Case study on tritium inventory in the fusion DEMO plant at JAERI

Fusion Engineering and Design 81 (2006) 1339–1345

Case study on tritium inventory in the fusion DEMO plant at JAERI Hirofumi Nakamura a,∗ , Shinji Sakurai b , Satoshi Suzuki c , Takumi Hayashi a , Mikio Enoeda c , Kenji Tobita d , Demo Plant Design Team a

Tritium Engineering Laboratory, JAERI, Tokai-mura, Ibaraki 319-1195, Japan Tokamak Program Division, JAERI, Naka-shi, Ibaraki-ken 311-0193, Japan Blanket Engineering Laboratory, JAERI, Naka-shi, Ibaraki-ken 311-0193, Japan d Reactor System Laboratory, JAERI, Naka-shi, Ibaraki-ken 311-0193, Japan b

c

Received 31 January 2005; received in revised form 21 October 2005; accepted 21 October 2005 Available online 4 January 2006

Abstract Case studies on tritium inventory and permeation in a fusion demo plant designed by JAERI (DEMO(J05)) have been carried out using several data sets on the tritium transport properties of the candidate materials of the plasma facing components. The results have been compared with the design guidelines for loss of tritium into the components or coolant for tritium accountancy (1.4 kg/y) and tritium concentration in the coolant (0.37 GBq/cm3 ) for DEMO(J05). The results revealed that tritium concentration and inventory in the coolant attributed to permeation from the blanket (plasma facing first wall and breeding region) is expected to be much higher both of these guidelines unless designated permeation barriers are used. Therefore, permeation must be reduced by a factor of 100 in order to satisfy the guidelines. Tungsten coating on the first wall and glassy coating on the blanket cooling tubes may be good countermeasures for permeation reduction in DEMO(J05). © 2005 Elsevier B.V. All rights reserved. Keywords: Fusion demo plant; Tritium inventory; Tritium permeation

1. Introduction Evaluation of tritium inventory and distribution in fusion power plants is essential to plant design, ∗ Corresponding author. Tel.: +81 29 282 6206; fax: +81 29 282 5917. E-mail addresses: [email protected], [email protected] (H. Nakamura).

0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.10.009

since it is directly related to: (1) tritium accountancy; (2) tritium source term for safety assessment; (3) the design basis of the tritium processing plants such as the water detritiation system (WDS). Extensive design studies of the fusion power plant (DEMO(J05) [1], formerly DEMO2001 [2]) have been performed at JAERI. The JAERI fusion power plants are characterized by: (1) use of pressurized water as the coolant (∼15 MPa, PWR equivalent or higher), (2) use of

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H. Nakamura et al. / Fusion Engineering and Design 81 (2006) 1339–1345

Fig. 1. Detail of classified in-vessel components in DEMO(J05) and analytical systems for the divertor, the first wall and the blanket. Here (a), (b) and (c) are the divertor, the first wall and the blanket, respectively.

F82H (a low-activation ferritic/martensitic steel) as the structure material, (3) use of a solid breeder and multiplier and (4) use of metallic material as the plasma facing component (PFC). As to the tritium inventory in the recent JAERI fusion power plant, there are few reports except [2,3], which have reported preliminary evaluations of tritium permeation and inventory in the blanket region of DEMO2001. While it is indicated that tritium permeation reduction is necessary from the viewpoint of tritium recovery from the blanket in DEMO2001 [3], the design criteria for tritium inventory or permeation have not been considered. In the design activity for DEMO(J05), two main guidelines for tritium safety and tritium accountancy have been considered. For tritium safety, we follow a guideline of maximum tritium concentration in the water of 0.37 GBq/cm3 based on the safe operation experience of the tritiated water in CANDU reactors [4]; for tritium accountancy, we follow a guideline for the loss of tritium into the components or the coolant (required to be less than 1% of tritium production amount in the blanket [5]) of ∼1.4 kg/y (110% of the tritium breeding ratio under 3 GW of the fusion power) in DEMO(J05). In this study, tritium inventory in the reactor components and tritium permeation into the coolant have been evaluated with several data sets on the tritium trans-

port properties and several boundary conditions under the tritium permeation reduction methods such as the tungsten-covered PFC and the glassy-material-coated blanket cooling tubes designated for DEMO(J05). Based on the results, the propriety of the current design of DEMO(J05) from the viewpoint of the above two tritium guidelines is discussed, and the key issues on the tritium inventory control are identified.

2. Evaluation method As in the case of ITER [6], four components were selected as the main reactor components of DEMO(J05): the divertor, the first wall as the PFC, the blanket, and the vacuum vessel. Those components were simplified into one-dimensional analytical systems as shown in Fig. 1. In Fig. 1(a), the divertor structure, which consists of tungsten monoblocks and penetrated F82H cooling tubes, is divided into two regions: a tungsten armor region (6.4 mm of averaged thickness) with the F82H cooling tubes (1 mm of thickness) and a region of tungsten armor only (21 mm of thickness) based on the current divertor design of DEMO(J05) [7]. Similarly, the first wall, which is a tungsten-coated F82H structure with cooling channels, is also divided into two regions: the structure with the cooling channel and that without the cooling channel.

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Table 1 Operating scenarios and conditions of the DEMO(J05) Thermal conditions

Divertor First wall Vacuum vessel Blanket a

Plasma side (heat flux (MW/m2 )/Temp. (K))

Coolant side (Temp. (K))

7.0/1273 0.6/773 0/333 0/643

473 603 333 643

Particle load (T atoms/m2 /s)

Surface area (m2 )

Material

1.5 × 1023a 1.5 × 1021a 1 Pa 1 Pa

200 700 800 8000

W/F82H W/F82H F82H Glass/F82H

Implantation range is assumed to be 25 nm.

Here, the thickness of the tungsten coating ranges from 0 to 2 mm in order to allow evaluation of the effect of tungsten coating on the tritium permeation reduction. Plasma facing tungsten armor on the divertor and the first wall surface is classified into two regions: the implantation region and the bulk region. Combined tritium transport and thermal transport analysis is carried out using the TMAP code [8] for simplified onedimensional analytical systems. Eqs. (1) and (2) represent tritium transport equations used in the TMAP code ∂C(x, t) ∂J(x, t) =− + γC(x, t) + S(x, t), ∂t ∂x 
∂C(x, t) C(x, t)Q∗ , + J(x, t) = −Deff (T ) ∂x kT 2

(1) (2)

where C(x, t), S, J(x, t), γ, Deff and Q* represent tritium concentration, tritium source, diffusion flux, decay constant, effective diffusivity and heat of transport, respectively. As to transport through the materials, the concentration of hydrogen isotope in material A and B is assumed to be expressed by Eq. (3) assuming a equilibrium between A and B, since chemical potential of hydrogen isotopes in materials can be defined by Sievert’s law CA CB = , KSB KSA

(3)

where KSA and KSB mean Sievert’s solution constant in A and B, respectively. Transport of hydrogen isotope across a material surface is assumed to be expressed by Eq. (4) at the plasma facing surface for the divertor and the first wall  Jm = Kdm Pm − krij Cj Ck , (4) j,k

where Jm is a flux of hydrogen isotopes molecule m from the material to the surface; Kdm , Pm , and krjk are the dissociation factor of the incident molecule from gas phase to the material surface, partial pressure of hydrogen isotope molecule m, and recombination coefficient of molecule m composed of atom j and k (protium and tritium in this evaluation); Cj and Ck are surface concentrations of atoms j and k, respectively. Since Pm is negligibly small in DEMO(J05), Kdm Pm can be considered to be zero. On the other hand, as to the material surfaces exposed to a gas phase or a down stream surface, surface concentration of hydrogen isotope on the material is assumed to be expressed by Sievert’s law at equilibrium between the hydrogen–metal system (case A) 1/2 C = KS Pm .

(5)

Additionally, we also evaluated an extreme case assuming no hydrogen isotope transport at the downstream surface, which gives maximum tritium inventory in the materials (case B). Thermal response of DEMO(J05) components is evaluated using Eq. (6) for each component material: ρCp

∂T = ∇ · (κ∇T ) + Sh , ∂t

(6)

where ρ, Cp , T, κ and Sh represent density, heat capacity, temperature, thermal conductivity, and heat source, respectively. For composite material, a gap conductance was considered as follows: q = hgap (TA − TB ),

(7)

where q, hgap , TA and TB are interface heat flux, gap thermal conductance, and surface temperature of A and B structures, respectively.

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Table 2 Tritium transport properties in the in-vessel component materials of DEMO(J05)

Deff JAERI

Ref.

KS kr

Tungsten

F82H

Beryllium

6.7 × 10−8 exp(−0.73/kT) [10]

1.07 × 10−7 exp(−0.15/kT)/(1 + 2.9 × 10−8 exp(0.58/kT)) [14]

8.0 × 10−7 exp(−0.73/kT) [15]

6.21 × 1023 exp(−0.28/kT) [14] 9.4 × 10−26 exp(−0.12/kT) [3]

–

4.1 × 10−7 exp(−0.37/kT) [11], 26 appm 1.46 eV trap [12] 1.83 × 1024 exp(−1.04/kT) [11] 5.8 × 10−18 exp(0.56/kT) [13]

–

Deff : effective diffusion coefficient (m2 /s), KS : solubility (atoms/m3 /Pa0.5 ), kr: recombination coefficient (m4 /s). Unit of the activation energy is eV. Deff in implantation region is assumed to be 10 times larger than that in bulk [16].

Operating scenarios for DEMO(J05) in-vessel components, such as estimated thermal load, temperature conditions and particle (tritium) load, are summarized in Table 1 based on the current design of DEMO(J05) [1] or DEMO2001 [2]. Additionally, surface area and basic structure materials of each component are also summarized in this table. Table 2 summarizes tritium transport properties in the in-vessel component materials used in this analysis [6,9–17]. The Q* in each material is assumed to be zero. These values were chosen on the basis of a previous review [9] and safety analysis of ITER [4,6]. Here, the effective diffusivity of tritium in tungsten is taken from the data obtained by the authors using pure tritium (JAERI-data) [10], which is one of the lowest diffusivities of hydrogen isotopes in tungsten, and the data set of the literatures’ data of the highest diffusivity [11] and the trap effects [12] are also used (Ref.-data), since the diffusivity of hydrogen isotopes in tungsten scatters widely. In this analysis, neutron irradiation effect on the tritium inventory was not considered. The thermal properties of the reactor components are taken from previous reports [6,17].

gin of tritium in the coolant through the components or the breakdown of tritium inventory in the components. Here, JAERI and Ref. mean the results with the JAERI-data and the Ref.-data shown in Table 2, respectively, and the -A and -B mean the results for the case A and case B, respectively. The results for tritium inventory and permeation in/from the vacuum vessel are not shown in Fig. 2, since those were negligibly small. It can be seen in Fig. 2 that tritium inventory in the components reaches saturation level (about 2 g) and tritium permeation into the coolant reaches steady state after

3. Results of evaluation Fig. 2 shows a time evolution of tritium inventory in the PFC (the divertor and the first wall) and tritium in the coolant permeated from the PFC during 5 years of DEMO(J05) operation for the standard design of DEMO(J05) (1 mm tungsten coated first wall), and the circles graphed in Fig. 2 show the breakdown of the ori-

Fig. 2. Time evolution of tritium inventory in the PFC and permeated tritium amount into the coolant during DEMO(J05) operation for 1 mm W coating on the first wall case. Here, JAERI-A (-B) and Ref.A (-B) mean JAERI-data and reference data under the A (realistic) or the B (conservative) cases, respectively.

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1 year of operation for both the JAERI-A and Ref.-A cases. Such early saturation in DEMO(J05) is attributed to long-term plasma operation, higher cooling temperature and higher diffusivity of tritium in F82H. The difference of tritium inventory in the PFC and coolant between JAERI-A and Ref.-A is less than twice. For case B, tritium inventory in the PFCs is expected to increase up to above 100 g after 5 years of operation. This result is not only much less than the guideline for tritium accountancy of 1.4 kg/y in DEMO(J05) even in the most conservative case, but also less than the expected tritium inventory in ITER PFCs (1 kg for safety assessment (∼350 g for the ambitious guideline [4,18])). Tritium permeation into the coolant is about 13 g/y (JAERI) and 20 g/y (Ref.) as shown in Fig. 2 for the 1 mm tungsten coated first wall case. Here, the first wall will be covered by the tungsten coating in the current design of the DEMO(J05). Since the thickness of tungsten coating on the first wall may vary due to unexpected events such as erosion or cracking, the effect of thickness of tungsten coating on tritium permeation reduction is evaluated. Fig. 3(a) shows effects of the tungsten coating on tritium permeation through the first wall for the JAERI-A and the Ref.-A cases, and Fig. 3(b) shows the tritium concentration profile in the first wall for the standard design. It is indicated that tritium permeation decreases by about three orders of magnitude for the 1 mm tungsten coating and by two orders of magnitude for the 0.1 mm coating on the first wall as shown in Fig. 3(a). Although the tritium permeation into the coolant cannot satisfy both guidelines when the permeation reduction by the tungsten coating cannot be expected (about 20 kg/y), it can satisfy both guidelines when the 0.1 mm of tungsten coating can be expected on the first wall. The permeation reduction by the tungsten coating on the first wall is attributed to the rate determining process, that is, the diffusion–diffusion control process by the Doyle–Brice model for implantation driven hydrogen permeation [13] as shown in Fig. 3(b). That is why the results of dif-

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Fig. 3. Effect of tungsten coating thickness on tritium permeation reduction through the first wall (a), and tritium concentration profile in the first wall (center region) (b).

ferent tritium transport properties such as JAERI-data and Ref.-data give almost identical permeation reduction, as shown in Fig. 3(a). Here, tritium inventory in the reactor components of DEMO(J05) is considered not only in the PFCs, but also in the blanket regions. Table 3 summarizes tritium inventory in the reactor components after 5 years of operation. Tritium inventory in the breeding region of the blanket is based on the results of DEMO2001 [2], and that in the beryllium multiplier region is evaluated by the same method as in the breeding region using the effective diffusivity in Table 2 [15]. Tritium inventory in the multiplier region is expected to be from 0.1 g to several hundreds of grams. This variance is attributed to the uncertainties regarding diffusivity and the grain size of the beryllium multiplier. Total tritium inventory in the reactor components of the DEMO(J05) will be less than 1 kg, which is also less than the guidelines for tritium accountancy (1.4 kg/y). Tritium inventory in the coolant is determined by the tritium permeation through the reactor components and the tritium recovery by the WDS from the coolant.

Table 3 Tritium inventory in the in-vessel components after 5 years of DEMO(J05) operation PFC

Blanket

Divertor

First wall

Cooling tubes

Breeder

Multiplier

2–46 g

0.2–75 g

0.7–1.4 g (PRF = 1)

1.2 g [2] (Li2 O case)

0.1–400 g (in Be)

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Fig. 4. Tritium concentration and tritium permeation amount in/into the coolant for various PRF in DEMO(J05) blanket. Here, water hold up is assumed to be 1000 m3 , and WDS scale of Darlington and ITER mean 360 kg/h [20] and 20 kg/h [4] of water processing rate, respectively.

As mentioned in Fig. 3, the tungsten coating can reduce tritium permeation by several orders of magnitude. In the blanket region, tritium permeation reduction is also expected due to a glassy coating (50 ␮m) of Cr2 O3 , SiO2 and P2 O5 compound developed by JAERI [19], employing glassy materials’ characteristics such as low diffusivity and solubility. Fig. 4 shows the correlation between the permeation reduction factor (PRF) of the glassy coating and tritium concentration in the coolant for the WDS recovery capacity for the ITER [4] and Darlington plants [20], and it also shows the dependence of tritium permeation amount into the coolant on the PRF. This is a result for the 1 mm tungsten coated first wall case under the JAERI-A case and the evaluation results by Kosaku et al. [3]. As shown in Fig. 4, tritium permeation into the coolant is expected to be ∼15 kg/y without permeation reduction in the blanket (breeding region). In this case, tritium concentration in the coolant is much higher than the guideline (0.37 GBq/cm3 ), and tritium permeation is also much larger than the tritium accountancy guideline (1.4 kg/y). The results indicate that a PRF of at least 100 is required in order to maintain the tritium concentration below 0.37 GBq/cm3 using the ITER scale WDS system (PRF = 10 for the Darlington scale WDS), and it can maintain a tritium permeation amount of less than 1.4 kg/y. According to the report by Nakamichi et al., the PRF of the glassy coating will be more than 1000 with the result of the preliminary permeation experiment

[19]. Therefore, designing for a glassy coating on the blanket cooling tubes can satisfy the requirements of both guidelines as long as long-term reliability of this permeation barrier can be ensured. Further R&D on this permeation barrier’s reliability is necessary, since it is directly related to the capacity of the tritium processing systems such as the WDS or tritium recovery system and the tritium source term for the safety evaluation in DEMO(J05). Based on the results of the case studies on tritium inventory and permeation for DEMO(J05), tritium inventory in the reactor components and tritium permeation into the coolant in the DEMO(J05) are identified, and the current design of the DEMO(J05) is shown to be appropriate from the viewpoint of the design guidelines for tritium accountancy and tritium concentration with the designated tritium permeation reduction methods. But, there are some uncertainties regarding the tritium transport properties such as neutron irradiation effect, or tritium transport mechanism at the plasma or coolant interface. When we obtain such knowledge, much more accurate estimates can be performed. Extensive experimental and modeling studies on these areas of uncertainty are now under way for the DEMO reactor design basis.

4. Summary Case studies on tritium inventory and permeation in a fusion demo plant designed by JAERI (DEMO(J05)) have been carried out with several data sets of the tritium transport properties in the candidate materials for the plasma facing components under conservative boundary conditions to reveal essential measures for tritium inventory control. Two design guidelines, for loss of tritium into the components or coolant for tritium accountancy, and for tritium concentration in the coolant, are considered for the tritium inventory in DEMO(J05) system. As the results, it is found that a PRF of more than 100 is necessary to maintain tritium concentration in the coolant and loss of tritium into the coolant by the permeation below the tritium concentration and tritium accountancy guidelines, respectively, with the design having no designated permeation barrier from the blanket (plasma facing first wall and breeding region) to the coolant. Tungsten coating on the first wall and glassy coating on the blanket cooling

H. Nakamura et al. / Fusion Engineering and Design 81 (2006) 1339–1345

tubes can be good measures for permeation reduction in DEMO(J05).

Acknowledgements The authors would like to thank Drs. S. Seki, H. Ninomiya, H. Takatsu, M. Kikuchi and M. Nishi for their encouragement on this study.

References [1] K. Tobita, S. Nishio, M. Enoeda, M. Sato, T. Isono, S. Sakurai, et al., Design study of fusion DEMO plant at JAERI, Fusion Eng. Design 81 (2006) 1151–1158. [2] M. Enoeda, Y. Kosaku, T. Hatano, T. Kuroda, N. Miki, T. Honma, et al., Design and technology development of solid breeder blanket cooled by supercritical water in Japan, Nucl. Fusion 43 (2003) 1837–1844. [3] Y. Kosaku, Y. Yanagi, M. Enoeda, M. Akiba, Evaluation of tritium permeation in solid breeder blanket cooled by super critical water, Fusion Sci. Technol. 41 (2002) 958–961. [4] ITER EDA documentation series no. 24, ITER technical basis, IAEA, Vienna, 2002. [5] M. Nishi, T. Yamanishi, T. Hayashi, DEMO Plant Design Team, Study on tritium accountancy in fusion DEMO plant at JAERI, Fusion Eng. Design 81 (2006) 745–751. [6] H. Nakamura, M. Nishi, Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties, JAERI-Research 2003-024, 2003. [7] S. Suzuki, Y. Ueda, K. Tokunaga, K. Sato, M. Akiba, Present research status on divertor and plasma facing components for fusion power plants, Fusion Sci. Technol. 44 (2003) 41–48.

1345

[8] G.R. Longhurst, D.F. Holland, J.L. Jones, B.L. Merrill, TMAP4 User’s Manual, EGG-FSP-10315, 1998. [9] R.A. Causey, Hydrogen isotope retention and recycling in fusion reactor plasma-facing components, J. Nucl. Mater. 300 (2002) 91–117. [10] H. Nakamura, T. Hayashi, M. Kakuta, T. Suzuki, M. Nishi, Tritium permeation behavior implanted into pure tungsten and its isotope effect, J. Nucl. Mater. 297 (2001) 285–291. [11] R. Frauenfelder, Solution and diffusion of hydrogen in tungsten, J. Vac. Sci. Technol. 6 (1969) 388–397. [12] R.A. Anderl, D.F. Holland, G.R. Longhurst, R.J. Pawelko, C.L. Trybus, C.H. Sellers, Deuterium transport and trapping in polycrystalline tungsten, Fusion Technol. 21 (1992) 745– 752. [13] B.L. Doyle, B.K. Brice, Steady state hydrogen transport in solid, Radiat. Eff. 89 (1985) 21–48. [14] E. Serra, A. Perujo, G. Benamati, Influence of traps on the deuterium behavior in the low activation martensitic steels F82H and Batman, J. Nucl. Mater. 245 (1997) 108–114. [15] E. Abramov, M.P. Riehm, D.A. Thompson, Deuterium permeation and diffusion in high-purity beryllium, J. Nucl. Mater. 175 (1990) 90–95. [16] H. Nakamura, W. Shu, T. Hayashi, M. Nishi, Tritium permeation study through tungsten and nickel using pure tritium ion beam, J. Nucl. Mater. 313–316 (2003) 670–675. [17] A.A.F. Tavassoli, J.W. Rensman, M. Schirra, K. Shiba, Materials design data for reduced activation martensitic steel type F82H, Fusion Eng. Design 61/62 (2002) 617–628. [18] G. Federici, C.H. Skinner, J.N. Brooks, J.P. Coad, C. Grisolia, A.A. Haasz, et al., Plasma material interactions in current tokamaks and their implications for next step fusion reactors, Nucl. Fusion 41 (2001) 1967–2118. [19] M. Nakamichi, H. Kawamura, T. Teratani, Fusion Sci. Technol. 41 (2002) 939–942. [20] J.M. Miller, Overview of Canadian activity in tritium, Fusion Sci. Technol. 41 (2002) 315–318.