CC BY
13
UDC 621.039
DOI: 10.18698/0236-3941-2020-3-94-104
HYBRID FUSION-FISSION SYSTEM WITH NEUTRON SOURCE BASED ON DEUTERIUM PLASMA
A.Yu. Chirkov chirkov@bmstu.ru
Bauman Moscow State Technical University, Moscow, Russian Federation
Abstract
Development of hybrid fusion-fission systems appears today as a promising area in practical use of thermonuclear fusion energy. Thermonuclear plasma in such systems is the source of fast neutrons with the power gain factor Q < 1 power amplification factor in plasma. Hybrid system high amplification is generally achieved through nuclear reactions in the subcritical blanket surrounding plasma. Not only power could be produced in such a blanket, but also nuclear fuel, and waste of the nuclear fuel cycle could be disposed. The problem of systems using the thermonuclear reaction between deuterium and tritium lies in the lack of tritium reserves and in the limited possibilities of its production. Therefore, the work considers organization of a fuel cycle based on the deuterium-deuterium reaction. Options of a neutron source based on toka-mak and linear system of the open trap type were examined. Magnitude of the so-called hybrid system electrical efficiency (ratio of the system electrical output to the blanket thermal power) was estimated. Calculations demonstrated fundamental possibility of realizing substantial neutron yield from deuterium plasma. To achieve acceptable performance, power gain in thermonuclear plasma should be Q = 0.5-1. In a tokamak of reasonable scale and when working on deuterium, the gain should be Q ~ 0.3. Potential advantages of linear systems associated with the possibility of high-pressure plasma confinement are presented
Keywords
Deuterium, fusion neutrons, hybrid reactor, magnetic confinement, plasma, energy balance
Received 15.10.2019 Accepted 30.10.2019 © Author(s), 2020
Introduction. It would seem that work on thermonuclear fusion came close to creating an industrial reactor. There is no doubt that ITER (International Thermonuclear Experimental Reactor) tokamak [1] under construction operating with a mixture of deuterium and tritium (D-T) would demonstrate modes with a plasma power gain factor of Q = 5-10 (Q = Pfus/Pext, where Pjus
is the released fusion power and Pext is the absorbed power of external heating). Despite this, realization of purely thermonuclear energy today arises certain pessimism, which is quite justified in the opinion, in particular, of the ITER project initiator, academician E.P. Velikhov [2]. Apparently, a more promising and, most importantly, faster way to practical use of thermonuclear fusion is connected not to a pure thermonuclear reactor, but to a symbiotic fusion-fission system combining thermonuclear and nuclear fission fuel cycles. Moreover, on the basis of such a hybrid system, it is possible to solve the problem of closed nuclear fuel cycle, which potentially is more efficient than involving fast neutron reactors. It should be noted that the idea of such use of thermonuclear reactions is not that new. The first US-Soviet symposium on hybrid reactors was held in Livermore (USA) in 1976 [3], then the second symposium was organized in Moscow in 1977 [4].
There are at least two most serious problems without solving which it is hardly possible to create an economically feasible commercial thermonuclear reactor. Firstly, the problem of materials, and it is unlikely to be resolved in the foreseeable future. Under conditions of neutron fluxes in the energy producing D-T reactor with Q = 5-10, lifetime of plasma facing components (the so-called first wall) will not apparently exceed 3-5 years even for the best existing radiation resistant materials.
Parameters of modern tokamaks are such that Q ~ 1 could be achieved by burning D-T fuel. Therefore, along with continuation of the ITER line in the form of a demonstration reactor (DEMO), today a strategy is being considered of using thermonuclear energy not as a separate area, but as part of the fusion-fission fuel cycle. Thermonuclear plasma with Q ~ 1 in hybrid systems is a source of fast neutrons for a subcritical blanket surrounding plasma. Neutrons are used in the blanket for energy generation, production of fissile isotopes (plutonium-239 and uranium-233) and transmutation of nuclear fuel cycle waste. Neutron fluxes' in a thermonuclear neutron source with Q ~ 1 is significantly lower than in the reactor with Q ~ 10. Despite the fact that the hybrid system concept is based on existing technologies, combination of many functions in one device presents considerable difficulties from the point of view of its designing. Nevertheless, even this complex system appears more suitable for power engineering than a pure thermonuclear reactor using D-T fuel [5].
Secondly, there is an acute problem of tritium reserves and possibilities of its accumulation in the amount required to launch large-scale energy projects [6, 7]. Scenarios were considered, where stages of operation on the D-T mixture are preceded by tritium production in the D-D reaction, i.e., the so-called D-D
start [8]. But such modes do not completely remove the need for the initial tritium stockpile even for a demonstration reactor (DEMO).
Deuterium is attractive primarily due to its relative availability. However, the D-D reaction low rate does not make it possible to Q ~ 10 in the tokamak reactor at reasonable parameters [9]. Conditions for sufficiently high efficiency of a thermonuclear system with D-D reaction require plasma confinement with high p value (P is the ratio of plasma pressure to magnetic pressure). In principle, this could be performed in an open trap, as well as in the field reversed configuration (FRC) and in a number of other systems [10-12].
This paper presents estimation of main parameters of hybrid systems with a neutron source based on the D-D reaction. Options of a thermonuclear plasma confinement magnetic system based on tokamak [13, 14] and linear magnetic configuration of the open trap type [10, 15-17] are considered. Cases of Maxwellian plasma [10,15] and nonequilibrium plasma heated by a powerful beam of neutral atoms (neutral beam injection or NBI) [16, 17] are examined. In the latter case, NBI maintains a significant population of fast particles in plasma, which provides an increase in the reaction rate and, consequently, a decrease in plasma confinement requirements.
In a tokamak with the aspect ratio A = R/a ~ 3 (R is the torus major radius, a is the minor radius) due to the low P values, it is problematic to achieve Q ~ 1 with the D-D fuel. Therefore, case of the reduced aspect ratio A = 2 is considered accepting P = 0.16. Examining tokamak makes it possible to estimate with a high degree of reliability, since they are based on the ITER Physics Basis (IPB) laws [1].
From the technical point of view, linear systems, such as open trap or FRC, are most attractive due to their design simplicity. From the point of view of required thermonuclear regimes readability, reliable physical justification of the concept based on open trap or FRC is hardly possible now, since parameters of today experimental facilities are relatively far from the Q ~ 1 regimes. Nevertheless, considering simple geometry is useful for estimating plasma parameters, sizes, power, level of neutron fluxes and a number of other important parameters of a hybrid system.
Hybrid system parameters. The following reactions are possible in deuterium plasma:
D + D я (2.45 MeV) + 3He (0.817 MeV) D + D —(3.02 MeV) + T (1.01 MeV) D + T -> n (14.1 MeV) + 4He (3.5 MeV)
(1) (2) (3)
D + 3He (14.68 MeV) + 4He (3.67 MeV) (4)
If tritium produced in reaction (2) can to react with deuterium, then the fast D-T neutrons with the energy of 14.1 MeV (reaction (3)) becomes noticeable. Due to differences in cross sections and reaction rates (3) and (4), a situation could occur, when significant part of tritium (possibly all tritium under certain conditions) has time to burn, and helium-3, on the contrary, has time to almost completely leave the trap not having time to react with deuterium. Such thermonuclear fuel cycle is called semi-catalyzed. This case of the D-D cycle is being considered further. Power fraction in fast neutrons with energy of 14.1 MeV in such a cycle is Pnu/Pjus ~ 0.56.
Total efficiency of the hybrid fusion-fission system is [18]:
TW = pnet fPfus = Tie (a„M + 1 - oc„ + 1/Q) - 1/(tvQ), (5)
where Pnet is the output (electric) power; Pfus is the fusion power; M is the blanket multiplication; r\e is the efficiency of heat into electricity conversion; r\d is the external heating system (driver) efficiency.
Expression (5) takes into account that the blanket thermal power is:
Pth = Pjus(anM + 1 - cc„ + 1 / Q), (6)
and the power consumed by external heating systems is:
Pd = Pjus/(Qr[d). (7)
Let us assume r\e ~ 0.35 for the analysis. Modern neutral beam injection (NBI) heating systems and electron cyclotron resonance (ECR) heating systems have the r\d ~ 0.4 efficiency. Blanket multiplication is connected to the keff effective neutron multiplication factor by the following relation:
M = ifiL k*ff , (8)
Ejus v(l-fcefl0
where Eju is the fission energy (Efls ~ 200 MeV); Efus is the fusion neutron energy; v is the number of fission neutrons per one fusion neutron (v « 3).
Electrical efficiency of the hybrid system is
r)* = Pnet/Pth=l/[l/r]e+l/(tle Tld Y\net Q)] • (9)
Figure 1 presents the electric efficiency values depending on the plasma power gain at different blanket multiplication values for systems with fusion neutron sources based on the D-T cycle and on the semi-catalyzed D-D cycle. Differences are due to the fact that the neutrons' energy fraction in D-T cycle is
0.
0.
0.
0.
0.
0.
0
0.1
10
100 Q
Fig. 1. Hybrid system electric efficiency depending on plasma power gain at various blanket multiplication values: solid lines is D-T fusion fuel cycle (an = 0.8), dashed lines is D-D cycle (an - 0.66)
higher. From the safety standpoint, the keff < 0.9-0.95 values are acceptable. The M ~ 50 blanket multiplication values satisfy this condition. The Q = 0.5-1 range could be considered acceptable. In particular, at Q =1, keff ~ 0.9, M « 35, system efficiency r\net ~ 5 and electrical efficiency r\* « 0.23.
It should be noted that in the case of a purely thermonuclear reactor (M «1) with Q = 10, r\e ~ 0.35 and r\d ~ 0.4, efficiency is цnet ~ 0.135, and electrical efficiency is only r|* « 0.12. Low efficiency is connected to the fact that a feature of thermonuclear power plants is significant internal energy consumption by the plasma heating systems [19]. According to estimates [20], the promising thermonuclear power plants efficiency could reach ц* « 0.23, and the cost of electricity production could be comparable to the level of today's renewable energy. Hybrid systems look more attractive from the economic point of view, while the reactor price is increasing approximately linearly with the size of the magnetic system [21]. Magnetic system accounts for up to 40 % of capital costs in construction of a power plant, which significantly exceeds other types of costs [20]. Therefore, it is important from the economic point of view to most efficiently use the magnetic field to confine plasma at the maximum acceptable pressure, which could be provided in systems alternative to tokamak with high p values.
Let us compare the main parameters of hybrid fusion-fission systems based on tokamaks and linear magnetic configuration systems. Figure 2 presents a scheme of the hybrid reactor based on an open trap. In principle, this schemealso corresponds to other linear magnetic traps, for example, FRC. Let us use the
Fig. 2. Scheme of a hybrid reactor based on the open trap (longitudinal section): 1 is plasma; 2 is first wall; 3 is blanket; 4 is radiation shield; 5 is casing structural elements;
6 is central section magnetic coils; 7 is magnetic mirror coils
previously developed energy balance models to calculate the plasma parameters [10-17], which detailed description would be omitted here.
The table shows values of plasma and hybrid systems main parameters: a is plasma columm radius; L is plasma columm length (linear systems); A = R/a is torus aspect ratio (tokamaks); P is relative plasma pressure; Bo is coils' magnetic field induction in vacuum; Ip is plasma current (tokamaks); Eo is injected atoms energy (NBI heating); T is volume average plasma temperature; Q is plasma power gain factor; te is energy confinement time; riD is deuterium average density; Vp is plasma volume; Vu is approximate blanket volume; Pjus is fusion power; Pn is neutron power; Pnu/Pfus is energy fraction in fast neutrons; Pb/Pfus is relative losses due to bremsstrahlung; Ps/Pfus is relative losses due to synchrotron radiation; Jn is neutron energy flux from plasma; Pbi is heat release volume density in blanket; Pth is blanket thermal power; Pnet is output electrical power of the hybrid system. Tokamak on equal component D-T fuel was presented as an example in Option 5.
Efficiency of a system based on a tokamak with D-D fuel (Option 1) is unacceptably low at low (3 values. Neutron fluxes are at the acceptable level in Options 2-4. In Option 3, the r|* value slightly exceeds the value for a purely thermonuclear reactor. It is supposed in Option 4 that 50 % of the produced tritium is not burned, and therefore, could be used in another reactor of comparable power with the D-T fuel neutron source (Option 5). Efficiency
of the latter is significantly higher, and the size is smaller. Such fuel cycle scheme could be more attractive from the economic point of view, than a scheme with full tritium utilization in the catalyzed D-D cycle.
Parameters of deuterium hybrid systems
Tokamak Open trap (linear)
Parameters [13,14] [15] [16]
Option
1 2 3 41 52 6 7 83 93 103
a, m 2 1.5 2 2 1 1 0.5 1 2 2
L, m - - - - - 10 10 4 4 4
А 3 2 2 2 3 - - - - -
Р 0.026 0.16 0.16 0.16 0.026 0.5 0.5 0.5 0.5 0.5
Во, Т 5.3 4.5 4.5 4.5 5.3 3.54 4.94 4.14 5.84 11.24
1Р, МА 15.8 36 48 48 7.9 - - - - -
Е0, MeV - - - - - - - 0.5 1 2
Г, keV 10 9 12 12.5 4.5 25 30 37 72 113
Q 0.050 0.25 0.39 0.25 1.7 1 1 0.084 0.23 0.53
Те, S 2.2 4.1 6.0 5.7 1.16 6.35 3.25 0.215 0.58s 0.745
По, 1020т~3 0.7 3.9 2.9 2.8 0.85 3.0 5.0 2.8 2.8 5.6
Vp, т3 800 330 790 790 100 31 8 12 50 50
Pjus, MW 10 72 170 110 50 20 19 9 76 520
Р„, MW 6.6 48 112 65 40 13 12.5 6 50 340
Pnu/Pfus 0.56 0.56 0.56 0.44 0.8 0.57 0.57 0.57 0.57 0.57
Pb/Pfrs 0.015 0.16 0.22 0.14 0.25 0.43 0.36 0.1 0.085 0.076
PJPjus 0.38 0.019 0.029 0.049 0.003 0.014 0.027 0.037 0.18 0.68
J„, MW/m2 0.015 0.16 0.22 0.14 0.25 0.2 0.4 0.25 0.99 10.0
Ры, MW/m3 - 15 15 12 10 9.5 15 - - -
Рл, MW - 3050 5590 4750 930 430 400 910 3420 14 500
Pnet, MW - 360 850 550 250 100 94 45 380 2600
Tf, % - 12 15 12 27 23 23.5 5 11 18
150 % of the tritium produced in the D-D reaction burns. 2 D-T fuel (50 % : 50 %). 3 Beam heating (NBI), significant non-Maxwellian population. 4 Central section. 5 Plasma thermal components required value.
High efficiency of systems based on the open trap (Options 6 and 7) is associated with a relatively high Q = 1. For this purpose, plasma confinement time in the open trap should be really high (te ~ 5 s); therefore, the need for complicating the system to reduce losses in plasma through the ends of an open trap is obvious. Let us note that significant progress was achieved in recent experiments on open traps [22].
Population of high-energy ions is generated in plasma when heated by the NBI (Options 8-10). Requirements to confinement time are reduced. With an increase in the energy of the injection energy Eo, Q also increases. On the other hand, plasma size and density should stay significant to ensure efficient capture of the beam. Therefore, power of such a system and neutron load on the walls could turn out to be unacceptably high.
The coils' limiting magnetic field is restricted by capabilities of existing superconductors. Maximum field on the superconductor surface does not exceed 20 T in all Options.
Parameters of a system based on the open trap (Options 6, 7) appear to be the most attractive. Decrease in neutron fluxes to the wall in the open trap is possible, when the wall is moved away from plasma. Power growth in a linear system is relatively easily achieved by proportional increase in its length. Heat release density level in the blanket makes it possible to consider cooling of the blanket with a gas coolant, for example, helium [23].
Single-circuit scheme based on a closed gas turbine installation could be used in heat conversion [15], its parameters are as follows: helium temperature at the compressor inlet Xi = 320 K, compressor outlet temperature X2 = 422 K, pressure increase level pilpi = 2, turbine inlet temperature T3 = 1000 K, turbine outlet temperature T4 = 758 K, heat recuperation degree 0.95, temperature difference in the heat exchanger-recuperator AT = 15 K, specific heat supplied per cycle qin = 134.5 kj/kg, recuperated heat qr = 165.3 kj/kg, cycle work lc = = 462.8 kj/kg. For given compressor efficiency r\ COmp = 0.8 and turbine efficiency T|turb = 0.9, the power installation efficiency is r\e = 0.35.
Conclusion. Potential possibility to use the D-D reaction is important from the prospects for large-scale energy production based on hybrid systems. Essential advantage of a hybrid system with neutron source based on the D-D reaction is that there is no need to produce tritium in the blanket. Tritium is produced only in the D-D reaction in plasma. This significantly simplifies a number of problems associated with the tritium cycle technology in hybrid reactors.
Using a gas-cooled blanket with solid components allows implementation of a single-circuit scheme with closed gas turbine installation similar to high-temperature gas-cooled reactors with spherical heat-generating elements.
It should be noted that high power in plasma heating is a characteristic feature of thermonuclear systems. Therefore, it is advisable to create a neutron source with Q » 1 or more; otherwise the hybrid system efficiency would be lower compared to nuclear reactors, for example.
Translated by D.L. Alekhin
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Chirkov A.Yu. — Dr. Sc. (Phys.-Math.), Head of Department of Thermal Physics, Bauman Moscow State Technical University (2-ya Baumanskaya ul. 5, str. 1, Moscow, 105005 Russian Federation).
Please cite this article as:
Chirkov A.Yu. Hybrid fusion-fission system with neutron source based on deuterium plasma. Herald of the Bauman Moscow State Technical University, Series Mechanical Engineering, 2020, no. 3, pp. 94-104. DOI: https://doi.org/10.18698/0236-3941-2020-3-94-104
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