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Design and analysis of mock-up of CFETR water cooled ceramic blanket for neutronics experiment
Fusion Engineering and Design 144 (2019) 18–22
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Design and analysis of mock-up of CFETR water cooled ceramic blanket for neutronics experiment
T
⁎
Wuhui Chena,b, Qingjun Zhua, , Songlin Liua a b
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031, China University of Science and Technology of China, Hefei, Anhui, 230027, China
A R T I C LE I N FO
A B S T R A C T
Keywords: CFETR Neutronics experiment Mock-up WCCB blanket TPR
The water cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR). The neutronics experiment of the WCCB blanket is to validate the neutron transport codes and nuclear data libraries. A scaled down mock-up of the WCCB blanket should be designed considering constraints of economic costs and experimental conditions before the neutronics experiment. In this work, the mockup was designed keeping neutronics and geometric characteristics of the original equatorial outboard blanket module of CFETR. The radial arrangement of the mock-up was designed firstly, including tungsten armor, first wall, breeding zone 1, beryllium layer 1, cooling plate 1, breeding zone 2, beryllium layer 2, cooling plate 2. Then the toroidal and poloidal dimensions of mock-up were optimized. Five mock-ups with the same radial arrangement and different toroidal and poloidal dimensions were selected as initial mock-ups. The nuclear response of these five mock-ups, including neutron intensity, neutron spectra and tritium production rate were analyzed. After the optimization, the size of the experimental mock-up was determined as 20.00 cm (Toroidal) × 20.00 cm (Poloidal) × 17.27 cm (Radial). The analyzed results showed that the neutronics performance of the mock-up were similar with the real blanket module and it can be employed to carry out neutronics experiments.
1. Introduction China Fusion Engineering Test Reactor (CFETR) is an ITER-like tokamak. Its main purpose is to bridge the gap between ITER and DEMO, and to demonstrate the engineering feasibility of continuous and stable operation. As one of the candidate blankets for CFETR, the water cooled ceramic breeder (WCCB) blanket [1] employs tungsten as an armor to protect the blanket from plasma erosion and corrosion, Li2TiO3/Be12Ti mixed pebble beds as tritium breeder and neutron multipliers. Besides, there are two layers of beryllium pebble bed serving as additional neutron multipliers. Reduced activation ferritic martensitic (RAFM, e.g. CLAM) steel is used as the structural material. The nuclear analyses of the WCCB blanket depend on the neutron transport code (e.g. Monte Carlo code) and related nuclear database (e.g. FENDL3.0). It is essential to carry out neutronics integral experiments to validate the neutron transport codes and nuclear data libraries, as the code had statistical error and most of the data in data libraries were obtained by theoretical process. The blanket mock-up used in the neutronics experiment needs to be scaled down and simplified because it is difficult to provide a system simulating a huge complex tokamak-type reactor [2]. The basic dimensions of the test
⁎
module in the neutronics experiments carried out by Nakamura, et al. were 63 cm in diameter and 61 cm in length [3]. The breeding material was Li2O and the first wall was composed of SS316 stainless steel and polyethylene. Klix, et al. carried out neutronics experiments for DEMO blanket [4–7]. The thickness of the first wall F82H steel, Li2TiO3 (6Lienriched 40 and 95%), second F82H and beryllium layer were 16, 12, 3, and 200 mm respectively. The size of the mock-up was 500 × 500 mm2 in area and 231 mm in thickness [5]. In Europe, Batistoni, et al. carried out helium cooled pebble bed (HCPB) neutronics experiment based on the ITER test blanket mock-up. The external dimension of the experimental mock-up was 310 mm × 310 mm × 290 mm [8]. With the development of the research on CFETR, neutronics experiments are planned to be carried out for the WCCB blanket. However, the original WCCB blanket module needs to be scaled down before the neutronics experiments considering the constraints of economic costs and experimental conditions. According to the equatorial outboard blanket of WCCB for CFETR, the radial arrangement of the mock-up was designed firstly, then the toroidal and poloidal dimensions of the mock-up were optimized. The neutronics performance for the designed mock-up was analyzed using MCNP4C and the radioactivity of the irradiated mock-up was analyzed using FISPACT II.
Corresponding author. E-mail address: [email protected] (Q. Zhu).
https://doi.org/10.1016/j.fusengdes.2019.04.056 Received 16 October 2018; Received in revised form 21 March 2019; Accepted 12 April 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 144 (2019) 18–22
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Fig. 1. Typical WCCB (#3) blanket for CFETR [9].
2. Design of WCCB mock-up 2.1. WCCB #3 blanket module CFETR blanket system has 16 sectors and each sector accounts for 22.5°. Along the toroidal direction, each sector is divided into 2 inboard and 3 outboard blanket segments. Poloidally, there are five blanket modules in each outboard (blanket modules #1-#5) and inboard (blanket modules #6-#10) segment [1]. The radial arrangement of the WCCB blanket for CFETR was optimized based on the neutron wall loading. The peak neutron wall loading is in the equatorial outboard blanket in CFETR, namely the #3 blanket [9] as seen in Fig. 1. The #3 WCCB blanket is divided into 9 main components according to the material and function. They are plasma-facing component (PFC), first wall (FW), mixed breeding zones (BZ), beryllium zones (Be), cooling plates (CP), black plates (BP), side walls (SW), cover plates and radialpoloidal stiffening plates (rpSPs).
Fig. 2. TPR with different 6Li enrichment in breeding zone (#3 blanket).
2.2. Radial design In the design process, the experimental mock-up should reflect the basic design characteristics of the original reference blanket module (#3 blanket), including the configuration and materials. The mock-up should meet the following principles: (1) the influence of the mock-up's size on the measuring region should be lower than 5%; (2) the main neutronics characteristics of the measuring region should be similar to the reference blanket module. In addition, it's important to make reasonable use of the existing experimental conditions so as to save the experimental cost and improve feasibility. In order to design a mock-up which has the similar neutronics performance with the #3 blanket of WCCB blanket, the nuclear response of the #3 blanket to the DT neutron generator used in the neutronics experiment should be investigated. The neutronics parameters of the #3 blanket with different 6Li enrichment were calculated using 14 MeV neutron source. The distance between the source and the blanket module was set as 2 cm. The results in Fig. 2 showed that tritium were mainly produced in breeding zone 1 (BZ1) and breeding zone 2 (BZ2) with breeding material in different 6Li enrichment, so breeding zone 3 (BZ3) and breeding zone 4 (BZ4) were excluded in the design of the neutronics experimental mock-up. When the initial enriched lithium (80% 6Li) was used in the #3 blanket, the contribution ratio of tritium production rate (TPR) from the nuclides 6Li, 7Li and Be was 99.2%, 0.4% and 0.4% respectively. If the natural lithium was used in the #3 blanket, the TPR contribution ratio of the nuclides 6Li, 7Li and Be was 94.0%, 3.4% and 2.6% respectively. The results shown in Fig. 3 indicate the neutron spectrum with different lithium enrichment in the high energy region was similar. Also the neutronics experiment in Japan [7] showed that the TPR
Fig. 3. Neutron spectrum with different 6Li enrichment in the breeding zone (#3 blanket).
distribution within the natural Li2TiO3 layer changes moderately, and the uncertainty caused by natural 6Li content would be negligibly small. So the natural lithium was used in the designed mock-up. The simplified experimental mock-ups were already used in the existing experiments carried out in Japan and Europe [4,8,10,11], in which the independent beryllium layers and tritium breeding layers were arranged instead of mixed pebble beds. In the design process, the independent layers were also used. Li2TiO3 was used as tritium breeding material, beryllium as neutron multiplier and polyethylene as coolant for the same hydrogen atom ratio as H2O. The nuclide density along the radial direction of the mock-up is consistent with the original 19
Fusion Engineering and Design 144 (2019) 18–22
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Table 1 The size of mock-ups. Symbol
Cross section size (In toroidal and poloidal plane)
V1 V2 V3 V4 V5 INF
10 cm × 10 cm 20 cm × 20 cm 30 cm × 30 cm 40 cm × 40 cm 50 cm × 50 cm Unlimited (Reference mock-up)
#3 blanket of CFETR to ensure the similar neutron characteristics. After optimization, the radial layout of the mock-up is tungsten layer, first wall, breeding zone 1 (BZ1), beryllium layer 1 (Be1), cooling plate 1 (CP1), breeding zone 2 (BZ2), beryllium layer 2 (Be2), cooling layer 2 (CP2). And the thickness of each region in radial direction of the mockup is 2.0 mm, 20.0 mm, 7.6 mm, 50.0 mm, 11.0 mm, 11.1 mm, 60.0 mm, 11.0 mm.
Fig. 5. Neutron flux in the central axis along the radial direction of different mock-ups.
one centimeter diameter cylinder of the central axis of the mock-up was taken as the place of interest to analyze the neutron parameters. As seen in Fig. 5, the neutron flux drops more slowly as the distance away from the neutron source gets farther. The influence of the size of the mock-up on its nuclear response is defined as edge effect to choose the most appropriate mock-up. The mock-up with infinite cross section was selected as the reference mockup. The neutron flux in the axis along radial direction was set as the evaluation parameter, the following formula was used to quantitatively describe the edge effect of a mock-up.
2.3. Optimizing the poloidal-toroidal dimension of the Mock-up It is important to have a balance between the cost and quality of the neutronics experiment. So the mock-up needs to be optimized to select the suitable dimension after considering the limitations of the economic and experimental conditions. The initial five mock-ups with the same radial arrangement but different toroidal and poloidal cross sections were used as the solution set for the optimized mock-up. A model with an infinite toroidal and poloidal cross section was set as a reference mock-up. The size of the six mock-ups is listed in Table 1. These mock-ups were set in the experimental scheme seen in Fig. 4. The target of the DT neutron generator was placed in the center of the experimental hall. The size of the hall is 21.6 m (length) × 15.7 m (width) × 10.0 m (height). The thickness of the concrete wall is 2.0 m. The target tube is associated with a Au-Si barrier detector to monitor the neutron intensity. The angle between the target tube and associated particle tube is 135 ° . The direction of the target tube is perpendicular to the intersecting surface of the experimental mock-up. The 600 kV Cockcroft-Walton accelerator [12] constructed at China Institute of Atomic Energy (CIAE) will be used in the up-coming neutronics experiment on WCCB of CFETR. The neutron intensity is 3 × 1011 n/s in DC mode with 3% uncertainty. The detail information of source neutrons was simulated by the Physikalisch-Technische Bundesanstalt (PTB) target codes. Then the information was set as the source term of SDEF card of MCNP code due to the neutron flux and energy spectrum's dependence on emission angle. The statistical uncertainty of the simulated results was lower than 5%. In this paper, the neutronics performances, including the neutron flux, neutron spectra and TPR of these five mock-ups, were evaluated. A
Deviation =
|Flux i − Flux 0 | , i = 1, 2, 3, 4, 5 Flux i
(1)
Among them, Flux 0 is the neutron flux from the infinite reference mock-up. Flux i (i = 1,2, 3,4, 5) is the neutron flux of the mock-up with the cross sectional size 10 cm × 10 cm, 20 cm × 20 cm, 30 cm × 30 cm, 40 cm × 40 cm, 50 cm × 50 cm. Fig. 6 shows the deviation value between the mock-up with different cross sectional dimensions (limited) and the mock-up with unlimited cross section size. The compared result shows that as the distance from the source increases, the deviation value increases. The deviation value becomes smaller as the toroidal-poloidal cross section size of the mock-up increasing. The deviation value of V2 in the area before Be2 (radial distance, 10.17 cm–16.17 cm) is less than 5%. Apart from V1, the deviation values of another four mock-ups are less than 5% in BZ1 and BZ2. For the smallest mock-up, V1, the deviation value of neutron flux is from 0 to 31% assuming the mock-up with unlimited cross section size as the reference mock-up and the deviation value is higher than 5% when the distance is farther than 3.52 cm where Be1 locates. The neutron flux with different energy shown in Fig. 7 indicate
Fig. 4. Experiment scheme. 20
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Table 2 Neutron flux and TPR in BZ1 and BZ2 of different mock-ups. Mock-up
INF V1 V2 V3 V4 V5
BZ1 FLUX(1/cm2/sn)
TPR(T/sn)
BZ2 FLUX(1/cm2/sn)
TPR(T/sn)
8.11E-02 7.87E-02 8.06E-02 8.09E-02 8.10E-02 8.10E-02
4.39E-04 3.80E-04 4.18E-04 4.35E-04 4.38E-04 4.39E-04
1.64E-02 1.26E-02 1.58E-02 1.64E-02 1.64E-02 1.64E-02
2.04E-04 7.89E-05 1.62E-04 1.98E-04 2.04E-04 2.06E-04
experiment in Japan [6]. First, the neutron flux of the experimental mock-up was calculated using MCNP4C with FENDL 3.0 database. The calculated neutron flux and material of the mock-up were input into the activation code FISPACT II with EAF2010 database, then the activation and dose results were obtained. The mock-up was assumed to be cooled for 6 days after irradiation. Based on the above discussion, it is preferred to arrange the activation foil on the central axis of BZ1. According to the simulated neutron spectrum, the activation foil to be selected are, Au, In, Ni, Mg, Al, Ti, Fe, Sc, Zr, Cu, V. The radioactivity of the experimental mock-up is 1.45 × 1010 Bq after 8 h irradiation and 1.07 × 108 Bq after 5 days pulse irradiation in Fig. 9. The main short-lived radioactive nuclides are 6He (T1/ 8 9 13 N (T1/ 2 = 806.70 ms), Li (T1/2 = 838ms), Li (T1/2 = 178.30 ms), 16 N (T1/2 = 7.13 s). The radioactive nuclides decay ra2 = 9.96 ms), pidly due to the short half-life. The main radioactive nuclides of the mock-up during cooling phase are 48Sc (T1/2 = 43.67 h), 47Sc (T1/ 3 H (T1/2 = 12.33y), 46Sc (T1/2 = 83.79d), 45Ca (T1/ 2 = 3.35d), 48 Sc, 47Sc, 3H account for about 60.78%, 2 = 162.61d). Among them, 27.55% and 7.33% respectively. 48Sc is mainly from the reaction between neutron and 48Ti, which is one of the nuclide of breeding material.47Sc is mainly from three reactions, 47Ti (n, p) 47Sc, 48Ti (n,np) 47 Sc and 48Ti (n,d) 47Sc. The account ratio of each reaction is 50.42%, 30.63% and 18.24%. The reactions, 6Li (n,t) 3H, 7Li (n,na) 3H and 9Be (n,t) 3H are the way to produce tritium in the mock-up. The contact dose rate of the experimental mock-up reaches 0.12 Sv/ h during the pulse irradiation. It drops to 6.98 × 10−4 Sv/h because the main contributor 16N reduces during the 6 days' cooling time. After cooling, the main nuclides affected the contact dose rate are 48Sc (83.60%), 46Sc (14.27%), 47Sc (1.41%). These results will give essential data for the radiation protection in the neutronics experiment.
Fig. 6. Deviation value in the central axis along the radial direction of different mock-ups.
that apart from V1 mock-up, the rest four mock-ups have the similar neutron spectrum, especially in high energy region. The integral value of the neutron flux and TPR were calculated to analyze the edge effect. The results were shown in Table 2. The difference of the values between the 5 mock-ups and reference mock-up will be reduced with the increasing of the mock-up's toroidal-poloidal cross section. The discrepancy of the flux is lower than TPR in the same region. The differences in BZ2 are greater than BZ1 for the same neutronics parameters. In BZ1 region, the biggest deviation value for the neutron flux in V1 is 3.03% but it reaches 15.36% for TPR. The deviation values of these two parameters are both less than 5% of V2 in BZ1. In BZ2 region, the deviation value of the neutron flux and TPR of V1 are larger than 5%. The deviation value for neutron flux of V2 is also less than 5% but larger than 5% for TPR. So V2 mock-up can meet the requirement. The blanket material, especially beryllium, is expensive and the resource is limited. Considering the economic costs and environmental constraints, V2 model was chosen as the final mock-up to carry out the up-coming neutronics experiment.
3. The designed mock-up for neutronics experiment Before starting the experiment, the activation and dose rate information should be obtained to prepare appropriate activation foils, reaction channels and protection. The radial layout and material of the designed mock-up V2 were shown in Fig. 8 and Table 3 respectively. It was assumed to be irradiated under the 14 MeV neutron source with 3 × 1011 n/s intensity for 5 days and 8 h per day so that the total neutron can be up to 1 × 1016 source neutrons based on the neutronics
4. Conclusions The present study focused on the design of the scaled down experimental mock-up, which can be concluded as follows:
Fig. 7. Neutron spectrum of BZ1 and BZ2 in the central axis along the radial direction of different mock-ups. 21
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Fig. 8. Radial layout of the designed experimental mock-up.
balance between the neutronics performance and the experimental cost. The result showed that the mock-up with 20.00 cm side length was the smallest size which can satisfy the requirement. (3) The radioactivity and dose rate of the mock-up were calculated for the radiation protection. According to the results, the experimental mock-up needs to be cooled for more than 5 days if it was irradiated for 5 days. And the main radioactive product is scandium from the breeding material after cooling.
Table 3 Material composition and mass of experimental mock-up. Part Tungsten armor (PFC) First Wall (FW) Breeding zone (BZ) Neutron multiplier (Be) Cooling Plate (CP)
Material composition
Mass(kg) 3
100% W (19.25 g/cm ) Polyethylene (0.96 g/cm3)+Steel (7.8 g/ cm3) Li2TiO3 (natural lithium, 3.43 g/cm3) Be (1.848 g/cm3) Polyethylene (0.96 g/cm3)+Steel (7.8 g/ cm3)
1.553 5.45 2.565 8.131 5.973
Acknowledgments The authors wish to acknowledge the financial support of National special project for magnetic confined nuclear fusion energy with Grant No.2015GB108002 for this research. References [1] S. Liu, et al., Conceptual design of the water cooled ceramic breeder blanket for CFETR based on pressurized water cooled reactor technology, Fusion Eng. Des. 124 (2017). [2] H. Maekawa, M.A. Abdou, Summary of experiments and analyses from the JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics, Fusion Eng. Des. 28 (1995) 479–491. [3] T. Nakamura, M.A. Abdou, Summary of recent results from the JAERI/U. S. Fusion Neutronics phase I experiments, Fusion Technol. 10 (1986) 541–548. [4] K. Ochiai, et al., Neutronics Experiment of 6Li-enriched Breeding Blanket With Li2TiO3/ Be/F82H Assembly Using D-T Neutrons, (2001), pp. 1147–1150. [5] S. Sato, et al., Neutronics experiments for DEMO blanket at JAERI/FNS, Nucl. Fusion 43 (2003) 527–530. [6] A. Klix, et al., Heterogeneous breeding blanket experiment with lithium titanate and beryllium, Fusion Eng. Des. 72 (2005) 327–337. [7] K. Kondo, et al., Measurement of TPR distribution in natural Li 2 TiO 3 /Be assembly with DT neutrons, Fusion Eng. Des. 85 (2010) 1229–1233. [8] P. Batistoni, et al., Neutronics experiment on a helium cooled pebble bed (HCPB) breeder blanket mock-up, Fusion Eng. Des. 82 (2007) 2095–2104. [9] X. Zhang, et al., Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR, Plasma Sci. Technol. 19 (2017) 92–100. [10] I. Kodeli, "Deterministic 3D Transport, Sensitivity and Uncertainty Analysis of TPR and Reaction Rate Measurements in HCPB Breeder Blanket Mock-up Benchmark,", (2006). [11] K. Ochiai, et al., Neutron Flux Measurements in ITER-TBM Simulating Assemblies by Means of Multi-Foil Activation Method vol. 1, (2011), pp. 142–145. [12] B. Jie, Study on Neutronics Integral Experiment, Measurement of Neutron Leakage Energy Spectrum of Be Plate Samples (in Chinese), China Institute of Atomic Energy, 2006.
Fig. 9. Radioactivity and dose rate of the designed experimental mock-up.
(1) The radial layers were determined based on the TPR in different breeding zones of WCCB equatorial outboard blanket module. The final radial layers were arranged as tungsten (2.0 mm), first wall (20.0 mm), BZ1 (7.6 mm), Be1 (50.0 mm), CP1 (11.0 mm), BZ2 (11.1 mm), Be2 (60.0 mm) and CP2 (11.0 mm). (2) Five mock-ups with the same radial arrangement but different toroidal and poloidal cross sections were evaluated to find the 22
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Design and analysis of mock-up of CFETR water cooled ceramic blanket for neutronics experiment
T
⁎
Wuhui Chena,b, Qingjun Zhua, , Songlin Liua a b
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031, China University of Science and Technology of China, Hefei, Anhui, 230027, China
A R T I C LE I N FO
A B S T R A C T
Keywords: CFETR Neutronics experiment Mock-up WCCB blanket TPR
The water cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR). The neutronics experiment of the WCCB blanket is to validate the neutron transport codes and nuclear data libraries. A scaled down mock-up of the WCCB blanket should be designed considering constraints of economic costs and experimental conditions before the neutronics experiment. In this work, the mockup was designed keeping neutronics and geometric characteristics of the original equatorial outboard blanket module of CFETR. The radial arrangement of the mock-up was designed firstly, including tungsten armor, first wall, breeding zone 1, beryllium layer 1, cooling plate 1, breeding zone 2, beryllium layer 2, cooling plate 2. Then the toroidal and poloidal dimensions of mock-up were optimized. Five mock-ups with the same radial arrangement and different toroidal and poloidal dimensions were selected as initial mock-ups. The nuclear response of these five mock-ups, including neutron intensity, neutron spectra and tritium production rate were analyzed. After the optimization, the size of the experimental mock-up was determined as 20.00 cm (Toroidal) × 20.00 cm (Poloidal) × 17.27 cm (Radial). The analyzed results showed that the neutronics performance of the mock-up were similar with the real blanket module and it can be employed to carry out neutronics experiments.
1. Introduction China Fusion Engineering Test Reactor (CFETR) is an ITER-like tokamak. Its main purpose is to bridge the gap between ITER and DEMO, and to demonstrate the engineering feasibility of continuous and stable operation. As one of the candidate blankets for CFETR, the water cooled ceramic breeder (WCCB) blanket [1] employs tungsten as an armor to protect the blanket from plasma erosion and corrosion, Li2TiO3/Be12Ti mixed pebble beds as tritium breeder and neutron multipliers. Besides, there are two layers of beryllium pebble bed serving as additional neutron multipliers. Reduced activation ferritic martensitic (RAFM, e.g. CLAM) steel is used as the structural material. The nuclear analyses of the WCCB blanket depend on the neutron transport code (e.g. Monte Carlo code) and related nuclear database (e.g. FENDL3.0). It is essential to carry out neutronics integral experiments to validate the neutron transport codes and nuclear data libraries, as the code had statistical error and most of the data in data libraries were obtained by theoretical process. The blanket mock-up used in the neutronics experiment needs to be scaled down and simplified because it is difficult to provide a system simulating a huge complex tokamak-type reactor [2]. The basic dimensions of the test
⁎
module in the neutronics experiments carried out by Nakamura, et al. were 63 cm in diameter and 61 cm in length [3]. The breeding material was Li2O and the first wall was composed of SS316 stainless steel and polyethylene. Klix, et al. carried out neutronics experiments for DEMO blanket [4–7]. The thickness of the first wall F82H steel, Li2TiO3 (6Lienriched 40 and 95%), second F82H and beryllium layer were 16, 12, 3, and 200 mm respectively. The size of the mock-up was 500 × 500 mm2 in area and 231 mm in thickness [5]. In Europe, Batistoni, et al. carried out helium cooled pebble bed (HCPB) neutronics experiment based on the ITER test blanket mock-up. The external dimension of the experimental mock-up was 310 mm × 310 mm × 290 mm [8]. With the development of the research on CFETR, neutronics experiments are planned to be carried out for the WCCB blanket. However, the original WCCB blanket module needs to be scaled down before the neutronics experiments considering the constraints of economic costs and experimental conditions. According to the equatorial outboard blanket of WCCB for CFETR, the radial arrangement of the mock-up was designed firstly, then the toroidal and poloidal dimensions of the mock-up were optimized. The neutronics performance for the designed mock-up was analyzed using MCNP4C and the radioactivity of the irradiated mock-up was analyzed using FISPACT II.
Corresponding author. E-mail address: [email protected] (Q. Zhu).
https://doi.org/10.1016/j.fusengdes.2019.04.056 Received 16 October 2018; Received in revised form 21 March 2019; Accepted 12 April 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 144 (2019) 18–22
W. Chen, et al.
Fig. 1. Typical WCCB (#3) blanket for CFETR [9].
2. Design of WCCB mock-up 2.1. WCCB #3 blanket module CFETR blanket system has 16 sectors and each sector accounts for 22.5°. Along the toroidal direction, each sector is divided into 2 inboard and 3 outboard blanket segments. Poloidally, there are five blanket modules in each outboard (blanket modules #1-#5) and inboard (blanket modules #6-#10) segment [1]. The radial arrangement of the WCCB blanket for CFETR was optimized based on the neutron wall loading. The peak neutron wall loading is in the equatorial outboard blanket in CFETR, namely the #3 blanket [9] as seen in Fig. 1. The #3 WCCB blanket is divided into 9 main components according to the material and function. They are plasma-facing component (PFC), first wall (FW), mixed breeding zones (BZ), beryllium zones (Be), cooling plates (CP), black plates (BP), side walls (SW), cover plates and radialpoloidal stiffening plates (rpSPs).
Fig. 2. TPR with different 6Li enrichment in breeding zone (#3 blanket).
2.2. Radial design In the design process, the experimental mock-up should reflect the basic design characteristics of the original reference blanket module (#3 blanket), including the configuration and materials. The mock-up should meet the following principles: (1) the influence of the mock-up's size on the measuring region should be lower than 5%; (2) the main neutronics characteristics of the measuring region should be similar to the reference blanket module. In addition, it's important to make reasonable use of the existing experimental conditions so as to save the experimental cost and improve feasibility. In order to design a mock-up which has the similar neutronics performance with the #3 blanket of WCCB blanket, the nuclear response of the #3 blanket to the DT neutron generator used in the neutronics experiment should be investigated. The neutronics parameters of the #3 blanket with different 6Li enrichment were calculated using 14 MeV neutron source. The distance between the source and the blanket module was set as 2 cm. The results in Fig. 2 showed that tritium were mainly produced in breeding zone 1 (BZ1) and breeding zone 2 (BZ2) with breeding material in different 6Li enrichment, so breeding zone 3 (BZ3) and breeding zone 4 (BZ4) were excluded in the design of the neutronics experimental mock-up. When the initial enriched lithium (80% 6Li) was used in the #3 blanket, the contribution ratio of tritium production rate (TPR) from the nuclides 6Li, 7Li and Be was 99.2%, 0.4% and 0.4% respectively. If the natural lithium was used in the #3 blanket, the TPR contribution ratio of the nuclides 6Li, 7Li and Be was 94.0%, 3.4% and 2.6% respectively. The results shown in Fig. 3 indicate the neutron spectrum with different lithium enrichment in the high energy region was similar. Also the neutronics experiment in Japan [7] showed that the TPR
Fig. 3. Neutron spectrum with different 6Li enrichment in the breeding zone (#3 blanket).
distribution within the natural Li2TiO3 layer changes moderately, and the uncertainty caused by natural 6Li content would be negligibly small. So the natural lithium was used in the designed mock-up. The simplified experimental mock-ups were already used in the existing experiments carried out in Japan and Europe [4,8,10,11], in which the independent beryllium layers and tritium breeding layers were arranged instead of mixed pebble beds. In the design process, the independent layers were also used. Li2TiO3 was used as tritium breeding material, beryllium as neutron multiplier and polyethylene as coolant for the same hydrogen atom ratio as H2O. The nuclide density along the radial direction of the mock-up is consistent with the original 19
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W. Chen, et al.
Table 1 The size of mock-ups. Symbol
Cross section size (In toroidal and poloidal plane)
V1 V2 V3 V4 V5 INF
10 cm × 10 cm 20 cm × 20 cm 30 cm × 30 cm 40 cm × 40 cm 50 cm × 50 cm Unlimited (Reference mock-up)
#3 blanket of CFETR to ensure the similar neutron characteristics. After optimization, the radial layout of the mock-up is tungsten layer, first wall, breeding zone 1 (BZ1), beryllium layer 1 (Be1), cooling plate 1 (CP1), breeding zone 2 (BZ2), beryllium layer 2 (Be2), cooling layer 2 (CP2). And the thickness of each region in radial direction of the mockup is 2.0 mm, 20.0 mm, 7.6 mm, 50.0 mm, 11.0 mm, 11.1 mm, 60.0 mm, 11.0 mm.
Fig. 5. Neutron flux in the central axis along the radial direction of different mock-ups.
one centimeter diameter cylinder of the central axis of the mock-up was taken as the place of interest to analyze the neutron parameters. As seen in Fig. 5, the neutron flux drops more slowly as the distance away from the neutron source gets farther. The influence of the size of the mock-up on its nuclear response is defined as edge effect to choose the most appropriate mock-up. The mock-up with infinite cross section was selected as the reference mockup. The neutron flux in the axis along radial direction was set as the evaluation parameter, the following formula was used to quantitatively describe the edge effect of a mock-up.
2.3. Optimizing the poloidal-toroidal dimension of the Mock-up It is important to have a balance between the cost and quality of the neutronics experiment. So the mock-up needs to be optimized to select the suitable dimension after considering the limitations of the economic and experimental conditions. The initial five mock-ups with the same radial arrangement but different toroidal and poloidal cross sections were used as the solution set for the optimized mock-up. A model with an infinite toroidal and poloidal cross section was set as a reference mock-up. The size of the six mock-ups is listed in Table 1. These mock-ups were set in the experimental scheme seen in Fig. 4. The target of the DT neutron generator was placed in the center of the experimental hall. The size of the hall is 21.6 m (length) × 15.7 m (width) × 10.0 m (height). The thickness of the concrete wall is 2.0 m. The target tube is associated with a Au-Si barrier detector to monitor the neutron intensity. The angle between the target tube and associated particle tube is 135 ° . The direction of the target tube is perpendicular to the intersecting surface of the experimental mock-up. The 600 kV Cockcroft-Walton accelerator [12] constructed at China Institute of Atomic Energy (CIAE) will be used in the up-coming neutronics experiment on WCCB of CFETR. The neutron intensity is 3 × 1011 n/s in DC mode with 3% uncertainty. The detail information of source neutrons was simulated by the Physikalisch-Technische Bundesanstalt (PTB) target codes. Then the information was set as the source term of SDEF card of MCNP code due to the neutron flux and energy spectrum's dependence on emission angle. The statistical uncertainty of the simulated results was lower than 5%. In this paper, the neutronics performances, including the neutron flux, neutron spectra and TPR of these five mock-ups, were evaluated. A
Deviation =
|Flux i − Flux 0 | , i = 1, 2, 3, 4, 5 Flux i
(1)
Among them, Flux 0 is the neutron flux from the infinite reference mock-up. Flux i (i = 1,2, 3,4, 5) is the neutron flux of the mock-up with the cross sectional size 10 cm × 10 cm, 20 cm × 20 cm, 30 cm × 30 cm, 40 cm × 40 cm, 50 cm × 50 cm. Fig. 6 shows the deviation value between the mock-up with different cross sectional dimensions (limited) and the mock-up with unlimited cross section size. The compared result shows that as the distance from the source increases, the deviation value increases. The deviation value becomes smaller as the toroidal-poloidal cross section size of the mock-up increasing. The deviation value of V2 in the area before Be2 (radial distance, 10.17 cm–16.17 cm) is less than 5%. Apart from V1, the deviation values of another four mock-ups are less than 5% in BZ1 and BZ2. For the smallest mock-up, V1, the deviation value of neutron flux is from 0 to 31% assuming the mock-up with unlimited cross section size as the reference mock-up and the deviation value is higher than 5% when the distance is farther than 3.52 cm where Be1 locates. The neutron flux with different energy shown in Fig. 7 indicate
Fig. 4. Experiment scheme. 20
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Table 2 Neutron flux and TPR in BZ1 and BZ2 of different mock-ups. Mock-up
INF V1 V2 V3 V4 V5
BZ1 FLUX(1/cm2/sn)
TPR(T/sn)
BZ2 FLUX(1/cm2/sn)
TPR(T/sn)
8.11E-02 7.87E-02 8.06E-02 8.09E-02 8.10E-02 8.10E-02
4.39E-04 3.80E-04 4.18E-04 4.35E-04 4.38E-04 4.39E-04
1.64E-02 1.26E-02 1.58E-02 1.64E-02 1.64E-02 1.64E-02
2.04E-04 7.89E-05 1.62E-04 1.98E-04 2.04E-04 2.06E-04
experiment in Japan [6]. First, the neutron flux of the experimental mock-up was calculated using MCNP4C with FENDL 3.0 database. The calculated neutron flux and material of the mock-up were input into the activation code FISPACT II with EAF2010 database, then the activation and dose results were obtained. The mock-up was assumed to be cooled for 6 days after irradiation. Based on the above discussion, it is preferred to arrange the activation foil on the central axis of BZ1. According to the simulated neutron spectrum, the activation foil to be selected are, Au, In, Ni, Mg, Al, Ti, Fe, Sc, Zr, Cu, V. The radioactivity of the experimental mock-up is 1.45 × 1010 Bq after 8 h irradiation and 1.07 × 108 Bq after 5 days pulse irradiation in Fig. 9. The main short-lived radioactive nuclides are 6He (T1/ 8 9 13 N (T1/ 2 = 806.70 ms), Li (T1/2 = 838ms), Li (T1/2 = 178.30 ms), 16 N (T1/2 = 7.13 s). The radioactive nuclides decay ra2 = 9.96 ms), pidly due to the short half-life. The main radioactive nuclides of the mock-up during cooling phase are 48Sc (T1/2 = 43.67 h), 47Sc (T1/ 3 H (T1/2 = 12.33y), 46Sc (T1/2 = 83.79d), 45Ca (T1/ 2 = 3.35d), 48 Sc, 47Sc, 3H account for about 60.78%, 2 = 162.61d). Among them, 27.55% and 7.33% respectively. 48Sc is mainly from the reaction between neutron and 48Ti, which is one of the nuclide of breeding material.47Sc is mainly from three reactions, 47Ti (n, p) 47Sc, 48Ti (n,np) 47 Sc and 48Ti (n,d) 47Sc. The account ratio of each reaction is 50.42%, 30.63% and 18.24%. The reactions, 6Li (n,t) 3H, 7Li (n,na) 3H and 9Be (n,t) 3H are the way to produce tritium in the mock-up. The contact dose rate of the experimental mock-up reaches 0.12 Sv/ h during the pulse irradiation. It drops to 6.98 × 10−4 Sv/h because the main contributor 16N reduces during the 6 days' cooling time. After cooling, the main nuclides affected the contact dose rate are 48Sc (83.60%), 46Sc (14.27%), 47Sc (1.41%). These results will give essential data for the radiation protection in the neutronics experiment.
Fig. 6. Deviation value in the central axis along the radial direction of different mock-ups.
that apart from V1 mock-up, the rest four mock-ups have the similar neutron spectrum, especially in high energy region. The integral value of the neutron flux and TPR were calculated to analyze the edge effect. The results were shown in Table 2. The difference of the values between the 5 mock-ups and reference mock-up will be reduced with the increasing of the mock-up's toroidal-poloidal cross section. The discrepancy of the flux is lower than TPR in the same region. The differences in BZ2 are greater than BZ1 for the same neutronics parameters. In BZ1 region, the biggest deviation value for the neutron flux in V1 is 3.03% but it reaches 15.36% for TPR. The deviation values of these two parameters are both less than 5% of V2 in BZ1. In BZ2 region, the deviation value of the neutron flux and TPR of V1 are larger than 5%. The deviation value for neutron flux of V2 is also less than 5% but larger than 5% for TPR. So V2 mock-up can meet the requirement. The blanket material, especially beryllium, is expensive and the resource is limited. Considering the economic costs and environmental constraints, V2 model was chosen as the final mock-up to carry out the up-coming neutronics experiment.
3. The designed mock-up for neutronics experiment Before starting the experiment, the activation and dose rate information should be obtained to prepare appropriate activation foils, reaction channels and protection. The radial layout and material of the designed mock-up V2 were shown in Fig. 8 and Table 3 respectively. It was assumed to be irradiated under the 14 MeV neutron source with 3 × 1011 n/s intensity for 5 days and 8 h per day so that the total neutron can be up to 1 × 1016 source neutrons based on the neutronics
4. Conclusions The present study focused on the design of the scaled down experimental mock-up, which can be concluded as follows:
Fig. 7. Neutron spectrum of BZ1 and BZ2 in the central axis along the radial direction of different mock-ups. 21
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Fig. 8. Radial layout of the designed experimental mock-up.
balance between the neutronics performance and the experimental cost. The result showed that the mock-up with 20.00 cm side length was the smallest size which can satisfy the requirement. (3) The radioactivity and dose rate of the mock-up were calculated for the radiation protection. According to the results, the experimental mock-up needs to be cooled for more than 5 days if it was irradiated for 5 days. And the main radioactive product is scandium from the breeding material after cooling.
Table 3 Material composition and mass of experimental mock-up. Part Tungsten armor (PFC) First Wall (FW) Breeding zone (BZ) Neutron multiplier (Be) Cooling Plate (CP)
Material composition
Mass(kg) 3
100% W (19.25 g/cm ) Polyethylene (0.96 g/cm3)+Steel (7.8 g/ cm3) Li2TiO3 (natural lithium, 3.43 g/cm3) Be (1.848 g/cm3) Polyethylene (0.96 g/cm3)+Steel (7.8 g/ cm3)
1.553 5.45 2.565 8.131 5.973
Acknowledgments The authors wish to acknowledge the financial support of National special project for magnetic confined nuclear fusion energy with Grant No.2015GB108002 for this research. References [1] S. Liu, et al., Conceptual design of the water cooled ceramic breeder blanket for CFETR based on pressurized water cooled reactor technology, Fusion Eng. Des. 124 (2017). [2] H. Maekawa, M.A. Abdou, Summary of experiments and analyses from the JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics, Fusion Eng. Des. 28 (1995) 479–491. [3] T. Nakamura, M.A. Abdou, Summary of recent results from the JAERI/U. S. Fusion Neutronics phase I experiments, Fusion Technol. 10 (1986) 541–548. [4] K. Ochiai, et al., Neutronics Experiment of 6Li-enriched Breeding Blanket With Li2TiO3/ Be/F82H Assembly Using D-T Neutrons, (2001), pp. 1147–1150. [5] S. Sato, et al., Neutronics experiments for DEMO blanket at JAERI/FNS, Nucl. Fusion 43 (2003) 527–530. [6] A. Klix, et al., Heterogeneous breeding blanket experiment with lithium titanate and beryllium, Fusion Eng. Des. 72 (2005) 327–337. [7] K. Kondo, et al., Measurement of TPR distribution in natural Li 2 TiO 3 /Be assembly with DT neutrons, Fusion Eng. Des. 85 (2010) 1229–1233. [8] P. Batistoni, et al., Neutronics experiment on a helium cooled pebble bed (HCPB) breeder blanket mock-up, Fusion Eng. Des. 82 (2007) 2095–2104. [9] X. Zhang, et al., Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR, Plasma Sci. Technol. 19 (2017) 92–100. [10] I. Kodeli, "Deterministic 3D Transport, Sensitivity and Uncertainty Analysis of TPR and Reaction Rate Measurements in HCPB Breeder Blanket Mock-up Benchmark,", (2006). [11] K. Ochiai, et al., Neutron Flux Measurements in ITER-TBM Simulating Assemblies by Means of Multi-Foil Activation Method vol. 1, (2011), pp. 142–145. [12] B. Jie, Study on Neutronics Integral Experiment, Measurement of Neutron Leakage Energy Spectrum of Be Plate Samples (in Chinese), China Institute of Atomic Energy, 2006.
Fig. 9. Radioactivity and dose rate of the designed experimental mock-up.
(1) The radial layers were determined based on the TPR in different breeding zones of WCCB equatorial outboard blanket module. The final radial layers were arranged as tungsten (2.0 mm), first wall (20.0 mm), BZ1 (7.6 mm), Be1 (50.0 mm), CP1 (11.0 mm), BZ2 (11.1 mm), Be2 (60.0 mm) and CP2 (11.0 mm). (2) Five mock-ups with the same radial arrangement but different toroidal and poloidal cross sections were evaluated to find the 22