- Email: [email protected]
Steady states and LOFA analyses of the updated WCCB blanket for multiple fusion power modes of CFETR
Fusion Engineering and Design 144 (2019) 23–28
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Steady states and LOFA analyses of the updated WCCB blanket for multiple fusion power modes of CFETR Xiaoman Chenga, Xuebin Mab, Xia Lia,c, Wenjia Wanga,c, Songlin Liua,
T
⁎
a
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031, China College of Physics and Optoelectronic Engineering, Shenzhen University, Shenzhen 518060, China c University of Science and Technology of China, Hefei, Anhui 230027, China b
A R T I C LE I N FO
A B S T R A C T
Keywords: CFETR Updated WCCB blanket Thermal hydraulics RELAP5
The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates of Chinese Fusion Engineering Test Reactor (CFETR). The core parameters of CFETR have been updated to major radius R = 7.2 m, minor radius a = 2.2 in 2018. Accordingly, the WCCB blanket design was also updated to satisfy multiple operation modes of CFETR with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. The WCCB blanket is characterized by 3 independent cooling systems (CSs), namely CS1 for the cooling of the First Wall (FW), CS2&3 for Breeding Zones (BZs). At 200 MW of low fusion power, CS3 is idle. At 500 MW to 1.5 GW of high fusion power, all 3 CSs are put into use to ensure adequate cooling. So it is indispensable to evaluate the feasibility of the updated WCCB blanket from the thermal hydraulic and safety point of view. A system analysis code RELAP5, was employed to model the blanket. The steady states were analyzed under 200 MW, 500 MW, 1.0 GW and 1.5 GW of fusion power. Then, the Loss of Flow Accident (LOFA) was simulated with different assumptions to verify the robustness of 3 cooling systems design under different initial postulations. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature, can basically meet the design criterion, which preliminarily verifies the feasibility of the WCCB blanket from the thermal hydraulic safety point of view. Comprehensive system design and analysis is under way.
1. Introduction The Chinese Fusion Engineering Test Reactor (CFETR) aims to bridge the gap between ITER and DEMO. It will operate in multiple modes with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. In 2018, CFETR has been changed from the previous small size machine (major radius R = 5.7 m, minor radius a = 1.8 m) to the present large size machine (R = 7.2 m, a = 2.2 m) [1]. Accordingly, challenging requirements are proposed to the blanket design, which demands that one set of blankets can satisfy all operation modes. The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for CFETR. The WCCB blanket employs pressurized water (15.5 MPa, 558 K–598 K) as coolant. Li2TiO3 and Be12Ti are selected as tritium breeder and neutron multiplier respectively. These two materials are arranged inside the blanket module in the form of mixed pebble beds. Reduced Activation Ferritic/Martensitic (RAFM) steel (e.g. CLF-1, CLAM) is used as structure material. Besides, 2 mm of tungsten is coated on the plasma-facing side of the First Wall (FW) to resist plasma erosion and corrosion. Temperature upper limits are 823 K for RAFM
⁎
steel and 1573 K for tungsten armor. As for the mixed pebble bed, the temperature is expected to be in the range of 673 K–1173 K for better performance of tritium release. The WCCB blanket design was updated according to the latest core parameters of CFETR [2]. In order to fulfill the challenging requirements of CFETR, the cooling scheme of the WCCB blanket was carefully designed with 3 independent cooling systems (CSs). Steady states under different fusion power conditions should be evaluated to validate the feasibility of the updated WCCB blanket design. In addition, accident scenarios also should be considered to verify the robustness of the blanket. From the thermal hydraulic point of view, the criteria are to meet the allowable temperature limits of materials. The pressure drop should be reasonably low at steady states. Under accident conditions, over-heating and over-pressurization of the blanket module should be prevented so as to ensure integrity of the blanket and confinement of radioactive materials. In this paper, the WCCB blanket was modeled using system analysis code RELAP5 which was originally developed for light water reactors. RELAP5 has also been widely used in thermal hydraulic safety analysis
Corresponding author. E-mail address: [email protected] (S. Liu).
https://doi.org/10.1016/j.fusengdes.2019.04.075 Received 28 March 2019; Received in revised form 19 April 2019; Accepted 19 April 2019 Available online 24 April 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 144 (2019) 23–28
X. Cheng, et al.
Fig. 1. Typical blanket module structure.
Fig. 2. Water coolant flow scheme.
2. Blanket structure and RELAP5 model
of fusion reactors. Numerical models in RELAP5 have been verified by a large number of experiments and calculations. Moreover, hydraulic components and heat structures in RELAP5 are flexible in use and can fulfill simulation of various kinds of structures. Therefore RELAP5 is suitable for the simulation of the WCCB blanket. The RELAP5 model of the updated WCCB blanket is presented in Section 2. In Section 3, Steady states analyses were carried out under the fusion power of 200 MW, 500 MW, 1.0 GW, 1.5 GW. Then Loss of Flow Accident (LOFA) was studied with different initial assumptions in Section 4. The last section is the conclusion.
2.1. Typical blanket structure and coolant flow scheme There are 16 blanket sectors in CFETR. Each sector contains 3 outboard blanket segments and 2 inboard segments in the toroidal direction. The WCCB blanket employs the multi-module design scheme. The blanket module on the outboard equatorial plane is taken as a typical example to evaluate the thermal hydraulic and safety performance. The blanket module structure is shown in Fig. 1. The blanket module is a RAFM steel box enclosed by FW, cover plates and back plates (BPs). Specifically, The FW is a U-shaped plate with 8 mm × 8 mm cooling channels along radial-toroidal-radial direction. The upper and lower cover plates comprise 5 mm × 5 mm coolant channels and purge gas channels. The BPs form the coolant and 24
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Fig. 4. Nuclear heating density distribution.
Table 1 Hydraulic parameters under different fusion powers. Fusion power
200 MW
500 MW
1 GW
1.5 GW
Nuclear heat Total heat
Heat (MW) 0.71 1.99
1.77 3.06
3.54 4.83
5.32 6.60
CS1 FW SP Cover CS2 CT1 CS3 CT2
Mss flow rate (kg/s) 7.90 9.10 6.88 7.93 0.63 0.725 0.387 0.445 2.40 2.05 – 4.20
11.5 10.0 0.915 0.56 4.10 8.40
13.80 1.20 1.10 0.671 6.10 12.30
CS1 CS2 CS3
Pressure drop (MPa) 0.063 0.073 0.008 0.008 – 0.022
0.118 0.025 0.080
0.167 0.046 0.162
Fig. 5. Temperature distribution in radial direction under different fusion power modes.
channels are used for strengthening the structure and dividing the module into 4 sub-modules. In each sub-module, 23 Cooling Tube Assemblies (CTA) are embedded in the mixed pebble beds to remove the nuclear heat. Each CTA consists of 3 Cooling Tubes (CTs, Di = 8 mm/Do = 13.5 mm) with multiple bends in the radial and toroidal direction. The CTs are connected by ribs which are applied to enhancement of the structure strength and improvement of heat
Fig. 3. Nodalization scheme of the WCCB blanket module.
purge gas distributing and collecting manifolds (MF) which are connected to feed pipes in the Back Support Structure (BSS). Inside the blanket module, 3 Stiffening Plates (SPs) with 5 mm × 5 mm curving 25
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Fig. 6. Coolant temperature of (a) FW and (b) CT1 under different fusion power modes.
Table 2 LOFA sensitivity study. Case No.
Fusion power
CS
Pump condition
Plasma shut down condition
Case Case Case Case Case Case Case Case Case Case
200 MW 500 MW 1.0 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW
CS1 CS1 CS1 CS1 CS2 CS1&2 CS1 CS1 CS1 CS1
Coast down Coast down Coast down Coast down Coast down Coast down, with 10% mass flow remained No coast down Coast down Coast down Coast down
Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, with disruption heat Shut down delay 10 s, with disruption heat Soft shut down in 60 s
1 2 3 4 5 6 7 8 9 10
2.2. RELAP5 model The typical blanket module was modeled using RELAP5 for the purpose of verifying the feasibility of the WCCB blanket design. The nodalization scheme is shown in Fig. 3. (1) Hydraulic components Fig. 3(a) is the nodalization of CS1 which provides coolant for the FW, SPs and cover plates. 160 coolant channels with regular length of the FW were divided into two groups (FW1/FW2) according to the flow direction. Besides, there are 8 irregular channels at the corner of two sides. As those channels can influence the flow distribution, they were also modeled but not shown in the figure. Three SPs have the same structure. Taking the middle SP2 as an example, 20 curving regular channels were lumped as 3 serial PIPE components. Similar to the FW, 2 irregular channels were modeled but not shown. As for the cover plates, the purge gas line was not considered in this paper. Water coolant channels were lumped separately in 4 sub-modules for the upper and lower cover plates. Fig. 3(b) and (c) are the nodalization of CS2 and CS3 respectively. CS2 and CS3 are similar. Only the CT shape has slight difference and the CT number ratio is 1:2. So CS2 is taken as an example to explain. In each sub-module, 23 CT of regular length were lumped as one PIPE component (CT1). 2 other irregular channels were simulated to but not shown in the figure.
Fig. 7. FW wall temperature during LOFA of CS1 under different fusion power modes.
transfer. Three independent CSs are designed and combined in use to fulfill the requirements of CFETR. The coolant flow scheme in the blanket module is shown in Fig. 2. CS1 provides cooling for the FW, SPs and cover plates. CS2 feeds water coolant into one CT of each CTA and removes the nuclear heat deposited in mixed pebble beds, while CS3 feeds the other two CTs of each CTA. At low fusion power of 200 MW, CS 1&2 are put into use and CS 3 is idle. At high fusion power (i.e. 500 MW, 1.0 GW and 1.5 GW), all three CSs operate so as to ensure adequate cooling of the blanket.
(2) Heat structures The three independent CSs together remove the nuclear heat of the blanket module. In RELAP5 modeling, the three CSs were related through Heat Structures (HSs). For example, Breeding Zone 1 (BZ1) is cooled by FW in CS1, CT1 in CS2 and CT2 in CS3. So BZ1 was divided 26
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Fig. 8. (a) FW wall temperature and (b) max. BZ temperature during LOFA of different CSs under 1.5 GW of fusion power.
and power [3].The physical properties of materials were consistent to previous reports [4]. Convective boundary conditions were applied. Other HSs were modeled in the similar way. (3) Heat sources The nuclear heating density of the typical blanket module was obtained from neutronic analyses under different fusion power levels [2] as shown in Fig. 4. As for the surface heat flux, as the core parameters of CFETR were just updated in 2018, there are still significant uncertainties concerning plasma physics, such as plasma heating, particle loading, etc. In addition, the blanket design also influences the heat flux, like the module size, FW position, etc. So the heat flux of plasma is a complex parameter to determine. Up to now, there is no accurate result for CFETR. So the uniform heat flux of 0.5 MW/m2 [5] was assumed for the purpose of preliminary assessment in this paper. In the future, simulations will be iterated and updated as soon as the accurate heat flux is determined. Fig. 9. FW wall temperature during LOFA with different pump conditions.
3. Steady states analysis under multiple fusion power modes Steady states were analyzed under the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. Hydraulic parameters are listed in Table 1. The total heat of the blanket module includes nuclear heat and surface heat. The ratio of surface heat to total heat is higher at lower fusion power levels, as the heat flux from the plasma was all assumed to be 0.5 MW/m2 for different fusion powers. Therefore, the mass flow rate of CS1 is relatively higher at lower fusion power. The pressure drop is low as the cooling channels are in parallel. The temperature distribution along the radial direction is shown in Fig. 5. Because the three CSs are designed taking 1.5 GW as rated power, the temperature distribution is ideal for tritium release under the fusion power of 1.5 GW. The average temperature is inevitably low for low fusion power modes. Fortunately, the coolant temperature is rather consistent under different fusion powers as shown in Fig. 6, which is in favor of the design of the power conversion system. 4. LOFA analysis Based on steady states, sensitivity study of LOFA was carried out with different initial postulations as listed in Table 2:
Fig. 10. FW wall temperature during LOFA with different plasma shutdown conditions.
1) 2) 3) 4)
into 4 HSs. The 4 HSs were connected to FW1/CT1, FW1/CT2, FW2/ CT1 and FW2/CT2 respectively, with 1/6, 1/3, 1/6 and 1/3 of nuclear heat of BZ1 according to the cooling channel number ratio. In addition, HSs for BZ1 were simulated with several layers of different materials
27
LOFA LOFA LOFA LOFA
of CS1 under different fusion power modes (Case 1–4); of different CSs (Case 4–6); with different pump conditions (Case 4, 7); with different plasma shutdown conditions (Case 4, 8–10).
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The LOFA was assumed to take place at time = 10 s. Then the first 10 s was steady state which was regarded as base line. When the CS mass flow was lower than 80% of normal level, the Fusion Power Transport System (FPTS) was triggered and the plasma shut down 3 s later at normal circumstances.
[7] for the simulations in this paper. In addition, there could be a soft plasma shutdown mode referring to EU DEMO safety analysis [8]. In the soft shutdown condition, the nuclear heating density was assumed to decrease linearly to decay heat in 60 s, while the plasma heat flux decreases linearly from 0.5 MW/m2 to zero in 60 s. As can be seen in Fig. 10, the plasma shutdown modes have strong impact on LOFA of CS1. The disruption heat flux causes more than 50 K of temperature rise of the FW in 1 s. After the plasma quench completely, the temperature begins to decline. So far, 10 s delay of shut down does not cause serious problem. However, the soft shutdown leads to the excess of RAFM steel temperature limit after the pump stops coast down. It is suggested that the pump coast down time should be compatible to the plasma soft shutdown duration. Then the structure temperature can be controlled smoothly in a desired range.
4.1. LOFA under different fusion power modes CS1 is the most challenged system in the updated WCCB blanket. Because CS1 should provide adequate cooling for the FW, SPs and cover plates in parallel. At the same time, the plasma heat flux is acting on the surface of the FW. Therefore, LOFA of CS1 was assumed under different fusion power modes in this section. The FW wall temperature is shown in Fig. 7. For all cases, the FW temperature drops fast after plasma quench, as the decay heat is as low as 1% of normal nuclear heat and no plasma heat flux is applied. Moreover, the pump is still coasting down. Then the FW temperature increases as a result of the loss of flow. The higher the fusion power is, the faster the temperature increases. At 1.0 GW and 1.5 GW of fusion power, the FW wall temperature rises to around 618 K and remains unchanged due to nucleate boiling of the coolant in the FW. Eventually, the FW wall temperature decreases gradually because of the negligible decay heat.
5. Conclusions The feasibility of the updated WCCB blanket design was verified under steady states of different fusion powers and LOFA with different assumptions. For steady states, the temperature distribution of the blanket module was ideal at high fusion power, while the coolant temperature was consistent for different fusion power modes. Then sensitivity study of LOFA was performed based on steady states. The results indicated that higher fusion power led to faster temperature increase of the FW. If natural circulation is established during LOFA, the accident consequence can be mitigated further. Moreover, pump conditions and plasma shutdown conditions had strong impact on the blanket temperature. Specially, the pump coast down time should be compatible to the plasma soft shutdown duration. The results preliminarily demonstrated the feasibility and robustness of the WCCB blanket design. In the next step, design of the Primary Heat Transfer System (PHTS) will be carried out for the WCCB blanket and comprehensive safety analysis will be performed.
4.2. LOFA of different cooling systems The accident consequence is worse at higher fusion power. Therefore, LOFA of different CSs was studied under 1.5 GW of fusion power. As mentioned in Section 2, CS2 and CS3 are similar. Here, LOFA of CS2 is presented. In Fig. 8(a), it is obvious that LOFA of CS2 has little impact on the FW temperature. In Fig. 8(b), the maximum BZ temperature declines continuously for all cases. If there is 10% of flow remained in the system, both the FW temperature and the maximum BZ temperature can be controlled to a much lower level. Therefore, natural circulation of CSs is desirable during LOFA to remove the heat more efficiently.
Acknowledgements This work was supported by National Key R&D Program of China [No. 2017YFE0300503] and the Chinese National Natural Science Foundation [No. 11775256].
4.3. LOFA with different pump conditions LOFA can be caused by multiple reasons. One of the main reasons is pump trip. In normal condition, the pump is assumed to coast down in 32 s [6] for the simulations in this paper. However, the pump will stop instantly in case of pump rotor seizure, making the situation worse. Fig. 9 indicates the FW wall temperature during LOFA with different assumptions of pump conditions. It can be seen that the FW temperature rises over 10 K in 3 s before the plasma shutdown in case of no pump coast down. Then the temperature decreases gradually as the decay heat is much lower. At last the temperature remains constant due to the nucleate boiling of the coolant in the FW.
References [1] Yuanxi Wan, Jiangang Li, et al., Present progresses and activities on the Chinese fusion engineering test reactor, Presented in the SOFT 2018 Conference, (2018). [2] Songlin Liu, et al., Updated design of water-cooled breeder blanket for CFETR, Fusion Eng. Des. (2019), https://doi.org/10.1016/j.fusengdes.2019.03.023. [3] Xiaoman Cheng, et al., Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR, J. Nucl. Sci. Technol. 53 (2016) 1673–1680. [4] Kecheng Jiang, et al., Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR, Fusion Eng. Des. 112 (2016) 81–88. [5] Mohamed Abdou, et al., Blanket/first wall challenges and required R&D on the patheway to DEMO, Fusion Eng. Des. 100 (2015) 2–43. [6] L. Topilski, Safety analysis data list v. 5.2.6. Cadarache (France), ITER IDM Doc, ITER_D_24LSAE, (2008). [7] N. Taylor, Accident analysis guidelines 4 v. 4.2.4. Cadarache (France), ITER IDM Doc, ITER_D_24TDZ8, (2007). [8] XueZhou Jin, BB LOCA analysis for the reference design of the EU DEMO HCPB blanket concept, Fusion Eng. Des. 136 (2018) 958–963.
4.4. LOFA with different plasma shutdown conditions The plasma disruption heat flux is not determined for the updated CFETR for now. In order to figure out the influence of different plasma shutdown conditions, the disruption heat flux referred to ITER reports
28
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Steady states and LOFA analyses of the updated WCCB blanket for multiple fusion power modes of CFETR Xiaoman Chenga, Xuebin Mab, Xia Lia,c, Wenjia Wanga,c, Songlin Liua,
T
⁎
a
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031, China College of Physics and Optoelectronic Engineering, Shenzhen University, Shenzhen 518060, China c University of Science and Technology of China, Hefei, Anhui 230027, China b
A R T I C LE I N FO
A B S T R A C T
Keywords: CFETR Updated WCCB blanket Thermal hydraulics RELAP5
The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates of Chinese Fusion Engineering Test Reactor (CFETR). The core parameters of CFETR have been updated to major radius R = 7.2 m, minor radius a = 2.2 in 2018. Accordingly, the WCCB blanket design was also updated to satisfy multiple operation modes of CFETR with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. The WCCB blanket is characterized by 3 independent cooling systems (CSs), namely CS1 for the cooling of the First Wall (FW), CS2&3 for Breeding Zones (BZs). At 200 MW of low fusion power, CS3 is idle. At 500 MW to 1.5 GW of high fusion power, all 3 CSs are put into use to ensure adequate cooling. So it is indispensable to evaluate the feasibility of the updated WCCB blanket from the thermal hydraulic and safety point of view. A system analysis code RELAP5, was employed to model the blanket. The steady states were analyzed under 200 MW, 500 MW, 1.0 GW and 1.5 GW of fusion power. Then, the Loss of Flow Accident (LOFA) was simulated with different assumptions to verify the robustness of 3 cooling systems design under different initial postulations. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature, can basically meet the design criterion, which preliminarily verifies the feasibility of the WCCB blanket from the thermal hydraulic safety point of view. Comprehensive system design and analysis is under way.
1. Introduction The Chinese Fusion Engineering Test Reactor (CFETR) aims to bridge the gap between ITER and DEMO. It will operate in multiple modes with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. In 2018, CFETR has been changed from the previous small size machine (major radius R = 5.7 m, minor radius a = 1.8 m) to the present large size machine (R = 7.2 m, a = 2.2 m) [1]. Accordingly, challenging requirements are proposed to the blanket design, which demands that one set of blankets can satisfy all operation modes. The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for CFETR. The WCCB blanket employs pressurized water (15.5 MPa, 558 K–598 K) as coolant. Li2TiO3 and Be12Ti are selected as tritium breeder and neutron multiplier respectively. These two materials are arranged inside the blanket module in the form of mixed pebble beds. Reduced Activation Ferritic/Martensitic (RAFM) steel (e.g. CLF-1, CLAM) is used as structure material. Besides, 2 mm of tungsten is coated on the plasma-facing side of the First Wall (FW) to resist plasma erosion and corrosion. Temperature upper limits are 823 K for RAFM
⁎
steel and 1573 K for tungsten armor. As for the mixed pebble bed, the temperature is expected to be in the range of 673 K–1173 K for better performance of tritium release. The WCCB blanket design was updated according to the latest core parameters of CFETR [2]. In order to fulfill the challenging requirements of CFETR, the cooling scheme of the WCCB blanket was carefully designed with 3 independent cooling systems (CSs). Steady states under different fusion power conditions should be evaluated to validate the feasibility of the updated WCCB blanket design. In addition, accident scenarios also should be considered to verify the robustness of the blanket. From the thermal hydraulic point of view, the criteria are to meet the allowable temperature limits of materials. The pressure drop should be reasonably low at steady states. Under accident conditions, over-heating and over-pressurization of the blanket module should be prevented so as to ensure integrity of the blanket and confinement of radioactive materials. In this paper, the WCCB blanket was modeled using system analysis code RELAP5 which was originally developed for light water reactors. RELAP5 has also been widely used in thermal hydraulic safety analysis
Corresponding author. E-mail address: [email protected] (S. Liu).
https://doi.org/10.1016/j.fusengdes.2019.04.075 Received 28 March 2019; Received in revised form 19 April 2019; Accepted 19 April 2019 Available online 24 April 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 144 (2019) 23–28
X. Cheng, et al.
Fig. 1. Typical blanket module structure.
Fig. 2. Water coolant flow scheme.
2. Blanket structure and RELAP5 model
of fusion reactors. Numerical models in RELAP5 have been verified by a large number of experiments and calculations. Moreover, hydraulic components and heat structures in RELAP5 are flexible in use and can fulfill simulation of various kinds of structures. Therefore RELAP5 is suitable for the simulation of the WCCB blanket. The RELAP5 model of the updated WCCB blanket is presented in Section 2. In Section 3, Steady states analyses were carried out under the fusion power of 200 MW, 500 MW, 1.0 GW, 1.5 GW. Then Loss of Flow Accident (LOFA) was studied with different initial assumptions in Section 4. The last section is the conclusion.
2.1. Typical blanket structure and coolant flow scheme There are 16 blanket sectors in CFETR. Each sector contains 3 outboard blanket segments and 2 inboard segments in the toroidal direction. The WCCB blanket employs the multi-module design scheme. The blanket module on the outboard equatorial plane is taken as a typical example to evaluate the thermal hydraulic and safety performance. The blanket module structure is shown in Fig. 1. The blanket module is a RAFM steel box enclosed by FW, cover plates and back plates (BPs). Specifically, The FW is a U-shaped plate with 8 mm × 8 mm cooling channels along radial-toroidal-radial direction. The upper and lower cover plates comprise 5 mm × 5 mm coolant channels and purge gas channels. The BPs form the coolant and 24
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Fig. 4. Nuclear heating density distribution.
Table 1 Hydraulic parameters under different fusion powers. Fusion power
200 MW
500 MW
1 GW
1.5 GW
Nuclear heat Total heat
Heat (MW) 0.71 1.99
1.77 3.06
3.54 4.83
5.32 6.60
CS1 FW SP Cover CS2 CT1 CS3 CT2
Mss flow rate (kg/s) 7.90 9.10 6.88 7.93 0.63 0.725 0.387 0.445 2.40 2.05 – 4.20
11.5 10.0 0.915 0.56 4.10 8.40
13.80 1.20 1.10 0.671 6.10 12.30
CS1 CS2 CS3
Pressure drop (MPa) 0.063 0.073 0.008 0.008 – 0.022
0.118 0.025 0.080
0.167 0.046 0.162
Fig. 5. Temperature distribution in radial direction under different fusion power modes.
channels are used for strengthening the structure and dividing the module into 4 sub-modules. In each sub-module, 23 Cooling Tube Assemblies (CTA) are embedded in the mixed pebble beds to remove the nuclear heat. Each CTA consists of 3 Cooling Tubes (CTs, Di = 8 mm/Do = 13.5 mm) with multiple bends in the radial and toroidal direction. The CTs are connected by ribs which are applied to enhancement of the structure strength and improvement of heat
Fig. 3. Nodalization scheme of the WCCB blanket module.
purge gas distributing and collecting manifolds (MF) which are connected to feed pipes in the Back Support Structure (BSS). Inside the blanket module, 3 Stiffening Plates (SPs) with 5 mm × 5 mm curving 25
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Fig. 6. Coolant temperature of (a) FW and (b) CT1 under different fusion power modes.
Table 2 LOFA sensitivity study. Case No.
Fusion power
CS
Pump condition
Plasma shut down condition
Case Case Case Case Case Case Case Case Case Case
200 MW 500 MW 1.0 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW 1.5 GW
CS1 CS1 CS1 CS1 CS2 CS1&2 CS1 CS1 CS1 CS1
Coast down Coast down Coast down Coast down Coast down Coast down, with 10% mass flow remained No coast down Coast down Coast down Coast down
Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, no disruption heat Shut down in time, with disruption heat Shut down delay 10 s, with disruption heat Soft shut down in 60 s
1 2 3 4 5 6 7 8 9 10
2.2. RELAP5 model The typical blanket module was modeled using RELAP5 for the purpose of verifying the feasibility of the WCCB blanket design. The nodalization scheme is shown in Fig. 3. (1) Hydraulic components Fig. 3(a) is the nodalization of CS1 which provides coolant for the FW, SPs and cover plates. 160 coolant channels with regular length of the FW were divided into two groups (FW1/FW2) according to the flow direction. Besides, there are 8 irregular channels at the corner of two sides. As those channels can influence the flow distribution, they were also modeled but not shown in the figure. Three SPs have the same structure. Taking the middle SP2 as an example, 20 curving regular channels were lumped as 3 serial PIPE components. Similar to the FW, 2 irregular channels were modeled but not shown. As for the cover plates, the purge gas line was not considered in this paper. Water coolant channels were lumped separately in 4 sub-modules for the upper and lower cover plates. Fig. 3(b) and (c) are the nodalization of CS2 and CS3 respectively. CS2 and CS3 are similar. Only the CT shape has slight difference and the CT number ratio is 1:2. So CS2 is taken as an example to explain. In each sub-module, 23 CT of regular length were lumped as one PIPE component (CT1). 2 other irregular channels were simulated to but not shown in the figure.
Fig. 7. FW wall temperature during LOFA of CS1 under different fusion power modes.
transfer. Three independent CSs are designed and combined in use to fulfill the requirements of CFETR. The coolant flow scheme in the blanket module is shown in Fig. 2. CS1 provides cooling for the FW, SPs and cover plates. CS2 feeds water coolant into one CT of each CTA and removes the nuclear heat deposited in mixed pebble beds, while CS3 feeds the other two CTs of each CTA. At low fusion power of 200 MW, CS 1&2 are put into use and CS 3 is idle. At high fusion power (i.e. 500 MW, 1.0 GW and 1.5 GW), all three CSs operate so as to ensure adequate cooling of the blanket.
(2) Heat structures The three independent CSs together remove the nuclear heat of the blanket module. In RELAP5 modeling, the three CSs were related through Heat Structures (HSs). For example, Breeding Zone 1 (BZ1) is cooled by FW in CS1, CT1 in CS2 and CT2 in CS3. So BZ1 was divided 26
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Fig. 8. (a) FW wall temperature and (b) max. BZ temperature during LOFA of different CSs under 1.5 GW of fusion power.
and power [3].The physical properties of materials were consistent to previous reports [4]. Convective boundary conditions were applied. Other HSs were modeled in the similar way. (3) Heat sources The nuclear heating density of the typical blanket module was obtained from neutronic analyses under different fusion power levels [2] as shown in Fig. 4. As for the surface heat flux, as the core parameters of CFETR were just updated in 2018, there are still significant uncertainties concerning plasma physics, such as plasma heating, particle loading, etc. In addition, the blanket design also influences the heat flux, like the module size, FW position, etc. So the heat flux of plasma is a complex parameter to determine. Up to now, there is no accurate result for CFETR. So the uniform heat flux of 0.5 MW/m2 [5] was assumed for the purpose of preliminary assessment in this paper. In the future, simulations will be iterated and updated as soon as the accurate heat flux is determined. Fig. 9. FW wall temperature during LOFA with different pump conditions.
3. Steady states analysis under multiple fusion power modes Steady states were analyzed under the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. Hydraulic parameters are listed in Table 1. The total heat of the blanket module includes nuclear heat and surface heat. The ratio of surface heat to total heat is higher at lower fusion power levels, as the heat flux from the plasma was all assumed to be 0.5 MW/m2 for different fusion powers. Therefore, the mass flow rate of CS1 is relatively higher at lower fusion power. The pressure drop is low as the cooling channels are in parallel. The temperature distribution along the radial direction is shown in Fig. 5. Because the three CSs are designed taking 1.5 GW as rated power, the temperature distribution is ideal for tritium release under the fusion power of 1.5 GW. The average temperature is inevitably low for low fusion power modes. Fortunately, the coolant temperature is rather consistent under different fusion powers as shown in Fig. 6, which is in favor of the design of the power conversion system. 4. LOFA analysis Based on steady states, sensitivity study of LOFA was carried out with different initial postulations as listed in Table 2:
Fig. 10. FW wall temperature during LOFA with different plasma shutdown conditions.
1) 2) 3) 4)
into 4 HSs. The 4 HSs were connected to FW1/CT1, FW1/CT2, FW2/ CT1 and FW2/CT2 respectively, with 1/6, 1/3, 1/6 and 1/3 of nuclear heat of BZ1 according to the cooling channel number ratio. In addition, HSs for BZ1 were simulated with several layers of different materials
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LOFA LOFA LOFA LOFA
of CS1 under different fusion power modes (Case 1–4); of different CSs (Case 4–6); with different pump conditions (Case 4, 7); with different plasma shutdown conditions (Case 4, 8–10).
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The LOFA was assumed to take place at time = 10 s. Then the first 10 s was steady state which was regarded as base line. When the CS mass flow was lower than 80% of normal level, the Fusion Power Transport System (FPTS) was triggered and the plasma shut down 3 s later at normal circumstances.
[7] for the simulations in this paper. In addition, there could be a soft plasma shutdown mode referring to EU DEMO safety analysis [8]. In the soft shutdown condition, the nuclear heating density was assumed to decrease linearly to decay heat in 60 s, while the plasma heat flux decreases linearly from 0.5 MW/m2 to zero in 60 s. As can be seen in Fig. 10, the plasma shutdown modes have strong impact on LOFA of CS1. The disruption heat flux causes more than 50 K of temperature rise of the FW in 1 s. After the plasma quench completely, the temperature begins to decline. So far, 10 s delay of shut down does not cause serious problem. However, the soft shutdown leads to the excess of RAFM steel temperature limit after the pump stops coast down. It is suggested that the pump coast down time should be compatible to the plasma soft shutdown duration. Then the structure temperature can be controlled smoothly in a desired range.
4.1. LOFA under different fusion power modes CS1 is the most challenged system in the updated WCCB blanket. Because CS1 should provide adequate cooling for the FW, SPs and cover plates in parallel. At the same time, the plasma heat flux is acting on the surface of the FW. Therefore, LOFA of CS1 was assumed under different fusion power modes in this section. The FW wall temperature is shown in Fig. 7. For all cases, the FW temperature drops fast after plasma quench, as the decay heat is as low as 1% of normal nuclear heat and no plasma heat flux is applied. Moreover, the pump is still coasting down. Then the FW temperature increases as a result of the loss of flow. The higher the fusion power is, the faster the temperature increases. At 1.0 GW and 1.5 GW of fusion power, the FW wall temperature rises to around 618 K and remains unchanged due to nucleate boiling of the coolant in the FW. Eventually, the FW wall temperature decreases gradually because of the negligible decay heat.
5. Conclusions The feasibility of the updated WCCB blanket design was verified under steady states of different fusion powers and LOFA with different assumptions. For steady states, the temperature distribution of the blanket module was ideal at high fusion power, while the coolant temperature was consistent for different fusion power modes. Then sensitivity study of LOFA was performed based on steady states. The results indicated that higher fusion power led to faster temperature increase of the FW. If natural circulation is established during LOFA, the accident consequence can be mitigated further. Moreover, pump conditions and plasma shutdown conditions had strong impact on the blanket temperature. Specially, the pump coast down time should be compatible to the plasma soft shutdown duration. The results preliminarily demonstrated the feasibility and robustness of the WCCB blanket design. In the next step, design of the Primary Heat Transfer System (PHTS) will be carried out for the WCCB blanket and comprehensive safety analysis will be performed.
4.2. LOFA of different cooling systems The accident consequence is worse at higher fusion power. Therefore, LOFA of different CSs was studied under 1.5 GW of fusion power. As mentioned in Section 2, CS2 and CS3 are similar. Here, LOFA of CS2 is presented. In Fig. 8(a), it is obvious that LOFA of CS2 has little impact on the FW temperature. In Fig. 8(b), the maximum BZ temperature declines continuously for all cases. If there is 10% of flow remained in the system, both the FW temperature and the maximum BZ temperature can be controlled to a much lower level. Therefore, natural circulation of CSs is desirable during LOFA to remove the heat more efficiently.
Acknowledgements This work was supported by National Key R&D Program of China [No. 2017YFE0300503] and the Chinese National Natural Science Foundation [No. 11775256].
4.3. LOFA with different pump conditions LOFA can be caused by multiple reasons. One of the main reasons is pump trip. In normal condition, the pump is assumed to coast down in 32 s [6] for the simulations in this paper. However, the pump will stop instantly in case of pump rotor seizure, making the situation worse. Fig. 9 indicates the FW wall temperature during LOFA with different assumptions of pump conditions. It can be seen that the FW temperature rises over 10 K in 3 s before the plasma shutdown in case of no pump coast down. Then the temperature decreases gradually as the decay heat is much lower. At last the temperature remains constant due to the nucleate boiling of the coolant in the FW.
References [1] Yuanxi Wan, Jiangang Li, et al., Present progresses and activities on the Chinese fusion engineering test reactor, Presented in the SOFT 2018 Conference, (2018). [2] Songlin Liu, et al., Updated design of water-cooled breeder blanket for CFETR, Fusion Eng. Des. (2019), https://doi.org/10.1016/j.fusengdes.2019.03.023. [3] Xiaoman Cheng, et al., Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR, J. Nucl. Sci. Technol. 53 (2016) 1673–1680. [4] Kecheng Jiang, et al., Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR, Fusion Eng. Des. 112 (2016) 81–88. [5] Mohamed Abdou, et al., Blanket/first wall challenges and required R&D on the patheway to DEMO, Fusion Eng. Des. 100 (2015) 2–43. [6] L. Topilski, Safety analysis data list v. 5.2.6. Cadarache (France), ITER IDM Doc, ITER_D_24LSAE, (2008). [7] N. Taylor, Accident analysis guidelines 4 v. 4.2.4. Cadarache (France), ITER IDM Doc, ITER_D_24TDZ8, (2007). [8] XueZhou Jin, BB LOCA analysis for the reference design of the EU DEMO HCPB blanket concept, Fusion Eng. Des. 136 (2018) 958–963.
4.4. LOFA with different plasma shutdown conditions The plasma disruption heat flux is not determined for the updated CFETR for now. In order to figure out the influence of different plasma shutdown conditions, the disruption heat flux referred to ITER reports
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