KR101440273B1 - Plutonium(Pu) amount determination System and Method in Uranium(U)/Transuranium(TRU) Ingot of Pyroprocessing product - Google Patents

Abstract

The present invention relates to a method for measuring plutonium (Pu) in U/TRU ingot as a product of a pyroprocess and a system using the same. Specifically, the method comprises a step of function calculating step (S100) which executes ORIGEN computation code with an input variable of spent fuel in U/TRAU ingot as a product of a pyroprocess in order to calculate a fitting function capable of measuring the mass of Pu by applying all degrees of burn up of the spent fuel; a total neutron counting step (S200) which counts total neutrons emitted from the spent fuel using a passive neutron counter; and a Pu measuring step (S300) which measures the mass of Pu in the spent fuel using the total neutron counting value of the spent fuel counted in (S200) as well as a preset curium (Cm) rate (Pu/Cm) of the spent fuel as variables of the fitting function calculated S100.

Description

Description of the Related Art Plutonium (Pu) amount determination system and method in Uranium (U) / Transuranium (TRU) Ingot of pyroprocessing product

The present invention relates to a method for measuring plutonium in a U / TRU ingot, which is a pyrolytic product, and a system using the same, wherein a total neutron emitted from a U / TRU ingot, which is a product of a pyrogenic process, is combined with a passive neutron counter The present invention relates to a method for measuring plutonium in a U / TRU ingot, which is a pyrolysis product that accurately counts the mass of Pu (plutonium, plutonium), which is one of the constituent elements of U / TRU.

In the chemical process of separating uranium (U) and other substances (TRU, RE), that is, pyrolysis, in the pyrolysis process, the fission products are separated first through the electrolytic reduction process of nuclear materials of spent nuclear fuel. Monitoring the amount of nuclear material per unit process is very important to ensure nuclear transparency and is a recommendation of the International Atomic Energy Agency (IAEA).

Generally, various nuclear fuel materials contain fissile materials that can cause fission. In order to make nuclear fuel that can be burned in nuclear reactors by using these nuclear fuel materials, it is necessary to accurately measure the content of fissile material contained in nuclear fuel material. The measurement methods used for this purpose are divided into a destructive method and a non - destructive method.

First, the destruction method is a chemical analysis method. However, it has a merit that the analysis accuracy is good, but it is economically disadvantageous because the analysis time is long and the use frequency is limited.

On the other hand, the non-destructive method is a method of measuring and analyzing gamma rays or neutrons. However, gamma-ray spectral analysis method has been used only for new nuclear fuel, and the fissionable material contained in spent nuclear fuel We use the method of quantifying Cm by measuring the neutrons emitted from the main neutron source, Cm, and quantifying U-235 or Pu based on this. At this time, a highly efficient neutron detector is required.

Gamma ray measurement When the non-destructive method is applied to spent nuclear fuel, it is difficult to detect plutonium isotopes because a large amount of gamma rays are emitted from rare earth elements as well as gamma rays emitted from uranium and plutonium isotopes.

A metal material composed of a mixture of U + TRU + RE separated by an electrolytic reduction process in this pyrolysis process is called a U / TRU ingot. The U / TRU material weighs 12kg in the fuel assembly after pressurized light water reactor (PWR) of about 500kgHM of Reference Engineering-scale Pyroprocessing Facility (REPF). In order to secure safety against critical danger, Two U / TRU ingots weighing 6 kg are produced.

The U / TRU ingot with a weight of 6 kg, manufactured using an input material with an initial concentration of 4.5%, a burning degree of 55,000 MWd / MtU, and a cooling period of 10 years as the reference of the REPF, the uranium (U) % Uranium, plutonium (Pu), and cuium (Cm), and about 15 wt% of rare earth elements. The Cm-244 contained in 6 kg of the U / TRU ingot thus constituted is about 24.6 g, and 2.95 x 108 neutrons are emitted from this Cm-244.

If the neutron multiplication of the induced fission neutrons emitted from the fissile material contained in the U / TRU ingot is taken into consideration, the number is doubled.

As described above, the non-destructive method is very useful. One recent example from the Los Alamos National Laboratory in the United States, for example,

In this method, generally, a Cd (cadmium) cylinder is added to a passive neutron counter to obtain a ratio between the case where the sample is surrounded by the Cd cylinder around the sample to be measured and the case where the sample is measured without the Cd cylinder . A proportional relationship between this ratio (Cd ratio) and the mass or Pu mass of the fissile material (U-235, Pu-239, etc.) present in the nuclear material is obtained to determine a fissile material or plutonium mass .

However, when this method is surrounded by a Cd cylinder, since the signal is weakened, the uncertainty of the Cd ratio becomes large and the measurement error becomes several% or more, and the measurement accuracy is low.

Table 1 below is a comparison table between the mass of actual plutonium and the mass of plutonium measured by this method.

Initial Enrichment (wt.%) Maximum error of plutonium mass
(Maximum Error in Pu mass) 2.5 5.89% 3.5 3.46% 4.5 2.14% Average 5.95%

At this time, the maximum error value of the plutonium mass can be obtained by the following equation.

Maximum error value of plutonium mass = (Pu '- Pu) / Pu

, Pu 'is the mass of Pu obtained, and Pu is the mass of the actual Pu.

As shown in the above table, it is possible to reduce the error up to 2.14%, but this is a specific case and when considering the condition of all spent fuel, it shows a very large error of 5.95% There is a disadvantage that utilization is difficult.

In domestic patent application No. 2005-0090199 (published on Sep. 13, 2005, entitled " Automatic Separation of Radionuclides and Automatic Separation Method of Plutonium Using the Same "), in the chemical pure separation of radionuclides present in environmental samples, The automated radionuclide separation apparatus and plutonium (Pu) automatic separation method developed to increase the efficiency of the analysis work by saving the time and manpower required for the analysis by automating the chemical separation have been disclosed.

However, since such a device and method can control the remote operation using a PC, it is only a device capable of separating the radionuclide while minimizing the radiation dose of the analyzer in the case of analyzing the sample with high radioactivity concentration. It has difficult problems to follow.

Korean Patent Laid-Open No. 10-2005-0090199 (published on Sep. 13, 2005, entitled: Automatic Separation of Radionuclides and Automatic Separation of Plutonium Using the Same)

SUMMARY OF THE INVENTION It is an object of the present invention to solve the above-mentioned problems, and it is an object of the present invention to provide a neutron sensor which accurately counts the total neutrons emitted from a U / TRU ingot, A method for measuring plutonium in a U / TRU ingot which is a pyrolysis product that can accurately measure the mass of Pu (plutonium), one of the constituent elements, and a system using the same.

The method of measuring plutonium in a U / TRU ingot, which is a pyrogen process product according to an embodiment of the present invention, is a method of measuring plutonium in a U / TRU ingot by performing an ORIGEN computation code using the spent fuel of a U / TRU ingot, A function calculating step (S100) of calculating a fitting function capable of measuring the mass of plutonium applied to all the degrees of burning of the spent nuclear fuel, a total neutron emitted from the spent nuclear fuel using the passive neutron counter (Plutonium) / Cm (cerium) ratio) of the spent fuel, and a total neutron count (S200) of the total spent fuel counted in the total neutron counting step (S200) The plutonium measurement step S300 of measuring the mass of plutonium in the spent nuclear fuel using the neutron coefficient value as a parameter of the fitting function calculated in the function calculating step S100 Characterized in that eojineun.

The plutonium measurement system in the U / TRU ingot, which is a pyrogen process product according to an embodiment of the present invention, can measure the mass of the plutonium contained in the spent fuel of the U / TRU ingot separated through the electrolytic reduction process of the pyro- The U / TRU ingot being a pyrolytic product, the system comprising: a function calculating unit for calculating a fitting function capable of measuring the mass of plutonium by applying to all the degrees of burning of the spent nuclear fuel, (Plutonium) / Cm (cerium) ratio of a predetermined neutron counting unit 20 for counting the total neutrons emitted from the spent nuclear fuel by using a passive neutron counter, ) And the total neutron count value counted by the neutron counting section (20) as a parameter of the fitting function calculated by the function calculating section (10) To the mass measurement, it characterized in that comprises a control unit 100.

The method of measuring plutonium in a U / TRU ingot, which is a pyrogenic product of the present invention, and a system using the same, accurately combine the total neutrons emitted from the U / TRU ingot, which is a product of the Pyro process, with a passive neutron counter, And can be easily used later.

Accordingly, a fitting function for measuring plutonium can be derived from a database in the range of 30 GWd / tU to 60 GWd / tU, and since the error of the fitting function at this time is within 0.2%, the TRU There is an advantage that the mass of Pu (plutonium, Plutonium) which is one of the constituent elements can be accurately measured within a total error of 2%.

In addition, based on this, it is expected that the future neutron multiplication changes depending on the morphology of U / TRU ingot, the direct measurement of plutonium in U / TRU ingot and the sensitivity of neutron counter depending on the inhomogeneity of fissile material included There is an advantage that can be utilized.

1 is a flowchart illustrating a method for measuring plutonium in a U / TRU ingot, which is a pyrolytic product according to an embodiment of the present invention.
FIG. 2 is a flowchart illustrating a method for measuring plutonium in a U / TRU ingot, which is a pyrolytic product according to an embodiment of the present invention.
3 is a block diagram illustrating a plutonium measurement system in a U / TRU ingot that is a pyrochrome product according to an embodiment of the present invention.
4 is a view illustrating an example of a passive neutron counter according to an embodiment of the present invention.

Hereinafter, a method for measuring plutonium in a U / TRU ingot, which is a Pyro process product according to an embodiment of the present invention, and a system using the same will be described in detail with reference to the accompanying drawings. The following drawings are provided by way of example so that those skilled in the art can fully understand the spirit of the present invention. Therefore, the present invention is not limited to the following drawings, but may be embodied in other forms. In addition, like reference numerals designate like elements throughout the specification.

In this case, unless otherwise defined, technical terms and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this invention belongs. In the following description and the accompanying drawings, A description of known functions and configurations that may unnecessarily obscure the description of the present invention will be omitted.

The Cd ratio using neutron counts measured in the absence of the Cd cylinder and the neutron counts measured in the surroundings of the Cd (cadmium) cylinder using the passive neutron counter and the fissile material (U-235, Pu -239, etc.), or the mass of Pu, and obtain the mass of the fissile material or Pu present in the nuclear material. As described above, conventionally, in the case of the neutron coefficient measured in the state surrounded by the Cd cylinder, since the signal is weakened by the Cd cylinder, the uncertainty of the Cd ratio itself becomes large, and the accuracy of the measurement becomes low.

On the other hand, the method of measuring plutonium in the U / TRU ingot, which is a pyrolytic product according to an embodiment of the present invention, and the system using the same, includes a passive neutron counter composed of a He- A passive neutron counter consisting of a fission chamber detector was combined to accurately count the total neutrons emitted from the TRU ingot and to measure the plutonium in the U / TRU ingot, which is a pyrogenic process product according to one embodiment of the present invention (Pu) contained in the spent nuclear fuel by using a total neutron coefficient value of the spent fuel and a predetermined Pu / Cm ratio as a parameter to a separate fitting function calculated in the system using the same Mass can be measured.

A method of measuring plutonium in a U / TRU ingot, which is a pyro-processed product according to an embodiment of the present invention, includes a function calculating step S100, a total neutron counting step S200, and a plutonium measuring step S300).

The function calculating step S100 is a step of calculating the function using the spent fuel of the U / TRU ingot, which is a product pyro process product, as an input variable and executing an ORIGEN computation code, that is, an ORIGEN computation program, To calculate a fitting function capable of measuring the mass of the plutonium.

More specifically, the function calculating step S100 may include a component measuring step S110, an initial function calculating step S120, a DB step S130, and a fitting function calculating step S140.

At this time, the plutonium measuring system in the U / TRU ingot, which is a pyrochrome product according to an embodiment of the present invention, may include a function calculating unit 10, a neutron counting unit 20 and a control unit 100 And the function calculating step S100 may be performed in the function calculating unit 10 of the plutonium measuring system in the U / TRU ingot, which is the pyro-processing product.

In the constituent component measuring step (S110), the ORIGEN computer code may be executed to measure a nuclear type constituent component in the spent nuclear fuel,

The initial function calculation step (S120) may calculate a separate initial function for measuring the mass of the plutonium contained in the spent fuel for each combustion degree of the spent fuel.

The DB step S130 is a step of calculating the nuclear component of the spent nuclear fuel measured in the constituent component measuring step S110 and the nuclear initial constituent of the spent fuel calculated in the initial function calculating step S200, Functions can be stored in a database.

The DB step S130 may store the data in a database part (not shown) of the plutonium measurement system in the U / TRU ingot which is the pyro-processing product.

Accordingly, the fitting function for measuring the mass of plutonium constituting the spent fuel at a later time can be easily calculated by referring to the data stored in the database unit.

The fitting function calculating step (S140) uses the data stored in the database unit in the DB step (S130) to calculate the mass of plutonium in the spent nuclear fuel by applying the data to all combustion degrees of the spent fuel The fitting function can be calculated.

At this time, the fitting function can be expressed by the following equation (1).

[Equation 1]

Pu = f (Cm ratio, total neutron count value of spent fuel)

That is, in the function calculation step S100, first, the initial function may be obtained for each combustion degree, and the fitting function capable of integrating them may be determined, and the general fitting function Can be calculated.

The above equation (1) can be converted into the following equation (2).

&Quot; (2) "

(Pu is the mass of plutonium,

CR is the total neutron count value of spent fuel,

R is the Cm ratio,

c1 is a first constant, the range of c1 is ~ 9.0E + 6,

c2 is a second constant, the range of c2 is -9.0E-1,

c3 is a third constant, the range of c3 is ~ 1.0E - 4,

c4 is a fourth constant, and the range of c4 is ~ 9.0E-10)

As shown in Equation (2), in order to measure the mass of the plutonium contained in the spent nuclear fuel using the fitting function, the total neutron count value and the Cm ratio of the spent nuclear fuel are used as variables.

In this case, c1 is set to 9.825576E + 06, c2 is set to -9.860544E + 01, c3 is set to 1.22218E-04, and c4 is set to 9.466833E-10 in the fitting function, The mass of the plutonium contained in the spent nuclear fuel can be more accurately measured when the total neutron count value of the spent fuel and the Cm ratio are set as the variable values.

The total neutron count step (S200) can count the total neutrons emitted from the spent nuclear fuel using a passive neutron counter, wherein the total neutron count step (S200) is performed using the U / TRU ingot In the neutron counting unit 20 of the plutonium measuring system in FIG.

More specifically, the passive neutron counter may be at least one of a passive neutron counter composed of a conventional He-3 detector and a passive neutron counter composed of a fission chamber detector. The spent nuclear neutron counter may include MCNPX (Monte Carlo N-Patricle Extended) code simulation can be applied to count the total neutrons emitted from the spent fuel.

However, when the passive neutron counter composed of the He-3 detector is used, the efficiency is high, and when the same measurement time is used, the coefficient value is about 100 times higher than that of the passive neutron counter composed of the fission detector, Will be reduced to 1/10. However, there is a problem that the gamma-ray-up signal, that is, the gamma ray signal due to the gamma-ray multiplexing collision, may be mixed with the neutron signal. Therefore, in the measurement of the nuclear material strongly emitting gamma rays,

When the passive neutron counter composed of the fission detector is used, the efficiency is about 1/100 as compared with the passive neutron counter composed of the He-3 detector, and the uncertainty is disadvantageous because the uncertainty is 10 times larger. However, There is an advantage of counting the exact total neutrons. However, in order to achieve the same efficiency as the passive neutron counter composed of the He-3 detector, the measurement time is required to be 100 times longer.

The total neutrons emitted from the spent nuclear fuel can be counted independently using a passive neutron counter composed of the He-3 detector and a passive neutron counter composed of the fission detector, but the total neutrons emitted from each passive neutron counter The present invention attempts to count the total neutrons most accurately by combining the advantages and disadvantages of a passive neutron counter composed of the He-3 detector and a passive neutron counter composed of the fission detector.

As shown in FIG. 4, a passive neutron counter composed of a He-3 detector and a passive neutron counter composed of a fission chamber detector may be combined. By using the passive neutron counter of FIG. 4, the MCNPX (Monte Carlo N-Patric Extended) code simulation can be applied to the spent fuel to count the total neutrons emitted from the spent fuel.

When the total number of neutrons emitted from the spent nuclear fuel is counted by using the passive neutron counter in which the passive neutron counter composed of the He-3 detector and the passive neutron counter composed of the fission detector are combined, As shown in the figure, the total neutron count value counted using the passive neutron counter composed of the He-3 detector is compared with the total neutron count value counted using the passive neutron counter composed of the fission detector, The total neutron count value of the spent nuclear fuel can be corrected and used.

In detail, the total neutron count value obtained by using the passive neutron counter composed of the He-3 detector is compared with the total neutron count value counted by using the passive neutron counter composed of the fission detector, and the total neutron count In this case, the total neutron count value counted using the passive neutron counter composed of the He-3 detector having a low uncertainty may be used as the final count value As a result,

And comparing the total neutron count value counted by each of the passive neutron counters with a total neutron count value counted by using a passive neutron counter composed of the He-3 detector as a passive neutron counter composed of the fission detector, In this case, the total neutron count value obtained by using the passive neutron counter composed of the fission detector is used as the value of the total neutron count, As a final count value, the uncertainty can be reduced by increasing the detection time.

The plutonium measurement step S300 may be performed using a predetermined ratio of Cm, that is, Pu (plutonium) / Cm (cerium) ratio constituting the spent nuclear fuel, and the total neutron count step S200 using the passive neutron counter The mass of the plutonium in the spent nuclear fuel can be measured by using the total neutron count value of the spent fuel as the parameter of the fitting function calculated in the function calculating step S100. The plutonium measurement step S300 may be performed in the control unit 100 of the plutonium measurement system in the U / TRU ingot, which is the pyro-processing product.

At this time, the Cm ratio can be calculated by executing the ORIGEN computer code to calculate the Cm ratio constituting the spent nuclear fuel, which is a well-known known technology.

In other words, the Cm ratio previously set in Equation (2) above is multiplied by the passive neutron counter according to the present invention in which the passive neutron counter composed of the He-3 detector and the passive neutron counter composed of the fission detector are counted The mass of the plutonium constituting the spent nuclear fuel can be easily measured by inputting the total neutron count value of the spent nuclear fuel as a variable.

Particularly, when the fitting function is derived using the data in the range of 30 GWd / tU to 60 GWd / tU, the fitting error of the fitting function is within 0.2%. Therefore, the neutron of the spent fuel is counted , The mass of plutonium in the spent fuel can be measured within a total error of 2% or less, assuming an error of about 1 to 1.5%.

That is, in other words, the method of measuring plutonium in the U / TRU ingot, which is a pyrolytic product according to an embodiment of the present invention, and the system using the same, has a predetermined Cm ratio and four constants c1 to c4 , The neutron count of the spent fuel is counted by using the passive neutron counter, and the coefficient of plutonium in the spent fuel is easily measured .

While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it is to be understood that the invention is not limited to the disclosed embodiments, but, on the contrary, And various modifications and changes may be made thereto by those skilled in the art to which the present invention pertains.

Therefore, it is to be understood that the subject matter of the present invention is not limited to the described embodiments, and all of the equivalents or equivalents of the claims are included in the scope of the present invention will be.

10: function calculating section
20: Neutron coefficient part
100:
S100 to S300: Each step of the method for measuring plutonium in the U / TRU ingot, which is a pyrolytic product according to the present invention

Claims (8)

A fitting function that can measure the mass of plutonium by applying the ORIGEN code to the spent fuel of the U / TRU ingot, which is the pyrogenic product, Calculating a function (S100);
A total neutron counting step (S200) of counting the total neutrons emitted from the spent nuclear fuel using a passive neutron counter; And
And a total neutron count value of the spent fuel counted in the total neutron counting step (S200) is calculated from the Cm ratio (Pu (plutonium) / Cm (urium ratio) of the predetermined spent fuel) A plutonium measuring step (S300) of measuring the mass of plutonium in the spent nuclear fuel using the calculated fitting function as a parameter of the fitting function;
Lt; / RTI >
The function calculating step (SlOO)
A constituent component measuring step (S110) of executing the ORIGEN code to measure a constituent component of the spent nuclear fuel;
An initial function calculation step (S120) of calculating an initial function for each combustion degree of the spent nuclear fuel;
A DB step (S130) of making the initial function of each combustion degree calculated in the initial component calculation step S120 into a database and storing the constituent components measured in the component measurement step (S110); And
A fitting function calculating step (S140) of calculating the fitting function capable of measuring the mass of plutonium in the spent fuel by applying the stored data in the DB step (S130) to all the degrees of burning of the spent fuel, );
And,
Wherein the fitting function is represented by the following equation: < EMI ID = 1.0 >

(Where Pu is the mass of the plutonium,
CR is the total neutron count value of spent fuel,
R is the Cm ratio,
c1 is a first constant, the range of c1 is ~ 9.0E + 6,
c2 is a second constant, the range of c2 is -9.0E-1,
c3 is a third constant, the range of c3 is ~ 1.0E - 4,
c4 is a fourth constant, and the range of c4 is ~ 9.0E-10.)
delete The method according to claim 1,
The total neutron counting step (S200)
A Monte Carlo N-Patricated Extended (MCNPX) code simulation is performed using any one of a passive neutron counter composed of a passive neutron counter composed of a He-3 detector and a passive neutron counter composed of a fission chamber detector, A method of measuring plutonium in a U / TRU ingot which is a pyrolytic product, characterized in that the total neutron number is counted.
The method of claim 3,
The total neutron count step (S200)
When both the passive neutron counter composed of the He-3 detector and the passive neutron counter composed of the fission detector are used,
The total number of neutrons of the spent nuclear fuel counted using a passive neutron counter composed of the He-3 detector,
Wherein the total number of neutrons of the spent fuel is calibrated by comparing the total number of neutrons of the spent nuclear fuel by using the passive neutron counter composed of the fission detector, Method of measuring plutonium in a TRU ingot.
In a plutonium measuring system in a U / TRU ingot, which is a pyrogenic product that measures the mass of plutonium contained in spent fuel of a U / TRU ingot separated through an electrolytic reduction process of a pyrogenic process,
A function calculating unit (10) for calculating a fitting function capable of measuring the mass of plutonium by applying an ORIGEN computer code using the spent fuel as an input variable to all combustion degrees of the spent fuel;
A neutron counting unit 20 for counting the total neutrons emitted from the spent nuclear fuel using a passive neutron counter; And
(Plutonium) / Cm (cerium) ratio) of the spent fuel, and the total neutron count value counted by the neutron counting unit 20 are calculated by the function calculating unit 10, A control unit (100) for measuring the mass of plutonium in the spent nuclear fuel by using as a parameter of a fitting function;
And,
Wherein the fitting function is represented by the following equation: < EMI ID = 1.0 >

(Where Pu is the mass of the plutonium,
CR is the total neutron count value of spent fuel,
R is the Cm ratio,
c1 is a first constant, the range of c1 is ~ 9.0E + 6,
c2 is a second constant, the range of c2 is -9.0E-1,
c3 is a third constant, the range of c3 is ~ 1.0E - 4,
c4 is a fourth constant, and the range of c4 is ~ 9.0E-10.)
delete 6. The method of claim 5,
The neutron counting section 20
A Monte Carlo N-Patricated Extended (MCNPX) code simulation is performed using at least one selected from a passive neutron counter composed of a passive neutron counter composed of a He-3 detector and a passive neutron counter composed of a fission chamber detector, Wherein the total number of neutrons in the U / TRU ingot is measured.
8. The method of claim 7,
The neutron counting section 20
When both the passive neutron counter composed of the He-3 detector and the passive neutron counter composed of the fission detector are used,
The total number of neutrons of the spent nuclear fuel counted using a passive neutron counter composed of the He-3 detector,
Wherein the total number of neutrons of the spent fuel is calibrated by comparing the total number of neutrons of the spent nuclear fuel by using a passive neutron counter composed of the fission detector, TRU In situ plutonium measurement system.

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