Spent fuel transport container shielding performance evaluation method

Abstract

The invention relates to a spent fuel transport container shielding performance evaluation method, which comprises the following steps: determining the total distribution condition of radionuclides to be loaded, and obtaining the activity proportions of the same radionuclide at different positions; determining a characteristic position for carrying out the external radiation level calculation of the transportation container; according to the feature position, single-nuclide shielding performance simulation calculation is carried out; according to the spent fuel transport container shielding performance evaluation method, the relationship between the activity of loaded radionuclides and the dose rate value of the corresponding position outside the container is obtained, and the dose rate contribution of the single radionuclide unit activity to the focus can be determined according to the radionuclide type and the spatial distribution condition thereof, so that the shielding performance of the spent fuel transport container can be evaluated. And the shielding performance of the spent fuel transportation container or the transportation container capable of loading multiple nuclides is determined, so that the shielding design of the container can meet the requirements of laws and regulations under the condition that the multiple nuclides are loaded.

Description

Technical Field

[0001] The invention belongs to the technical field of radioactive material transportation safety, and in particular relates to a method for evaluating shielding performance of a spent fuel transportation container. Background Art

[0002] Radioactive material transport containers are important equipment to ensure the safety of transportation activities. They have the functions of containing radioactive materials, providing sufficient shielding capabilities, and preventing critical and thermal damage. In the past, when designing and evaluating the shielding performance of containers, especially when the contents to be loaded are multiple nuclides and the spatial distribution is uneven, the shielding performance evaluation needs to be calculated for each working condition separately, resulting in a large workload, and it is also impossible to form an envelope evaluation of the loading volume and type.

[0003] In the transportation of medical radioactive sources, transportation safety is of vital importance, so safe and reliable transport containers are required for transportation. According to regulations, radioactive materials should be transported in dedicated radioactive material transport packaging containers, and the design and manufacture of radioactive material transport containers should comply with national radioactive material transportation safety standards.

[0004] In the past, when designing and evaluating the shielding performance of containers, on the one hand, due to the fact that the contents to be loaded are multi-nuclides and the spatial distribution is uneven, the shielding performance evaluation needs to be calculated separately for each working condition, which requires a lot of calculation work and cannot form an envelope evaluation of the loading volume and type. On the other hand, due to the calculation method, the calculation results may deviate greatly, the calculation efficiency is very low, and the credibility of the results is not high. Summary of the invention

[0005] In view of the defects existing in the prior art, the object of the present invention is to provide a method for evaluating the shielding performance of a spent fuel transport container, which can determine the shielding performance of a spent fuel transport container or a transport container capable of loading multiple nuclides.

[0006] To achieve the above purpose, the technical solution adopted by the present invention is: a method for evaluating the shielding performance of a spent fuel transport container, comprising the steps of: determining the total distribution of radioactive nuclides to be loaded, and obtaining the activity proportions of the same nuclide at different positions; determining a characteristic position for calculating the external radiation level of the transport container; performing a single nuclide shielding performance simulation calculation based on the characteristic position; and obtaining the relationship between the activity of the loaded radioactive nuclides and the dose rate value at the corresponding position outside the container.

[0007] Furthermore, the relationship between the activity of the loaded radionuclides and the dose rate value at the corresponding position outside the container is specifically:

[0008]

[0009] Among them, D iis the radiation level of the transport container at point i; Q i,j,s is the dose rate value at transport container i when the nuclide j of unit activity is loaded at position s in the transport container; P j,s is the activity proportion of radionuclide j at position S; A j It is the total activity of the radionuclides to be loaded in the transport container.

[0010] Furthermore, the single-nuclide shielding performance simulation calculation is specifically performed as follows: solving the dose rate value at the transport container i when the nuclide j with unit activity is loaded at the position s in the transport container.

[0011] Furthermore, the calculation process of the dose rate value at the transport container i is specifically as follows: based on the shielding structure and material properties of the transport container, a shielding calculation model of the transport container is established, and the Monte Carlo method is used to calculate the dose rate value at the transport container i when the nuclide j of unit activity is loaded at position s in the transport container.

[0012] Furthermore, the characteristic positions include the side center, top center, bottom center and shielding weak points of the transport container.

[0013] Furthermore, the shielding weak points include the surface of the transport container, 1 meter from the surface and 2 meters from the surface.

[0014] Furthermore, the obtaining of the activity proportions of the same nuclide at different positions is specifically: according to various fuel consumptions, operating histories and cooling times, the activity proportions of the same nuclide at different positions in the component are given.

[0015] Furthermore, the total distribution of the radionuclides to be loaded specifically refers to: the spatial distribution of various radionuclides in the transport container.

[0016] The effect of the present invention is that by determining the contribution of the unit activity of a single nuclide to the dose rate of the point of interest, and then determining the shielding performance of the spent fuel transport container or the transport container that can be loaded with multiple nuclides according to the type of nuclides and their spatial distribution, it is ensured that the container shielding design can meet the regulatory requirements when loading multiple nuclides. It avoids the situation that when the contents to be loaded are multiple nuclides and the spatial distribution is uneven, the shielding performance evaluation needs to be calculated separately for each working condition, which requires a lot of calculation work and cannot form an envelope evaluation of the loading amount and type. BRIEF DESCRIPTION OF THE DRAWINGS

[0017] Figure 1 The present invention is a flowchart of the steps of a method for evaluating the shielding performance of a spent fuel transport container. DETAILED DESCRIPTION

[0018] The present invention is further described below in conjunction with the accompanying drawings and specific embodiments.

[0019] like Figure 1 As shown, the present invention proposes a method for evaluating shielding performance of a spent fuel transport container, comprising the steps of:

[0020] S1, determine the total distribution of the radionuclides to be loaded, and obtain the activity proportion of the same nuclide at different locations;

[0021] Specifically, determine the spatial distribution of various radionuclides j in the transport container, P j,s Since the fission products in the fuel, the activation products of the fuel cladding, and the various nuclides corresponding to the assembly shelf positions and their distribution are consistent in the spent fuel assembly, only the activities are different. Therefore, according to the various burnup, operation history and cooling time, the activity proportion of the same nuclide in the assembly at different positions s is given.

[0022] S2, determine the characteristic position for calculating the external radiation level of the transport container;

[0023] Specifically, since the dose rate limits on the surface, 1 meter from the surface, and 2 meters from the surface of the spent fuel transport container should comply with regulatory standards and design requirements, it is necessary to calculate the dose rate at this feature point when conducting shielding design and evaluation to determine whether the maximum radiation level at the corresponding position meets regulatory and standard requirements. Based on engineering experience and container structure, a suitable feature position group i is selected, generally the side center, top center, bottom center, and other possible shielding weak points (surface, 1 meter from the surface, and 2 meters from the surface at various positions of the container), so that the corresponding detection points can be selected in the next step of conducting single-nuclide shielding performance simulation calculations.

[0024] S3, simulate and calculate the shielding performance of a single nuclide based on the characteristic position;

[0025] Specifically, the dose rate value Q at transport container i is calculated when the unit activity nuclide j is loaded at position s in the transport container. i,j,s Based on the shielding structure and material properties of the transport container, a shielding calculation model for the transport container is established. The Monte Carlo method is used to calculate the dose rate value Q at the transport container i when the nuclide j of unit activity is loaded at the position s in the transport container. i,j,s .

[0026] S4, obtaining the relationship between the activity of the loaded radioactive nuclide and the dose rate value at the corresponding position outside the container;

[0027] Specifically, the dose rate calculation formula at point i of the transport container is given. Based on the solutions of the above three steps, the functional relationship expression between the activity of various radionuclides loaded and the dose rate value at point I outside the container can be established:

[0028]

[0029] Among them, D i is the radiation level of the transport container at point i, μSv·h -1 ;Q i,j,s is the dose rate value at transport container i when the nuclide j of unit activity is loaded at position s in the transport container, μSv·h -1 ·Bq -1 ;P j,s is the activity proportion of radionuclide j at position S, dimensionless; A j is the total activity of the radionuclides to be loaded in the transport container, Bq.

[0030] According to the established functional relationship expression, the shielding performance of the spent fuel transport container or the transport container capable of carrying multiple nuclides is determined.

[0031] It can be seen from the above embodiments that the present invention determines the dose rate contribution of the unit activity of a single nuclide to the point of interest, and then determines the shielding performance of the spent fuel transport container or the transport container that can carry multiple nuclides according to the type of nuclide and its spatial distribution, so as to ensure that the container shielding design can meet regulatory requirements when carrying multiple nuclides.

[0032] The present invention is not limited to the embodiments described in the specific implementation manner. Those skilled in the art may derive other implementation manners based on the technical solution of the present invention, which also fall within the technical innovation scope of the present invention.

Claims

1. A method for evaluating shielding performance of a spent fuel transport container, characterized in that: Includes steps: Determine the overall distribution of the radionuclides to be loaded and obtain the activity percentage of the same nuclide at different locations; Identify characteristic locations for calculation of radiation levels outside the transport container; According to the characteristic position, the shielding performance simulation calculation of single nuclide is carried out; Obtain the relationship between the activity of the loaded radionuclide and the dose rate value at the corresponding position outside the container.

2. A spent fuel transport container shielding performance evaluation method according to claim 1, characterized in that: The relationship between the activity of the loaded radionuclides and the dose rate value at the corresponding position outside the container is specifically: Among them, D i is the radiation level of the transport container at point i; Q i,j,s is the dose rate value at transport container i when the nuclide j of unit activity is loaded at position s in the transport container; P j,s is the activity proportion of radionuclide j at position S; A j It is the total activity of the radionuclides to be loaded in the transport container.

3. A spent fuel transport container shielding performance evaluation method as claimed in claim 2, characterized in that: The simulation calculation of the shielding performance of a single nuclide is specifically as follows: solving the dose rate value at the transport container i when the nuclide j with unit activity is loaded at the position s in the transport container.

4. A spent fuel transport container shielding performance evaluation method as claimed in claim 3, characterized in that: The calculation process of the dose rate value at the transport container i is specifically as follows: based on the shielding structure and material properties of the transport container, a shielding calculation model of the transport container is established, and the Monte Carlo method is used to calculate the dose rate value at the transport container i when the nuclide j of unit activity is loaded at the position s in the transport container.

5. A spent fuel transport container shielding performance evaluation method as claimed in claim 1, characterized in that: The characteristic locations include the side center, top center, bottom center and shielding weak points of the transport container.

6. A spent fuel transport container shielding performance evaluation method as claimed in claim 5, characterized in that: The shielding weak points include the surface of the transport container, 1 meter from the surface and 2 meters from the surface.

7. A spent fuel transport container shielding performance evaluation method according to claim 1, characterized in that: The obtaining of the activity proportion of the same nuclide at different positions is specifically: according to various fuel consumption, operation history and cooling time, the activity proportion of the same nuclide at different positions in the component is given.

8. A spent fuel transport container shielding performance evaluation method according to claim 1, characterized in that: The total distribution of the radionuclides to be loaded specifically refers to: the spatial distribution of various radionuclides in the transport container.

Images