CZ20014161A3 - Energy obtained from fission of used nuclear waste - Google Patents

Abstract

A linear accelerator, preferably of the monochromatic type, accelerates electrons to an energy of about 10 MeV which are directed onto a high Z target such as tungsten to generate gamma rays which are directed onto the fuel material such as U<238> which results in the ( gamma <i>f</i>) reaction, thus releasing about 200 MeV. A reactor built according to this principle requiring an accelerator driven by 1 MW will develop about 20 MW of power. The reaction is not self-sustaining and stops when the beam is turned off. This accelerator driven reactor may be used to "burn-up" spent fuel from fission reactors, if simply operated at 10 MeV. The photo-fission results in typical spent fuel waste products such as Cs<137> and Sr<90> which undergo photodisintegration by the ( gamma ,n) reaction resulting in short lived or stable products.

Description

Energy from fission of spent nuclear waste

Technical field

The invention relates to the treatment of spent nuclear waste. In particular, it involves the generation of electrical energy from fission reactions occurring in spent nuclear waste, and also the efficient storage of such waste.

BACKGROUND OF THE INVENTION

The currently known, for example 1000 MW pressurized-water nuclear reactors, when operating at 75% of maximum power, produce about 21 tonnes of spent fuel at a specific consumption of 43 GWe / t. 21 tonnes of spent fuel contained in 42 fuel elements of the pressurized water reactor, with a total volume of 11 m3, will produce energy of 6.6 TWh. To produce the same energy it would be necessary to burn about 2 million tons of coal, in the production of electricity in a conventional thermal power plant, which will produce 120 thousand. tons of ash, 5.4 million tons of carbon monoxide and 50 ths. tons of sulfur dioxide.

Spent nuclear fuel extracted from conventional reactors contains about 96% uranium by weight 238. For light water reactors, the most commonly used reactors, including pressurized water reactors, the spent fuel contains about 0.90% uranium 235, with natural uranium it contains only about 0.70% of this isotope. Plutonium represents about 1% by weight of spent fuel. Plutonium is a fissionable material, which means that it is useful as a fuel for nuclear reactors. Light actinides make up about 0.1% by weight of spent fuel. They contain about 50% Np, 47% Am and 3% Cm, these elements being highly radioactive. Fission products such as iodine, technetium, neodymium, zirconium, molybdenum, cerium, cesium, ruthenium, palladium, etc., make up about 2.9% by weight of the spent fuel.

The two fission products of essential importance, with regard to their thermal action in the repositories and in contrast to the health risk reduction criteria, are strontium 90 and cesium 137. These radionuclides play a decisive role in the release of heat from spent fuel in particular. Cesium 137 is also the main source •

Penetrating radiation emitted by spent fuel. The main fission products causing health risks are technetium 99 and iodine 129. Their danger lies in the long half-life, their relatively high proportion in fission products, and their solubility in terms of geological storage conditions, resulting in a possible relatively high velocity penetration into groundwater.

Long-term spent fuel toxicity is a dominant feature of actinides such as Np 237, U 234, U 236 and Pu 239, Pu 240 and Pu 242. One of the desirable alternative approaches to addressing the problem of highly hazardous waste disposal is the transmutation of long-lived nuclides to stable or short-lived nuclides stimulated by nuclear reactions.

There are around 300 different types of radioactive elements that arise during the operation of a nuclear reactor, mainly as a result of neutron absorption and neutron initiated fission. Impact of different radionuclides varies due to differences in their own chemical behavior and intensity of their radioactive radiation. The main danger of radionuclides lies in the risks of their deposition in geological repositories. The most widespread release mechanism and subsequent exposure involves groundwater contact with waste, followed by slow dissolution, transport of radionuclides to surrounding areas, dissemination in the biosphere, and possible carry-over from food and water. Of the hundreds of isotopes contained in spent fuel and waste derived from this fuel, only a few are of significant importance. Substantial isotopes in spent fuel, resp. in nuclear waste, light-water reactors are Cs 137, Sr 90, I 129 and Te 99 in relation to their heat produced, groundwater solubility and health risks.

Spent fuel management should ensure the protection of the biosphere under economically acceptable conditions, without undesirable side effects and impacts, and the public must have confidence in the effectiveness of the measures being implemented. Given that spent fuel contains radionulides with a very long half-life, some protection must be provided for at least 100,000 years. There are two ways to proceed.

According to one strategy, it is possible to wait for the natural disintegration of radioactive elements, under their physical isolation from the biosphere, by creating effective • • · ·

• Barriers at the appropriate depth in the ground. This strategy is referred to as deep geological storage.

According to a second strategy, nuclear reactions can be used which transmute wastes with a very long half-life to less radioactive products with a shorter half-life. This strategy is referred to as transmutation.

The problem with the disposal of nuclear waste underground is that there is no material available that fully resists radioactive waste that continuously produces heat, releases gases such as hydrogen and helium, as well as other unstable products.

Nuclear, transuranic wastes produced in the defense industry are now disposed of in 23 repositories throughout the United States, while mixed radioactive and hazardous wastes are disposed of at 40 sites throughout this territory. There are now around 114 nuclear reactors in the US and around 400 commercial nuclear power plants are in operation worldwide, with around 120 GWe being power plants in Western Europe and 45 GWe in power plants in the former USSR and Eastern Europe. In the USA alone, 34 thousand have already accumulated. tonnes of nuclear waste. The current production of highly active nuclear waste, ie primarily spent fuel, is estimated at 3,000. tonnes per year. The average commercial nuclear power plant annually stores 60 fuel cells in temporary repositories, respectively. storage, and it is envisaged to continue this scheme until 2000, where the waste is to be transferred to long-term repositories, designated DOE. The statistics do not include lower-level radioactive waste, such as gloves, filters, tools, suits, etc., originating from nuclear power stations, research centers, or hospitals where radioactive materials are used. This generates 1.6 trillion cubic feet of such waste per year.

At present, the planned costs of the US Environmental Program are about $ 7.5 trillion per year. In this area, the costs of paper-based projects account for about 20% of the total costs. According to the Environmental Management Report, the liquidation costs for the Nuclear Weapons Program are $ 230 trillion for 75 years, including $ 50 trillion for the Hanford Program.

At present, most commercial nuclear reactors are designed as large capacity light water reactors with an output of around 400 MW. Boiling water reactors, known as one of the kinds of light water reactors, have a pressure vessel and a reactor core housed in this pressure vessel. The reactor core then comprises a plurality of fuel cells. They are provided at the reactor core

• · · ·

Control rods to control reactor power by inserting these rods into the core. Boiling water reactors have a circulating system for circulating the coolant through the reactor core, where the system also serves to fine-tune the performance of the nuclear reactor. The steam generated in the pressure vessel of the nuclear reactor is introduced into the steam turbine, propelled and subsequently condensed in a condenser. The condensate is then returned as a coolant to the pressure vessel.

Another typical case of a light water reactor is a pressurized water reactor consisting of a pressurized vessel comprising a multi-cell reactor core, a steam generator and a closed loop primary cooling system comprising a pressurized vessel and a steam generator. The heated coolant, after being heated in the reactor core, is fed through the primary coolant line to the steam generator for replacement or replacement. to transfer heat to the feed water introduced into the steam generator. The coolant, whose temperature has been reduced as a result of heat exchange, is returned from the steam generator to the pressure vessel via the primary cooling system piping. On the other hand, due to heat exchange, the feed water evaporates and becomes steam. The steam is then fed to the turbine, driving it, and then condensing in the condenser. The condensate is then returned as feed water to the steam generator.

The cleavage of heavy element cores due to the absorption of electromagnetic radiation, the so-called photo-cleavage, was first predicted by Bohr and Wheeler in their famous 1939 paper. In 1941, Haxby, Shoupp, Stephens and Wells were the first to achieve gamma-ray cleavage. . A review of literature sources suggests that the study of photonuclear reactions in actinide nuclei has been investigated in several laboratories over the past 40 years using different types of gamma radiation sources. The main aim of these studies was to obtain nuclear information on excitatory energies, in the field of gigantic dipole resonance, and also in the field of low energy, near photo cleavage and photoneutron thresholds. Gigantic dipole resonance, also referred to as GDR, has been found to substantially increase the cross section available for the photon strikes to the target nucleus.

Using a quasi-monochromatic photon beam obtained from annihilation during monochromatic positron flight, Bowman was the first to observe the characteristic dispersion of gigantic dipole resonance in two core components, a phenomenon observed similarly to other permanently deformed nuclei. However, it was found that the photon-induced Γη / ff ratio was strongly energy-dependent, a result completely contrary to the data found under conditions of neutron-induced fission, flux induced fission and charged particle induced fission.

Systematics of gigantic dipole resonance, characterizing the absorption of electromagnetic radiation by nuclei in the energy range of about 5 to 30 MeV, has been of interest since the discovery of gigantic resonance itself. Over the years, photoneutron cross sections for many cores have been measured in several laboratories. All measured values are reported in the nuclear data tables. For most of the cases studied, the correlation of results is remarkably good.

The classical description of the dipole photon absorption process states that for spherical nuclei the overall cross section of photon absorption is characterized by the Lorenz equation in the form σ (Ε _γ ) = σ, / [1 + (Ε _γ ² - Em ² ) ² Z Εγ ² Γ ² ] (Eq. 1) where Om is the peak cross-section, E _m is the resonant energy and Γ is full width at half maximum. For deformed, spheroid nuclei, their common image shows the division of gigantic resonance into two components, corresponding to oscillations in a direction parallel and perpendicular to the axis of symmetry of the nucleus.

For medium and heavy nuclei, the Coulomb barrier prevents the emission of charged particles, with energy at the level of gigantic resonance, and the cross-section of photon roptyl is a good approximation to the overall effective photon absorption cross-section.

The internal four-pole moment Q _o for the deformed core can be calculated from the expression in the form

Qo = 2/5 ZR ² s = 2/5 ZR ² (η-1) η ^2/3 (Eq.2) where the radius of the core R = RoA ^1/3 , Z and A are atomic number and atomic mass, ε is the eccentricity of the core and the parameter η is the ratio of the large axis to the small axis, for the elongated core, given by the expression

E _m (2) / E _m (1) = 0.911 η + 0.089 (Eq.3) where E _m (1) and E _m (2) are lower and upper resonant energies of two components

Lorenz curves relevant to giant resonance.

The energy of the dipole resonance is so low that in the area of gigantic resonance, simple processes of the type (γ, η), (γ, ρ), (γ, 2ρ) and the type of photocleavage reactions occur. The process-to-process ratio corresponds to common statistical • · · ·

«· ·« Conditions for de-excitation of composite nuclei, so in practice the neutron emission predominates in practice. Of particular interest is the characteristic of gigantic dipole resonance for actinide nuclei. For such nuclei, with a high Z number and a high Coulomb barrier value, the total cross-section of photon absorption is equal to the sum of the photoneutron and photo-fission cross-sections. The total photoneutron cross-section is the sum of the following cross-sections of the reaction σ (γ, η _ίο ι) = σ (γ, η) + 2σ (γ, 2η) + υσ (γ, ζ) (Eq.4) where υ is neutron multiplicity . The total cross-section of neutron production is then σ _γΝ = σ _γη + υσ _γζ (Eq.5)

The distribution between neutron emission and fission can be expressed as r _n / r _f (E) = σ (γ, η) / σ (γ, ί) (Ε) (Eq.6)

The value of r _n / Tf decreases exponentially with the fission of the nucleus. The theoretical expression for r _n / Tf, explaining the behavior of the nucleus for neutron emission and fission distribution, is derived from the relations describing the constant temperature of the nucleus for densities of a certain level and is expressed as r _n / r _f = 2 TA ^273/10 exp {(E ^f '- Bn') / T (Eq.7) where (E ^f '- Bn') are the effective thresholds for the respective photonuclear processes and T is the nuclear temperature.

The fact that more than one neutron is emitted when fissile isotopes such as Th ²³² , U ²³³ , U ²³⁵ , U ²³⁸ and Pu ²³⁹ leads to the possibility of a chain reaction at a certain mass of fissile material. The ratio between neutron production during fission and neutron loss during various processes, such as extracellular neutron capture, primary (η, γ) system reactions, and neutron leakage through the system surface, depends on whether the chain reaction remains stable or whether this reaction is getting stronger or whether it's in decline.

Energy is released with an intensity of 200 MeV per fission of one atom, or with an intensity of about 23.10 ⁶ kWh per fission of one kilogram of uranium U ²³⁵ . Fission fragments carry 82% of the energy in the form of kinetic energy. Instant neutrons carry an additional 2.5% energy, instant gamma radiation carries 3.5% energy, beta decomposition carries 4%, delayed gamma radiation 3%, and neutrino carries the remaining 5% energy. Neutrinos and their energies are lost because the probability of collision with neutrinos is very low. Some fission occurs when fast neutrons hit U ²³⁸ . So

Thus, when fuel is already consumed, plutonium is produced, and a large percentage of the energy is now produced from the cleavage of Pu ²³⁹ plutonium atoms. About 80% of the neutron absorption range in U ²³⁸ results in fission, the remaining 20% are (η, γ) reactions. The average kinetic energy released by Th ²³² photo cleavage is about 0.8 of the energy released from slow-neutron fission of uranium U ²³⁸ , or can be said to be about 160 MeV.

SUMMARY OF THE INVENTION

Following the described state of the art, the problems outlined above are solved by the present invention, the object of which is to provide a nuclear reactor filled with depleted or non-fissionable radioactive material, such as spent fuel from conventional nuclear reactors.

Another object of the invention is to provide a nuclear reactor with increased safety by using fuel in a subcritical amount.

Yet another object is to provide a nuclear reactor with small dimensions and small capacity.

According to the present invention, designed to achieve the foregoing, is a system in which electrons are accelerated to energy of at least 6 MeV by means of an accelerator, and wherein accelerated electrons or gamma photons resulting from the accelerated electron impact on the target are introduced into the fuel in a nuclear reactor, where they cause fuel fission in the nuclear reaction process, known as photo-fission. Photo fission of fuel produces heat and neutrons as in a conventional nuclear reactor, although the fission reaction does not proceed spontaneously due to the fact that a subcritical amount of fuel is used and / or non-fissile elements are used as fuel. Thus, the cleavage process is interrupted immediately with interruption of electron beam irradiation. The heat and neutrons produced in this process are usable as in a conventional nuclear reactor, i.e. for generating electricity.

The present invention is a method and apparatus for producing nuclear power from heavy elements but not fission elements. The reaction is not initiated by a known spontaneous chain reaction, U ²³⁵ , but is initiated by an accelerator. The fuel for this type of reactor, with the accelerator-initiated reaction, may be spent fuel from fission reactors. The mechanism by which • 9

9 · «·

Nuclear energy is released from non-fissile material, known as photo-fission, where photons or gamma radiation are introduced at a level above the photo-fission threshold energy, resulting in fission of the target core. For example, in uranium U ^{235, the} threshold energy of photo cleavage is about 6 MeV and results in nuclear fission of this atom, with an energy release of about 200 MeV.

BRIEF DESCRIPTION OF THE DRAWINGS

The invention is illustrated in more detail by way of example, with reference to the accompanying drawings, in which: FIG. 1 is a schematic representation of the entire device, by means of a block diagram; FIG. 2 is a schematic diagram of an exemplary electron accelerator; fotonukleárního cross section for uranium ²³⁸ U, showing reactions (γ, η), (γ, 2η) and (γ, ί) and cross section (y.total), then the graph of Figure 4 is photonuclear cross sections for TH ²³² , U ²³⁸ and Np ^237, Figure 5 is a graph fotonukleárního cross section for Pu ^239, and Figure 6 is a graph of the gamma spectroscopic analysis of the resulting products of the fission of uranium ²³⁸ U at a level of 30 MeV.

DETAILED DESCRIPTION OF THE INVENTION

As can be seen in FIG. 1, an exemplary embodiment of the invention utilizes a linear accelerator 1, preferably of the monochromatic type, which accelerates electrons to energy in the range of 5 to 30 MeV, preferably to 10 MeV, which then bombs the high atomic Z converter. such as tungsten, generating gamma radiation which is directed to a fuel material such as uranium U ²³⁸ present in reactor 2, where a (γ, ί) type reaction, also known as photo cleavage, then occurs, releases energy around 200 MeV. The coolant is pumped through the reactor 2 by means of the primary cooling pump 3, thereby transferring heat from the reactor to the heat exchanger 4, where the heat is transferred to the secondary cooling system where circulation is effected by the secondary cooling pump 5. in turbine 6,

whereupon the refrigerant passes through the condenser 8 and recirculates. The turbine 6 drives a generator 7 which generates electricity.

A reactor constructed in accordance with the present invention using a 1 MW accelerator 7 will have a power of 20 MW. The photo-fission reaction taking place in reactor 2 does not proceed spontaneously, but stops when irradiation with accelerator 1 is switched off. The present invention is applicable to the "incineration" of already spent fuel from conventional light-water fission reactors, only if the fuel is treated with an energy of 10 to 20 MeV. The process of the invention involves treating typical waste products of the nature of the consumed fuel, such as Cs ¹³⁷ and Sr ⁹⁰ , that undergo photodesintegration, by (γ, η) reaction in reactor 2, whereby products with short half-life are already formed or stable products. Chemical separation of spent fuel is also not necessary. It will be understood that more than one accelerator 1 may be used to intensify the &quot; combustion &quot; process when initiating the work of reactor 2, thereby also increasing the performance of the plant. Ideally, four large accelerators with a power input of about 4.8 MW would be used to initiate the operation of the reactor with a power output of about 100 MW. The fact that the reaction does not proceed spontaneously is a safety factor allowing immediate reactor shutdown in the event of any problems.

The primary nuclear reactions that are important for reactor 2 operation are shown in Table 1.

Table I - Reactions in an Accelerator-Driven Reactor

Photo cleavage reactions (γ, ί): U ²³⁸ , Th ²³² , Pa ²³¹ , Np ²³⁷ , Np ²³⁸ , Pa ²³² , Pu ²³⁹ , Pu ²⁴¹ , Th ²²⁷ , U ²³¹ , U ²³³ , U ²³⁵ .

Fission reactions (one of several possible reactions)

92U ²³⁵ + 0n ^1- > 38Sr ⁹⁴ + 54Xe ¹⁴⁰ + 2 on ¹ ₃₈ Sr ^94- > ₃₉ Y ⁹⁴ + β (stable) ₅₄ Xe ¹⁴⁰ ^ 55Cs ¹⁴⁰ + p -> 56Β3 ^14θ + β -> 37La ¹⁴⁰ + β - > 58Ce ^14t) + β (stable) (γ, η) - enrichment:

g ₂ U ²³⁶ 4-y -> 92U ²³⁵ + οΠ ¹ ₉₂ ^{κ 23θ} + γ -> 92υ ²³⁶ + 20η ¹ + γ -> 92U ²³⁵ 4- 0η ¹ ₉₅ Αιτι ²⁴¹ + γ -> 95Απί ²⁴ ° + 0η ¹ + ^ e ⁰ -> 94Pu ²⁴⁰ ₉₃ 237ρ ²³⁷ + γ -> 93Νρ ²³⁵ 4- 20η ¹ + .-, θθ -> 92U ²³⁵ ₉₂ U ²³⁸ 4- γ -> 92U ²³⁷ 4- 0η ¹ 4- γ -> 92υ ^23θ 4- οΠ ¹ 4- γ-> 92U ²³⁵ 4- οΠ ¹ ₉₄ Pu ²³⁸ 4- γ-> 94Pu ²³⁶ 4- 20η ¹ 4- γ-> 94Pu ²³⁶ 4- 0η ¹ 4- .-, θθ -> 93Νρ ²³⁵ + .ιβ ° -4> 92U ²³⁵ 92U ²³⁸ 4-γ - 92 92υ ²³⁸ 4- 0η ¹ -> 93Νρ ²³⁷ + β4-γ -> 93Np ²³⁵ 4- 20n ¹ + -1e ° -> ₉₂ U ²³⁵ g4Pu ²³⁹ 4 -y- »94Pu ²³⁷ 4-20η ¹ 4-γ-> 94Pu ²³⁵ + 20n ¹ 4-gn ⁰ 93Np ²³⁵ 4-. _d n ° ₉₂ U ²³⁵ ₉₄ Pu ²³⁹ 4-y -> ₉₂ U ²³⁵ + a

Neutron enrichment

9 ₂ U ²³⁴ + 0n ^1- > 92U ²³⁵ 4-y ₉₂ U ²³⁸ 4- on ¹ -> g2U ²³⁹ 4- y - »g3Np ²³⁹ 4- β -» g ₄ Pu ²³⁹ 4- β ₉₄ Th ²³² 4- 0n ¹ -> 90Th ²³³ 4-y 91-8 ²³³ 4-β -> 92υ ²³³ + β (γ, η) - neutralization ₁ Η ³ 4-γ - »-ιΗ ² 4- ₀ n ¹ (stable) ₆ C ¹⁴ 4- y -> 6C ¹³ 4-0n ¹ (stable) ₃₉ Y ⁹⁰ 4- y -> 39Y ⁸⁹ 4 · 0n ¹ (stable)

28Ni ⁶³ 4- y -> 2eNi ⁶² 4- ₀ n ¹ (stable) ₃₆ Kr ³⁵ 4-y -> 36Kr ³⁴ 4- 0n ¹ (stable)

27Co ⁶⁰ 4- y 27CO ⁵⁹ 4- 0n ¹ (stable) siTI ²⁰⁴ 4- y -> 8-iTI ²⁰³ 4- 0n ¹ (stable) ₃₈ Sr ⁹⁰ 4- y -> 38Sr ⁸⁸ 4- 2 0n ¹ (stable) ₈₃ Bi ²¹⁰ 4- y -> 83BI ²⁰⁹ 4- ₀ n ¹ (stable)

56Ba ¹³³ 4- y 56Ba ¹³² 4- 0n ¹ (stable) ₈₂ Pb ²¹⁰ 4- y -> 82Pb ²⁰⁸ 4- 2 0n ¹ (stable) ₃₈ Sr ⁹⁰ 4- y -> 38Sr ⁸⁹ 4- 0n ¹ -> _3g Y ⁸⁹ 4- β (stable)

• · ·

53 < ¹²⁹ - γ ->53> ¹²⁸ * ΟΠ ¹ -> 54Χθ ¹²⁸ - β (stable) 55Cs ¹³⁷ 4- γ -> 55Cs ¹³⁸ 4- 0n ¹ - + seBa ¹³⁶ 4- β (stable) 26Fe ⁶⁰ 4- γ -> 26Fe ⁵⁹ 4- 0n ¹ - + 27Co ³⁹ 4- β (stable) 82Pb ²¹⁰ 4- γ -> 82Pb ²⁰⁹ 4- 0n ¹ -> 83Bi ²⁰⁹ 4- β (stable)

Neutron neutralization

43TC ⁹⁹ 4- 0n ¹ - ^ · 43Tc ¹⁰⁰ 4- γ-> 44Ru ¹⁰⁰ 4- β (stable) 53I ¹²⁹ on ¹ -> 53l ¹³⁰ -Ξ-γ - »54Xe ¹³⁰ + β (stable) s / W -> 53I ¹²⁸ γ -> 54Χθ ¹²⁸ * β (stable) iiNa ²² 4- _o n ¹ - + 11 Na ²³ 4-γ (stable)

The energy thresholds for photocleavage and neutron fission of the respective isotopes are given in Table II.

Table II - Threshold energy for cleavage of selected nuclides photo cleavage threshold (MeV) threshold neutron Am ²⁴¹ 6.0 Am ²⁴² - 6.4 Th ²³² 5.8 1.3 Np ²³⁷ 5.6 0.4 Np ²³⁸ - 6.0 u ²³³ 5.7 0,025 u ²³⁴ 6.0 0.4 u ²³⁵ 5.3 0,025 u ²³⁸ - 0.8 u ²³⁷ - 6.3 u ²³⁸ 5.8 1,2 Pu ²³⁹ 5.8 0,025

• ·

The device according to the invention requires a high-performance low-energy electron accelerator 1 for 10 MeV energy or a Linac emitter for generating gamma radiation, respectively. 2 for the initiation of reactor 2, the preferred embodiment of which is shown in FIG. The current state of the art offers the use of an electron accelerator type TWRR, traveling wave resonant ring powered by two CW 11 Clystrons 11, a continuous wave with a power input of 1.2 MW, operating preferably at a frequency of 1249 MHz, where an electron beam of 10 MeV and a current of 100 mA can be produced.

The TWRR type was chosen to achieve a beam current threshold and to achieve a high accelerator efficiency, which is achieved due to the low attenuation constant value and the high field multiplication factor, which are the advantages of the TWRR type. The advantages of using the TWRR type compared to the stable wave guidance in the accelerator are as follows. The simplicity of the cavity structure in the accelerator, the larger aperture size, the ease of manufacture, and the simple mechanical separation from the recirculation wave guide, all of which simplify operation under strong radiation field conditions.

The klystrons 11 are preferably powered by a 90 kV DC power source with a radiated power of 1.2 MW. This power is fed to the TWRR accelerator via differential couplers 12. The injector then consists of a 200 kV DC electron gun, two magnetic lenses, a mechanical radiation splitter 14, a pre-cluster 15 and a 16-sum electron gun. a mechanical beam splitter 14 and a cluster 16 require a peak current of 400 mA, with a beam energy of 200 keV.

The accelerator 1 consists of seven acceleration lines 18. Each of these units in the respective section of the accelerator forms an accelerator of the TWRR type. Each of the accelerator lines 18, 1.2 m long, contains 13 cavities with a 2π / 3 module and two connection cavities. All acceleration lines 18 are of the constant gradient structure type, under beam-insertion conditions with a current value of 100 mA.

The first klystron powers the cluster and the three acceleration lines, while the second klystron powers the remaining four acceleration lines. The radiation energy supplied to the cluster and to each acceleration line is 200, respectively. 250 kW.

U ²³⁸ itself is usable as a gamma converter as well as a photo-cleavage target, i.e., to eliminate a separate electron at a gamma converter such as • · · ·

tungsten, and the use of uranium U ²³⁸ itself as a target material forming a source of X-rays. The advantage here is the extraction of heat commonly hidden in the converter, which makes up about 70% of the beam energy.

It is important to note here that although the reactor operates under subcritical conditions, it is driven or driven by the reactor. initiated by gamma rays, the produced neutrons still induce both fast and slow neutron fission, as in a conventional nuclear reactor. These neutron reactions result in an additional enegetic yield, thus increasing the power to power ratio from 1/20 to that of a particular design.

Fig. 3 shows the photonuclear cross-section of uranium U ²³⁸ in connection with the respective reactions (γ, 2η), (y, f), and (y.total). Here it is also seen that at the preferred level of 10 MeV, the two photonuclear reactions are (y, n) and (y, f) type reactions.

Figure 4 shows the total photonuclear cross sections for Np ²³⁷ , U ²³⁸ and Th ²³² , while Figure 5 shows the total photonuclear cross sections for Pu ²³⁹ . When comparing the cross-sections in Figs. 4 and 5, it is evident that the cross-sections are almost identical, with differences of only about 10%. This is very important as the photon source cannot distinguish between the fuel sources used. Thus, the power will be the same regardless of which element of the range U ²³⁸ , Th ²³² , Np ²³⁷ and Pu ²³⁹ , or other fissile material, will be used as a fuel source.

FIG. 6 also shows gamma spectroscopic analysis of the products resulting from the uranium photo ²³⁸ cleavage described herein. The only products with a long half-life are Na ²² , Kr ⁸⁵ and Cs ¹³⁵ , which are disintegrated upon exposure to sustained photon flux.

Calculations show that effective incineration type (y, n) for fission waste products requires a gamma radiation flux of 10 ¹⁸ y / cm ² s to achieve a 180 fold acceleration of disintegration time.

The number of nuclei (y, n) reacting during irradiation can be determined by the following differential equation:

d, Vi / dt = - (λ + σιφ) Νί + Σ% (λ] ί + σ _Β φ) Ν, (Eq.8) i = 1,2 ..... Na, where N, = number i -th nuclei, λ = i-th nucleus decay constant, σ, - total photonuclear cross section of i-th nucleus, • ·

λ], = decay constant of the j-th nucleus that is transmuted to the i-th nucleus, φ = radiation flux γ rays

N _a = number of cores considered in the model.

Using matrix expression, the equation (Eq.8) has the following form:

DN / dt = Y · N (Eq.9) where

- (λί + σιφ)

A = *! λρ + σ ^ φ (i Φ j).

The matrix of nuclei N at time t = At can be obtained by Taylor extension:

N (t + A t) = N (t) + Z _{r = 1} (A t) ^r / r! dN ⁽ⁿ⁾ (t) / dt (Eq.10) where dN ⁽ⁿ⁾ (t) / dt is the rth derivative of N (t).

By combining equations (Eq.9) and (Eq.10), N (t + At) can be obtained as follows:

N (t + A t) = N (t) + Z _r = i (A t) 7r! A ^r N (t). (Eq.11)

Matrix A contains two kinds of data, namely decay constants and photonuclear cross sections.

Also, fast neutrons in the treatment of nickel result in the formation of Compton scattered gamma radiation by the reaction of the Ni ⁵⁸ (η, γ) Ni ^{59 type} , where gamma radiation is generated with energy of 5 to 9 MeV. In the present example, if the reactor contains a conversion material such as nickel, it is as if it were at each bombardment, respectively. nickel irradiation used "repeater station". For example, if the reactor contains layers of nickel and uranium U ²³⁸ , the following occurs:

1. The accelerator produces 10 MeV gamma rays acting on the uranium target, followed by a photo-fission process with the formation of fast neutrons,

2. Damping of gamma rays when passing through the first uranium layer reduces the energy of these gamma rays to a value below the threshold for initiation (γ, η) of the reaction,

3. Absorption of fast neutrons in the first nickel layer results in gamma rays of 5 to 9 MeV energy, which then cause photo-cleavage in the next layer of uranium.

Using this layer concept, there are no limitations on the reactor core size since the accelerator beam is continuously amplified with passage through each nickel layer.

In addition, optionally as an alternative to nickel with the "repetition" function described above, other elements, compounds or mixtures, such as sulfur, dysprosium, yttrium, calcium, titanium, beryllium, magnesium, lead, aluminum and copper may be used.

Although the present invention has been described above in an exemplary particular embodiment, using certain materials, formulations and kits, it is to be understood that this does not limit the scope of the claimed protection to only those embodiments, but that the subject matter covers all equivalents, within the scope of protection provided by the appended claims.

Claims (7)

PATENT CLAIMS 1. A heat source, characterized in that it comprises a nuclear reactor (2) operating in a subcritical mode and containing radioactive isotopes, and further comprising an electron accelerator (1), with an acceleration to an energy of 5 to 30 MeV, (1) is directed so that electrons irradiate radioactive isotopes in a nuclear reactor (2). Heat source according to claim 1, characterized in that the radioactive isotopes are spent nuclear fuel. Heat source according to claim 1, characterized in that the radioactive isotopes are selected from the group consisting of Np ²³⁷ , Np ²³⁸ , Pa ²³¹ , Pa ²³² , Pu ²³⁸ , Pu ²³⁹ , Pu ²⁴¹ , Th ²²⁷ , Th ²³² , U ²³¹ , U ²³³ , U ²³⁵ and U ²³⁸ . The heat source of claim 1, wherein the radiation of radioactive isotopes results in the emission of nucleons from these isotopes, thereby producing isotopes of lower atomic mass. 5. The heat source of claim 1, wherein the irradiation of radioactive isotopes results in the emission of fragments occurring during photo cleavage. Heat source according to claim 1, characterized in that the nuclear reactor (2) comprises a converter material for the conversion of fast neutrons produced by the isotope irradiation in the reactor, by conversion to gamma radiation, by reaction of the (γ, η) type. * · 4 «·· ·· ·· · 7. The heat source of claim 6 wherein the converter material is selected from the group consisting of nickel, sulfur, dysprosium, yttrium, calcium, titanium, beryllium, magnesium, lead, iron, aluminum, and copper. A method of producing heat, characterized in that the radioactive isotopes are charged into the nuclear reactor (2), the reactor (2) is operated in a subcritical mode, such that the electrons are accelerated to energy between 5 and 30 MeV, in the nuclear reactor (2). ) with accelerated electrons irradiate the radioactive isotopes, the heat being produced in the radioactive isotopes, whereupon the heat is removed by heat exchange. 20. The method of producing heat of claim 19 wherein the radioactive isotopes are spent nuclear fuel. The method of producing heat according to claim 19, wherein the radioactive isotopes are selected from the group consisting of Np ²³⁷ , Np ²³⁸ , Pa ²³¹ , Pa ²³² , Pu ²³⁸ , Pu ²³⁹ , Pu ²⁴¹ , Th ²²⁷ , Th ²³² , U. ²³¹ , U ²³³ , U ²³⁵ and U ²³⁸ . 22. The method of producing heat as set forth in claim 19 wherein irradiating radioactive isotopes with accelerated electrons results in emission of nucleons from said isotopes, thereby producing isotopes of lower atomic mass. The method of producing heat according to claim 19, characterized in that irradiation of the radioactive isotopes results in the emission of fragments occurring in the photo cleavage of these isotopes. 24. The method of producing heat as set forth in claim 19, wherein the radiation of radioisotopes with electrons causes rapid neutrons to be emitted from said isotopes, wherein a converter material is converted into the nuclear reactor (2) to convert the rapid neutrons generated during isotope irradiation. reactor, by conversion to gamma radiation, reaction type (γ, η). 25. The heat production method of claim 24 wherein the converter material is selected from the group consisting of nickel, sulfur, dysprosium, yttrium, calcium, titanium, beryllium, magnesium, lead, iron, aluminum, and copper.