical stress obtained at the maximal temperature location, this stress is related to a maximal allowable temperature and from the thermal transient to a time t2 when this temperature is reached. • The minimum of t1 and t2 gives the time t0 = min(t1 , t2 ) when the HCLL-TBM box will start not to fulfill the SDC-IC criteria (primary stress upper than the maximal allowable stress).
Fig. 2. Temperature field in the beryllium layer under nominal conditions.
The loss of helium cooling transient is then simulated for 100 s. The thermal field is calculated as a function of time. As soon as the helium flow is stopped, the FW temperature tends toward an homogeneous thermal field. The temperature exceeds 700 ◦ C, 15 s after the transient initiation then reaches some 1000 ◦ C at 50 s. At the end of simulation (100 s) the FW temperature is about 1300 ◦ C (Fig. 3). In the frame of this study, the consequences of a plasma shutdown with disruption and the potential to remove decay heat by thermal radiation were also assessed. The assumption is that the plasma shutdown creates a disruption with a peak heat flux (5.5 MW m−2 during 100 ms) adding an additional heat loading that must be taken into account in the analysis. As an example, a plasma shutdown has been simulated when the FW temperature reaches 1115 ◦ C (approximately melting temperature of the beryllium layer), i.e. at 65 s after the transient initiation. The thermal computation shows that the increment of temperature caused by the disruption is weak (only 8 ◦ C). After the plasma shutdown, the thermal radiation allows the temperature to be decreased rapidly (Fig. 4). For the mechanical assessment, since the LOCA is classified as a level D event, the mechanical analysis deals only with the primary stresses. Consequently, the stress field does not depend on the thermal field. The equivalent primary stresses (Von Misès) are displayed in the meshing (Fig. 5). The calculations lead to a maximal stress intensity of 324 MPa.
5. Thermal and mechanical assessment First, the permanent thermal field before the accidental transient is calculated (Fig. 2). At this stage, the temperature fields under nominal working conditions are as follows: • • • • •
361 ◦ C ≤ Tberyllium layer ≤ 572 ◦ C (see Fig. 2), 310 ◦ C ≤ Tfirst wall–side wall ≤ 563 ◦ C, 429 ◦ C ≤ Tstiffening plate vertical ≤ 486 ◦ C, 390 ◦ C ≤ Tcooling plate ≤ 509 ◦ C, 369 ◦ C ≤ Tlithium–lead ≤ 534 ◦ C.
Fig. 3. FW temperature evolution following a loss of helium cooling.
C. Girard et al. / Fusion Engineering and Design 83 (2008) 1738–1741
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pressurisation, is the location where structural damages will appear first. 6. Conclusion The thermo-mechanical study presented in this paper deals with the HCLL-TBM behaviour under an in-TBM LOCA event without active plasma shutdown leading to: • Total loss of helium cooling, • TBM box pressurisation at 8.0 MPa. The thermal simulations performed have shown that:
Fig. 4. Decay heat removal by thermal radiation after plasma shutdown.
• The FW temperature increases rapidly up to 1300 ◦ C after 100 s, not very far from the EUROFER melting temperature which is around 1500 ◦ C, • Plasma shutdown simulation indicated that the disruption effect is not significant causing an additional heating of only 8 ◦ C and that the thermal radiation phenomena after shutdown is able to remove the decay heat. The mechanical analysis has been particularly performed at two places: • Over-stress place at the junction FW–SW, • Hot-spot place on the FW.
Fig. 5. Equivalent primary stress under TBM pressurisation at 8.0 MPa.
As seen before (Section 4.2) two locations have been analysed: • The location of the maximal stress, • The location of the hottest spot where the temperature is maximal. At the location of the maximal stress, the analysis shows that the ITER safety criteria are fulfilled as long as the temperature stays below 634 ◦ C. From the thermal transient calculations, this time is reached in a time t1 equals to 70 s. At the location of the maximal temperature the stress analysis shows that the criteria are fulfilled when the temperature is lower than 681 ◦ C. The time to reach this temperature is derived from the thermal transient calculations and it is found that this time, t2 , equals 15 s. Since this time is lower than the time obtained at the maximal stress location, it is concluded that the hottest spot in case of a combined loss of cooling and in-TBM LOCA producing a 8.0 MPa
The results show that the hottest spot of the first wall was the most challenged location and that the structure can withstand such accidental conditions without plasma shutdown up to 15 s. Afterwards, the structural design criteria defined by ITER for invessel components is not any more fulfilled. This does not necessary mean that a lithium–lead leakage can occur in the vacuum vessel. More research and development are needed to have a clear understanding of crack propagation and break size in a material like EUROFER where few data are available, particularly at the temperature reached in this accident. Finally as a conclusion of the studies presented in this paper, it can be stated that, although very conservative assumptions were taken (particularly the instantaneous total loss of heat removal), the time span is large enough to trigger a plasma shutdown in order to avoid more severe operational and safety consequences of this accidental situation. References [1] Cast3M FE code, CEA/DEN/DM2S/SEMT, Industrial Version 2000 for Windows