KR-102963429-B1 - An apparatus and method for measuring spent fuel burn-up using a He-3 neutron detector

KR102963429B1KR 102963429 B1KR102963429 B1KR 102963429B1KR-102963429-B1

Abstract

A burnup measurement method may include the step of a burnup measurement device measuring the number of neutrons emitted from the target spent nuclear fuel; the step of the burnup measurement device calculating the total neutron reaction rate for the spent nuclear fuel of a neutron detector; the step of the burnup measurement device correcting the calculated total neutron reaction rate; and the step of the burnup measurement device inputting the corrected total neutron reaction rate into a relationship equation between the total neutron reaction rate of the neutron detector, the measurement distance, the initial enrichment of the spent nuclear fuel, and the cooling period to measure the burnup of the target spent nuclear fuel.

Inventors

박광헌
이운장
송양수
차소희
김문오
박경태
임예전
성진현

Assignees

경희대학교 산학협력단
(주) 코네스코퍼레이션
주식회사 오리온이엔씨

Dates

Publication Date: 20260513
Application Date: 20231220

Claims (20)

A step in which a burnup measuring device measures the number of neutrons emitted from the target spent nuclear fuel; The step of the burnup measuring device calculating the total neutron reaction rate for the spent nuclear fuel of the neutron detector; A step in which the above burnup measuring device corrects the calculated total neutron reaction rate; The burnup measuring device comprises the step of measuring the burnup of the target spent nuclear fuel by inputting the corrected total neutron reaction rate into the relationship between the total neutron reaction rate and measurement distance of the neutron detector, the initial enrichment and cooling period of the spent nuclear fuel; After the step of correcting the calculated total neutron reaction rate, A burnup measurement method comprising further including the step of determining whether the measured number of neutrons and the corrected total neutron reaction rate match the burnup measuring device.
In paragraph 1, A burnup measurement method in which the step of correcting the calculated total neutron reaction rate is to multiply the calculated total neutron reaction rate by a correction factor of less than 1.
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In paragraph 1, If the measured number of neutrons matches the corrected total neutron reaction rate, the corrected total neutron reaction rate is substituted into the above relationship, and A burnup measurement method that recalculates the total neutron reaction rate for spent nuclear fuel in a neutron detector if the measured number of neutrons does not match the corrected total neutron reaction rate.
In paragraph 1, A burnup measurement method in which the step of determining whether the measured number of neutrons matches the corrected total neutron reaction rate is determined to be consistent if the following equation is satisfied. Measured number of neutrons = Calculated total neutron reaction rate * Correction factor (0 < Correction factor < 1).
In paragraph 5, A burnup measurement method in which the step of determining whether the measured number of neutrons matches the corrected total neutron reaction rate is determined to be consistent if any one of the following equations is satisfied. Total counts of measured neutrons = Calculated total neutron reaction rate * 0.41 Peak counts of measured neutrons = Calculated total neutron reaction rate * 0.15
A step in which a burnup measuring device measures the number of neutrons emitted from the target spent nuclear fuel; The step of the burnup measuring device calculating the total neutron reaction rate for the spent nuclear fuel of the neutron detector; A step in which the above burnup measuring device corrects the calculated total neutron reaction rate; The burnup measuring device comprises the step of measuring the burnup of the target spent nuclear fuel by inputting the corrected total neutron reaction rate into the relationship between the total neutron reaction rate and measurement distance of the neutron detector, the initial enrichment and cooling period of the spent nuclear fuel; A burnup measurement method comprising the step of determining whether the measured number of neutrons matches the corrected total neutron reaction rate, measuring the number of neutrons emitted from the entire target spent nuclear fuel, converting the measured value into a value per ton of uranium of the target spent nuclear fuel, and determining whether the converted value matches the corrected total neutron reaction rate.
A step in which a burnup measuring device measures the number of neutrons emitted from the target spent nuclear fuel; The step of the burnup measuring device calculating the total neutron reaction rate for the spent nuclear fuel of the neutron detector; A step in which the above burnup measuring device corrects the calculated total neutron reaction rate; The burnup measuring device comprises the step of measuring the burnup of the target spent nuclear fuel by inputting the corrected total neutron reaction rate into the relationship between the total neutron reaction rate and measurement distance of the neutron detector, the initial enrichment and cooling period of the spent nuclear fuel; A burnup measurement method for a target spent nuclear fuel, wherein the burnup of the spent nuclear fuel is measured using the following equation (1). BU = a1(RR) b1 (IE) c1 e d1t e f1x --- Equation (1) Here, BU is the burnup (GWD/MTU) of the target spent nuclear fuel, RR is the total neutron reaction rate (reaction rate, Reaction Rates/sec·MTU) of the target spent nuclear fuel for the neutron detector, IE is the initial enrichment (%) of the target spent nuclear fuel, t is the cooling period (years) of the target spent nuclear fuel, x is the measurement distance (cm) of the target spent nuclear fuel for the neutron detector, and a1 is 0.462±0.03, b1 is 0.28±0.005, c1 is 0.325±0.02, d1 is 0.0103±0.000125, and f1 is 0.0527±0.001.
In paragraph 8, Before the step of measuring the number of neutrons, The step of the burnup measuring device acquiring information regarding the sample spent nuclear fuel; and The method further includes the step of deriving the above formula (1) from the information regarding the above sample spent nuclear fuel, and The step of obtaining information regarding the above-mentioned sample spent nuclear fuel is A burnup measurement method comprising the step of the burnup measurement device using ORIGEN to calculate the neutron generation rate (S N ) of actinide, spontaneous fission nuclides, and sample spent nuclear fuel (Fs), using actinide and spontaneous fission as neutron sources for MCNP to calculate the reaction rate per neutron (R N ) of a neutron detector, and calculating the total neutron reaction rate (RR) of a neutron detector as the product of the calculated neutron generation rate (S N ) and the reaction rate per neutron (R N ) of the neutron detector.
In paragraph 1, The step of measuring the burnup of the above spent nuclear fuel is a burnup measurement method in which Cm-244 is the main neutron-generating nuclide during the cooling period of the entire cooling period.
A detector that measures the number of neutrons emitted from the target spent nuclear fuel; A memory unit for storing information about a sample of spent nuclear fuel; and A calculation unit that calculates the total neutron reaction rate for spent nuclear fuel of a neutron detector, corrects the calculated total neutron reaction rate, and measures the burnup of the target spent nuclear fuel using a relationship between the corrected total neutron reaction rate, the measurement distance, the initial enrichment of the spent nuclear fuel, and the cooling period; The above-mentioned operation unit determines whether the measured number of neutrons matches the corrected total neutron reaction rate, a burnup measuring device.
In Paragraph 11, A burnup measuring device in which the above-mentioned operation unit corrects the calculated total neutron reaction rate by multiplying the calculated total neutron reaction rate by a correction factor of less than 1.
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In Paragraph 11, If the measured number of neutrons matches the corrected total neutron reaction rate, the above operation unit substitutes the corrected total neutron reaction rate into the above relationship, and A burnup measuring device that recalculates the total neutron reaction rate for spent nuclear fuel from a neutron detector if the measured number of neutrons does not match the corrected total neutron reaction rate.
In Paragraph 11, A burnup measuring device in which the above-mentioned operation unit determines that the measured number of neutrons and the corrected total neutron reaction rate match if the following equation is satisfied. Measured number of neutrons = Calculated total neutron reaction rate * Correction factor (0 < Correction factor < 1).
In paragraph 15, A burnup measuring device in which the above-mentioned operation unit determines that the measured number of neutrons and the corrected total neutron reaction rate match if any one of the following equations is satisfied. Total counts of measured neutrons = Calculated total neutron reaction rate * 0.41 Peak counts of measured neutrons = Calculated total neutron reaction rate * 0.15
A detector that measures the number of neutrons emitted from the target spent nuclear fuel; A memory unit for storing information about a sample of spent nuclear fuel; and A calculation unit that calculates the total neutron reaction rate for spent nuclear fuel of a neutron detector, corrects the calculated total neutron reaction rate, and measures the burnup of the target spent nuclear fuel using a relationship between the corrected total neutron reaction rate, the measurement distance, the initial enrichment of the spent nuclear fuel, and the cooling period; A burnup measuring device that measures the number of neutrons emitted from the entire target spent nuclear fuel, converts the measured value into a value per ton of uranium of the target spent nuclear fuel, and determines whether the converted value matches the corrected total neutron reaction rate.
A detector that measures the number of neutrons emitted from the target spent nuclear fuel; A memory unit for storing information about a sample of spent nuclear fuel; and A calculation unit that calculates the total neutron reaction rate for spent nuclear fuel of a neutron detector, corrects the calculated total neutron reaction rate, and measures the burnup of the target spent nuclear fuel using a relationship between the corrected total neutron reaction rate, the measurement distance, the initial enrichment of the spent nuclear fuel, and the cooling period; The above calculation unit is a burnup measuring device that measures the burnup of spent nuclear fuel using the following equation (1). BU = a1(RR) b1 (IE) c1 e d1t e f1x --- Equation (1) Here, BU is the burnup (GWD/MTU) of the target spent nuclear fuel, RR is the total neutron reaction rate (reaction rate, Reaction Rates/sec·MTU) of the target spent nuclear fuel for the neutron detector, IE is the initial enrichment (%) of the target spent nuclear fuel, t is the cooling period (years) of the target spent nuclear fuel, x is the measurement distance (cm) of the target spent nuclear fuel for the neutron detector, and a1 is 0.462±0.03, b1 is 0.28±0.005, c1 is 0.325±0.02, d1 is 0.0103±0.000125, and f1 is 0.0527±0.001.
In Paragraph 18, The above computational unit obtains information regarding the sample spent nuclear fuel before measuring the number of neutrons, derives the above equation (1) from the information regarding the sample spent nuclear fuel, and A burnup measuring device comprising the steps of: the above-mentioned computational unit calculating the neutron generation rate (S N ) of actinide, spontaneous fission nuclides, and sample spent nuclear fuel (Fs) using ORIGEN; calculating the reaction rate per neutron (R N ) of a neutron detector using actinide and spontaneous fission as neutron sources of MCNP; and calculating the total neutron reaction rates ( RR ) of a neutron detector as the product of the calculated neutron generation rate (S N ) and the reaction rate per neutron (R N) of the neutron detector.
In Paragraph 11, The above-mentioned operation unit is a burnup measuring device that measures the cooling period during which Cm-244 is the main neutron-generating nuclide during the entire cooling period.

Description

An apparatus and method for measuring spent fuel burn-up using a He-3 neutron detector The present invention relates to an apparatus and method for measuring the burnup of spent nuclear fuel using a He-3 neutron detector. The burn-up of spent nuclear fuel must be accurately known to ensure safety when operating storage and transport vessels, and a method for verifying this is also required. Currently, the burn-up of spent nuclear fuel is estimated by the fuel manufacturers and power companies utilizing the fuel based on the fuel's burn-up history. However, to ensure safety, it is necessary to verify the estimated burn-up when moving spent nuclear fuel within or outside the power plant, or when transferring it from the storage pool to re-store it. Currently, the burnup of spent nuclear fuel is measured by detecting neutrons and gamma rays emitted from the actinides and fission products remaining in the fuel. Gamma ray measurement estimates burnup by measuring gamma rays from major sources such as Cs-137, Cs-134, and Eu-154 to estimate the amount of fission products produced. On the other hand, burnup measurement using neutrons is more complex than the estimation method based on gamma rays because it is linked to the amount of actinides produced; furthermore, the precise estimation method is not publicly disclosed as the manufacturers of the burnup measurement equipment keep it confidential. Therefore, it is necessary to enhance the safety and economic efficiency of the transportation, storage, and management of spent nuclear fuel within nuclear power plants by measuring its burnup to accurately determine its combustion history. Furthermore, conventional spent nuclear fuel management is performed by assuming the fuel is new, disregarding its burnup; however, management costs can be reduced by accurately measuring the burnup. Figure 1 shows a combustibility measuring device. Figure 2 shows the total neutron generation rate by radionuclide according to the cooling period. Figure 3 shows the energy spectrum of 252 Cf measured with a He-3 neutron detector. Figure 4 shows the 252 Cf neutron measurement values of a He-3 neutron detector according to the thickness of the moderator. Figure 5 shows the neutron calculation values using MCNP and the 252 Cf neutron measurement values of the He-3 neutron detector according to the moderator thickness. Figure 6 shows the neutron calculation values using MCNP according to the moderator thickness and the Total Counts or Peak Counts measured values of the He-3 neutron detector. Figure 7 shows the neutron calculation values using MCNP according to the moderator thickness and the correction coefficients for Total Counts or Peak Counts of the He-3 neutron detector. Figure 8 shows the current status of the initial enrichment and burnup of spent nuclear fuel. The present disclosure is described below with reference to embodiments illustrated in the attached drawings. The described embodiments are not limited to those described herein and may take different forms. Accordingly, the embodiments are described below with reference to the drawings only to explain aspects and features of the present invention. The present disclosure includes various embodiments and modifications, and specific embodiments are illustrated in the drawings and described in the description below. However, the present disclosure is not limited to the embodiments described above and includes all modifications, equivalents, and substitutions that fall within the spirit and scope of the present invention. When an element or layer is referred to as being "on" another element or layer, or being "connected" or "combined" to another element or layer, said element or layer may be directly connected or combined to the other element or layer. Or one or more additional elements or layers may exist. When an element or layer is referred to as being "immediately on" another element or layer, or being "directly connected" or "directly combined," there may not be any other intermediate elements or intermediate layers between them. For example, where it is stated that a first element is "combined" or "connected" to a second element, the first element may be directly combined or connected to the second element, or the first element may be indirectly combined or connected to the second element through one or more intermediate elements. In the drawings, dimensions of various elements, layers, etc., may be exaggerated for the sake of clarity of example. Identical reference numerals may denote identical elements. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. Additionally, when describing embodiments of the present disclosure, the use of “may” relates to “one or more embodiments of the present disclosure.” Expressions such as “at least one” and “any one” may modify the entire list of elements when preceding a list of elements, but may not modify the individual eleme