- Email: [email protected]
Development of water-cooled blanket concept with pressure tightness against in-box LOCA for JA DEMO
Fusion Engineering and Design 146 (2019) 894–897
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Development of water-cooled blanket concept with pressure tightness against in-box LOCA for JA DEMO
T
⁎
Youji Someyaa, , Hironobu Kudob, Kenji Tobitaa, Ryoji Hiwataria, Yoshiteru Sakamotoa, Joint Special Design Team for Fusion DEMO a b
National Institutes for Quantum and radiological Science and Technology, Rokkasho Aomori, Japan Metal Technology Co. Ltd., Mito Ibaraki, Japan
A R T I C LE I N FO
A B S T R A C T
Keywords: JA DEMO Breeding blanket Pressure tightness Mixed pebbles and tritium breeding ratio
A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 ºC-325 ºC. In this design, the breeding area of the module is divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module is designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 ºC and 0.32 MPa, respectively. It was found that the self-sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.
1. Introduction JA DEMO blanket concepts are based on a water-cooled solid breeder (WCSB). In this design, the self-sufficient production of tritium and thermal power extraction for electricity generation are required to sustain DEMO operation. The cooling water for DEMO is operated at the Pressurized Water Reactor (PWR) water conditions of 15.5 MPa and 290 °C–325 °C. In addition, an overall Tritium Breeding Ratio (TBR) ≥ 1.05 is required, which allow the self-sufficient production of tritium to sustain its own operation. The original conceptual blanket was designed to have simple interior for mass production and to arrange a mixed bed of Li2TiO3 pebbles and Be12Ti ones in the blanket module. In the blanket concept, 3D neutronics analysis for overall TBR ensures that the TBR reaches the TBR target. In recent years, the blanket conceptual design has focused on pressure tightness of blanket module against inbox LOCA (loss of coolant accident). 2. Update of blanket design The previous concept of the DEMO blanket was based on a mixed breeder [1,2] with a simplified pebble bed distribution as shown in Fig. 1. A key point of the concept is to use chemically stable Be12Ti
⁎
pebbles as neutron multiplier [3]. The breeding zone is composed of a mixed pebbles bed of Li2TiO3 and Be12Ti, cooling pipes and supporting plates. Since the cooling pipes are welded to coolant headers located in the back of the blanket modules, irradiation degradation of welding lines is alleviated. On the other hand, it is noted that the cooling pipe positioning requires sub-mm-accuracy in the breeding area, as 5 mm of shift in the positioning implies 100 °C of temperature increase. Therefore, the supporting structure of cooling pipes was considered in the blanket module, which was arranged in the form of plates in the radial direction. With 5 supporting plates, it is possible to arrange piping alignment within 1 mm or less as shown in Fig. 1. The calculated overall TBR was 1.16 when the TBR was estimated from the number of tritium production over that of initial neutrons for each blanket. The problem of the previous blanket concept is it does not withstand inner pressurization by in-box LOCA, namely, ingress of cooling water due to a break of the cooling pipe inside the blanket module. The previous blanket design is being updated from the point of view of the over-pressure assurance in which the blanket module is divided into discrete cells with rib structure. In the concept, the operating pressure of the coolant is assumed to 15.5 MPa. Since the pressure drop of the primary cooling water system with breeding blanket modules was 1.2 MPa [4], the design pressure of the coolant is set to be 17.2 MPa
Corresponding author. E-mail address: [email protected] (Y. Someya).
https://doi.org/10.1016/j.fusengdes.2019.01.107 Received 8 October 2018; Received in revised form 21 January 2019; Accepted 22 January 2019 Available online 14 March 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 146 (2019) 894–897
Y. Someya, et al.
Fig. 1. Previous Breeding Blanket Concept for JA DEMO.
Fig. 2. Update of the breeding blanket concept for JA DEMO.
Fig. 3. Distribution of the nuclear heating of the breeding blanket concept for JA DEMO.
with the margin of 0.5 MPa. The dimension of each cell of the breeding area is 0.1 m in width and 0.1 m in height, and the structural material of rib is selected to be Reduced Activation Ferritic Martensitic (RAFM) steel of F82H as shown in Fig. 2. The dimension of the updated blanket modules is typically 1.44 m in width and 0.73 m in height. The first wall (FW) of the blanket is covered with 0.5 mm-thick tungsten (W) to suppress erosion by physical sputtering. A square cooling channel structure (8 mm × 8 mm) is arranged in the FW of thickness in 18 mm. The thickness of the FW with W coating is totally 18.5 mm. The distance of each cooling pipes in the breeding area and the rib structure is designed to satisfy the operation temperature of the materials. The blanket module is fabricated by hot isostatic pressing (HIP) using F82H to form the 0.1-m-squared cells as shown in Fig. 2.
Fig. 4. Temperature distribution of the F82H of the structure material (a), and of the Li2TiO3 and Be12Ti of the mixed tritium breeder (b).
3. Basic performance for JA DEMO blanket The JA DEMO reactor must demonstrate self-sufficient production of tritium, sufficient plant availability, and sufficient power generation. In the JA DEMO blanket concept with the self-sufficient production of tritium, a heat removal for power generation, a robustness of the module structure and a consistency for plasma stability are required to sustain any DEMO operation. The main parameters of the JA DEMO reactor are the plasma major radius of ∼8.5 m and fusion output of ∼1.5 GW [5]. 895
Fusion Engineering and Design 146 (2019) 894–897
Y. Someya, et al.
Table 1 Estimation of the stress limits for the updated blanket concept. Target
Service condition
Estimation of stresses
Allowable stress
Pressure resistance in the box
Accident such as in-box LOCA (Condition D)
General primary membrane stress (Through-thickness direction) Local primary membrane stress (The same as the above) Primary membrane stress and bending stress (The same as the above) Primary membrane stress and bending stress (The same as the above) Fatigue assessment (Surface stress with peak stress)
Min (2.4Sm, 2/3Su)
311 MPa
1.5× Min (2.4Sm, 2/3Su) 1.5× Min (2.4Sm, 2/3Su) 3Sm
467 MPa
Thermal stress
Study state operation (Condition A and B)
467 MPa 468 MPa
Cumulative fatigue damage＜1.0
Fig. 7. Overall TBR vs the ratio of Li2TiO3 regarding types of the coolant water in the breeding blanket. Fig. 5. MCNP calculation model for the DEMO blanket.
Table 2 Calculation condition of the MCNP model. Plasma Wall
W coating：500μm
First wall / side wall
FW:18 mm/SW:22 mm, F82 H(68%)+H2O(32%) UW and BW:22 mm, F82 H(93%)+H2O(7%) Ex. Ribs:18 mm, F82 H(93%)+H2O(7%) Mixed breeder：74%, He-purge gas:19%
Upper and bottom wall Rib structures Breeding area w/o coolant pipe Coolant pipe in the breeding area Mixed Breeding material
Fig. 8. Tritium concentration in the primary cooling loop when heavy water was selected to coolant for JA DEMO.
F82 H(3%)+H2O(4%)
3.1. Heat removal for power generation Li2TiO3(30%) / Be12Ti(70%)
The maximum neutron wall load is 1.66 MW/m2 and the heat wall load due to radiation from the plasma is 0.5 MW/m2, when the fusion power is 1.5 GW and the major radius is 8.2 m. The coolant condition was assumed to be PWR water conditions of 15.5 MPa and ΔT = 35 °C (290 °C–325 °C). The upper coolant velocity was limited to 5 m/s and the outlet temperature was less than 325 °C. In the tritium breeding areas and rib structures in the blanket, the distance between neighbouring cooling pipes is designed to satisfy the operational temperature window of the structural and functional materials. The operation temperature of mixed pebbles of Li2TiO3 and Be12Ti, and RAFM steel of F82H was limited to 900 °C and 550 °C, respectively. The temperature of the using materials such as the mixed pebbles and the RAFM steel of F82H evaluated by a distribution of the nuclear heating rate in the breeding blanket module. The distribution of the nuclear heating rate without W coating in the blanket is shown in Fig. 3. The nuclear heating of the W coating is also 76.8 MW/m3 during 0 to 0.5 mm radius. Temperature distributions of the F82H for structure, and of the mixed pebbles for tritium breeding were shown in Fig. 4. In Fig. 4, The maximum temperature of the mixed pebbles and structure materials are 846 °C and 544 °C, respectively. It was found that the cooling pipes in
Fig. 6. Overall TBR vs the thickness of blanket in the radial direction.
896
Fusion Engineering and Design 146 (2019) 894–897
Y. Someya, et al.
Stress limits of the updated blanket concept were assessed based on 1) study-state conditions A and B; 2) accident condition D (i.e. in-box LOCA), respectively reported in Table 1. A classification of the blanket concept is assumed to be a Class 1 component. With material design data of F82H, the minimum ultimate tensile strength as Su and the primary membrane stress intensity as Sm are assumed to be 467 MPa and 156 MPa, respectively [7]. As a result, maximum general primary membrane stress, and maximum primary membrane stress and bending stress are 303 MPa and 454 MPa. It was found that a rib with the thickness of 0.015 m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. A stress analysis for thermal stress ensures that maximum stress value is less than allowable value based on the temperature distribution in Section 3.1.
sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with the idea of improving TBR. On the other hand, there is a concern for safety that the tritium concentration in the primary coolant water may increase because of generating the tritium by the D(n,γ)T reaction. In the JA DEMO condition, the following tritium permeation rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region, 1.6 g/day in the divertor and totally 5.7 g/day [10]. In Fig. 8, the tritium permeation from the in-vessel components such as the blanket and divertor dominate in the tritium concentration of the primary coolant loop as compared with nuclear reaction in the heavy water. In the JA DEMO, the tritium concentration is assumed to be 1 T Bq/kg like CANDU reactor [11]. The tritium concentration can be managed when 93.7 kg/hr of the feed water is processed in a water detritiation system (WDS) with an extraction efficiency of 0.96 as shown in Fig. 8. The existing WDS (e.g., a Wolsong plant in Korea [12]) will be applicable to JA DEMO to maintain the management concentration. Therefore, the tritium generation in the primary coolant water by the D(n,γ)T reaction have little affect on JA DEMO design. However, there are critical issues that purification process of the heavy water with management of the tritium concentration and increasing a power consumption must be considered in the future work.
3.3. Tritium breeding ratio
5. Conclusion
In the 3D neutronic analysis for the blanket, the MCNP-5 code [8] with the nuclear library FENDL-3 [9] was used. The blanket coverage depletion in the divertor is 7.5%, depletion for the ports is 0.8% and depletion for the gap is 3.7%. The MCNP calculation model includes the gap, frame and ribs as shown in Fig. 5. The overall TBR is estimated to be 1.05, satisfying the overall TBR target of 1.05 for the blanket coverage of 91.7%. The calculation conditions are as follows: [6] Li enrichment of 90%, pebbles packing ratio of 80%. The tritium breeding area is filled with the mixture of tritium breeder of Li2TiO3 and neutron multiplier of Be12Ti pebbles. The neutronics analysis result indicates that the TBR becomes a maximum value when the ratio of the Li2TiO3 and Be12Ti pebbles is near the desired ratio of 3:7 (see Fig. 7). Table 2 shows the parameter of the MCNP model. Fig. 6 shows that the TBR as a function of the rib thickness. It was found that the overall TBR is anticipated to be 1.05 when the radial thickness of the blanket and rib in thickness are 1.2 m and 12 mm, respectively. On the other hand, a rib with the thickness of 0.015 m was needed to withstand the pressure of 17.2 MPa. In addition, the conducting shell needs to be located at rwall/ap ≤ 1.35 for plasma positional stability and high beta access. where ap is the plasma minor radius and rwall is the distance between the center of the plasma and the center of conducting shell. The design target is to satisfy the overall TBR of ≥1.05 using the blanket with the thickness of 0.7 m. From the 3D neutronics analysis results, the TBR of the updated design seems insufficient to meet the self-sufficiency of tritium. Because of overall TBR in this design doesn’t have enough margin for a self-sufficient supply of tritium. In order to enhance tritium production capability, further improvement of the blanket concept is needed.
A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. In this design, the breeding area of the module was divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3 pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015 m was needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module was designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 °C and 0.32 MPa, respectively. In addition, an overall TBR ≥ 1.05 is required, which allow the self-sufficient production of tritium to sustain its own operation. It was found that the self-sufficient production of tritium was likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.
the tritium breeding area and the rib structure were designed to satisfy the operational temperature window of the tritium breeding material and structural materials. In the Computational Fluid Dynamics analysis, the outlet coolant temperature and pressure drop are 321 °C and 0.356 MPa for the blanket concept. From this result, we find the overall outlet coolant temperature to be in the acceptable range. 3.2. Robustness of the module structure
Acknowledgments This work was carried out in the framework of the activity of the Joint Special Design Team for Fusion DEMO and partly by the DEMO Design Activity under the Broader Approach. References [1] [2] [3] [4] [5] [6]
4. Discuss to improve tritium breeding ratio In the viewpoints of neutronics characteristics, it is proposed that the cooling water is changed from light water to heavy water in order to improve the tritium productivity. As a result of neutronics analysis, heavy water as cooling water improved the overall TBR by about 0.05 due to the (n, 2n) reaction of the deuterium and the less attenuation effect on the neutron energy compared to light water as shown in Fig. 7. It was found that the self-
[7] [8] [9] [10] [11] [12]
897
Y. Someya, et al., Fusion Eng. Des. 86 (2011) 2269. Y. Someya, et al., Fusion Eng. Des. 98-99 (2015) 1872. M. Nakamichi, et al., Fusion Eng. Des. 69 (2003) 257. Y. Miyoshi, et al., Fusion Eng. Des. 126 (2018) 110. K. Tobita, et al., Fusion Sci. Technol. 72-4 (2017) 537. Japan Society for Mechanical Engineers, JSME S NC1–2014, Codes for Nuclear Power Generation Facilities, (2019). A.-A.F. Tavassoli, et al., Fusion Eng. Des. 61-62 (2002) 617. X-5 Monte Carlo Team, MCNP – A General Monte Carlo N-Particle Transport Code, version 5, LANL report, LA-CP-03-0245, April 2003 (rev. March 2005) (2019). R.A. Forrest, FENDL-2 Library: Summary Documentation, IAEA Report INDC(NDS)0628 (2012). K. Katayama, et al., Fusion Sci. Technol. 71 (2017) 261. S.K. SOOD, et al., Fusion Eng. Des. 12 (1990) 356. K.M. SONG, et al., Fusion Sci. Technol. 48 (2005) 290.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Development of water-cooled blanket concept with pressure tightness against in-box LOCA for JA DEMO
T
⁎
Youji Someyaa, , Hironobu Kudob, Kenji Tobitaa, Ryoji Hiwataria, Yoshiteru Sakamotoa, Joint Special Design Team for Fusion DEMO a b
National Institutes for Quantum and radiological Science and Technology, Rokkasho Aomori, Japan Metal Technology Co. Ltd., Mito Ibaraki, Japan
A R T I C LE I N FO
A B S T R A C T
Keywords: JA DEMO Breeding blanket Pressure tightness Mixed pebbles and tritium breeding ratio
A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 ºC-325 ºC. In this design, the breeding area of the module is divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module is designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 ºC and 0.32 MPa, respectively. It was found that the self-sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.
1. Introduction JA DEMO blanket concepts are based on a water-cooled solid breeder (WCSB). In this design, the self-sufficient production of tritium and thermal power extraction for electricity generation are required to sustain DEMO operation. The cooling water for DEMO is operated at the Pressurized Water Reactor (PWR) water conditions of 15.5 MPa and 290 °C–325 °C. In addition, an overall Tritium Breeding Ratio (TBR) ≥ 1.05 is required, which allow the self-sufficient production of tritium to sustain its own operation. The original conceptual blanket was designed to have simple interior for mass production and to arrange a mixed bed of Li2TiO3 pebbles and Be12Ti ones in the blanket module. In the blanket concept, 3D neutronics analysis for overall TBR ensures that the TBR reaches the TBR target. In recent years, the blanket conceptual design has focused on pressure tightness of blanket module against inbox LOCA (loss of coolant accident). 2. Update of blanket design The previous concept of the DEMO blanket was based on a mixed breeder [1,2] with a simplified pebble bed distribution as shown in Fig. 1. A key point of the concept is to use chemically stable Be12Ti
⁎
pebbles as neutron multiplier [3]. The breeding zone is composed of a mixed pebbles bed of Li2TiO3 and Be12Ti, cooling pipes and supporting plates. Since the cooling pipes are welded to coolant headers located in the back of the blanket modules, irradiation degradation of welding lines is alleviated. On the other hand, it is noted that the cooling pipe positioning requires sub-mm-accuracy in the breeding area, as 5 mm of shift in the positioning implies 100 °C of temperature increase. Therefore, the supporting structure of cooling pipes was considered in the blanket module, which was arranged in the form of plates in the radial direction. With 5 supporting plates, it is possible to arrange piping alignment within 1 mm or less as shown in Fig. 1. The calculated overall TBR was 1.16 when the TBR was estimated from the number of tritium production over that of initial neutrons for each blanket. The problem of the previous blanket concept is it does not withstand inner pressurization by in-box LOCA, namely, ingress of cooling water due to a break of the cooling pipe inside the blanket module. The previous blanket design is being updated from the point of view of the over-pressure assurance in which the blanket module is divided into discrete cells with rib structure. In the concept, the operating pressure of the coolant is assumed to 15.5 MPa. Since the pressure drop of the primary cooling water system with breeding blanket modules was 1.2 MPa [4], the design pressure of the coolant is set to be 17.2 MPa
Corresponding author. E-mail address: [email protected] (Y. Someya).
https://doi.org/10.1016/j.fusengdes.2019.01.107 Received 8 October 2018; Received in revised form 21 January 2019; Accepted 22 January 2019 Available online 14 March 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 146 (2019) 894–897
Y. Someya, et al.
Fig. 1. Previous Breeding Blanket Concept for JA DEMO.
Fig. 2. Update of the breeding blanket concept for JA DEMO.
Fig. 3. Distribution of the nuclear heating of the breeding blanket concept for JA DEMO.
with the margin of 0.5 MPa. The dimension of each cell of the breeding area is 0.1 m in width and 0.1 m in height, and the structural material of rib is selected to be Reduced Activation Ferritic Martensitic (RAFM) steel of F82H as shown in Fig. 2. The dimension of the updated blanket modules is typically 1.44 m in width and 0.73 m in height. The first wall (FW) of the blanket is covered with 0.5 mm-thick tungsten (W) to suppress erosion by physical sputtering. A square cooling channel structure (8 mm × 8 mm) is arranged in the FW of thickness in 18 mm. The thickness of the FW with W coating is totally 18.5 mm. The distance of each cooling pipes in the breeding area and the rib structure is designed to satisfy the operation temperature of the materials. The blanket module is fabricated by hot isostatic pressing (HIP) using F82H to form the 0.1-m-squared cells as shown in Fig. 2.
Fig. 4. Temperature distribution of the F82H of the structure material (a), and of the Li2TiO3 and Be12Ti of the mixed tritium breeder (b).
3. Basic performance for JA DEMO blanket The JA DEMO reactor must demonstrate self-sufficient production of tritium, sufficient plant availability, and sufficient power generation. In the JA DEMO blanket concept with the self-sufficient production of tritium, a heat removal for power generation, a robustness of the module structure and a consistency for plasma stability are required to sustain any DEMO operation. The main parameters of the JA DEMO reactor are the plasma major radius of ∼8.5 m and fusion output of ∼1.5 GW [5]. 895
Fusion Engineering and Design 146 (2019) 894–897
Y. Someya, et al.
Table 1 Estimation of the stress limits for the updated blanket concept. Target
Service condition
Estimation of stresses
Allowable stress
Pressure resistance in the box
Accident such as in-box LOCA (Condition D)
General primary membrane stress (Through-thickness direction) Local primary membrane stress (The same as the above) Primary membrane stress and bending stress (The same as the above) Primary membrane stress and bending stress (The same as the above) Fatigue assessment (Surface stress with peak stress)
Min (2.4Sm, 2/3Su)
311 MPa
1.5× Min (2.4Sm, 2/3Su) 1.5× Min (2.4Sm, 2/3Su) 3Sm
467 MPa
Thermal stress
Study state operation (Condition A and B)
467 MPa 468 MPa
Cumulative fatigue damage＜1.0
Fig. 7. Overall TBR vs the ratio of Li2TiO3 regarding types of the coolant water in the breeding blanket. Fig. 5. MCNP calculation model for the DEMO blanket.
Table 2 Calculation condition of the MCNP model. Plasma Wall
W coating：500μm
First wall / side wall
FW:18 mm/SW:22 mm, F82 H(68%)+H2O(32%) UW and BW:22 mm, F82 H(93%)+H2O(7%) Ex. Ribs:18 mm, F82 H(93%)+H2O(7%) Mixed breeder：74%, He-purge gas:19%
Upper and bottom wall Rib structures Breeding area w/o coolant pipe Coolant pipe in the breeding area Mixed Breeding material
Fig. 8. Tritium concentration in the primary cooling loop when heavy water was selected to coolant for JA DEMO.
F82 H(3%)+H2O(4%)
3.1. Heat removal for power generation Li2TiO3(30%) / Be12Ti(70%)
The maximum neutron wall load is 1.66 MW/m2 and the heat wall load due to radiation from the plasma is 0.5 MW/m2, when the fusion power is 1.5 GW and the major radius is 8.2 m. The coolant condition was assumed to be PWR water conditions of 15.5 MPa and ΔT = 35 °C (290 °C–325 °C). The upper coolant velocity was limited to 5 m/s and the outlet temperature was less than 325 °C. In the tritium breeding areas and rib structures in the blanket, the distance between neighbouring cooling pipes is designed to satisfy the operational temperature window of the structural and functional materials. The operation temperature of mixed pebbles of Li2TiO3 and Be12Ti, and RAFM steel of F82H was limited to 900 °C and 550 °C, respectively. The temperature of the using materials such as the mixed pebbles and the RAFM steel of F82H evaluated by a distribution of the nuclear heating rate in the breeding blanket module. The distribution of the nuclear heating rate without W coating in the blanket is shown in Fig. 3. The nuclear heating of the W coating is also 76.8 MW/m3 during 0 to 0.5 mm radius. Temperature distributions of the F82H for structure, and of the mixed pebbles for tritium breeding were shown in Fig. 4. In Fig. 4, The maximum temperature of the mixed pebbles and structure materials are 846 °C and 544 °C, respectively. It was found that the cooling pipes in
Fig. 6. Overall TBR vs the thickness of blanket in the radial direction.
896
Fusion Engineering and Design 146 (2019) 894–897
Y. Someya, et al.
Stress limits of the updated blanket concept were assessed based on 1) study-state conditions A and B; 2) accident condition D (i.e. in-box LOCA), respectively reported in Table 1. A classification of the blanket concept is assumed to be a Class 1 component. With material design data of F82H, the minimum ultimate tensile strength as Su and the primary membrane stress intensity as Sm are assumed to be 467 MPa and 156 MPa, respectively [7]. As a result, maximum general primary membrane stress, and maximum primary membrane stress and bending stress are 303 MPa and 454 MPa. It was found that a rib with the thickness of 0.015 m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. A stress analysis for thermal stress ensures that maximum stress value is less than allowable value based on the temperature distribution in Section 3.1.
sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with the idea of improving TBR. On the other hand, there is a concern for safety that the tritium concentration in the primary coolant water may increase because of generating the tritium by the D(n,γ)T reaction. In the JA DEMO condition, the following tritium permeation rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region, 1.6 g/day in the divertor and totally 5.7 g/day [10]. In Fig. 8, the tritium permeation from the in-vessel components such as the blanket and divertor dominate in the tritium concentration of the primary coolant loop as compared with nuclear reaction in the heavy water. In the JA DEMO, the tritium concentration is assumed to be 1 T Bq/kg like CANDU reactor [11]. The tritium concentration can be managed when 93.7 kg/hr of the feed water is processed in a water detritiation system (WDS) with an extraction efficiency of 0.96 as shown in Fig. 8. The existing WDS (e.g., a Wolsong plant in Korea [12]) will be applicable to JA DEMO to maintain the management concentration. Therefore, the tritium generation in the primary coolant water by the D(n,γ)T reaction have little affect on JA DEMO design. However, there are critical issues that purification process of the heavy water with management of the tritium concentration and increasing a power consumption must be considered in the future work.
3.3. Tritium breeding ratio
5. Conclusion
In the 3D neutronic analysis for the blanket, the MCNP-5 code [8] with the nuclear library FENDL-3 [9] was used. The blanket coverage depletion in the divertor is 7.5%, depletion for the ports is 0.8% and depletion for the gap is 3.7%. The MCNP calculation model includes the gap, frame and ribs as shown in Fig. 5. The overall TBR is estimated to be 1.05, satisfying the overall TBR target of 1.05 for the blanket coverage of 91.7%. The calculation conditions are as follows: [6] Li enrichment of 90%, pebbles packing ratio of 80%. The tritium breeding area is filled with the mixture of tritium breeder of Li2TiO3 and neutron multiplier of Be12Ti pebbles. The neutronics analysis result indicates that the TBR becomes a maximum value when the ratio of the Li2TiO3 and Be12Ti pebbles is near the desired ratio of 3:7 (see Fig. 7). Table 2 shows the parameter of the MCNP model. Fig. 6 shows that the TBR as a function of the rib thickness. It was found that the overall TBR is anticipated to be 1.05 when the radial thickness of the blanket and rib in thickness are 1.2 m and 12 mm, respectively. On the other hand, a rib with the thickness of 0.015 m was needed to withstand the pressure of 17.2 MPa. In addition, the conducting shell needs to be located at rwall/ap ≤ 1.35 for plasma positional stability and high beta access. where ap is the plasma minor radius and rwall is the distance between the center of the plasma and the center of conducting shell. The design target is to satisfy the overall TBR of ≥1.05 using the blanket with the thickness of 0.7 m. From the 3D neutronics analysis results, the TBR of the updated design seems insufficient to meet the self-sufficiency of tritium. Because of overall TBR in this design doesn’t have enough margin for a self-sufficient supply of tritium. In order to enhance tritium production capability, further improvement of the blanket concept is needed.
A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. In this design, the breeding area of the module was divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3 pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015 m was needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module was designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 °C and 0.32 MPa, respectively. In addition, an overall TBR ≥ 1.05 is required, which allow the self-sufficient production of tritium to sustain its own operation. It was found that the self-sufficient production of tritium was likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.
the tritium breeding area and the rib structure were designed to satisfy the operational temperature window of the tritium breeding material and structural materials. In the Computational Fluid Dynamics analysis, the outlet coolant temperature and pressure drop are 321 °C and 0.356 MPa for the blanket concept. From this result, we find the overall outlet coolant temperature to be in the acceptable range. 3.2. Robustness of the module structure
Acknowledgments This work was carried out in the framework of the activity of the Joint Special Design Team for Fusion DEMO and partly by the DEMO Design Activity under the Broader Approach. References [1] [2] [3] [4] [5] [6]
4. Discuss to improve tritium breeding ratio In the viewpoints of neutronics characteristics, it is proposed that the cooling water is changed from light water to heavy water in order to improve the tritium productivity. As a result of neutronics analysis, heavy water as cooling water improved the overall TBR by about 0.05 due to the (n, 2n) reaction of the deuterium and the less attenuation effect on the neutron energy compared to light water as shown in Fig. 7. It was found that the self-
[7] [8] [9] [10] [11] [12]
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