The Possibility of Tritium Self-Sufficiency in a Fusion Reactor

Abstract—

Based on modern data, analysis was carried out of the possibility of tritium self-sufficiency in a fusion D–T reactor. Taking into account the real losses of the tritium isotope during the operation of the reactor shows that at modern technologies and results of materials science studies, the self-sufficiency is impossible even at the maximum tritium breeding ratio ${{K}_{{\text{T}}}} = 1.9$.

1 INTRODUCTION

For future production of at least a part of the energy humankind requires, a possibility is considered to use fusion reactors that operate on a deuterium–tritium mixture. A drawback of this method of energy production is the absence of tritium in nature and the limited ability of fission reactors to produce it.

Because of this, a possibility is discussed to increase the production of tritium in the fusion reactor itself [1–3] (self-sufficiency), meaning that the tritium produced by the reactor should compensate all losses of this isotope in the reactor, including losses due to radioactive decay (5.47% per year).

For self-sufficiency of the reactor, it is proposed to use tritium produced as a result of interaction of fusion neutrons with lithium isotopes

$${}^{6}{\text{Li}} + n \to {\text{T}} + \alpha + 4.78\;{\text{MeV}},$$

(1)

$${}^{7}{\text{Li}} + n \to {\text{T}} + \alpha + n{\kern 1pt} '\; - 2.47\;{\text{MeV}}.$$

(2)

The natural lithium contains 7.52% of the $^{6}{\text{Li}}$ isotope and 92.48% of the $^{7}{\text{Li}}$ isotope.

To increase the efficiency of tritium production, it is proposed to use breeding of neutrons due to the interaction between fusion neutrons with beryllium or heavy elements, e.g., lead

$${}^{9}{\text{Be}} + n \to 2\alpha + 2n{\kern 1pt} '\; - 2.5\;{\text{MeV,}}$$

(3)

$${}^{{208}}{\text{Pb}} + n \to {}^{{207}}{\text{Pb}} + 2n{\kern 1pt} '\; - 7.4\;{\text{MeV}}.$$

(4)

To produce tritium, the chamber of the fusion reactor is covered by a blanket that holds a lithium-containing working substance. This substance can be either in liquid (melted LiPb mixture) or solid state (lithium ceramics Li₄SiO₄, Li₂ZrO₃, Li₂TiO₃, etc.) [4, 5].

To estimate the efficiency of blanket operation, the value called tritium breeding ratio is introduced

$${{K}_{{\text{T}}}} = {{f}_{{\text{T}}}}{\text{/}}{{f}_{{\text{b}}}},$$

(5)

where ${{f}_{{\text{T}}}}$ is the amount of tritium produced in the blanket and ${{f}_{{\text{b}}}}$ is the amount of tritium burned during the same time in D–T reactions. Usually, the value ${{f}_{{\text{b}}}}$ is measured in percent of the tritium that fills the reactor. Under different conditions, ${{f}_{{\text{b}}}}$ varies from 0.36 [6] to 10% [7]. The most frequently used value of ${{f}_{{\text{b}}}}$ is 5%.

In the ideal case of a blanket with infinite thickness, ${{K}_{{\text{T}}}} = 1.9$ [4]. Decreasing the blanket thickness leads to a decrease in ${{K}_{{\text{T}}}}$. Thus, at blanket thickness of 55 cm, ${{K}_{{\text{T}}}} = 1.55$ and at blanket thickness of 45 cm, ${{K}_{{\text{T}}}} = 1.3$ [8]. Since inside the blanket, construction elements are located, they further decrease ${{K}_{{\text{T}}}}$. Most often, the theoretical values of the tritium breeding ratio are between 1.1–1.2, but they can also reach 1.42 [9]. This value is considered sufficient for tritium self- sufficiency of the reactor.

Until now, the possibility of tritium self-sufficiency of the D–T reactor was usually discussed without taking into account the real losses of this isotope [1–3, 8].

The qualitative analysis of the possibility of tritium self-sufficiency of the reactor carried out in [10] allows its authors to state that such self-sufficiency is impossible.

In this work, we consider this question using the existing experimental and theoretical data about the losses of this isotope in the D–T fusion reactor.

2 TRITIUM CYLCE OF THE REACTOR

A scheme of the tritium cycle in the D–T reactor is shown in Fig. 1. Fusion neutrons produced in the chamber as a result of D–T reactions travel to blan-ket 1, where tritium is generated. Tritium is recovered from the blanket and, together with helium and other admixtures, it is transferred into the purification system 2 and after purification, into accumulating unit 3. At the same time, tritium, helium and other residual gases are evacuated from the reactor chamber into system 4, in which tritium is separated from its chemical compounds and after purification in system 5, it also is supplied to accumulation unit 3. From the accumulating unit, tritium is transferred back to the working chamber.

Fig. 1.

Tritium cycle of a fusion reactor: 1—lithium blanket, 2—system for tritium purification, 3—accumulating unit, 4—system for tritium extraction from chemical compounds, and 5—system for tritium purification.

For simplicity, we ignore all losses of tritium whose values are unknown, i.e., we will estimate only the minimum value of losses. In addition, we will consider the ideal blanket with ${{K}_{{\text{T}}}} = 1.9$, i.e., use the maximum possible amount of tritium produced in the blanket.

All losses of tritium can be separated into two parts: losses inside and outside the reactor chamber.

1. Losses inside the chamber. It is assumed [4, 11] that a reactor with power of 1 GW burns 55–56 kg of tritium per year. This means that, e.g., in ITER reactor, during continuous operation at power of 0.4 GW, 20–25 kg of tritium will be burned.

Below, we consider a hypothetical ITER-type reactor in which in one year, 20 kg of tritium is burned and ${{f}_{{\text{b}}}} = 5\%$. This means that in the ideal case, in the lithium blanket, 38 kg of this isotope are produced. However, the chamber contains ports, and neutrons falling into ports do not produce tritium. In ITER, the area of the blanket is ${{f}_{{\text{S}}}} = 85\%$ of the area of the working chamber [12]. The efficiency of tritium extraction from the blanket whose working substance is the PbLi melt is 80% [13, 14] and that from the solid-body L-i₄SiO₄ blanket exceeds 95% [15]. Further, we will ignore the nonideality of the blanket and assume that all lithium produced in the blanket is extracted from it (38 kg).

During the transfer of tritium to the accumulating unit, it is adsorbed by the walls of the duct and the purification unit. To decrease tritium losses via diffusion through the walls and protect the environment, it is proposed to make the walls double [16]. Since the losses of tritium at this stage are unknown, we will ignore them as well.

From the accumulating unit, tritium should be supplied to the working chamber. To date, four systems of fuel injection into the reactor exist [17].

(a) Injection by supersonic molecular flow. The maximum efficiency of this method is 60% [17]. Thus, from the initial 38 kg, only 22.8 kg of tritium can be injected into the reactor by this method.

(b) Injection through a gas valve. The efficiency of this method is 20% [16], and consequently, only 7.6 kg of tritium can be injected into the reactor.

(c) Pellet injection. Its efficiency is 80%, however, during the formation and acceleration of pellets, 30% of the working substance is lost [17]. Consequently, by this method, 21.3 kg of fuel can be injected into the reactor. Tritium losses in the injector can be decreased if the so-called shoedpellets (pellets in sabots) are used, i.e., pellets enclosed in a plastic cover [18].

(d) Injection by accelerated neutral atoms. The source of fast neutral atoms consists of an ion source, the transformer of positive ions into negative and a neutralizer of fast ions. The sequence of the source operation is the following. Tritium molecules are injected into the ion source and ionized there. The gas efficiency (the ratio of the quantity of produced ions to the quantity of molecules of the initial gas) is about 30% [19]. The produced ions are extracted from the source. The ion flow after the source is composed mainly of ions ${{{\text{T}}}^{ + }}\,(81.5\% ),$ ${\text{T}}_{2}^{ + }\,{\text{(11.5}}\% )$, and ${\text{T}}_{3}^{3}\,(7\% )$ (the percentages are given for protons) [20]. The probability of the protons transformation into negative ions is 60% [21]. The probability of their neutralization is 60% [22]. If we assume that the efficiency of plasma capture of neutral atoms is 100% [16], then by this method, only 3.4 out of 38 kg tritium can be injected into the reactor.

In [23], it is noted that the ion source can be of interest if the fraction of atomic ions in the extracted ion beam is ~90% and the gas efficiency exceeds 50%. The losses of tritium in the pumping system are unknown, therefore, we will ignore them.

Thus, out the modern systems of fuel injection into the reactor chamber, the most efficient are the injection of supersonic molecular flow and pellet injection. However, at ${{K}_{{\text{T}}}}\sim 1.3$, these methods do not even allow to compensate the tritium burned in the fusion reaction.

Note that most of the data given above was obtained by numerical simulations. It is also noted in [24] that the error of initial data required for the simulations is 5–10%, which affects the accuracy of the estimates.

2. Losses inside the chamber. Let us consider the tritium that fills the working chamber of the reactor. Part of this tritium burns in the D–T reactions (in the case under study this part is 5%). Part of the tritium is absorbed by the chamber wall. When the chamber wall is bombarded by neutral particles during the operation of the reactor, part of the adsorbed molecules are knocked out of the wall surface and return to the plasma (recycling). Tritium has a high penetrating ability and it implants into the wall material and accumulates there. Part of tritium diffuses through the walls. In the processes of accumulation and diffusion, a major role is played by the irradiation of the walls by neutrons and fast particles [25]. For example, the concentration of hydrogen atoms in beryllium after irradiation of the walls by helium atoms with energy of 2.8 MeV increases 2.8 times compared to the nonirradiated samples (synergism).

The time of tritium accumulation in the chamber walls is approximately 100 s [26].

The amount of gas that can be accumulated in the wall is limited. Therefore, after some time from the start of the reactor operation, the walls will saturate and further losses due to this process can be ignored, leaving only diffusion through the walls. Irradiation by neutrons leads to an increase of the diffusion rate (radiation-stimulated diffusion). The gas diffused through the walls can be collected and reused.

The bombardment of the walls also leads to their sputtering and generation of dust. The dust constantly gathers during the operation of the reactor. It is expected that up to 1 t of dust can gather in ITER in one year of operation [27]. Since the total surface of dust grains is relatively large, dust can adsorb a large quantity of tritium.

Tritium losses caused by the interaction of the plasma with the chamber walls are determined not only by the construction and materials of the walls but also by the temperature of the intrachamber details.

In [3], it is stated that materials with large atomic numbers should be used for the reactor walls, which do not sputter chemically under the action of hydrogen isotopes. Studies carried out at the JET installation showed [28, 29] that in an installation with carbon walls, approximately 10–100 times more tritium is absorbed than in the chamber whose walls are made from a material similar to the one planned from the ITER tokamak (Be + W). Moreover, it was noted in [30] that, due to significant difficulties caused by the construction specificities of ITER that do not allow one to determine the locations of dust accumulation and its local temperature, a reliable estimate of tritium accumulation in this installation is quite difficult.

Thus, it is seen that experimental data obtained at different installations differs substantially. One can hope that the operating parameters of the blanket and tritium losses will be refined during the planned D–T experiments at the ITER [31] and JET [32] installations.

To decrease the influence of the walls on tritium confinement, it is proposed to use the so-called “liquid” walls, i.e., walls covered in liquid easily-melted metal such as lithium, tin or gallium [3, 29].

Since dust is produced constantly during the discharge, the losses of tritium also increase when the duration of the discharge pulse is increased.

Another process that leads to loss of tritium is the production of chemical compounds containing this isotope. These compounds are the molecules of I₂ gases (where I is the mixture of hydrogen isotopes, T₂, TD, TH, DH, D_2, and H₂), water I₂O, hydrocarbonates C_xI_y, ammonia NI₃, and other highly volatile compounds. They are evacuated from the chamber, and part of tritium is lost during its detachment from chemical compounds, purification and transfer of gases. All these processes are determined by the materials of the construction elements of the chamber, its temperature, plasma temperature, and composition of residual gases [33]. Due to the complexity of these processes, it is impossible to determine the tritium losses that they cause in advance.

In [26], the term “accumulation factor” is introduced, which is the ratio of the amount of accumulated tritium to the amount of tritium used in experiments. At the Tore Supra installation, this factor was equal to 50%.

During D–T experiments at the JET installation, 30% of tritium was “stuck” in the chamber, and after the end of its operation, only 20% of these isotopes was recovered [31]. At the TFTR installation, this amount was of 51%, at JT-60 it was 70–80%, and at DIII-D it was 10–20% of hydrogen isotopes [32]. Thus, it was shown experimentally that the known intrachamber losses of tritium at different installations differ substantially and significantly exceed the amount of tritium that can be produced in the blanket.

Currently, no technologies exist that would allow one to extract tritium absorbed by the walls and by dust during reactor operation [26]. These days, tritium is extracted during the intervals between discharge pulses. Experiments show that at existing technologies, 10% of tritium is lost irreversibly [31].

The decrease of intrachamber losses also requires developing special technologies and materials science research.

Estimates of minimum tritium losses ${{f}_{{\max }}}$ admissible for self-sufficiently also differ in different works. These values vary from 0.02 in [30] to 1% in [4]. To obtain this estimate, the following considerations can be used. The amount of tritium generated in the blanket is

$${{N}_{{{\text{bl}}}}} = 56P{{K}_{{\text{T}}}}{{f}_{{\text{S}}}}\;{\text{kg,}}$$

(6)

where P is the reactor power in GW. If ${{f}_{{\max }}}$ is the fraction of all tritium losses in the tritium cycle, then in total, the losses in the installation are

$${{N}_{{{\text{loss}}}}} = 56P\frac{{{{f}_{{\text{b}}}} + {{f}_{{\max }}}}}{{{{f}_{{\text{b}}}}}}.$$

(7)

The fraction of maximum tritium losses is determined from the condition of equality of the amount of tritium produced in the blanket to the amount of tritium lost in the reactor, ${{N}_{{{\text{bl}}}}} = {{N}_{{{\text{loss}}}}}$.

Thus, from (6) and from (7) we find that the maximum tritium losses in the cycle at which self-sufficiency is possible are equal to

$${{f}_{{\max }}} = ({{K}_{{\text{T}}}}{{f}_{{\text{S}}}} - 1){{f}_{{\text{b}}}}.$$

(8)

From (8), it is seen that even in the ideal case (${{K}_{{\text{T}}}} = 1.9$, ${{f}_{{\text{S}}}} = 0.85\%$) and for ${{f}_{{\text{b}}}} = 5\%$, the maximum admissible losses are ${{f}_{{\max }}} = 3\%$. Consequently, even if all losses other than losses due to the natural radioactive decay (5.76% per year) are absent, under modern conditions, self-sufficiency cannot be organized.

Self-sufficiency can be ascertained only if the tritium breeding ratio K_T and the burning fraction of this isotope ${{f}_{{\text{b}}}}$ are increased and all other losses in the tritium cycle are minimized.

Tritium production in the blanket (${{K}_{{\text{T}}}}$) can be increased if fissionable substances, e.g., uranium are used [4]. However, this possibility is not considered in this work.

3 CONCLUSIONS

Experimental and theoretical data, including those that were obtained in D–T experiments conducted at the JET and TFTR installations, show that currently existing technologies do not allow one to realize tritium self-sufficiency of the fusion reactor by tritium production in its lithium blanket.

Nevertheless, tritium produced in the blanket can be the major fraction of this isotope required from the external source to compensate its losses in the tritium cycle of a fusion reactor.