RU2344500C2 - Nuclear reactor with fissible region representing salt melt - Google Patents

Abstract

FIELD: physics, nuclear physics.

SUBSTANCE: nuclear reactor with fissible region representing a salt melt comprises a lead heat-carrier. The heat-carrier directly contacts aforesaid fissible region and circulates in primary heat exchange circuit. Initial nuclear fuel consists of potassium chloride and uranium tetrachloride. Tetrachloride is uranium-235-enriched to 24%. Internal chamber of the fissible region is, preferably, cylindrical changing over into tapered shape. The aforesaid fissible region aluminIum nitride wall is enclosed by a cylindrical layer of aforesaid heat carrier fed in tangentially to directly contact fuel salt melt at the fissible region bottom. The thick-walled cylinder of liquid lead makes a part of radiation protection.

EFFECT: higher reactor safety and efficiency with heat carrier temperature increasing, reduced fuel circulation time, decreased radioactive waste volume and radioactivity.

Description

Application area

The invention relates to new generations of fast neutron power reactors (BR), which it is high time to displace existing thermal reactors with solid-state fuel elements (TVEL) and fuel assemblies (FA). Moreover, the whole hitherto truly unclosed nuclear fuel cycle (NFC) is being simplified radically. The search for the safest, most environmentally friendly and economical concepts of nuclear power reactors is the proposed solution - the device BR.

State of the art

Already at the Second International Conference on the Peaceful Use of Atomic Energy (Geneva, 1958), it was published on the development of thermal reactors with an active zone (AZ) in the form of molten fluorides (report by H. MacPherson et al. No. 605). In 1965, the MSRE-10 thermal neutron reactor with AZ in the form of Li ⁷ F-BeF ₂ -ThF ₄ -UF ₄ melt was critical at the Oak Ridge National Laboratory. The same melt circulated as a coolant. The very useful long-term operating experience of this reactor convincingly confirmed many of the advantages of using salt melts as nuclear fuel (NF).

The main advantages:

* unlimited radiation resistance of a fully ionized nuclear fuel melt. Even the most severe β- and γ-radiation can cause only some labile structural changes that do not significantly change their properties,

* internal safety due to a negative reactivity coefficient with increasing temperature,

* the possibility of periodic optimization of the isotopic and mass composition of nuclear fuel without stopping the reactor,

* the pressure above the liquid salt nuclear fuel differs little from atmospheric,

* there is no need to produce fuel elements and fuel assemblies,

* the reproduction rate of fissile nuclei (HF), which was 1.05 - 1.07, was sufficiently high for the MSRE-10 nuclear reactor;

A serious drawback of this reactor is the use of a fluoride melt containing fissile material as a heat transfer medium. In this case, the circulating nuclear fuel fills the communications and pumps. Its content in heat exchange systems may be 2-3 times higher than that necessary to maintain a chain reaction in AZ. The high radioactivity of the salt melt sent to the heat exchange circuit requires additional protection measures, including the flow of some fraction of delayed neutrons leaving the AZ.

With this in mind, you can give preference to a reactor in which salt fuel is only in the AZ. The design of such a reactor was patented by Bettis E.S. in 1966 (ES Bettis, Fused - Salt - Fueled, Molten - Metal - Cooled Power breeder system, patent US 3262856, 26 Juli, 1966). This is a thermal breeder reactor, in which the heat carrier - liquid lead is in direct contact with the AZ material (Li ⁷ F - 68%, BeF ₂ - 31%, UF ₄ - 1%) and the reproduction zone (Li ⁷ F -71%, ThF ₄ - 29%). On the surface of liquid lead, "AZ""separated" between themselves and the reproduction zone (SV). For heat removal, lead is pumped through both zones. The design of the reactor is greatly complicated by the significant amount in the AZ of a graphite moderator and tubes for the circulation of lead coolant.

The disadvantages of the described solution (prototype) can be avoided by switching to the use of fast neutrons. However, a high concentration of fluorine ions in the melt will lead to a significant deceleration of neutrons so that the neutron spectrum can hardly be attributed to fast ones.

Disclosure of invention

The NR proposed by us should work on chloride fuel and fast neutrons with the energy of their formation in fission reactions of uranium-235, i.e. more than one MeV. The use of chlorides as nuclear fuel is advisable because chlorine isotopes, in comparison with fluorine, slow down neutrons much worse, which will lead to a greater conservation of the spectrum of fast neutrons and ultimately to an increase in the reproduction coefficient of fissile plutonium nuclei - 239. The use of chlorides also gives other benefits: chloride melts are a ready-made electrolyte, and subsequently it can be implemented in a short fuel cycle (CTC) for processing molten circulating nuclear fuel (SNF) using a variety of fairly well-developed anhydrous high-temperature methods thermal electrochemistry, reducing the volume and activity of waste.

As an example of a constructive solution, we offer a schematic description of a fast experimental reactor with a capacity of 400 MW _e (BORS-400) with the initial filling of AZ with KCl-UCl ₄ salt melt. In contact with the heat carrier - lead, an equilibrium is established: 2U ⁴⁺ + Pb = 2U ³⁺ + Pb ²⁺ , a

more precisely: 2UCl ₆ ^2- + Pb + 2Cl ^- = 2UCl ₅ ^2- + PbCl ₄ ^2- , as a result of which the salt system becomes four-component: KCl-UCl ₄ -UCl ₃ -PbCl ₂ , which contains at 24% uranium enrichment ( wt.%): K - 4.7; U ²³⁵ -11.5; U ²³⁸ - 36.7; Pb - 9.6; Cl - 37.5. At a temperature in AZ 750 ° C, the density of this ionic melt will be 3.1 g / cm ³ .

Approximate calculations showed that the reactor will achieve criticality, i.e. a chain reaction can begin with a diameter and height of AZ equal to two meters. The mass of the AZ will be 19,500 kg. Of these, 9400 kg of uranium (of which U ²³⁵ = 2246 kg). If we assume that the conversion of heat into electrical energy will be with efficiency = 34%, the energy intensity in the AZ will be 187.4 kW / liter. This is a relatively modest amount. In some existing reactors, it reaches 500-700 kW / l. The internal safety of the reactor is ensured by the form of its AZ. From cylindrical, it passes into a conical one. In case of unauthorized “acceleration” (for example, due to insufficient intensity of heat removal) and an increase in the volume of the melt, it will fill the volume of the conical part and the reactivity due to the fact that more neutrons will leave the AZ will decrease.

The reactor should be made of high-strength materials with radiation resistance, and where they are in contact with molten salts of AZ and lead coolant, good compatibility when heated to 1000 ° C. For lining the internal cavities of the BRS, we propose a new material - aluminum nitride, the production technology of which was developed at USTU - UPI. AlN has good thermal conductivity, stable up to 2400 ° C.

Another feature of the BRS is that there are two heat removal circuits: the main one is a lead coolant path with an inlet temperature of 400 ° С and an outlet temperature of 750 ° С, the auxiliary one is a nitrogen path circulating in the provided “jacket” with an inlet temperature of 200 ° C and at the exit of 700 ° C. This path is also used for the initial heating of the AZ by supplying nitrogen heated to 700 ° C. In the heat and power part, heat taken from the carriers can be used together.

AZ is replenished periodically with nuclear fuel in the form of chlorides no more than a month later. Since the CV is greater than unity, U ²³⁸ must be introduced into the composition of nuclear fuel in the form of depleted (dump) uranium.

The consumption of U ²³⁵ will be approximately 1200 g / day, and the reproduction of Pu ²³⁹ more than 1300 g / day.

The selection of fission radionuclides (RND) begins with the start of the reactor. The isotopes of krypton and xenon are removed together with nitrogen filling the space above the AZ mirror. The mixture is cooled; by known methods of thermal diffusion, krypton and xenon are isolated from it, fixing them on cryosorbents for storage until activity decreases.

а это значит, что ионы элементов, потенциал выделения которых электроположительнее потенциала перезаряда четырехвалентного урана до трехвалентного, восстанавливаются до металлического состояния и растворяются в свинце либо образуют в нем суспензию. К таким элементам среди РНД относятся изотопы: ниобия, молибдена, технеция, рутения, родия, палладия и серебра. Доля изотопов этих металлов составляет около 25 мас.% от всех продуктов деления. Их переход из солевой фазы в металлическую значительно снижает уровень радиоактивности ядерного топлива АЗ и позволяет извлекать эти ценные металлы при периодическом рафинировании свинцового теплоносителя.Important electrochemical processes occurring at the boundary of the AZ – lead melt are strictly determined by the redox potential

and this means that ions of elements whose release potential is more electropositive than the potential of the recharging of tetravalent uranium to trivalent, are restored to a metallic state and dissolve in lead or form a suspension in it. Such elements among RNDs include isotopes: niobium, molybdenum, technetium, ruthenium, rhodium, palladium and silver. The proportion of isotopes of these metals is about 25 wt.% Of all fission products. Their transition from the salt phase to the metallic phase significantly reduces the level of radioactivity of nuclear fuel AZ and allows the extraction of these valuable metals during periodic refining of lead coolant.

As for the electronegative RND remaining in the salt phase, their long delay in intense neutron fluxes of AZs (≥4 · 0 ¹⁵ neutrons / cm ² · s) is useful. Capturing neutrons, after one or two β decays, they become stable. The most interesting example of the above: ₅₅ Cs ¹³⁷ , having a half-life of more than 30 years, acquiring a neutron, goes into ₅₅ Cs ¹³⁸ , which has a half-life of only 33 minutes and which transmutes to stable ₅₆ Ba ¹³⁸ . Similar transmutations occur with ₃₈ Sr ⁹⁰ , ₄₀ Zr ⁹⁵ , ₅₃ J ¹²⁹ and other isotopes - RND. Transmutation will significantly reduce the level of waste radioactivity (RW) during nuclear fuel reprocessing.

Task to be solved

A sharp improvement in economic indicators of the energy use of fissile nuclei due to an increase in the degree of burnout and installed capacity utilization factor (KIUM) of the BR due to the absence of the need for stops to replace nuclear fuel.

Technical Results

Improving the safety of nuclear weapons; increase in efficiency with increasing coolant temperature; reduction of nuclear fuel turnover time (it does not “cool down” and is not maintained for years in pools, dry storage facilities, etc.); reduction of volume and radioactivity of waste.

Appendix 1. The scheme of the proposed device salt reactor for fast neutrons (BRS)

Designations for the BRS scheme:

AZ - core (melt of the chloride system)

TH - lead coolant

1 - lining made of aluminum nitride

2 - sealed gateway for loading nuclear fuel

3 - hermetic gateway for SNF output

4 - output of gases (nitrogen, krypton, xenon)

5 - coolant supply - lead

6 - output ТН to heat exchangers

7 - supply of pure nitrogen to the space above the AZ

8 - input and output of pure nitrogen

9 - removal of technical nitrogen to heat exchangers

10 - radiation protection and thermal insulation

11 - reactor vessel

12 - supply of technical nitrogen for cooling (heating)

13 - foundation

14 - selection of lead for refining

Appendix 2. Comparison of the prototype with the proposed solution

Type of invention: nuclear reactor - a device for converting the energy of fission of uranium nuclei into heat.

Prototype: E.S. Bettis, US patent 3262856, 07.26.1966

Sign Prototype Proposed solution 1. Neutron energy Thermal Fast 2. The composition of nuclear fuel (NF) A mixture of lithium fluorides (Li ⁷ ) and uranium A mixture of potassium and uranium chlorides 3. Temperature in AZ 400 - 500 ° C 700 - 800 ° C 4. The composition of the zone
reproduction
(ST) A mixture of lithium fluorides (Li ⁷ ) and thorium Not provided 5. Cavity lining material Nickel Based Alloys Aluminum nitride 6. Primary circuit coolant Lead Lead 7. The method of heat removal Circulation through tubes and direct contact of AZ fluoride melt with lead Circulation of lead in the cavity, direct contact of the AZ chloride melt with lead, and heat removal with nitrogen

Claims (4)

1. A nuclear reactor with a molten salt core with a lead coolant in direct contact with the core and circulating in the primary heat exchange circuit, characterized in that the initial nuclear fuel consists of potassium chloride and uranium tetrachloride enriched in uranium-235 to 24% . 2. The reactor according to claim 1, characterized in that the internal cavity of its active zone is cylindrical, turning into a cone-shaped. 3. The reactor according to claim 1, characterized in that the active zone is encircled through the wall of aluminum nitride by a cylindrical layer of a coolant introduced tangentially and coming into direct contact with the fuel salt melt in the lower part of the active zone. 4. The reactor according to claim 1, characterized in that the thick-walled cylinder of liquid lead is part of radiation protection.