JP2003028976A - Fuel for molten-salt reactor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a fuel for a fluoride molten-salt reactor which absorbs fewer neutrons and has higher solubility of the trifluoride of a transuranic element. SOLUTION: A fuel for a molten-salt reactor, which has higher solubility of the trifluoride of transuranic element at a liquid-phase temperature of 550 deg.C or lower and absorbs few neutrons, is provided by using as a solvent a molten salt, made of a mixture of a lithium fluoride (<7> LiF) formed by enriching and isotope whose nucleon number is 7 and a beryllium difluoride (BeF2 ), adding a zirconium tetrafluoride (ZrF4 ) to the solvent, and limiting the composition of the solvent within the range of a quadrangle, formed by consecutively connecting four points of 77-3-20 mole%, 75-5-20 mole%, 69-31-0 mole% and 71-29-0 mole% with each other on a phase digram of a three constituent system of LiF-BeF2 -ZrF4 .

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a lithium fluoride ( ⁷ LiF) and a beryllium difluoride (B) obtained by enriching a core made of a graphite moderator with an isotope having a mass number of 7.
eF ₂₎ a molten salt comprising a mixture of a solvent, fission cross-sectional area and 3 fluorides 3 fluoride and fission products elements transuranic elements circulate molten salt reactor fuel as a solute is relatively It relates to the composition of a molten salt reactor fuel suitable for a reactor for effectively fissioning low transuranium elements, and determines the solubility of transuranic element trifluoride and fission product element trifluoride in a solvent. Higher, thus increasing the concentration of these solutes that can be dissolved in the solvent, facilitating the operation of nuclear reactors with high nuclear reactivity and high burn rates, and further reducing the possibility of solute precipitation / precipitation during operation. To improve safety,
The present invention relates to a molten salt reactor fuel which allows the presence of fission products to reduce the need for chemical treatment of the molten salt reactor fuel and to improve economic efficiency.

[0002]

2. Description of the Related Art A molten salt reactor is a chemically stable salt mixture composed of elements having a small parasitic neutron capture cross section and elements that propagate fissile elements as necessary.
A solvent is a substance that melts at a temperature at least 50 ° C. lower than the minimum operating temperature of a nuclear reactor, has a low vapor pressure in a molten state, and has physical properties suitable as a heat medium such as viscosity and specific heat, as a solvent,
If necessary, the same salt of fission element and fission product element is dissolved as a solute to fill the molten fuel salt into the core where the neutron moderator is placed, and the fission progresses in the core. The elevated molten fuel salt is guided to a heat exchanger outside the core to transfer heat to the secondary coolant, and the molten fuel salt whose temperature has dropped is returned to the core for circulation.

In the first molten salt nuclear reactor, uranium ( ²³³ U) with a mass number of 233 is fissionable as a fissionable material in the core, and thorium ( ²³² Th) as a breeding material is a mass number of 233 which is generated by absorbing neutrons. Protactinium (
It was developed at Oak Ridge National Laboratory in the United States for the purpose of decaying ²³³ Pa) and propagating ^{23 3} U more than the fissionable amount.

In the molten salt reactor, the fuel is in a molten state, so that the fuel salt can be freely taken in and out, and ²³³ Pa is continuously separated from the fuel salt to be converted into ²³³ U, or a nuclear fission that absorbs neutrons. It was possible to remove the product, which was advantageous for establishing a breeder reactor. The background and technical contents of this development are Nuclear Applications & Technology, Vol.8,
Published in February 1970 as eight reports.

Fuel salts used in a 7.3 MW (t) molten salt experimental reactor constructed and operated for molten salt reactor research gave compounds with a low neutron capture cross section and a low melting point. lithium mass number 7 is a component, beryllium, and mixtures fluorides zirconium ⁽⁷ LiF-BeF
_2- ZrF ₄ ) as a solvent, and the composition thereof is a condition that gives a low liquidus temperature (that is, a temperature at which a solid phase is first precipitated when cooled) and viscosity are 65-30-5 mol.
e% and the fuel material is uranium tetrafluoride (UF ₄ ) 0.9
Mole% ( ²³³ UF ₄ is 0.3 mol%) and the liquidus temperature was 434 ° C. The fission cross section of ²³⁵ U at the average operating temperature of 645 ° C is 329 × 1
0 is ^-24 cm ^2. The macroscopic fission cross section that governs the nuclear reactivity of the core is determined by the density of solvent molecules in a unit volume of 3.3.
It is 0.033 cm ^−1, which is obtained by multiplying the microscopic nuclear fission cross section by ²³⁵ U atom density 0.99 × 10 ²⁰ atoms / cm ³ obtained from × 10 ²² atoms / cm ³ .

Further, the composition of the molten salt solvent in a typical fuel salt proposed as a fuel for a 1000 MW (e) molten salt breeder reactor is lithium and beryllium having a mass number of 7 in order to satisfy the above requirements. , And a mixed fluoride of thorium ( ⁷ LiF-BeF ₂ -ThF ₄ ) as a solvent,
Its composition was 72-16-12 mole%, the fuel material was uranium tetrafluoride ( ²³³ UF ₄ ) 0.4 mole%, and the liquidus temperature was 500 ° C. The fission cross section of ²³³ U at average operating temperature of 637 ° C is 308 ×
It is 10 ^-24 cm ² . The macroscopic fission cross section that governs the nuclear reactivity of the core is the density of solvent molecules in a unit volume.
It is 0.039 cm ^−1, which is obtained by multiplying the microscopic nuclear fission cross section by ²³³ U atom density 1.28 × 10 ²⁰ atoms / cm ³ obtained from 2 × 10 ²² atoms / cm ³ .

As a method for removing rare earth fission products having a large neutron capture cross section from molten salt reactor fuel, the solubility of rare earth trifluoride is measured in order to utilize the phenomenon that rare earth trifluoride precipitates as a solid solution. I was broken. Further, in the development of pursuing the possibility of operating the plutonium separated from spent LWR fuel before increasing the concentration of ^{233 U} in the molten salt breeder reactor as fuel, LiF-BeF ₂ -
Plutonium trifluoride (Pu) in ThF ₄ type solvent
Solubility measurements of F ₃₎ is performed. The dissolution behavior of plutonium and rare earth element trifluorides, which may form a solid solution, is similar, and the smaller the size of the cation, the higher the solubility of fluoride, and the solubility of each trifluoride in the mixture of trifluorides. Is a numerical value obtained by multiplying the solubility of a single trifluoride by the concentration fraction, and it has been found that the solubilities of transuranium element trifluorides are equal to each other and such regularity is found. In addition, the effects of solvent composition and temperature on the solubility of trifluoride have been studied,
Solubility is 67-33 m in LiF-BeF ₂ system solvent
It has been shown that it becomes minimum near the composition of ole% and increases with temperature.

Of the plutonium separated from the spent LWR fuel, 58.8% was fissionable ²³⁹ Pu,
If ²⁴¹ Pu are to be 12.1%, the average operating temperature 637 ° C. of the molten salt breeder reactor, the effective fission cross section plutonium is 317 × ^{10 -2} 4 cm ^2,
The concentration of PuF ₃ should be 0.39 mole% in order to make the macroscopic cross section equal to that of the case of operating at ²³³ U. The fuel salt temperature of the molten salt breeder reactor is 566 ° C, which is the lowest at the entrance to the core.
If the solubility of PuF ₃ at 0 ° C. is 0.39 mole% or more, it is considered that PuF ₃ can be used as the initial operation fuel of the molten salt breeder reactor. Indeed, _LiF-BeF 2 -Th a solvent of the molten salt breeder reactor fuel salt
F ₄ PuF ₃ in (72-16-12mole%)
Had a solubility of 0.68 mole% at 516 ° C.

Recently, since molten salt reactors can freely move fuel in and out, V. Prusakov, P. Aledsee
v, A. Dudnikov, S. Subbotin, R. Zakirov, V. Lelek,
andI. Peka, “Concept of the Demonstration Molten
Salt Unit for the Transuranium Elements Transmuta
tions, ”ADTTA'99, June 6-11, 1999, Praha, Czec
As published in h Republic, We-I-14, Np, P
In the thermal neutron spectrum of u, Am, Cm, etc., it is recognized as a value for use in burning out transuranium elements including nuclides having a small fission cross section by fission. In this case, since the nuclides with a large fission cross section preferentially fission, the effective fission cross section becomes smaller as the combustion progresses, and in order to give the macroscopic fission cross section necessary to keep the core critical. It is necessary to increase the concentration of transuranium trifluoride in the fuel salt. Further, if the fuel salt can be operated with the fission product contained, the chemical treatment cost can be saved. However, the presence of trifluoride, a rare earth fission product, lowers the permissible concentration of transuranium element trifluoride. It is necessary to evaluate the solubility of trifluoride and rare earth fission product trifluoride in a comprehensive manner, and higher solubility of trifluoride is required as compared with the solubility of trifluoride in the conventional molten salt breeder reactor. It

[0010]

The object of the present invention is to provide ⁷ L
By adding the third component to the molten salt solvent consisting of a mixture of iF and BeF ₂ and adjusting the composition of the molten salt solvent to give the desired liquidus temperature, the transuranium trifluoride trifluoride in the molten salt solvent is adjusted. Of Fluoride and Fission Product Trifluoride at a constant temperature, and melting of the transuranium trifluoride and Fission Product Trifluoride in appropriate concentrations required for molten salt reactor operation To provide salt reactor fuel.

[0011]

In order for a nuclear reactor to be nuclearly established, the number of neutrons generated in the core by nuclear fission must be larger than the number of neutrons lost by absorption and leakage from the core. The number of neutrons generated depends on the fission density (fiss).
ions / cm ³ −s) and the number of neutrons generated per fission, and the fission density is proportional to the macroscopic fission cross section which is the product of the fission cross section and the fission nuclide density.

The absorption of neutrons in the core of a molten salt nuclear reactor is performed by fission reaction and capture reaction of fissionable nuclides, capture reaction of fission product atoms, capture reaction by constituent atoms of fuel salt solvent, and capture by graphite moderator atoms. There is absorption due to reaction. All of these are proportional to the macroscopic cross section, which is the product of fission or capture cross section and constituent atom density. The reaction cross section of the neutron decreases in inverse proportion to the square root of the absolute temperature. For example, the reaction cross section at 650 ° C. is the thermal neutron (293
K) Calculated by multiplying the reaction cross section by (293/923) ^1/2 , and if necessary, add 5% of the resonance absorption cross section by epithermal neutrons.

The ratio of the neutron generation rate to the neutron absorption rate is a neutron multiplication coefficient. If it is greater than 1, it becomes critical and the nuclear reactor is established. The leakage of neutrons is small if the core system is large, but it is assumed that the reactor will be established when the neutron multiplication factor becomes 1.04 with a probability of 4%.

Of fissile materials used in conventional nuclear reactors
Fission cross section at 650 ° C (including 5% of resonance absorption
If)²³⁵U; 342 × 10^-24cm^Two,
²³³U; 337 × 10^-24cm^Two,²³⁹Pu; 4
33 x 10^-24cm^TwoBut fission in the present invention
The target is a transuranic element such as Np, Pu, Am, Cm.
Isotope mixture with a large fission cross section
The body preferentially fission and is lost. Therefore, effective
The fission cross section depends on the isotopic composition and is 120 × 10
^-24cm^TwoThrough 13 × 10^-24cm^TwoIs. this
Holds the criticality in the fission of materials with small fission cross section
To do this requires high concentrations of fissile material.

Uranium in the molten fluoride salt is stable in the tetravalent valence state and exists as UF ₄ , and there is no problem of solubility, but transuranic elements such as Np, Pu, Am and Cm are trivalent. Has a stable valence state and a small neutron capture cross-section, so it is important as a solvent for fluoride molten salt nuclear reactor fuel in LiF-BeF ₂ binary mixed fluoride solvent.
The solubility of fluorides is particularly limited.

[0016] LiF-BeF ₂ is the dissolution behavior of transuranic elements 3 fluoride and rare earth elements 3 fluoride in the binary system mixed fluorides solvent known regularity as follows. (1) Trifluorides of the same crystal form form a solid solution and have similar dissolution behavior. Both transuranium trifluoride and rare earth trifluoride give crystals belonging to the hexagonal system. (2) In a mixture of trifluorides forming a solid solution, the solubility of a specific trifluoride is obtained by multiplying the solubility when the trifluoride is present alone by the composition fraction in the solid solution. (3) The smaller the diameter of the cation, the greater the solubility of trifluoride. Since the ion diameter of the rare earth element in the trivalent valence state is smaller than the ion diameter of the transuranium element in the trivalent valence state, the solubility of the rare earth element trifluoride is equal to the solubility of the transuranium element trifluoride. If anything, it is modest. (4) Solubility of trans-uranium element trifluoride is almost equal. (5) The solubility of PuF ₃ and CeF ₃ is LiF-BeF ₂
It becomes a minimum in the composition of (63-37 mole%),
When LiF is increased to 67 mole% or more, it is rapidly increased, but even when BeF ₂ is increased to 37 mole% or more, the increase is small.

The invention of claim 1 is derived from the following regularity in addition to the regularities of (1) to (5) above. (6) ZrF the BeF ₂ of LiF-BeF ₂ based mixture ₄
Substitution with 3 increases the solubility of the trifluoride. LiF-B
This is because the ion potential ⁴ of Zr ⁴⁺ is closer to the ion potential ³ of Pu ³⁺ and Ce ³⁺ rather than the ion potential (valence / ion radius) of Be ²⁺ of 6 in the eF ₂ mixture. Conceivable. (7) 2Li with _LiF-BeF 2 -ZrF ₄ based mixture
3 depending on the amount of free fluorine ion (F ⁻ ) when it is assumed that the F-BeF ₂ complex and the Li ₂ ZrF ₆ complex are present.
The solubility of fluoride increases.

The molten salt reactor fuel according to the invention of claim 1 is a molten salt solvent consisting of ⁷ LiF and BeF ₂ in ZrF ₄
Is added, and the composition of the three components is adjusted so that the liquidus temperature of the solvent is lower than 500 ° C., so that the transuranium element trifluoride and the fission product trifluoride are heated at 500 ° C. which is not lower than the liquidus temperature of the solvent. It is composed of a molten salt solvent having a high solubility of the compound, and the composition of the solvent is ⁷ LiF-BeF ₂ -Zr.
75-5-20mo on the phase diagram of the F ₄ ternary system
le%, 61-11-20 mole%, 67-33-0
It is within the range of a quadrangle connecting the points of mole% and 69-31-0 mole% in sequence, the liquidus temperature is lower than 500 ° C, and the working temperature is higher than 500 ° C.

[0019] Since the liquidus temperature LiF composition increases as shown in the phase diagram of the _LiF-BeF 2 -ZrF ₄ ternary system of FIG. 1 is increased, the upper limit of LiF composition allowed by the operating conditions of the reactor It is determined by the maximum liquidus temperature of the fuel salt. When the minimum operating temperature of the fuel salt circulating in the core is 550 ° C, the maximum liquidus temperature of the fuel salt allowed is 500 ° C, so the upper limit of the LiF composition gives a liquidus temperature lower than 500 ° C. Must. Specifically, the upper limit of the LiF composition differs depending on the composition of ZrF ₄ , and is 69 mo
From le% to 0.33 mole% per 1 mole% increase of ZrF ₄ composition. The lower limit of the LiF composition is not particularly limited, and the LiF composition is about 50 mol.
If it decreases to about e%, the liquidus temperature of the LiF—BeF ₂ system mixture decreases to 400 ° C. or lower, but since the viscosity of the molten salt increases with the increase of the BeF ₂ composition, the lower limit is 67 mole% as the LiF composition. To do.

Further, as shown in FIG. 1, when the ZrF ₄ composition increases above a certain level, the liquidus temperature rises, and Li
It should give a liquidus temperature below 500 ° C. as well as the upper limit of F composition. Specifically, ZrF ₄
The upper limit of the composition is 20 mole%. The lower limit of ZrF ₄ should be one which gives a liquidus temperature below 500 ° C., corresponding to the LiF composition. Specifically, ZrF increases with increasing LiF composition from 69 mole% to 1 mole%.
The lower limit of the _four compositions increases by 3 mole%.

The thermal neutron capture cross section (0.18 × 10 ⁻²⁴ cm ² ) of ZrF ₄ substituting BeF ₂ is compared with the thermal neutron capture cross section of BeF ₂ (0.01 × 10 ⁻²⁴ cm ² ). 18 times higher and standard ⁷ LiF-BeF
₂ (67-33 mole%) for the thermal neutron capture cross section (0.025 × 10 ⁻²⁴ cm ² ) for all B
If eF ₂ is replaced with ZrF ₄ , the thermal neutron capture cross section (0.082 × 10 ⁻²⁴ cm ² ) is increased by 3.32 times. For this reason, the upper limit of the ZrF ₄ composition is determined in terms of the thermal neutron capture cross-section allowed as a solvent for the molten salt reactor fuel. Furthermore, the molecular volume of BeF ₂ (23.67
cm ³ / mole) compared to the molecular volume of ZrF ₄ (3
7.75 cm ³ / mole) is as large as 1.6 times, and all of BeF ₂ is added to ZrF with respect to the standard ⁷ LiF-BeF ₂ (67-33 mole%) molecular volume (14.41 cm ³ / mole). If replaced by ₄ , the molecular volume (1
(9.05 cm ³ / mole) is 1.3 times larger, and the number of molecules contained in a unit volume of the fuel salt solvent is 0.76 times lower. This is ⁷ LiF-BeF ₂ (67-33 mol
e%) Even at the same mole concentration as in the solvent, it means that the number of solute molecules in the unit volume of ⁷ LiF-ZrF ₄ (67-33 mole%) solvent is 0.67 times smaller.

Due to the restrictions relating to the coexistence of molten salt reactor fuel and reactor constituent materials, the maximum allowable temperature of fuel salt is about 700.
℃. The minimum liquidus temperature of the fuel salt solvent and the solubility evaluation temperature of transuranium trifluoride of 500 ° C allow the minimum temperature of the fuel salt to 550 ° C with a margin, and the high temperature side and the low temperature side of the heat exchanger. The temperature of 150 ° C. can be applied on the side, and the thermal efficiency of the power generation equipment can reach 44.4%.

The invention of claim 2 is derived from the following regularity in addition to the regularities of (1) to (7). (8) In the composition of LiF-BeF ₂ (67-33 mole%), the heat of solution of transuranium trifluoride is 14.2 k.
cal / mole, temperature is from 500 ° C to 550 ° C
The solubility of trans-uranium element trifluoride increases 1.5 times as much as.

The molten salt nuclear reactor fuel according to the invention of claim 2 is a molten salt solvent composed of ⁷ LiF and BeF ₂ in ZrF ₄
Is added and the composition of the three components is adjusted so that the liquidus temperature of the solvent is lower than 550 ° C., so that the transuranic element trifluoride and the fission product trifluoride are heated at 550 ° C. which is not lower than the liquidus temperature of the solvent. It is composed of a molten salt solvent having a high solubility of the compound, and the composition of the solvent is ⁷ LiF-BeF ₂ -Zr.
77-3-20mo on the phase diagram of the F ₄ ternary system
le%, 61-11-20 mole%, 67-33-0
It is within the range of a quadrangle connecting the points of mole% and 71-29-0 mole% in sequence, the liquidus temperature is lower than 550 ° C, and the working temperature is higher than 550 ° C.

[0025] Since the liquidus temperature LiF composition increases as shown in the phase diagram of the _LiF-BeF 2 -ZrF ₄ ternary system of FIG. 2 is high, the upper limit of LiF composition allowed by the operating conditions of the reactor Determined by the maximum allowable liquidus temperature of the fuel salt When the minimum operating temperature of the fuel salt circulating in the core is 600 ° C., the maximum liquidus temperature of the fuel salt is allowed up to 550 ° C., and the upper limit of the LiF composition gives the liquidus temperature up to 550 ° C. Good. Specifically, the upper limit of the LiF composition is about 2 mole% higher than that of the invention of claim 1.
Although there is no particular reason for limiting the lower limit of the LiF composition, the lower limit is 67 mole% as in the first aspect of the invention.

Further, as shown in FIG. 1, when the ZrF ₄ composition increases to a certain extent or more, the liquidus temperature rises.
When the iF composition is 67 to 69 mole%, the upper limit of the ZrF ₄ composition is 25 in order to give a liquidus temperature up to 550 ° C.
It is allowed up to mole%. However, the upper limit of the ZrF ₄ composition is limited for the same reason as in the invention of claim 1. Further, for the same reason as in the invention of claim 1, the lower limit of the ZrF ₄ composition must be such that the liquidus temperature is lower than 550 ° C. at the upper limit of the LiF composition.

Due to the limitation on the coexistence of the molten salt reactor fuel and the constituent materials of the reactor, the maximum allowable temperature of the fuel salt is about 700.
℃. The minimum liquidus temperature of the fuel salt solvent and the solubility evaluation temperature of transuranium trifluoride of 550 ° C allow the minimum temperature of the fuel salt to be 600 ° C with a margin, and the high temperature side and low temperature of the heat exchanger. The temperature of 100 ° C. can be applied on the side, and the thermal efficiency of the power generation equipment can reach 42%.

The invention of claim 3 is based on the above (1) to (5).
In addition to the regularity of (8), it derives from the following regularity. (9) the BeF ₂ of LiF-BeF ₂ two-component mixtures T
Substitution with hF ₄ increases the solubility of trifluoride. Li
Th, rather than 6, which is the ionic potential (valence / ion radius) of Be ^{2+ in} the F-BeF ₂ binary mixture.
⁴ which is the ion potential of ⁴⁺ is Pu ³⁺ or Ce.
It is considered that this is because it is close to 3 which is the ion potential of ³⁺ . (10) The solubility of trifluoride depends on the amount of free fluoride ion (F ⁻ ) when it is assumed that 2LiF-BeF ₂ complex and LiThF ₅ complex are present in the LiF-BeF ₂ -ThF ₄ ternary mixture. To increase.

The molten salt nuclear reactor fuel according to the invention of claim 3 is a molten salt solvent consisting of ⁷ LiF and BeF ₂ in ThF ₄
Is added, and the composition of the three components is adjusted so that the liquidus temperature of the solvent is in the range of 550 ° C. and 500 ° C. High solubility of the product trifluoride, ²³
It is composed of a molten salt solvent capable of continuously replenishing ³ U as a fissile material, and the composition of the solvent is ⁷ LiF-BeF.
76-6- on the phase diagram of the _2- ThF ₄ ternary system
18 mole%, 69-31-0 mole%, and 71
It is characterized in that it is within the range of a triangle connecting the points of -29-0 mole% sequentially, the liquidus temperature is within the range of 500 ° C and 550 ° C, and the operating temperature is higher than 550 ° C.

As shown in the phase diagram of the LiF-BeF ₂ -ThF ₄ ternary system in FIG. 3, the liquidus temperature increases as the LiF composition increases, so the upper limit of the LiF composition is allowed under the operating conditions of the reactor. It is determined by the maximum liquidus temperature of the fuel salt. When the minimum operating temperature of the fuel salt circulating in the core is 600 ° C, the maximum liquidus temperature of the fuel salt is allowed up to 550 ° C.
The upper limit of the iF composition may give a liquidus temperature of up to 550 ° C. Specifically, the upper limit of the LiF composition is ThF.
Depends on the composition of No. ₄ , from 71 mole% to ThF
0.22 mole per 1 mole% increase of ₄ compositions
Increase by%. In the invention of claim 3, LiF-Be
Since the liquidus temperature of the F ₂ -ThF ₄ ternary mixture is limited to the range between 550 ° C. and 500 ° C., the lower limit of the LiF composition is limited to 69 mole%.

Further, as shown in FIG. 3, when the ThF ₄ composition increases above a certain level, the liquidus temperature rises.
When the iF composition is 69 to 75 mole%, the upper limit of the ThF ₄ composition is 18 in order to give a liquidus temperature up to 550 ° C.
It is allowed up to mole%. The lower limit of ThF ₄ should be one which gives a liquidus temperature below 550 ° C., corresponding to the LiF composition. Specifically, the LiF composition is 71 mol
The lower limit of the ThF ₄ composition increases by 4.5 mole% per 1 mole% increase from e%.

ThF_FourIs simply BeF_TwoTo replace trifluoride
Not only to increase the solubility of things,^{Two 32}Th is thermal neutron
Capture (thermal neutron capture cross section 7.4 x 10^-24cm
^Two)do it²³³Generate Pa,²³³Pa decays by β
²³³It becomes U and is a proliferative substance because it undergoes nuclear fission.
²³³U is UF_FourDissolves in a solvent as
It does not affect the solubility. Molten salt breeder fuel
Then ThF_Four12 mole% was proposed as the composition
But did not aim for proliferation²³²Generate from Th
²³³ThF as a nuclear reactor that burns U without excess or deficiency_Four
The composition may be 10%. Under this condition,
Do²³³U amount burns in the reactor²³³U amount and fishing
Therefore, the transuranic element, which is the field of application of the present invention, can be efficiently used.
It is inconvenient as a fuel salt for nuclear fission reactors. Shi
However, transuranic nuclei with large fission cross section and easy fission
The fission section that the species precedes fission and occupies in the absorption cross section
Transuranic nuclides with reduced product ratio are fission-induced
In a nuclear reactor that reduces to a high concentration²³²From Th
Generated²³³The neutrons generated by the fission of U
It can be used effectively.

BeF_TwoReplacing ThF_FourThermal neutron capture
Cross-sectional area of catch (7.4 × 10^-24cm ^Two) Is BeF_TwoHeat of
Neutron capture cross section (0.01 × 10^-24cm^Two) To
740 times higher than the standard⁷LiF-BeF
_TwoThermal neutron capture cross section of (67-33 mole%)
(0.025 x 10^-24cm^Two) Of 10%
BeF _TwoZrF_FourThermal neutron capture cross section
Product (0.76 × 10^-24cm ^Two) Increases about 30 times
It High neutron capture cross section is high²³³U yield means
And equilibrate²³³It is possible to burn transuranium elements under U concentration.
Disappear. On the other hand, equilibrium^{23 Three}Furnace until U concentration is reached
Construct trifluoride to compensate for parasitic neutron capture in the heart
The load concentration of transuranic elements must be further increased
Absent. Therefore, it is acceptable as a solvent for molten salt reactor fuel.
From the point of thermal neutron capture cross section_FourThe upper limit of composition is controlled
Limited Furthermore, BeF_TwoMolecular volume of (23.67c
m^Three/ Mole) compared to ThF_FourMolecular volume of (4
4.27 cm^Three/ Mole) is as large as 1.87 times,
Quasi⁷LiF-BeF_Two(67-33 mole%)
Molecular volume (14.41 cm^Three/ Mole) for 10
mole% BeF_TwoThF_FourIf you replace with
Child volume (16.47 cm^Three/ Mole) is 1.14 times
The number of molecules contained in a large unit fuel salt solvent is 0.
875 times. this is,⁷LiF-BeF_Two(6
7-33 mole%) At the same mole concentration as in the solvent
Even of unit volume⁷LiF-BeF_Two-ThF_Four(67
The number of solute molecules in the solvent is 0.8.
It means 75 times less.

Due to the restrictions relating to the coexistence of molten salt reactor fuel and reactor constituent materials, the maximum allowable temperature of fuel salt is about 700.
℃. The minimum liquidus temperature of the fuel salt solvent and the solubility evaluation temperature of transuranium trifluoride of 550 ° C allow the minimum temperature of the fuel salt to be 600 ° C with a margin, and the high temperature side and low temperature of the heat exchanger. The temperature of 100 ° C. can be applied on the side, and the thermal efficiency of the power generation equipment can reach 42%.

The molten salt reactor fuel according to the invention of claim 4 is the molten salt reactor fuel according to the invention of claim 3,
Th should be contained in an appropriate amount to effectively burn transuranium elements.
The molten salt raw material according to the invention of claim 3, which is made in view of the fact that when the addition amount of F ₄ is reduced, the addition amount of ThF ₄ necessary for increasing the solubility of transuranium trifluoride is not reached. ZrF _{4 is used as} a part of ThF _{4 in} the reactor fuel.
Is not lower than the liquidus temperature of the solvent by replacing with 5
It consists of a molten salt solvent capable of reducing the amount of ThF ₄ and limiting the replenishment amount of ²³³ U to increase the burning rate of transuranium element while maintaining high solubility of transuranic element trifluoride at 50 ° C. is there.

As a result, it is possible to increase the transuranic element concentration in the fuel salt and increase the fission ratio of the transuranic element to the ²³³ U fission. In addition, excess ThF ₄
By substituting ZrF ₄ for ZrF ₄ , the capture cross section of the solvent is reduced, the neutron multiplication factor is increased, and it becomes possible to include fission products in the fuel salt.

[0037]

BEST MODE FOR CARRYING OUT THE INVENTION Embodiment 1. Mode for Carrying Out the Invention
The molten salt reactor fuel according to state 1 is LiF-B shown in FIG.
eF _Two-ZrF_FourThe composition of the ternary system, the liquidus temperature and the contour line
The liquidus temperature on the partial phase diagram shown is lower than 500 ° C.
The composition range is higher than 450 ° C. Its composition range
Mole% set of each component for the sample collected from the box
Formation, thermal neutron capture cross section of solvent, Li_TwoBeF_FourAnd Li
_TwoZrF₆, The number of free fluorine ions obtained as
Molecular volume occupied by solvent of 1 gram molecule at 500 ℃
Value, Pu concentration ratio in solution [Pu] / [Li + Be + Zr
+ Pu] Mole% Pu at 500 ° C.
F_ThreeAnd the molecular weight of Pu gram in a unit volume of solvent
PuF at 500 ° C defined_ThreeSolubility
This is shown in Table 1. Figure 1 shows the sample numbers corresponding to each composition in Table 1.
Shown in.

[0038]

[Table 1]

The calculation results under each condition will be described below.

[0040]Condition 1-1 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0141
× 10^-24cm^Two] [3.66 × 10²²atom /
cm^Three] [0.1] = 0.000052cm^-1)
The effective molten fission cross section is 1 in the molten salt reactor
19 x 10^-24cm^Two, Effective capture cross section is 169 × 1
0^-24cm^TwoAnd the neutron yield is 2.96
When the uranium element undergoes fission, fission products are always removed.
If it is removed and does not exist, the transuranic element concentration is 0.
6 x 10²⁰atoms / cm^Three(1.00 x 10^-4
mole / cm^Three) The macroscopic fission cross section of the core
Is 0.00071 cm^-1, The macroscopic capture cross section is 0.0
0101 cm^-1And the neutron multiplication factor is 1.04
Reaches criticality.

This transuranium trifluoride concentration is the sample number
1 67% LiF-33% BeF_Two-0% ZrF_FourMelting
Resolution 1.15 × 10^-4mole / cm^Three(0.69 x
10 ²⁰atoms / cm^Three) Can be satisfied.

[0042]Condition 1-2 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0248
× 10^-24cm^Two] [3.22 × 10²²atom /
cm^Three] [0.1] = 0.000080 cm^-1)
The effective molten fission cross section is 1 in the molten salt reactor
19 x 10^-24cm^Two, Effective capture cross section is 169 × 1
0^-24cm^TwoAnd the neutron yield is 2.96
When uranium element undergoes nuclear fission, effective capture cross section 6 × 1
0^-24cm^TwoFission products of the core have a core heat power density of 19.
5 W / cm^Three(5.93 x 10¹¹missions /
cm^Three-S) 1.0 × 10⁸accumulated by driving
2.51 × 10 in fuel salt (including outside the core)²⁰
atoms / cm^ThreeThen the core macroscopic capture cross section is
0.000151 cm^-1And the transuranium element concentration is
0.95 x 10²⁰atoms / cm^Three(1.58 x 1
0^-4mole / cm^Three) Macroscopic nuclear fission in
Cross-sectional area is 0.00113 cm^-1, The macroscopic capture cross section is
0.00161 cm^-1And the neutron multiplication factor is 1.
It becomes 04 and reaches criticality.

7.53 × 10 ¹⁹ at for fission products
^{oms / cm 3 (1.25 × 10} - 4 mole / c
m ³ ), which includes rare earth element trifluoride, and the total amount of transuranium element trifluoride is 1.70 × 10 ²⁰ atoms /
cm ³ (2.83 × 10 ⁻⁴ mole / cm ³ ) is included in the trifluoride. This trifluoride concentration is sample number 7
Of _LiF-BeF 2 -ZrF 4 _(72-18-10mo
Le%) solubility 3.05 × 10 ⁻⁴ mole / cm ³
Can be satisfied by

[0044]Condition 1-3 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0344
× 10^-24cm^Two] [3.02 × 10²²atom /
cm^Three] [0.1] = 0.000104cm^-1)
The effective molten fission cross section is 1 in the molten salt reactor
19 x 10^-24cm^Two, Effective capture cross section is 169 × 1
0^-24cm^TwoAnd the neutron yield is 2.96
When uranium element undergoes nuclear fission, effective capture cross section 6 × 1
0^-24cm^TwoFission products of the core have a core heat power density of 19.
5 W / cm^Three(5.93 x 10¹¹missions /
cm^Three-S) 1.5 × 10⁸accumulated by driving
And the fuel salt (including outside the core) concentration is 3.76 × 10²⁰
atoms / cm^ThreeThen the core macroscopic capture cross section is
0.000226 cm^-1And the transuranium element concentration is
1.15 x 10²⁰atoms / cm^Three(1.91 x 1
0^-4mole / cm^Three) Macroscopic nuclear fission in
Cross-sectional area is 0.00137 cm^-1, The macroscopic capture cross section is
0.00194 cm^-1And the neutron multiplication factor is 1.
It becomes 04 and reaches criticality.

1.13 × 10 ²⁰ at for fission products
^{oms / cm 3 (1.87 × 10} - 4 mole / c
m ³ ), which contains rare earth element trifluoride, and the total amount of transuranium element trifluoride is 2.28 × 10 ²⁰ atoms /
cm ³ (3.78 × 10 ⁻⁴ mole / cm ³ ) will be included. This trifluoride concentration is sample number 8
Of _LiF-BeF 2 -ZrF 4 _(75-5-20mol
e%) solubility of 3.11 × 10 ⁻⁴ mole / cm ³ is also unsatisfactory.

According to the first embodiment, the following effects can be obtained. (1) When the minimum operating temperature of the nuclear reactor is about 550 ° C. and the maximum allowable liquidus temperature of the fuel salt is 500 ° C., ZrF ₄ is added to the fuel salt solvent, and transuranium element 3 at 500 ° C. By providing a fuel salt with high fluoride solubility and increasing the transuranic element concentration in the fuel salt, it is possible to limit the nuclear reactor that is composed of transuranic elements having a certain low fission cross section and contains some fission products. Can be reached or the reactivity can be increased. (2) It is possible to select a fuel salt that is less affected by the increase in the thermal neutron capture cross section due to the addition of ZrF ₄ .

Embodiment 2. According to the second embodiment of the present invention
The molten salt nuclear reactor fuel is LiF-BeF shown in FIG. _Two−
ZrF_FourContour line of ternary composition and liquidus temperature
The liquidus temperature on the typical phase diagram is lower than 550 ° C.
It is shown in the composition range higher than 0 ° C. Taken from its composition range
The mole% composition of each component and the solvent
Thermal neutron capture cross section, Li_TwoBeF_FourAnd Li_TwoZrF₆
The number of free fluorine ions determined to be generated by 1 gram
Molecular volume value occupied by the child solvent at 550 ° C, in solution
Concentration ratio [Pu] / [Li + Be + Zr + Pu] m of
trans-uranium element 3 at 550 ° C defined as ole%
Fluoride solubility, Pu-gram molecular weight in unit volume of solvent
Of transuranium trifluoride at 550 ° C
The solubilities are shown in Table 2, respectively. Corresponds to each composition in Table 2
The sample numbers are shown in FIG.

[0048]

[Table 2]

The calculation results under each condition will be described below.

[0050]Condition 2-1 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0192
× 10^-24cm^Two] [3.54 × 10²²atom /
cm^Three] [0.1] = 0.000068cm^-1)
The effective molten fission cross section is 1 in the molten salt reactor
19 x 10^-24cm^Two, Effective capture cross section is 169 × 1
0^-24cm^TwoAnd the neutron yield is 2.96
When uranium element undergoes nuclear fission, effective capture cross section 6 × 1
0^-24cm^TwoFission products of the core have a core heat power density of 19.
5 W / cm^Three(5.93 x 10¹¹missions /
cm^Three-S) 1.5 × 10⁸accumulated by driving
And the fuel salt (including outside the core) concentration is 3.76 × 10²⁰
atoms / cm^ThreeThen the core macroscopic capture cross section is
0.000226 cm^-1And the transuranium element concentration is
1.10 x 10²⁰atoms / cm^Three(1.83 x 1
0^-4mole / cm^Three) Macroscopic nuclear fission in
Cross section is 0.00131cm^-1, The macroscopic capture cross section is
0.00186 cm^-1And the neutron multiplication factor is 1.
It becomes 04 and reaches criticality.

1.13 × 10 ²⁰ at for fission products
^{oms / cm 3 (1.87 × 10} - 4 mole / c
m ³ ), which contains the rare earth element trifluoride, and the total amount of the transuranium element trifluoride is 2.23 × 10 ²⁰ atoms /
cm ³ (3.70 × 10 ⁻⁴ mole / cm ³ ) will be included. This trifluoride concentration is sample number 1
2 of _LiF-BeF 2 -ZrF 4 _(67-28-5mo
le%) solubility 4.70 × 10 ⁻⁴ mole / cm ^3.
Can be satisfied by

[0052]Condition 2-2 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0146
× 10^-24cm^Two] [3.58 × 10²²atom /
cm^Three] [0.1] = 0.000052cm^-1)
In the molten salt reactor produced, the effective fission cross section is 7
6.4 x 10^-24cm^Two, The effective capture cross section is 111.
6 x 10^-24cm^TwoAnd the neutron yield is 2.96
When a transuranium element undergoes fission, fission products
If not present, transuranic element concentration is 1.10 x 1
0²⁰atoms / cm^Three(1.83 × 10^-4mol
e / cm^Three), The macroscopic nuclear fission cross section of the core is 0.
00084 cm^-1, The macroscopic capture cross section is 0.0012
3 cm^-1And the neutron multiplication factor is 1.04.
Reach the world.

This transuranium element concentration is L of sample number 11.
_{iF-BeF 2} -ZrF 4 (71-29-0mole
%) Solubility of 3.33 × 10 ⁻⁴ mole / cm ³ is satisfactory.

[0054]Condition 2-3 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0197
× 10^-24cm^Two] [3.55 × 10²²atom /
cm^Three] [0.1] = 0.000070 cm^-1)
In the molten salt reactor produced, the effective fission cross section is 7
6.4 x 10^-24cm^Two, The effective capture cross section is 111.
6 x 10^-24cm^TwoAnd the neutron yield is 2.96
Effective capture cross section when a transuranium element undergoes fission
6 x 10^-24cm^TwoFission products are core heat power density
19.5 W / cm^Three(5.93 x 10¹¹fissio
ns / cm^Three-S) 1.5 × 10⁸-S (4.76
Year) operation and accumulated fuel salt (including outside the core)
Concentration is 3.76 × 10²⁰atoms / cm^ThreeIf
Core macroscopic capture cross section is 0.000226 cm^-1And
The transuranic element concentration is 1.85 × 10²⁰atoms
/ Cm^Three(3.07 x 10^-4mole / cm ^Three)
And the macroscopic nuclear fission cross section of the core is 0.00141, macroscopic
Cross section is 0.00206 and neutron multiplication factor
Becomes 1.04 and reaches criticality.

1.13 × 10 ²⁰ at for fission products
^{oms / cm 3 (1.87 × 10} - 4 mole / c
m ³ ), which contains rare earth element trifluoride, and the total amount of transuranium element trifluoride is 2.98 × 10 ²⁰ atoms /
cm ³ (4.95 × 10 ⁻⁴ mole / cm ³ ) will be included. This trifluoride concentration is sample number 1
5 of _LiF-BeF 2 -ZrF 4 _(70-25-5mo
le%) solubility 5.90 × 10 ⁻⁴ mole / cm ^3.
Can be satisfied by

[0056]Condition 2-4 Average temperature 650 with graphite moderator and core volume ratio of 10%
℃ fuel salt (solvent salt macroscopic capture cross section [0.0349
× 10^-24cm^Two] [3.37 × 10²²atom /
cm^Three] [0.1] = 0.000118 cm^-1)
The effective molten fission cross section is 1 in the molten salt reactor
3.1 x 10^-24cm^Two, The effective capture cross section is 27.9
× 10^-24cm^TwoAnd the neutron yield is 3.42
Transuranic elements undergo fission and fission products are constantly removed.
If not present, transuranic element concentration is 1.60
× 10²¹atoms / cm^Three(2.66 × 10^-3m
ole / cm^Three), The macroscopic fission cross section of the core is
0.00210 cm^-1, Macroscopic capture cross section is 0.00
446 cm^-1And the neutron multiplication factor is 1.04.
Reach criticality.

This transuranium element trifluoride concentration is
No. 20 LiF-BeF_Two-ZrF _Four(71-29-0
Mole%) solubility 1.12 × 10^-3mole / c
m^ThreeIs not satisfied even by.

According to the second embodiment, the following effects can be obtained. (1) When the minimum operating temperature of the reactor is about 600 ° C. and the maximum allowable liquidus temperature of the fuel salt is 550 ° C., ZrF ₄ is added to the fuel salt solvent, and transuranium element 3 at 550 ° C. Reactor composed of transuranic elements with a fairly low fission cross section by providing a fuel salt with a high solubility of fluoride and increasing the transuranic element concentration in the fuel salt and containing a significant amount of fission products. To reach criticality,
Alternatively, the reactivity can be increased. (2) It is possible to select a fuel salt that is less affected by the increase in the thermal neutron capture cross section due to the addition of ZrF ₄ .

Third Embodiment According to the third embodiment of the present invention
The molten salt reactor fuel is LiF-BeF shown in FIG. _Two−
ThF_FourContour line of ternary composition and liquidus temperature
The liquidus temperature given on the typical phase diagram is lower than 550 ℃.
The composition range is higher than 500 ° C. Its composition range
Mole% set of each component for the sample collected from the box
Sung, Li_TwoBeF_FourAnd LiThF₅Is generated as
Number of free fluorine ions, 1g molecule of solvent is 550 ℃
Volume value in the solution, Pu concentration ratio in the solution [P
u] / [Li + Be + Th + Pu] mole%
Solubility of transuranium trifluoride at 550 ° C,
Defined as Pu-gram molecular weight in a unit volume of solvent 55
Solubility of transuranium trifluoride at 0 ℃
It shows in Table 3. The sample number corresponding to the specific composition in Table 3
As shown in FIG. Sample numbers 21 to 29 are comparative samples,
This is outside the range of the third embodiment.

The calculation results under each condition will be described below.

[0061]

[Table 3]

[0062]Condition 3-1 Volume ratio to graphite moderator is 10% and Th is 6 mole%
F_FourMolten salt reactor fuel (at 650 ° C)
²³²Neutron capture cross section containing 5% of Th resonance integral cross section
Product = 0.580 × 10^-24cm^Two, Macroscopic capture of the core
Cross-sectional area = 0.00190cm^-1) Composed of
At²³²Th captures and produces neutrons
²³³UF_FourThe equilibrium concentration of is about 0.3 mole%.
At the average operating temperature of 650 ° C²³³U fission
Cross-sectional area is 299 × 10^-24cm^TwoAnd the neutron capture
Cross-sectional area is 56.5 × 10^-24cm^TwoIs. Unit volume
Solvent molecular density in 3.28 × 10²²atoms / c
m^ThreeSought from²³³U atom density 9.83 × 10¹⁹a
toms / cm^Three(1.63 x 10^-4mole / cm
^Three) And the macroscopic core content of the core
Cleavage cross-sectional area is 0.00294 cm^-1And the core macro
Neutron capture cross section is 0.000555 cm^-1And
It²³³Neutron yield due to fission of U is 2.49
Therefore, the neutron multiplication factor is 1.297.

The effective fission cross section is 13.1 × 10 ^−24.
cm ² , effective trapping cross section is 27.9 × 10 ⁻²⁴ cm ^2.
In and, _LiF-BeF 2 -ThF 4 of transuranic neutron yield is 3.42 is the sample number 33 (73-21
5.93 × 10 ⁻⁴ mole / cm ³ (3.57 × 10) which is a solubility of transuranium trifluoride of −6 mole%).
^When present at a concentration equal to or lower than ²⁰ atoms / cm ³ ), the macroscopic nuclear fission cross section of the core is 0.000468c.
m ^-1 , the macroscopic capture cross section of the core is 0.000996c
m ⁻¹ , and the neutron multiplication factor becomes 1.255 or more, which is critical.

²³³ U macroscopic fission cross section of the core
00294 cm ⁻¹ and macroscopic fission cross section of transuranium element 0.000468 cm ⁻¹ are 86.3% and 13.7%
Equivalent to. 1 GWt-y fission heat output is about 300 k
g fissile material, but in this case,
With the consumption of 259 kg of ²³³ U, 41 kg of transuranic element is consumed.

[0065]Condition 3-2 The volume ratio to graphite moderator is 10% and Th is 3 mole%.
F_FourMolten salt reactor fuel with addition of
Neutron capture cross section including 0.2% of the sound integration cross section = 0.26
7 x 10^-24cm^Two, Macroscopic capture cross section in the core
= 0.000945cm^-1)
And²³²Th captures and produces neutrons²³³U
F_FourEquilibrium concentration of about 0.15 mole% (5.31 x 1
0¹⁹atoms / cm^Three). At average operating temperature
At 650 ° C²³³U fission cross section is 299 ×
10^-24cm^TwoAnd the neutron capture cross section is 56.5.
× 10^-24cm^TwoIs. In the core²³³U giant
Cross section of visual fission is 0.00159cm^-1And is huge
Visual neutron capture cross section is 0.00030 cm^-1And
It²³³Neutron yield due to fission of U is 2.49
Therefore, the neutron multiplication factor is 1.283. Effective core
The fracture cross section is 13.1 × 10^-24cm^Two, Effective capture cross section
Product is 27.9 × 10^-24cm^TwoAnd the neutron yield is
The transuranium element of 3.42 is LiF- of the sample number 33.
BeF_Two-ThF_FourOver (73-21-6 mole%)
Uranium element trifluoride solubility 3.57 × 10²⁰a
toms / cm^ThreeMacroscopic of the core, if present below the concentration of
Nuclear fission cross section is 0.000468 cm^-1, The core giant
Visual capture cross section is 0.000996 cm^-1And inside
The sexual multiplication factor becomes 1.222 or more and becomes critical.

²³³ U macroscopic fission cross section of the core
Macroscopic fission cross section of 00159 and transuranium element 0.0
[00468] corresponds to 77.3% and 26.7%. 1G
The Wt-y fission heat output is generated by about 300 kg of fissile material, in this case 232 kg of fissile material.
With the consumption of ²³³ U, 68 kg of transuranium element is consumed. Here, _LiF-BeF 2 -ThF a solvent
₄ (72-25-3 mole%) at 3 at 550 ° C
The solubility of fluoride is 3.0 × 10 ²⁰ atoms / cm ³
It is estimated to be (5 × 10 ⁻⁴ mole / cm ³ ), and the concentration of transuranium element that can be actually loaded in the molten salt reactor fuel is reduced to 84% of the case where fission products are not included. Therefore, the combustion ratio of transuranium element is reduced to 22.4%.

[0067]Condition 3-3 Volume ratio to graphite moderator is 10% and Th is 2 mole%
F_FourMolten salt reactor fuel with addition of
Neutron capture cross section including 0.2% of the sound integration cross section = 0.24
4 x 10^-24cm^Two, Macroscopic capture cross section in the core
= 0.000644cm^-1)
And²³²Th captures and produces neutrons²³³U
F_FourEquilibrium concentration of about 0.1 mole% (3.54 × 10
¹⁹atoms / cm^Three). Average operating temperature
At 650 ° C²³³U fission cross section is 299 × 1
0^-24cm^TwoAnd the neutron capture cross section is 56.5 ×
10^-24cm^TwoIs. In the core²³³U macro
Nuclear fission cross section is 0.001058cm^-1And is huge
Visual neutron capture cross section is 0.000200 cm^-1And
It²³³Neutron yield due to fission of U is 2.49
Therefore, the neutron multiplication factor is 1.224. Effective core
The fracture cross section is 13.1 × 10^-24cm^Two, Effective capture cross section
Product is 27.9 × 10^-24cm^TwoAnd the neutron yield is
The transuranium element of 3.42 is LiF- of the sample number 33.
BeF_Two-ThF_FourOver (73-21-6 mole%)
Uranium element trifluoride solubility 3.57 × 10²⁰a
toms / cm^ThreeMacroscopic of the core, if present below the concentration of
Nuclear fission cross section is 0.000468 cm^-1, The core giant
Visual capture cross section is 0.000996 cm^-1And inside
The sexual multiplication factor becomes 1.171 or more and becomes critical.

²³³ U macroscopic fission cross section of the core
001058 cm ⁻¹ and the macroscopic fission cross section of transuranium element 0.000468 cm ⁻¹ are 69.3% and 31.7%.
Equivalent to%. The nuclear fission heat output of 1GWt-y is about 300.
It is generated by kg of fissile material, in which case 92 kg of transuranic element is consumed with 208 kg of ²³³ U consumed. Here, LiF- which is a solvent
BeF ₂ -ThF ₄ (71.5-25.5-2mole
%) Has a solubility of trifluoride at 550 ° C. of 2.4 × 1
0 ²⁰ atoms / cm ³ (4 × 10 ⁻⁴ mole / c
m ³ ), and the transuranic element concentration that can be actually loaded in the molten salt reactor fuel is 67 when the fission products are not included.
%. Therefore, the combustion ratio of transuranium element is 2
It drops to 1.3%.

[0069]Condition 3-4 Under the condition 3-3, fission products are generated in the molten salt reactor fuel.
Not included. Fission cross section is 299 × 10^-24cm
^TwoAnd the neutron capture cross section is 56.5 × 10 ^-24c
m^TwoAnd the neutron yield is 2.49²³³U is
²³³UF_FourAs 0.1 mole% (3.54 × 10
¹⁹atoms / cm^Three) Present and effective fission
Section area is 13.1 × 10^-24cm^Two, Effective capture cross section
Is 27.9 x 10^-24cm^TwoAnd the neutron yield is
Transuranic element trifluoride of 3.42 is 3.57 × 10
²⁰atoms / cm^ThreeExisting in the concentration of neutron multiplier
The number was 1.171. Furthermore, the effective neutron capture cross section
Product is 6 × 10^-24cm^TwoFission products of molten salt combustion
7.60 × 10 in the fee²⁰atoms / cm^ThreeAt the concentration of
Even if it exists, the neutron multiplication factor is 1.04 and the criticality is maintained.
Wear. The amount of this fission product is the core power density 19.5 W /
cm^Three(Fuel salt fission density 5.93 × 10¹²fiss
ion / cm^Three1.50 × 10 in −s)⁸-S
(4.76 years) Equivalent to driving and accumulating. here
The fission product is 2.28 × 1 rare earth element trifluoride.
0²⁰atoms / cm^ThreeTo contain molten salt reactor fuel
The medium is at least 5.85 × 10 as total trifluoride.²⁰a
toms / cm^Three(9.71 x 10^{− Four}mole / cm
^Three). Where is the solvent
LiF-BeF_Two-ThF_Four(72-26-2 mole
%) Has a solubility of trifluoride at 550 ° C. of 2.4 × 1
0²⁰atoms / cm^Three(4 x 10^-4mole / c
m^Three) And can actually be loaded into the molten salt reactor fuel
Transuranic element concentration is 41 when fission products are not included.
%. Therefore, the combustion ratio of transuranium element is 13
%.

According to the third embodiment, the following effects can be obtained. (1) When the minimum operating temperature of the reactor is about 600 ° C. and the maximum allowable liquidus temperature of the fuel salt is 550 ° C., ThF ₄ which is a breeding material is added to the fuel salt solvent, and the temperature at 550 ° C. At the same time as providing a fuel salt having a high solubility of transuranium trifluoride, neutrons generated in fission of ²³³ U produced can be used to burn transuranic elements having a remarkably low fission cross section with high efficiency. (2) It is possible to provide a molten salt reactor fuel having a high solubility of transuranium trifluoride that can burn a transuranic element having a small nuclear fission cross section.

Fourth Embodiment The calculation results under each condition will be described.

[0072]Condition 4-1 The volume ratio to graphite moderator is 10%, L of sample number 33
iF-BeF_Two-ThF_Four(73-21-6 mole
%) Solvent ThF_Four4 mole% of which is ZrF _FourSet in
2 mole% ThF_FourMolten salt reactor fuel containing
(Neutron containing 5% of resonance integral cross section at 650 ° C
Capture cross section = 0.187 × 10^-24cm ^Two, In the core
Macroscopic capture cross-section = 0.000663cm^-1)so
In the constructed reactor,²³²Th captures neutrons
And generate²³³UF_FourEquilibrium concentration of about 0.1 mol
e% (3.54 × 10¹⁹atoms / cm^Three)
It At the average operating temperature of 650 ° C²³³U core
Cross-section area is 299 × 10^{-2 Four}cm^TwoAnd the neutron
Capture cross section is 56.5 × 10^-24cm^TwoIs. Core
In²³³U macroscopic fission cross section is 0.0010
58 cm^-1And the macroscopic neutron capture cross section is 0.0
00200 cm^-1Is.²³³Due to U fission
Since the sexual yield is 2.49, the neutron multiplication factor is 1.
224. Effective fission cross section is 13.1 × 10
^-24cm^Two, The effective capture cross section is 27.9 × 10^-24
cm^TwoAnd a transuranium element with a neutron yield of 3.42
Element is 3.57 × 10²⁰atoms / cm^ThreeBelow the concentration
, The macroscopic fission cross section of the core is 0.000
468 cm^-1, The core macroscopic capture cross section is 0.000
996 cm^-1And the neutron multiplication factor is 1.165 or more.
Becomes critical.

Of the core²³³U macroscopic fission cross section 0.
001058 cm^-1Fission fragmentation of uranium and transuranium elements
Area 0.000468 cm^-1Are 69.3% and 31.7
Equivalent to%. The nuclear fission heat output of 1GWt-y is about 300.
generated by kg of fissile material, in this case
208 kg²³³Over 92kg with U consumption
Uranium element is consumed. Here, LiF- which is a solvent
BeF_Two-ThF_Four-ZrF_Four(73-21-2-4m
ole%) has a solubility of trifluoride at 550 ° C. of 3.
57 x 10²⁰atoms / cm^Three(5.93 x 10
^-4mole / cm ^Three) Is actually in the molten salt reactor fuel
3.57 × 10²⁰atoms / cm^ThreeSuper uranium yuan
Since the element can be loaded, the burning ratio of transuranium element is 31.
It will be 7%.

[0074]Condition 4-2 The volume ratio to the graphite moderator is 10%, and the sample number 36L
iF-BeF_Two-ThF_Four(77-5-18 mole
%) Solvent ThF_Four16 mole% of which is ZrF_Fourso
Replaced with 2 mole% ThF_FourMolten salt nuclear reactor fuel containing
Material (neutral including 5% of resonance integral cross section at 650 ° C)
Cross-sectional area of child capture = 0.220 x 10^-24cm^Two, In the core
Macroscopic capture cross-section in = 0.000736cm^-1)
In a reactor composed of²³²Th captures neutrons
Catch and generate²³³UF_FourEquilibrium concentration of about 0.1mo
le% (3.54 × 10¹⁹atoms / cm^Three)
It At the average operating temperature of 650 ° C²³³U core
Cross-section area is 299 × 10^{− 24}cm^TwoAnd the neutron
Capture cross section is 56.5 × 10^-24cm^TwoIs. Core
In²³³U macroscopic fission cross section is 0.0010
58 cm^-1And the macroscopic neutron capture cross section is 0.0
00200 cm^-1Is.²³³Due to U fission
Since the sexual yield is 2.49, the neutron multiplication factor is 1.
173. Effective fission cross section is 13.1 × 10
^-24cm^Two, The effective capture cross section is 27.9 × 10^-24
cm^TwoAnd a transuranium element with a neutron yield of 3.42
Element is 3.57 × 10²⁰atoms / cm^ThreeBelow the concentration
, The macroscopic fission cross section of the core is 0.000
468 cm^-1, The core macroscopic capture cross section is 0.000
996 cm^-1And the neutron multiplication factor is 1.142 or more.
Becomes critical.

Further, the effective neutron capture cross section is 6 × 10.
^-24cm^TwoFission products of 6.3 in molten salt reactor fuel
0x10²⁰atoms / cm^ThreeEven if present at a concentration of
The sexual multiplication factor is 1.04 and the criticality can be maintained. This nucleus
The amount of fission products is the core power density 19.5 W / cm.^Three(fuel
Salt fission density 5.93 × 10¹²missions /
cm^Three-S) at 1.25 × 10⁸-S (3.95
It is equivalent to driving and accumulating. Where fission
The product is rare earth element trifluoride 1.89 × 10 ²⁰a
toms / cm^ThreeThe solvent of molten salt reactor fuel is 3 because it contains
At least 5.46 × 10 as fluoride²⁰atoms
/ Cm^Three(9.07 × 10^-4mole / cm^Three) Melting
Must have a degree of solution. Here, as a solvent, LiF-
BeF_Two-ThF_Four(77-5-18 mole%) T
hF_FourPart of ZrF_FourLiF-BeF substituted with_Two−
ThF_Four-ZrF_Four(77-5-2-16 mole%)
The solubility of trifluoride at 550 ° C is LiF-BeF
_Two-ThF_FourSame as (77-5-18 mole%)
6.63 x 10²⁰atoms / cm^Three(11.0 x 1
0^-4mole / cm^Three) And is actually a molten salt furnace
The transuranic element concentration that can be loaded in the fuel is satisfied.

²³³ U macroscopic fission cross section of the core
Macroscopic fission cross section of 001058 and transuranium element.
000468 is equivalent to 69.3% and 31.7%. 1
The fission heat output of GWt-y is generated by about 300 kg of fissile material.
With the consumption of ²³³ U, 92 kg of transuranic element is burned.

According to the fourth embodiment, the following effects can be obtained. (1) When the minimum operating temperature of the reactor is about 600 ° C. and the maximum allowable liquidus temperature of the fuel salt is 550 ° C., ThF ₄ and ZrF ₄ are added to the fuel salt solvent and the temperature exceeds 550 ° C. At the same time as increasing the solubility of uranium trifluoride,
ThF ₄ , which is a necessary and sufficient breeding material, can be present to burn transuranium elements having a significantly low fission cross section with high efficiency by utilizing the neutrons generated in the generated ²³³ U fission. (2) It is possible to provide a molten salt nuclear reactor fuel having a high solubility of transuranium trifluoride and rare earth trifluoride that can effectively burn transuranium elements having a small nuclear fission cross section.

The phase diagrams of FIGS. 1 and 2 are RE T
homa, H. Insley, HA Friedman and GM Hebert,
“Equilibrium Phase Diagram of the Lithium Fluorid
e-Beryllium Fluoride-Zirconium Fluoride Systems ”,
Journal of Nuclear Materials Vol. 27 (1968), p.16
The phase diagram of Fig. 3 is from W.
R. Grimes, “Molten-Salt Reactor Chemistry”, Nucl
ear Applications and Technology Vol.8, February 19
70, pp.137-155.

[0079]

According to the invention of claim 1, the temperature is kept at 500 ° C.
And high concentration of transuranium element 3 in molten salt reactor fuel
To a very low fission cross section (about 120
× 10 ^-24cm^Two) Composed of transuranic elements
Fission products containing rare earth elements in molten salt reactor fuel
Nuclear reactors that do not allow the existence of
Well, with a margin of 50 ° C, a minimum operating temperature of 550 ° C is acceptable.
The molten salt reactor fuel suitable for operation with excellent thermal efficiency.
Can be provided.

According to the invention of claim 2, the transuranic element trifluoride is dissolved in the molten salt reactor fuel at a high concentration at 550 ° C., and the nuclear fission cross section is further reduced (about 76 × 1).
0 ^{- 24 cm 2)} provide sufficient nuclear reactivity against the reactor to allow for the presence of fission products containing a rare earth element in the molten salt reactor fuel consists of transuranium elements with claim 1 Although the thermal efficiency is inferior to that of the invention, it is possible to provide a molten salt reactor fuel which has a margin of 50 ° C. and is suitable for operation in which a minimum operating temperature of 600 ° C. is allowed.

According to the invention of claim 3, the odor at 550 ° C.
Trans-uranium trifluoride at high concentration in molten salt reactor fuel
The effect obtained by the invention of claim 2 by dissolving the substance
And the added ThF_FourBy the neutron capture of
Generate²³³UF_FourOf high fission cross section
Remarkably low fission cross section (about 13 × 10^-24cm
^TwoMolten salt reactor composed of transuranium element
Does not allow the existence of fission products containing rare earth elements in the material
Give enough nuclear reactivity to the reactor,²³³U is burning
The amount burned is the combustion of transuranium elements per thermal output of the reactor.
Although the amount is small, it has a margin of 50 ° C and a maximum of 600 ° C.
Molten salt reactor fuel suitable for operation where low operating temperature is allowed
Can be provided.

According to the invention of claim 4, the transuranic element trifluoride is dissolved in the molten salt reactor fuel at a high concentration at 550 ° C., and the effect obtained by the invention of claim 3 is produced and added. that part of the ZrF ₄ of ThF ₄
By substituting with, the amount of ThF ₄ used as a breeding material is limited to reduce the amount of ²³³ U burned while keeping the solubility of trans-uranium trifluoride high, and even if the existence of fission products is allowed, The burning rate of uranium element can be increased to the theoretical value of 31.7%.

[Brief description of drawings]

FIG. 1 is a LiF—BeF ₂ —ZrF ₄ ternary phase diagram showing a solvent salt composition that is a preferred embodiment of the present invention.

FIG. 2 is a graph showing a composition of a solvent salt according to another embodiment of the present invention.
_iF-BeF 2 -ZrF a ₄ ternary phase state diagram.

FIG. 3 is an L showing another example of the solvent salt composition of the present invention.
_iF-BeF 2 -ThF a ₄ ternary phase state diagram.

Claims (4)

[Claims] 1. Lithium enriched with an isotope having a mass number of 7
Fluoride (⁷LiF) and beryllium difluoride (Be)
F_Two) And zirconium tetrafluoride (ZrF_Four) Mixed with
In the solvent of the molten salt composed of a substance, trifluoride of transuranium element
It contains trifluoride of the fission product element as a solute and is a solvent.
The composition of⁷LiF-BeF_Two-ZrF _FourPhase of ternary system
75-5-20 mole%, 61-11
-20 mole%, 67-33-0 mole%, and 6
Range of quadrangle connecting 9-31-0 mole% points sequentially
Liquid phase temperature, which is the temperature at which the solid phase precipitates for the first time
Lower than 500 ° C and operating temperature higher than 500 ° C
Characterized molten salt reactor fuel. 2. Lithium enriched with an isotope having a mass number of 7.
Fluoride (⁷LiF) and beryllium difluoride (Be)
F_Two) And zirconium tetrafluoride (ZrF_Four) Mixed with
In the solvent of the molten salt composed of a substance, trifluoride of transuranium element
It contains trifluoride of the fission product element as a solute and is a solvent.
The composition of⁷LiF-BeF_Two-ZrF _FourPhase of ternary system
77-3-20 mole%, 61-11
-20 mole%, 67-33-0 mole%, and 7
Range of a quadrangle connecting 1-29-0 mole% points sequentially
Liquid phase temperature, which is the temperature at which the solid phase precipitates for the first time
Lower than 550 ° C and operating temperature higher than 550 ° C
Characterized molten salt reactor fuel. 3. A lithium fluoride ( ⁷ LiF) enriched with an isotope having a mass number of 7 and a beryllium difluoride (Be).
F ₂ ) and a thorium tetrafluoride (ThF ₄ ) mixture of a molten salt solvent containing a transuranium element trifluoride and a fission product element trifluoride as solutes. _{^{7 LiF-BeF 2 -ThF 4 76-6-18mole}} % in the phase diagram the state of a three-component system, 69-31-0
mole% and 71-29-0 mole% points are sequentially connected within a triangular range, and the liquidus temperature at which the solid phase precipitates for the first time is within the range of 500 ° C and 550 ° C. Is higher than 550 ° C. Molten salt nuclear reactor fuel. 4. ThF in a solvent obtained by substituting a part of thorium tetrafluoride (ThF ₄ ) in the molten salt reactor fuel according to claim 3 with zirconium tetrafluoride (ZrF ₄ ).
Molten salt reactor fuel characterized by having a reduced content of ₄ .