JPH0453397B2 - - Google Patents

Description

Detailed Description of the Invention [Technical Field of the Invention] The present invention relates to a method for non-destructive measurement of fuel assemblies for evaluating the nuclear properties of fuel assemblies by non-destructive measurement, and in particular to a method for evaluating the nuclear properties of a fuel assembly using active neutron measurement. This invention relates to a method for non-destructive measurement of spent fuel assemblies to determine various combustion parameters of the assemblies.

[Technical background of the invention and its problems] In general, spent fuel taken out from a nuclear reactor is stored in a fuel storage pool for a certain period of time to attenuate its radioactivity, which has a short half-life, and then is stored in a transport container and recycled. Transported to processing plants or long-term storage facilities.

Understanding the combustion characteristics of spent fuel is extremely important in ensuring criticality safety during transportation, storage, and reprocessing, and in terms of nuclear material management.

Parameters that indicate the combustion characteristics of spent fuel include burnup, residual fission nuclide concentration, multiplication factor, total plutonium concentration, etc. A method for quantifying these parameters is gamma ray measurement, which measures the gamma ray spectrum emitted from spent fuel. Spectral measurement method, passive neutron measurement method that measures neutrons emitted from spent fuel, and active neutron measurement method that measures neutron flux formed by injecting neutrons emitted from an external neutron source into the spent fuel. There is a law. The active neutron measurement method can also be called a multiplied neutron measurement method or a multiplication factor measurement method.
The most important parameters in terms of ensuring criticality safety or safeguards during spent fuel transportation, storage, and reprocessing are the multiplication factor and total fission nuclide concentration (hereinafter referred to as Fiss), but conventional measurement methods There were disadvantages such as the measuring device was large-scale and it was not easy to determine the absolute values of the combustion parameters.

[Object of the invention] The present invention has been made to address the above problems, and
A neutron source is placed on one side of a fuel assembly placed underwater, and a neutron detector is placed on the opposite side,
The effective neutron multiplication factor (hereinafter referred to as Keff) is determined by the active neutron measurement method, which measures the neutron flux (hereinafter referred to as φ) formed by injecting neutrons emitted from a neutron source into a fuel assembly. The purpose is to provide a method for determining combustion parameters such as Fiss, burnup, and plutonium concentration from the value of Keff.

[Summary of the Invention] That is, the present invention relates to (a) an active neutron measurement method in which a neutron beam is opposed to one side of a fuel assembly disposed underwater, and a neutron detector is arranged on the other side; , the neutron flux measured by the neutron detector is between (1-
(b) determining the axial position of the fuel assembly which is inversely proportional to Keff) by experiment or analysis, placing the neutron detector at the position and measuring the neutron flux φ by an active neutron measurement method;
a step of calculating an effective neutron multiplication factor Keff from the neutron flux φ obtained in step (a), using the correlation between the neutron flux φ and the effective neutron multiplication factor Keff, which was obtained by conducting a simulation experiment or analysis in advance; (c) From the effective neutron multiplication factor Keff obtained in step (b), use the correlation between the effective neutron multiplication factor Keff obtained in advance, the infinite neutron multiplication factor _k (d) step (c) to obtain the multiplication factor k _∞ and the total fissile nuclide concentration Fiss;
From the values of the infinite neutron multiplication factor k _∞ and the total fissile nuclide concentration Fiss obtained in
The method is characterized by comprising the step of determining the burnup indicating the combustion characteristics of the spent fuel assembly and other parameters using correlation with Pu/U, ²³⁵ U concentration, ²³⁹ Pu concentration or ²⁴¹ Pu concentration. This is a non-destructive measurement method for spent fuel assemblies.

[Embodiment of the Invention] The present invention will be described in detail below with reference to an embodiment shown in the drawings.

The inventors of the present invention have been conducting research on active neutron measurement methods and devices for spent fuel assemblies for a long time, and have made many proposals. Through these processes, it was discovered that φ can be expressed by the following formula.

φ=(A/(1−Keff))·F(Keff) Here, A is a constant, and F(Keff) is a correction factor for locally introducing a neutron source from the outside.
When a neutron source is uniformly distributed within the aggregate without placing an external neutron source, F (Keff) = 1, but if a neutron source is locally introduced from the outside,
Generally, F(Keff) does not equal 1. F (Keff) is
Keff is expressed as a linear approximation by the following equation.

F(Keff)=1+B(1-Keff) However, B is a constant.

Therefore, in the correlation between φ and Keff, there are two unknowns, A and B, and in order to create a calibration curve of φ versus Keff, two fuel assemblies with known Keff are required. However, if B=0, the only unknown quantity is A, and φ and
Keff's correlation is more simplified.

In one embodiment of the present invention shown in the flowchart of FIG. 1, first, a measurement system in which the value of F (Keff) is approximately 1.0 regardless of the value of Keff is determined by a simulation experiment or analysis.

In other words, a neutron source is placed on one side of a fuel assembly disposed underwater, and the neutron flux and (1 Find the position where the reciprocal of (-Keff) is proportional (or nearly proportional).
Next, as shown in Figure 2, Keff is known as 1.
Using two standard fuel assemblies to create a calibration curve of φ vs. Keff in a measurement system of F (Keff) = 1, and measuring φ of the fuel assembly to be measured under the same conditions as the measurement system, Keff is determined from the calibration curve of φ vs. Keff.

Furthermore, we calculated the correlation between Fiss and Keff as shown in Figure 3 and the correlation between infinite neutron multiplication factor (hereinafter referred to as K∞) and Keff as shown in Figure 4, and added these to the calibration curve. Use as Fiss
and determine K∞.

Furthermore, as shown in Figures 3 to 6, K∞ or Fiss and burnup, total plutonium to total uranium concentration ratio (hereinafter referred to as Pu/U) and K∞ or burnup, and K∞ and ²³⁵ The correlation between the U concentration, ²³⁹ Pu concentration, and ²⁴¹ Pu concentration is calculated separately, and these are used as a calibration curve to determine the burnup, Pu/U, ²³⁵ U concentration, ²³⁹ Pu concentration, and ²⁴¹ Pu concentration.

[Effects of the Invention] As is clear from the above explanation, according to the present invention, the criticality safety or security device during transportation, storage, and reprocessing of spent fuel assemblies is the most important parameter. The multiplication factor and Fiss can be directly determined, and the burnup, Pu/U,
Uranium concentration, plutonium concentration, etc. can also be determined indirectly.

Note that the present invention is not limited to spent fuel assemblies, but can of course be applied to new fuel assemblies or irradiated fuel assemblies once taken out of the reactor during use.

[Brief explanation of drawings]

FIG. 1 is a flowchart showing an embodiment of the present invention, FIG. 2 is a graph showing a calibration curve of Keff for φ, and FIG. 3 is a graph showing a calibration curve of Keff for Keff or K∞.
Graph showing the calibration curves of ²³⁵ U concentration, ²³⁹ Pu concentration, ²⁴¹ Pu241 concentration, Figure 4 is a graph showing the calibration curve of K∞ against Keff, Figure 5 is a graph showing the calibration curve of burnup against K∞, Figure 4 is a graph showing the calibration curve of K∞ against Keff, Figure 6 is a graph showing a calibration curve of Pu/U with respect to burnup.

Claims (1)

[Scope of Claims] 1 (a) In an active neutron measurement method in which a neutron beam is placed on one side of a fuel assembly disposed in water and a neutron detector is placed on the other side, the neutron detector The axial position of the fuel assembly is determined in advance by experiment or analysis such that the neutron flux measured by the effective neutron multiplication factor Keff is inversely proportional to (1-Keff), and the neutron detector is placed at that position. (b) measuring the neutron flux φ using the active neutron measurement method;
(c) Calculating the effective neutron multiplication factor Keff obtained in advance from the effective neutron multiplication factor Keff obtained in step (b). Magnification Keff
and (d) calculating the infinite neutron multiplication factor k _∞ and the total fissile nuclide concentration Fiss using the correlation between the infinite neutron multiplication factor k _∞ and the total fissile nuclide concentration Fiss; These values, burnup, total plutonium to total uranium concentration ratio Pu/U, ²³⁵ U concentration _, ²³⁹ Pu concentration or ²⁴¹ A method for non-destructive measurement of spent fuel assemblies, comprising the steps of: determining the burnup indicating the combustion characteristics of the spent fuel assemblies and other parameters using the correlation with the Pu concentration;