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Study on lithium rod test module and irradiation method for tritium production using high temperature gas-cooled reactor
Fusion Engineering and Design xxx (xxxx) xxx–xxx
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Study on lithium rod test module and irradiation method for tritium production using high temperature gas-cooled reactor ⁎
Yuki Kogaa, , Hideaki Matsuuraa, Yuma Idaa, Ryo Okamotoa, Kazunari Katayamab, Teppei Otsukac, Minoru Gotod, Shigeaki Nakagawad, Satoru Nagasumid, Etsuo Ishitsukad, Yosuke Shimazakid a
Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka Nishi-ku, Fukuoka-shi, Fukuoka, 319-0395 Japan Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasugakouen Kasuga-shi Fukuoka, 816-0811 Japan c Department of Electrical and Electrtonic Engineering, Kindai University, 3-4-1 Kowakae, HigashiOsaka city Osaka, 577-0818 Japan d Japan Atomic Energy Agency, 4002 Narita-cho Oarai-machi Higashi-Ibaraki-gun, Ibaraki, 311-1393 Japan b
A R T I C LE I N FO
A B S T R A C T
Keywords: Fusion reactor Tritium production High temperature gas-cooled reactor Li rod HTTR Irradiation test
Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by 6Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO2, alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod’s tritium production and containment performance was presented.
1. Introduction Fusion reactor R&D has advanced to full-scale validation of both nuclear burning and reactor engineering. A fusion reactor is assumed to use T(d,n)α reaction. Deuterium exists in nature and can be produced from water; however, tritium is a very rare radioactive nuclide, so it should be produced artificially. A fusion reactor with 3 GW thermal output power must burn approximately 400 g of tritium in a day and more tritium is needed in the tritium circulation system and initial tritium inventory. Furthermore, technical tests for tritium circulation and blanket system must be conducted before starting up the DEMO fusion reactor. Because of these reasons, it is necessary to have sufficient amounts of tritium. For example, it was reported that the initial tritium amount required for DEMO was several 100 g ∼ approximately 20 kg [1,2]. We proposed a tritium production method that worked by loading a Li compound as a burnable poison (BP) into a high temperature gas-cooled reactor (HTGR) [3]. The HTGR is a fourth generation reactor types [4]. High outlet coolant temperatures enable the HTGR to operate with high-efficiency and allows hydrogen production. A HTGR has the following tritium production advantages. First, the HTGR moderator/ structural material consists of graphite and its coolant is He gas, both of
⁎
which are chemically stable and do not significantly react with Li compounds. When we consider tritium production in a light water reactor (LWR), separating tritium that has leaked into water is difficult. Second, a large amount of Li compounds without a 6Li concentration can be loaded. This is because a HTGR reactor core is larger than those of other fission reactor types with the same thermal output (e.g., LWR or fast breeder reactors). Third, it is easily to replace B4C with a Li compound without significantly changing the structural core design, and it may be possible to replace all the burnable poison (BP) depending on the core design. Furthermore, B4C pellets as BP can be loaded into a core using a separate module from the fuel rod; therefore, tritium produced from a Li compound can be recovered without polluting the reactor. Fourth, as HTGRs are generally designed to have high burnup, their nuclear transformation performance tends to better than low burnup reactors (e.g., LWRs). In the first study we considered the use of Li particles covered with layers of carbon, PyC (pyrolytic carbon), and SiC [3]. A tritium diffusion simulation using permeability data obtained by Katayama et al. [5] revealed that most of the tritium was lost from the Li particle due to its large surface area. We continued the study on the Li rod, assuming a cylindrical rod based on the design used in the Watt Bar nuclear plant [6]. We performed the core calculation in the case that Li rods were
Corresponding author at: Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka Nishi-ku, Fukuoka-shi, Fukuoka, 319-0395 Japan. E-mail address: [email protected] (Y. Koga).
https://doi.org/10.1016/j.fusengdes.2018.03.029 Received 29 September 2017; Received in revised form 26 February 2018; Accepted 12 March 2018 0920-3796/ © 2018 Elsevier B.V. All rights reserved.
Please cite this article as: Koga, Y., Fusion Engineering and Design (2018), https://doi.org/10.1016/j.fusengdes.2018.03.029
Fusion Engineering and Design xxx (xxxx) xxx–xxx
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loaded in BP holes of the Gas Turbine High Temperature Reactor 300 (GTHTR300), [7] a commercial HTGR designed by the Japan Atomic Energy Agency. The simulation showed that 500–800 g of tritium could be produced during one year of operation and that the tritium leakage from the Li rods could be suppressed to less than 1% of the amount of the tritium produced [8]. In the simulation we assumed a temperature of 800 K at the Li rod region; however, from an electric power generation efficiency perspective, it is important to prevent tritium leakage from the rod in lower levels at a higher temperature range, e.g., 1100–1200 K. It is possible to reduce tritium leakage by adding Zr layers [9] for high temperature operation. To advance to the new phase, it is necessary to demonstrate tritium production and confirm containment performance. We considered a demonstration test using irradiation facilities or the High Temperature engineering Test Reactor (HTTR) [10]. In this paper, we designed a Li rod suitable for use in the HTTR, and exhibited an irradiation test scenario in the HTTR.
nuclear burning calculation using the Monte Carlo neutron transport code MVP-BURN [12] and JENDL-4.0 [13]. We assumed a HTTR core system with Li rods loaded into all BP holes and a one-year operation period. We set the Li rod temperature to 1170 K and a thermal output to 30 MW. This rod is used in a Li-rod capsule and irradiated in the irradiation test. On the other hand, to design Li-rod capsules for the irradiation test, we assumed a 30 day the HTTR operation with an irradiation column neutron flux of 4.2 × 1017 n/m2s [14]. We set the maximum temperature at 1070 K. We evaluated the neutron flux in the LiAlO2 in the Li-rod capsule using this irradiation flux and statistical processing by MVP code [12] and JENDL-4.0 [13]. We assumed that the Li-rod capsule was irradiated uniformly in the irradiation model that has a vacuum system with white reflector boundary. In this model, 2,160,000 neutrons are generated in the vacuum area as a fixed source. We evaluated one-day tritium production from the LiAlO2 in the Li-rod capsule using the following equation:
2. Calculation and evaluation models
where Nt, N6Li, ϕ, σ6Li+n, and t are the number of T, number of 6Li atoms, the neutron flux in LiAlO2, the microscopic cross-section of the 6 Li + n reaction, and irradiation time, respectively.
Nt = N6Li [1 − exp(−ϕσ6Li + nt )],
2.1. Irradiation equipment in the HTTR
(1)
2.3. Tritium diffusion model
Fig. 1 shows a schematic diagram of the HTTR core. The HTTR has a thermal output of 30 MW and an average moderator temperature of 1170 K. We conduct the irradiation test in a graphite hexagonal column with a 360 mm facing distance and a height of 580 mm. The height and diameter of the Li rods are 450 mm and 14–30 mm, respectively. Fig. 2(a) and (b) show example Li rods. One Li rod consists of a 1–20 mm alumina layer, two 0.02–1.0 mm Zr layers, a 0.5–4.0 mm LiAlO2 layer as a Li compound with a theoretical density of 85% [11], and a hollow portion. These layers have a cylindrical configuration and the design of their thicknesses are discussed in Section 3.1. We also prepare a Li rod without the Zr layers to confirm the effect of leakage prevention due to the Zr layers. The irradiation capsules composing the Li rods have a cylindrical configuration with a diameter of 120 mm, a height of 480 mm, and a 20-mm thick alumina cover, (inner radius is 40 mm) to prevent tritium leakage (Fig. 2c). He gas is circulated through the inner or outer hollow portion in the Li-rod capsule to recover tritium in the real test, but we did not consider this in this paper. This mechanism is set to be designed after determining the experiment execution. The irradiation test is conducted in a graphite irradiation block because aside from its external form, the irradiation block can be designed at will and is capable of having other devices set on it to maintain temperature and measure and recover tritium.
We used the tritium diffusion equation in a cylindrical coordinate system to evaluate tritium leakage and tritium absorption by Zr.
∂C D ∂ ⎛ ∂C ⎞ = r , ∂t r ∂r ⎝ ∂r ⎠
(2)
where c, D, and r are the tritium concentration, diffusion coefficient, and radius direction distance, respectively. We calculated the tritium leakage amount from alumina using the steady solution:
G=
2πhK P (mol/ s ), ln r2 − ln r1
(3)
where r1, r2, h, K, and P are the inner radius and outer radius of the alumina layer, the height of the alumina layer, the permeability coefficient, and the partial pressure in the hollow portion, respectively. We evaluated tritium absorbed into Zr using the cylindrical transient diffusion Eq. (2) and Sievert’s law:
C = S P.
(4)
We assumed that the Zr layers absorbed tritium quickly and that tritium partial pressure decreased. Fig. 3 shows the tritium leakage simulation model for the Li-rod capsule. For conservative, we evaluated tritium leakage from the Li rod by the tritium partial pressure in the inner hollow portion. Tritium leakage from the Li rod determines tritium partial pressure in the outer
2.2. Nuclear calculation To design a Li rod suited for use in the HTTR, we conducted a
Fig. 1. Schematic diagram of reactor core of the HTTR.
2
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Fig. 2. Schematic diagram of Li rods and rod cover.
3. Result and discussion 3.1. Evaluation of tritium production and leakage We conducted a nuclear burning calculation to determine an outer radius and a thickness of LiAlO2 in Li rod suited for use in the HTTR. BP holes in the HTTR have a height of 500 mm and a radius of 7.5 mm. We set the Li rod radius to 7.0 mm, it is the original BP design of the HTTR, and set its height to 480 mm. We fixed the outer radius of the LiAlO2 and the alumina layer thicknesses to 5.0 mm and 2.0 mm. Fig. 4 shows tritium production and the effective multiplication factor after a 360day operation. The amount of tritium produced is proportional to the LiAlO2 layer thickness, because the LiAlO2 thickness increases the amount of loaded 6Li. The effective multiplication factor decreases with increasing tritium production. It was revealed that Li rods with 3.0 mm LiAlO2 layers could produce 30.9 g of tritium and had an effective multiplication factor of 1.02 at the end of operation. We achieved 30 g of tritium production per year in the following simulations and designed the outer radius and thickness of LiAlO2 in the standard rod to be 5.0 mm and 3.0 mm, respectively. Because the tritium production amount and the effective multiplication factor are not influenced by alumina layer thickness and the existence of Zr layers, we carried out the diffusion and absorption calculations to determine alumina and Zr layer thicknesses assuming a constant tritium production value. Fig. 5 shows the alumina layer thickness transition of the tritium leakage rate from the alumina and Zr layers. In the previous study, the total amount of tritium that leaked from the rod was estimated to be almost 1% of the produced amount under low-temperature conditions [8]. Thus, in this paper, we also
Fig. 3. Schematic diagram of tritium leakage from the capsule.
hollow portion and tritium leakage through the alumina cover each second. We calculated the amount of tritium absorbed into the Zr in the Li rod at each second before the tritium leakage calculation. In the diffusion and absorption calculation about Li rod, we considered the calculation of the inner hollow portion, the Zr layers and the alumina layer. We considered the decrease of tritium leakage from the Li-rod capsule by tritium recovery using He circulation about the capsules, especially for the capsules without Zr because a large quantity of tritium leaks from them at high temperature. In this calculation, we determined the ratio of tritium recovered by flowing He at each second. In addition to leakage from the side wall, we considered leakage from the upper and lower covers, but these were much smaller than leakage from the side wall. The alumina transmission coefficient for tritium is
K = 1.96 × 10−12 exp(− 4.55 × 10 4 RT ) [mol/m1 2⋅kg1 2] ,
(5)
where R and T are the gas constant and temperature, respectively. The alumina transmission coefficient for tritium was obtained by Katayama et al. [5], and the Zr diffusion coefficient and solubility product for hydrogen were obtained by Okamoto. et al. [15]. The calculation for tritium absorption into Zr can be estimated based on the assumption that the S and D for tritium are approximately the same as those of hydrogen. From the viewpoint of tritium pollution in the HTTR reactor in the irradiation test, we set the criterion of tritium leakage from Li-rod capsules to 0.1 μg, which is the peak tritium amount in the HTTR primary helium gas [16].
Fig. 4. Tritium production and effective multiplication factor after 360-day operation from the thickness of LiAlO2.
3
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Fig. 5. Alumina layer thickness transition of the tritium leakage rate with and without Zr layers.
Fig. 8. Amount of tritium leakage by irradiation time in capsules C and D.
carried out simulations to suppress the amount of tritium that leaked from the Li rods to less than 1%. For this purpose, we included two 0.1mm thick Zr layers in the rod. The Zr layer’s tritium absorption speed is not proportional to its thickness but the surface area; amount of the two Zr layers in the Li rod was sufficient to absorb tritium. We designed the alumina and Zr layer thicknesses to be 2.0 mm and 0.1 mm, respectively, and this rod was defined as the standard rod. It can produce 30.9 g of tritium per year with an anount of leaked tritium of less than 1%.
Table 1 Design and expected experimental values of Li-rod capsules. Capsule
A
B
C
D
Hollow radius(mm) Zr (mm) LiAlO2 (mm) Alumina (mm) T2 production (g) T2 leakage (g) Inner pressurea (Pa)
1.9 0.1 3.0 1.9 3.17 × 10−3 3.57 × 10−7 1.87 × 10−1
3.4 0.1 1.5 1.9 2.32 × 10−3 2.16 × 10−7 0
2.0 – 3.0 2.0 3.17 × 10−3 6.45 × 10−4 3.68 × 105
2.0 – 3.0 10 3.18 × 10−3 2.10 × 10−4 4.34 × 105
3.2. Irradiation test a
Estimated value.
In the irradiation test, we intended to confirm the standard rod’s performance (capsule A) by looking at tritium production and the containment property. At the same time, we wanted to confirm how performance was influenced by LiAlO2, Zr, and alumina layer thicknesses. We designed capsules B, C, and D for the comparison experiment to confirm these performances in experiments I, II, and III. Table 1 shows the designs and expected experimental values of capsules A, B, C, and D. We evaluated Li-rod capsules for three experiments considering tritium production or leakage by changing Zr, LiAlO2, and alumina layer thicknesses. We assumed that He gas was circulated through the outer hollow portion of the capsules to recover and measure tritium leakage. 3.2.1. Experiment I In experiment I, we confirm the change of tritium production due to LiALO2 thickness and the tritium containment performance of Zr layers. We compare the amount of tritium produced and leaked from capsules A and B. The LiALO2 layer thickness in capsule B is half of that in capsule A and capsule B’s tritium production is about 73% of that of capsule A. The amount of tritium produced in capsules A and B can be measured using Zr (as a sum of the absorbed in it) and He gas (leaked tritium). We estimated the tritium leakage from the rods in capsules A and B, and the results are shown in Fig. 6 as a function of irradiation time. We evaluated the tritium leakage from the Li rod in capsule B as 61% of that of capsule A.
Fig. 6. Tritium leakage by irradiation time in capsules A and B.
3.2.2. Experiment II We confirm the tritium leakage decrease due to the use of the Zr in experiment II. We measure and compare tritium leakage from capsules A and C, and the results of which are shown in Fig. 7 as a function of irradiation time. We expected that the tritium leakage of capsule A was about 0.055% of that of capsule C. Fig. 7. Tritium leakage by irradiation time course about capsule A and C.
3.2.3. Experiment III In experiment III we confirm the tritium containment performance of the alumina layer. We irradiate capsules C and D to compare tritium 4
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Fig. 9. Schematic diagram of the irradiation block.
tritium production and leakage. In experiment II, we confirm the tritium leakage decrease due to the effect of Zr layers by tritium leakage. In experiment III, we measure tritium leakage to confirm the effects of the alumina layer. In the irradiation test, we can not confirm the effect of the Zr phase change at more than 1130 K because the irradiation region’s maximum temperature is 1070 K. Tritium leakage from all the tested capsules exceeds the leakage criterion, thus we must consider ways to circulate He through the capsules to recover the tritium. We also conduct other countermeasures for all the capsules to decrease tritium leakage as low as reasonably achievable. The present irradiation test is important for establishing a technique to produce 100 g tritium orders in Japan.
leakage and examine alumina layer performance in preventing tritium leakage. The alumina layer in capsule D is five times thicker than in capsule C. Fig. 8 shows the tritium leakage of capsules C and D as a function of irradiation time, and shows that the tritium leakage from the rod in capsule D is 33% of that in capsule C at 30 days of irradiation time. 3.2.4. Irradiation block The rod in capsule A can produce 30.9 g of tritium in a year if loaded into all BP holes in the HTTR. This tritium production is not simply from the tritium production in capsule A seen in the irradiating test in experiment I. This is because the considered irradiation region is the edge of HTTR core, which has a different condition from that of the BP holes in the core’s center region. An irradiation block with four holes is needed because four Li-rod capsules should be irradiated simultaneously in one irradiation block. Therefore, we designed an irradiation block as shown in Fig. 9. The tritium leakage from all the capsules exceeds the 0.1 μg tritium leakage criterion. Tritium leakage can be decreased sufficiently below this criterion sufficiently using a He circulation system flow of 132 L/min.
Acknowledgments This work is supported by JSPS KAKENHI Grant-in Aid for Scientific Research (B) JP15H04230. References [1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13] [14] [15] [16]
4. Concluding remarks We designed a Li rod suited for use in the HTTR that produced 30 g of tritium in a year of operation, and we proposed a Li rod irradiation test to demonstrate tritium production and confirm tritium containment performance. We exhibited four Li-rod capsule designs, the irradiation condition, and the irradiation block to test tritium production and leakage in the HTTR. In addition, we evaluated the expected data, amount of tritium produced and leaked, and the inner pressure during the test at the irradiation region in the HTTR. The irradiation block has four holes, allowing irradiation of four capsules simultaneously. In experiment I, we confirm the effects of the LiAlO2 and Zr layers by the
5
R. Hiwatari, to be published in J Atomic Energy Society Japan [in Japanese]. M. Nishikawa, et al., Fusion Eng. Des. 87 (2012) 466–470. H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95–101. https://www.gen-4. org/gif/jcms/c_9334/origins (GEN IV international Forum). K. Katayama, et al., Fusion Sci Tech. 68 (3) (2015) 662–668. K.A. Burn, et al., (2012). TTQP-1-015, 19, PNNL-22086. T. Nakata, et al., JAERI-Tech (2002) 2002-066 (in Japanese). H. Nakaya, et al., Nucl. Eng. Des. 292 (2015) 277–282. S. Nagasumi, et al., Fusion Sci tech. 72 (2017) 753–759. S. Saito, et al., JAERI-1332 (1994). Y. Kawamura, et al., Journal of Nucl. Sci Tech. 29 (5) (1992) 436–444. Y. Nagaya, et al., JAERI 1348 (2005). K. Shibata, et al., J. Nucl Sci. Tech. 48 (1) (2011) 1–30. T. Shibata, et al., JAERI-Tech (2002) 2002-097 (in Japanese). R. Okamoto, et al., Fall Meeting of AESJ 3O13, (2017). L. Arnoldus, et al., Annals of Nucl Energy 88 (2016) 126–134.
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Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Study on lithium rod test module and irradiation method for tritium production using high temperature gas-cooled reactor ⁎
Yuki Kogaa, , Hideaki Matsuuraa, Yuma Idaa, Ryo Okamotoa, Kazunari Katayamab, Teppei Otsukac, Minoru Gotod, Shigeaki Nakagawad, Satoru Nagasumid, Etsuo Ishitsukad, Yosuke Shimazakid a
Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka Nishi-ku, Fukuoka-shi, Fukuoka, 319-0395 Japan Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasugakouen Kasuga-shi Fukuoka, 816-0811 Japan c Department of Electrical and Electrtonic Engineering, Kindai University, 3-4-1 Kowakae, HigashiOsaka city Osaka, 577-0818 Japan d Japan Atomic Energy Agency, 4002 Narita-cho Oarai-machi Higashi-Ibaraki-gun, Ibaraki, 311-1393 Japan b
A R T I C LE I N FO
A B S T R A C T
Keywords: Fusion reactor Tritium production High temperature gas-cooled reactor Li rod HTTR Irradiation test
Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by 6Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO2, alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod’s tritium production and containment performance was presented.
1. Introduction Fusion reactor R&D has advanced to full-scale validation of both nuclear burning and reactor engineering. A fusion reactor is assumed to use T(d,n)α reaction. Deuterium exists in nature and can be produced from water; however, tritium is a very rare radioactive nuclide, so it should be produced artificially. A fusion reactor with 3 GW thermal output power must burn approximately 400 g of tritium in a day and more tritium is needed in the tritium circulation system and initial tritium inventory. Furthermore, technical tests for tritium circulation and blanket system must be conducted before starting up the DEMO fusion reactor. Because of these reasons, it is necessary to have sufficient amounts of tritium. For example, it was reported that the initial tritium amount required for DEMO was several 100 g ∼ approximately 20 kg [1,2]. We proposed a tritium production method that worked by loading a Li compound as a burnable poison (BP) into a high temperature gas-cooled reactor (HTGR) [3]. The HTGR is a fourth generation reactor types [4]. High outlet coolant temperatures enable the HTGR to operate with high-efficiency and allows hydrogen production. A HTGR has the following tritium production advantages. First, the HTGR moderator/ structural material consists of graphite and its coolant is He gas, both of
⁎
which are chemically stable and do not significantly react with Li compounds. When we consider tritium production in a light water reactor (LWR), separating tritium that has leaked into water is difficult. Second, a large amount of Li compounds without a 6Li concentration can be loaded. This is because a HTGR reactor core is larger than those of other fission reactor types with the same thermal output (e.g., LWR or fast breeder reactors). Third, it is easily to replace B4C with a Li compound without significantly changing the structural core design, and it may be possible to replace all the burnable poison (BP) depending on the core design. Furthermore, B4C pellets as BP can be loaded into a core using a separate module from the fuel rod; therefore, tritium produced from a Li compound can be recovered without polluting the reactor. Fourth, as HTGRs are generally designed to have high burnup, their nuclear transformation performance tends to better than low burnup reactors (e.g., LWRs). In the first study we considered the use of Li particles covered with layers of carbon, PyC (pyrolytic carbon), and SiC [3]. A tritium diffusion simulation using permeability data obtained by Katayama et al. [5] revealed that most of the tritium was lost from the Li particle due to its large surface area. We continued the study on the Li rod, assuming a cylindrical rod based on the design used in the Watt Bar nuclear plant [6]. We performed the core calculation in the case that Li rods were
Corresponding author at: Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka Nishi-ku, Fukuoka-shi, Fukuoka, 319-0395 Japan. E-mail address: [email protected] (Y. Koga).
https://doi.org/10.1016/j.fusengdes.2018.03.029 Received 29 September 2017; Received in revised form 26 February 2018; Accepted 12 March 2018 0920-3796/ © 2018 Elsevier B.V. All rights reserved.
Please cite this article as: Koga, Y., Fusion Engineering and Design (2018), https://doi.org/10.1016/j.fusengdes.2018.03.029
Fusion Engineering and Design xxx (xxxx) xxx–xxx
Y. Koga et al.
loaded in BP holes of the Gas Turbine High Temperature Reactor 300 (GTHTR300), [7] a commercial HTGR designed by the Japan Atomic Energy Agency. The simulation showed that 500–800 g of tritium could be produced during one year of operation and that the tritium leakage from the Li rods could be suppressed to less than 1% of the amount of the tritium produced [8]. In the simulation we assumed a temperature of 800 K at the Li rod region; however, from an electric power generation efficiency perspective, it is important to prevent tritium leakage from the rod in lower levels at a higher temperature range, e.g., 1100–1200 K. It is possible to reduce tritium leakage by adding Zr layers [9] for high temperature operation. To advance to the new phase, it is necessary to demonstrate tritium production and confirm containment performance. We considered a demonstration test using irradiation facilities or the High Temperature engineering Test Reactor (HTTR) [10]. In this paper, we designed a Li rod suitable for use in the HTTR, and exhibited an irradiation test scenario in the HTTR.
nuclear burning calculation using the Monte Carlo neutron transport code MVP-BURN [12] and JENDL-4.0 [13]. We assumed a HTTR core system with Li rods loaded into all BP holes and a one-year operation period. We set the Li rod temperature to 1170 K and a thermal output to 30 MW. This rod is used in a Li-rod capsule and irradiated in the irradiation test. On the other hand, to design Li-rod capsules for the irradiation test, we assumed a 30 day the HTTR operation with an irradiation column neutron flux of 4.2 × 1017 n/m2s [14]. We set the maximum temperature at 1070 K. We evaluated the neutron flux in the LiAlO2 in the Li-rod capsule using this irradiation flux and statistical processing by MVP code [12] and JENDL-4.0 [13]. We assumed that the Li-rod capsule was irradiated uniformly in the irradiation model that has a vacuum system with white reflector boundary. In this model, 2,160,000 neutrons are generated in the vacuum area as a fixed source. We evaluated one-day tritium production from the LiAlO2 in the Li-rod capsule using the following equation:
2. Calculation and evaluation models
where Nt, N6Li, ϕ, σ6Li+n, and t are the number of T, number of 6Li atoms, the neutron flux in LiAlO2, the microscopic cross-section of the 6 Li + n reaction, and irradiation time, respectively.
Nt = N6Li [1 − exp(−ϕσ6Li + nt )],
2.1. Irradiation equipment in the HTTR
(1)
2.3. Tritium diffusion model
Fig. 1 shows a schematic diagram of the HTTR core. The HTTR has a thermal output of 30 MW and an average moderator temperature of 1170 K. We conduct the irradiation test in a graphite hexagonal column with a 360 mm facing distance and a height of 580 mm. The height and diameter of the Li rods are 450 mm and 14–30 mm, respectively. Fig. 2(a) and (b) show example Li rods. One Li rod consists of a 1–20 mm alumina layer, two 0.02–1.0 mm Zr layers, a 0.5–4.0 mm LiAlO2 layer as a Li compound with a theoretical density of 85% [11], and a hollow portion. These layers have a cylindrical configuration and the design of their thicknesses are discussed in Section 3.1. We also prepare a Li rod without the Zr layers to confirm the effect of leakage prevention due to the Zr layers. The irradiation capsules composing the Li rods have a cylindrical configuration with a diameter of 120 mm, a height of 480 mm, and a 20-mm thick alumina cover, (inner radius is 40 mm) to prevent tritium leakage (Fig. 2c). He gas is circulated through the inner or outer hollow portion in the Li-rod capsule to recover tritium in the real test, but we did not consider this in this paper. This mechanism is set to be designed after determining the experiment execution. The irradiation test is conducted in a graphite irradiation block because aside from its external form, the irradiation block can be designed at will and is capable of having other devices set on it to maintain temperature and measure and recover tritium.
We used the tritium diffusion equation in a cylindrical coordinate system to evaluate tritium leakage and tritium absorption by Zr.
∂C D ∂ ⎛ ∂C ⎞ = r , ∂t r ∂r ⎝ ∂r ⎠
(2)
where c, D, and r are the tritium concentration, diffusion coefficient, and radius direction distance, respectively. We calculated the tritium leakage amount from alumina using the steady solution:
G=
2πhK P (mol/ s ), ln r2 − ln r1
(3)
where r1, r2, h, K, and P are the inner radius and outer radius of the alumina layer, the height of the alumina layer, the permeability coefficient, and the partial pressure in the hollow portion, respectively. We evaluated tritium absorbed into Zr using the cylindrical transient diffusion Eq. (2) and Sievert’s law:
C = S P.
(4)
We assumed that the Zr layers absorbed tritium quickly and that tritium partial pressure decreased. Fig. 3 shows the tritium leakage simulation model for the Li-rod capsule. For conservative, we evaluated tritium leakage from the Li rod by the tritium partial pressure in the inner hollow portion. Tritium leakage from the Li rod determines tritium partial pressure in the outer
2.2. Nuclear calculation To design a Li rod suited for use in the HTTR, we conducted a
Fig. 1. Schematic diagram of reactor core of the HTTR.
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Fig. 2. Schematic diagram of Li rods and rod cover.
3. Result and discussion 3.1. Evaluation of tritium production and leakage We conducted a nuclear burning calculation to determine an outer radius and a thickness of LiAlO2 in Li rod suited for use in the HTTR. BP holes in the HTTR have a height of 500 mm and a radius of 7.5 mm. We set the Li rod radius to 7.0 mm, it is the original BP design of the HTTR, and set its height to 480 mm. We fixed the outer radius of the LiAlO2 and the alumina layer thicknesses to 5.0 mm and 2.0 mm. Fig. 4 shows tritium production and the effective multiplication factor after a 360day operation. The amount of tritium produced is proportional to the LiAlO2 layer thickness, because the LiAlO2 thickness increases the amount of loaded 6Li. The effective multiplication factor decreases with increasing tritium production. It was revealed that Li rods with 3.0 mm LiAlO2 layers could produce 30.9 g of tritium and had an effective multiplication factor of 1.02 at the end of operation. We achieved 30 g of tritium production per year in the following simulations and designed the outer radius and thickness of LiAlO2 in the standard rod to be 5.0 mm and 3.0 mm, respectively. Because the tritium production amount and the effective multiplication factor are not influenced by alumina layer thickness and the existence of Zr layers, we carried out the diffusion and absorption calculations to determine alumina and Zr layer thicknesses assuming a constant tritium production value. Fig. 5 shows the alumina layer thickness transition of the tritium leakage rate from the alumina and Zr layers. In the previous study, the total amount of tritium that leaked from the rod was estimated to be almost 1% of the produced amount under low-temperature conditions [8]. Thus, in this paper, we also
Fig. 3. Schematic diagram of tritium leakage from the capsule.
hollow portion and tritium leakage through the alumina cover each second. We calculated the amount of tritium absorbed into the Zr in the Li rod at each second before the tritium leakage calculation. In the diffusion and absorption calculation about Li rod, we considered the calculation of the inner hollow portion, the Zr layers and the alumina layer. We considered the decrease of tritium leakage from the Li-rod capsule by tritium recovery using He circulation about the capsules, especially for the capsules without Zr because a large quantity of tritium leaks from them at high temperature. In this calculation, we determined the ratio of tritium recovered by flowing He at each second. In addition to leakage from the side wall, we considered leakage from the upper and lower covers, but these were much smaller than leakage from the side wall. The alumina transmission coefficient for tritium is
K = 1.96 × 10−12 exp(− 4.55 × 10 4 RT ) [mol/m1 2⋅kg1 2] ,
(5)
where R and T are the gas constant and temperature, respectively. The alumina transmission coefficient for tritium was obtained by Katayama et al. [5], and the Zr diffusion coefficient and solubility product for hydrogen were obtained by Okamoto. et al. [15]. The calculation for tritium absorption into Zr can be estimated based on the assumption that the S and D for tritium are approximately the same as those of hydrogen. From the viewpoint of tritium pollution in the HTTR reactor in the irradiation test, we set the criterion of tritium leakage from Li-rod capsules to 0.1 μg, which is the peak tritium amount in the HTTR primary helium gas [16].
Fig. 4. Tritium production and effective multiplication factor after 360-day operation from the thickness of LiAlO2.
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Fig. 5. Alumina layer thickness transition of the tritium leakage rate with and without Zr layers.
Fig. 8. Amount of tritium leakage by irradiation time in capsules C and D.
carried out simulations to suppress the amount of tritium that leaked from the Li rods to less than 1%. For this purpose, we included two 0.1mm thick Zr layers in the rod. The Zr layer’s tritium absorption speed is not proportional to its thickness but the surface area; amount of the two Zr layers in the Li rod was sufficient to absorb tritium. We designed the alumina and Zr layer thicknesses to be 2.0 mm and 0.1 mm, respectively, and this rod was defined as the standard rod. It can produce 30.9 g of tritium per year with an anount of leaked tritium of less than 1%.
Table 1 Design and expected experimental values of Li-rod capsules. Capsule
A
B
C
D
Hollow radius(mm) Zr (mm) LiAlO2 (mm) Alumina (mm) T2 production (g) T2 leakage (g) Inner pressurea (Pa)
1.9 0.1 3.0 1.9 3.17 × 10−3 3.57 × 10−7 1.87 × 10−1
3.4 0.1 1.5 1.9 2.32 × 10−3 2.16 × 10−7 0
2.0 – 3.0 2.0 3.17 × 10−3 6.45 × 10−4 3.68 × 105
2.0 – 3.0 10 3.18 × 10−3 2.10 × 10−4 4.34 × 105
3.2. Irradiation test a
Estimated value.
In the irradiation test, we intended to confirm the standard rod’s performance (capsule A) by looking at tritium production and the containment property. At the same time, we wanted to confirm how performance was influenced by LiAlO2, Zr, and alumina layer thicknesses. We designed capsules B, C, and D for the comparison experiment to confirm these performances in experiments I, II, and III. Table 1 shows the designs and expected experimental values of capsules A, B, C, and D. We evaluated Li-rod capsules for three experiments considering tritium production or leakage by changing Zr, LiAlO2, and alumina layer thicknesses. We assumed that He gas was circulated through the outer hollow portion of the capsules to recover and measure tritium leakage. 3.2.1. Experiment I In experiment I, we confirm the change of tritium production due to LiALO2 thickness and the tritium containment performance of Zr layers. We compare the amount of tritium produced and leaked from capsules A and B. The LiALO2 layer thickness in capsule B is half of that in capsule A and capsule B’s tritium production is about 73% of that of capsule A. The amount of tritium produced in capsules A and B can be measured using Zr (as a sum of the absorbed in it) and He gas (leaked tritium). We estimated the tritium leakage from the rods in capsules A and B, and the results are shown in Fig. 6 as a function of irradiation time. We evaluated the tritium leakage from the Li rod in capsule B as 61% of that of capsule A.
Fig. 6. Tritium leakage by irradiation time in capsules A and B.
3.2.2. Experiment II We confirm the tritium leakage decrease due to the use of the Zr in experiment II. We measure and compare tritium leakage from capsules A and C, and the results of which are shown in Fig. 7 as a function of irradiation time. We expected that the tritium leakage of capsule A was about 0.055% of that of capsule C. Fig. 7. Tritium leakage by irradiation time course about capsule A and C.
3.2.3. Experiment III In experiment III we confirm the tritium containment performance of the alumina layer. We irradiate capsules C and D to compare tritium 4
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Fig. 9. Schematic diagram of the irradiation block.
tritium production and leakage. In experiment II, we confirm the tritium leakage decrease due to the effect of Zr layers by tritium leakage. In experiment III, we measure tritium leakage to confirm the effects of the alumina layer. In the irradiation test, we can not confirm the effect of the Zr phase change at more than 1130 K because the irradiation region’s maximum temperature is 1070 K. Tritium leakage from all the tested capsules exceeds the leakage criterion, thus we must consider ways to circulate He through the capsules to recover the tritium. We also conduct other countermeasures for all the capsules to decrease tritium leakage as low as reasonably achievable. The present irradiation test is important for establishing a technique to produce 100 g tritium orders in Japan.
leakage and examine alumina layer performance in preventing tritium leakage. The alumina layer in capsule D is five times thicker than in capsule C. Fig. 8 shows the tritium leakage of capsules C and D as a function of irradiation time, and shows that the tritium leakage from the rod in capsule D is 33% of that in capsule C at 30 days of irradiation time. 3.2.4. Irradiation block The rod in capsule A can produce 30.9 g of tritium in a year if loaded into all BP holes in the HTTR. This tritium production is not simply from the tritium production in capsule A seen in the irradiating test in experiment I. This is because the considered irradiation region is the edge of HTTR core, which has a different condition from that of the BP holes in the core’s center region. An irradiation block with four holes is needed because four Li-rod capsules should be irradiated simultaneously in one irradiation block. Therefore, we designed an irradiation block as shown in Fig. 9. The tritium leakage from all the capsules exceeds the 0.1 μg tritium leakage criterion. Tritium leakage can be decreased sufficiently below this criterion sufficiently using a He circulation system flow of 132 L/min.
Acknowledgments This work is supported by JSPS KAKENHI Grant-in Aid for Scientific Research (B) JP15H04230. References [1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13] [14] [15] [16]
4. Concluding remarks We designed a Li rod suited for use in the HTTR that produced 30 g of tritium in a year of operation, and we proposed a Li rod irradiation test to demonstrate tritium production and confirm tritium containment performance. We exhibited four Li-rod capsule designs, the irradiation condition, and the irradiation block to test tritium production and leakage in the HTTR. In addition, we evaluated the expected data, amount of tritium produced and leaked, and the inner pressure during the test at the irradiation region in the HTTR. The irradiation block has four holes, allowing irradiation of four capsules simultaneously. In experiment I, we confirm the effects of the LiAlO2 and Zr layers by the
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