JPH04249797A - Burnup measurement method of irradiated fuel assembly - Google Patents

Abstract

PURPOSE:To precisely find proportion constant when the burnup of an irradiated fuel assembly is obtained by the use of a gamma radiation spectrum analysis method. CONSTITUTION:A neutron radiation rate radiated from a specified position of a standard irradiated fuel assembly 1 in which initial concentration has been known and irradiation history and a cooling period have been known or the cooling period of at least 1.5 years, normally 2 years or more has passed is measured 2. And the neutron radiation rate is found from long half life nuclide except for the neutron radiation rate based on Ca-242 of the standard irradiation fuel assembly and the burnup 3 is determined from the neutron radiation rate. On the other hand, a gamma radiation spectrum 4 radiated from a specified position of the standard irradiation fuel assembly 1 is measured to determine the proportion factor 6 with the burnup of gamma radiation strength on the basis of Cs-137.

Description

[Detailed description of the invention]

[Object of the invention]

0002

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for measuring the burnup of an irradiated fuel assembly for accurately and efficiently measuring the burnup of fuel irradiated in a nuclear reactor.

0003

2. Description of the Related Art Methods for measuring the burn-up of nuclear fuel irradiated in a nuclear reactor can be broadly classified into destructive measurement and non-destructive measurement. In these cases, whatever method is used, it is necessary to determine the relationship between the measured quantity and burnup.

In destructive measurements, Nd-14, which is a fission product of stable nuclides (hereinafter referred to as FP) accumulated in the fuel,
A method of measuring .

On the other hand, known non-destructive measurements include a method of measuring gamma rays or neutrons emitted from nuclear fuel using a radiation detector, and a method of measuring the calorific value of the fuel using a thermometer.

[0006] In the method of measuring burnup by measuring gamma rays emitted by fuel (gamma ray spectrum analysis method), when the gamma ray spectrum is determined using a germanium (Ge) semiconductor detector etc. with high energy resolution, the measured By analyzing the obtained spectral data, it is possible to determine the intensity of single-energy gamma rays uniquely emitted for each FP nuclide. If the cooling period from the end of irradiation in the reactor to the measurement of the target measurement fuel is more than one year, most of the nuclides with short half-lives among the FP accumulated in the fuel have decayed. Therefore, only Cs-137 (Ba-13), which has a somewhat long half-life, can be measured.
7), Cs-134, Eu-154, Ce
-144 (Pr -144 ) and Ru -106
(Rh-106) and other limited nuclides.

Among these, Cs-137 is mainly used for U-235 and Pu-235, which cause nuclear fission in light water reactors.
Since the fission yield of 239 is about the same and the half-life is long, about 30 years, there is a very good proportional relationship with the burnup of the fuel. Figure 4 shows the calculation of compositional changes due to combustion in nuclear fuel.
Cs -137 calculated by "ORIGEN" code
, Cs -134 and Eu -154 according to their burn-up. In the figure, BWR is a boiling water reactor, Ei is initial enrichment (wt%), and PD is power density/
The output per weight and Tc indicate the cooling time (in days).

The “ORIGEN” code is a combustion simulation code developed in the United States.
) is used to calculate the change in the production rate.

However, when trying to determine the burnup by measuring the gamma ray intensity of Cs -137 alone,
It is necessary to accurately determine the proportionality constant between the count rate and burnup of gamma rays emitted by Cs −137, which corresponds to the gamma ray detection efficiency determined by the measurement system or the shape of the fuel. In conventional technology, this proportionality constant is determined by calculations based on the fuel shape and measurement system, such as the method disclosed in Japanese Patent Application Laid-Open No. 61-262693, or by calculating the burnup value of the measured fuel and Cs - Cs-137 can be calculated from the measured gamma ray count value of Cs-137, or measured by burn-up destruction analysis of the measured fuel.
A method of calculating from counting can be used. The method of determining the proportionality constant by calculation requires accuracy in the measurement system and calculation conditions, and it is also necessary to confirm the accuracy. In the method of using calculated burnup values, it is also necessary to confirm the accuracy of the calculated values. Although the burn-up measurement method using destructive analysis has high reliability, it requires a great deal of time and effort.

On the other hand, Cs -134 /Cs -137
A known method is to determine the burnup by measuring the intensity ratio of Eu -154 /Cs -137. In this method, by measuring the ratio of gamma ray intensities, factors such as measurement system or fuel shape in gamma ray detection efficiency are canceled out, and only the energy dependence of detection efficiency according to gamma ray energy needs to be taken into account.

Cs-134 and Eu-154
The process of production in a nuclear reactor is not due to the decay of nuclides produced directly by nuclear fission, as in Cs -137, but by the absorption of neutrons by nuclides primarily produced by nuclear fission. Specifically, Cs −134 is generated by neutron absorption of Cs −133 , which is a primary FP, and Eu −154 is generated by neutron absorption of Eu −153 . Therefore, since these nuclides are produced through two neutron absorptions, one during nuclear fission and the other after that, the amount produced is not linear with respect to the neutron irradiation amount, but as shown in Figure 4. It increases almost quadratically with respect to the neutron irradiation dose. On the other hand, Cs -137 increases linearly with the neutron irradiation amount, so Cs -134 or Eu -1
The ratio of 54 and Cs -137 has a relatively linear relationship to the neutron irradiation dose. Further, it can be considered that the amount of neutron irradiation and the burnup are approximately proportional. Therefore, C changes like a quadratic curve with respect to burnup.
The value obtained by dividing the s -134 intensity or the Eu -154 intensity by the Cs -137 intensity, which is proportional to the burnup, changes approximately linearly with the burnup. Cs-134
/Cs −137 or Eu −154 /Cs
The method of measuring burnup by measuring the -137 ratio takes advantage of this property.

However, when trying to calculate the burnup from these intensity ratios, it is greatly influenced by the specifications of the target fuel or the irradiation history in the reactor. information is required, which is a major drawback.

[0013] When the power density of the reactor differs, the time required to reach the same burnup differs, but Cs -13
4 has a not-so-long half-life of about two years, so this difference in time affects the balance between production and decay. Eu
-154 has a half-life of about 8.5 years, so the influence of power density is relatively small.

[0014] Furthermore, Cs-134 and Eu-154
Since the fission yields of U-234 and Pu-239 differ considerably, both of them are affected by the initial enrichment of the fuel, the neutron energy spectrum, etc.

For these reasons, Cs-134
/Cs-137 intensity ratio and Eu-154 /C
When attempting to determine the burnup by measuring the intensity ratio of s-137, first obtain detailed information such as the initial enrichment of the target fuel or the history of the power density during irradiation, and then calculate the It is necessary to relate it to burnup. This is useful, for example, when measuring the burnup of fuel during transport from a nuclear reactor to another storage or reprocessing facility, or when measuring burnup when receiving fuel at a storage or reprocessing facility. When a large amount of fuel must be measured, such as when performing a impede implementation.

In addition to the above-mentioned gamma ray spectrum analysis method, a method for measuring the burnup of nuclear fuel is a method of measuring neutrons emitted from the fuel, for example, Japanese Patent Laid-Open No. 1986-61-
The neutron emission rate method disclosed in Japanese Patent No. 262689 is also known. In this method, although it is known that the relationship between the amount of neutrons emitted by the fuel and the burnup is almost linear within a certain range of logarithmic values of both, there is an influence of the initial enrichment. , and the influence of the measurement system or fuel shape must be understood in some way.

[0017]

Problem to be Solved by the Invention When measuring the burnup of an irradiated fuel assembly using a gamma ray spectrum, Cs
In order to determine the burnup from the gamma ray intensity of -137 Cs -137
It is necessary to find the proportionality constant between the gamma ray count and the burnup. In conventional techniques, when determining the constant of proportionality by calculation, a means is required to confirm its accuracy, and when determining the constant of proportionality by performing destructive analysis, a very labor-intensive task is required. On the other hand, Cs −
134 /Cs-137 intensity ratio or Eu-15
Determining the burnup from the intensity ratio of 4/Cs -137 requires a detailed irradiation history of the measured fuel, which makes it difficult to perform efficient measurements when trying to measure a large number of fuels. Become.

The present invention has been made to solve the above-mentioned problems, and consists of (1) calculating the above-mentioned proportionality coefficient;
By compensating for the shortcomings of the three methods: 2) Determining by destructive analysis and (3) Determining by using irradiation history and radioactivity intensity ratio, burnup can be measured accurately and efficiently from gamma ray spectrum measurement. An object of the present invention is to provide a method for measuring burnup of an irradiated fuel assembly. [Structure of the invention]

[0019]

[Means for Solving the Problems] The present invention provides a method for measuring the burnup of an irradiated fuel assembly in which burnup is measured by measuring the intensity of gamma rays emitted from the irradiated fuel assembly irradiated in a nuclear reactor. measuring neutrons emitted from a standard irradiated fuel assembly of known enrichment and known irradiation history and cooling period or a cooling period of at least 1.5 years;
Curium 242 (Cm-) of the standard irradiated fuel assembly
The neutron emission rate excluding the neutron emission rate based on 242) is determined, and the burnup is determined from this neutron emission rate.On the other hand, the gamma ray spectrum emitted from the same height (level) in the axial direction of the standard irradiated fuel assembly is calculated. Measure cesium 1
37 (Cs -137 ), and determine the proportional coefficient between this gamma ray intensity and burnup, and for other irradiated fuel assemblies other than the standard irradiated fuel assembly,
Cs-137 without knowing the initial concentration and irradiation history.
It is characterized by measuring burnup from gamma ray measurement values based on .

[0020]

[Operation] In the method for measuring the burnup of an irradiated fuel assembly according to the present invention, the initial enrichment is known, and the irradiation history and cooling period are known or the cooling period is at least 1.5 years, usually 2 years.
The neutron emission rate emitted from a predetermined position of a standard fuel assembly (hereinafter referred to as a standard irradiated fuel assembly) that has been in use for more than a year is measured. And Cm of the standard irradiated fuel assembly
The neutron emission rate from long half-life nuclides excluding the neutron emission rate based on -242 is determined, and the burnup is determined from this neutron emission rate. On the other hand, measuring the gamma ray spectrum emitted from the predetermined position of the standard irradiated fuel assembly,
The gamma ray intensity based on s −137 is determined, and the proportionality coefficient between this gamma ray intensity and burnup is determined.

[0021] However, in other irradiated fuel assemblies having almost the same structure as the standard irradiated fuel assembly, the burnup can be determined from the measured value of gamma ray intensity based on Cs -137 without knowing the initial enrichment and irradiation history. can be determined.

[0022]

[Embodiment] An embodiment of the method for measuring burnup of an irradiated fuel assembly according to the present invention will be described with reference to FIGS. 1 to 3.

FIG. 1 shows the basic configuration of the present invention, FIG. 2 shows the gamma ray spectrum analysis method in FIG. 1, and FIG. 3 shows the neutron emission rate measurement method in FIG. 1 in block diagrams.

That is, the standard irradiated fuel assembly 1 irradiated in FIG. 1 is subjected to neutron emission rate measurement 2, and the burnup 3 is determined by taking into account initial enrichment, irradiation history or cooling period, shape effect, etc. be done. On the other hand, as shown in Figure 2, in the gamma ray spectrum analysis method, the spectrum measurement
The Cs -137 gamma ray intensity 5 is obtained, and the burnup degree 7 is obtained via the proportionality coefficient 6. A proportional coefficient for implementing the skim shown in FIG. 2 is determined. In this way, the neutron emission rate is measured by neutron emission rate measurement 2, and burnup 3 is determined by incorporating the initial concentration, irradiation history or cooling time, and shape effect, and from this and the Cs-137 gamma ray intensity, the proportional coefficient 6 is required. The basic flow shown in FIG. 2 is disclosed in Japanese Patent Laid-Open No. 61-262693, and the basic flow shown in FIG. 3 is disclosed in Japanese Patent Laid-Open No. 61-262689.

The basic flow of the neutron emission rate measurement method will be briefly explained with reference to FIG. That is, one standard irradiated fuel assembly 1 is placed in water, and the neutron flux is measured 8 at its outer periphery.
I do. In the case of gamma ray measurement, gamma rays emitted from inside the irradiated fuel assembly are blocked by the fuel assembly located on the outer periphery, and the response to the gamma ray detector is small; however, in the case of neutron flux measurement 8, such shielding The phenomenon is significantly mitigated, resulting in a neutron flux that on average characterizes the irradiated fuel assembly. Once the neutron flux is obtained, the next step is to derive the total neutron emission rate 10, but since the neutron flux is affected by the multiplication effect, in order to obtain the neutron emission rate (also called the primary neutron emission rate), In this case, multiplication effect correction 9 must be performed. Since such a correction method utilizes a well-established neutron transport/diffusion code, measurement conditions, that is, shape effects, are required. In this way, the neutron emission rate emitted from all nuclides (total neutron emission rate 10
) is required. Total neutron emission rate 10
The nuclides contributing to are Cm -244 and Cm -242
, Pu -238 , Pu -240 , Am -24
1 is representative. Until about 3 months after the end of irradiation, the high-energy gamma rays (2.53 MeV) released following the β-decay of La-140 produced by the β-decay of the fission product (FP) Ba-140 are It reacts with hydrogen nuclei and releases neutrons, but after that it can be completely ignored. Of the five nuclides mentioned above, Cm -242 has the shortest half-life at 163 days, while Cm -244 has the next shortest half-life at 18 years. 2-3 after irradiation
As years pass, the neutron emission rate based on Cm -242 becomes almost negligible, and after a cooling period of one year, the neutron emission rate based on Cm -242 and the neutron emission rate based on the other four nuclides become comparable in size. There are many things. Therefore, during the cooling period of 1 to 2 years, the neutron emission rate based on Cm -242 must be removed (numeral 11 in the figure).
). The present inventors have disclosed a method of carrying out measurements by changing the cooling period twice or more by taking advantage of the difference in half-life in Japanese Patent Application Laid-Open No. 53-22993. If the cooling period is 1 to 2 years, even if it is removed by calculation, it is unlikely to cause a large error. The effect of Cm −242 is thus removed. The main component of this neutron emission rate is usually Cm -244 , and in a normally burned fuel assembly, it accounts for about 90% or more of the value after removing the Cm -242 component. Tokukai Showa 6
Publication No. 1-262689 mainly discloses a method of extracting only Cm -244 and calculating the burnup from the correlation between it and the burnup, but the inventors have also discovered that other than Cm -244 may also be included. etc. are disclosed. Since the neutron emission rate obtained in this way is not significantly influenced by the irradiation history, correction for the influence of the irradiation history is often not necessary. The influence of the initial enrichment (εi) is clearly visible. That is, the neutron emission rate (S) excluding the Cm −244 component
The theoretically calculated value of can be fitted using the following equation with burnup (Bu) and εi as variables. That is, (Bu)=exp [1/A(lnS-B)]・
F (A, B are constants) A=A2 εi
2 +A1 εi +A0 (A2 , A1
, A0 is a constant) B=B2 εi
2 +B1 εi 2 +B0 (B2 , B
1, B0 is a constant) F=F2 ε
i 2 +F1 εi 2 +F0 (F2,
(F1 and F0 are constants) When this equation is actually calculated numerically, it can be seen that in an actual fuel assembly, even if there is a 10% error in the neutron emission rate (measured value), the error in the burnup will be It has the excellent property that the propagation of is reduced to about 2 to 3%, and burnup can be determined with high accuracy. Thus, the burnup degree 13 is obtained by adding the correction 12 for the initial enrichment (εi) to the Cm -242 removal/release rate 11. If the value of this burnup 13 is significantly different from the burnup value assumed in the multiplication effect correction 9, the burnup is multiplied by repeated calculations similar to those disclosed in JP-A No. 61-262689. The burnup that correctly corrects the effects is required.

When the burnup 3 shown in FIG. 1 is determined in this way, the gamma ray spectrum is measured at that measurement position, the obtained data is analyzed, and Cs-137
Since the gamma ray intensity (relative intensity may also be used) is determined, the proportionality coefficient 6 is determined from the ratio of the two. Once the proportionality factor 6 has been determined, an irradiated fuel assembly of the same size or design as the standard irradiated fuel assembly 1 is analyzed by gamma ray spectrometry without detailed information on the irradiation history using the procedure shown in Figure 2. , and the burnup can be determined without information on the initial enrichment. The fissile nuclide concentration, Pu concentration, etc. can be derived from the burnup.

[0027]

According to the present invention, by first determining the proportionality constant between Cs -137 intensity and burnup in an irradiated fuel assembly for which detailed irradiation history information is clear, the irradiated fuel assembly to be subsequently measured is determined. It becomes possible to accurately and efficiently measure burnup without knowing irradiation history information.

[Brief explanation of the drawing]

FIG. 1 is a block diagram illustrating an embodiment of a method for measuring burnup of an irradiated fuel assembly according to the present invention.

FIG. 2 is a block diagram for explaining the gamma ray spectrum measurement method in FIG. 1.

FIG. 3 is a block diagram for explaining the neutron emission rate measurement method in FIG. 1.

[Fig. 4] Cs -137, C for explaining the conventional example
FIG. 2 is a characteristic diagram showing burnup changes in gamma ray intensity emitted by s -134 and Eu -154.

[Explanation of symbols]

1...Standard irradiated fuel assembly, 2...Neutron emission rate measurement, 3...
Burnup, 4... Gamma ray spectrum measurement, 5... Cs −1
37 Gamma ray intensity, 6... proportionality coefficient, 7... burnup, 8...
Neutron flux measurement, 9... Multiplication effect correction, 10... Total neutron emission rate, 11... Cm -242 removal emission rate, 12... Initial enrichment effect correction, 13... Burnup.

Claims (1)

[Claims] Claim 1: A method for measuring the burnup of an irradiated fuel assembly, which measures the burnup by measuring the intensity of gamma rays emitted from the irradiated fuel assembly irradiated in a nuclear reactor, wherein the initial enrichment is known, and Neutrons emitted from a standard irradiated fuel assembly with a known irradiation history and cooling period or a cooling period of at least 1.5 years are measured to determine the neutrons based on curium 242 (Cm -242 ) of the standard irradiated fuel assembly. The neutron emission rate excluding the emission rate is determined, and the burnup is determined from this neutron emission rate. On the other hand, the gamma ray spectrum emitted from the same height (level) in the axial direction of the standard irradiated fuel assembly is measured and cesium 137 ( Cs-137
), and determine the proportional coefficient between this gamma ray intensity and the burnup. 1. A method for measuring burnup of an irradiated fuel assembly, characterized in that burnup is measured from a measured value of gamma ray intensity based on -137.