A control rod burnup assessment method for advanced nuclear reactors

Abstract

A control rod burnup assessment method for advanced nuclear reactors first divides the control rods in the core into a fine burnup assessment rod group and a direct burnup assessment rod group, and then performs a burnup calculation analysis over the core life. In each burnup step, the average neutron flux density of each segment of the control rod assembly is obtained based on the results of the steady-state neutron transport calculation. Next, for the direct burnup assessment rod group, only the average neutron flux density of each segment is accumulated to calculate the total neutron flux. For the fine burnup assessment rod group, after obtaining the average neutron flux density, the absorber is immediately solved. 10 B's burnup equation, and the calculated control rod 10 The B nucleon density is fed back to the neutron transport calculation of the next burnup step. When all burnup step calculations are completed, the total neutron flux during the lifetime is used to calculate the direct burnup assessment rod group. 10 The burnup equation of B is used to obtain the material composition of this set of control rods after burnup.

Description

Technical Field

[0001] The present invention relates to the fields of nuclear reactor core design and nuclear reactor physics calculation, and in particular to a control rod burnup assessment method applied to advanced nuclear reactors. Background Art

[0002] The reactivity control design of nuclear reactors is crucial to their safe operation. Soluble boron control and control rod control are the primary means of reactivity control. Currently, most advanced nuclear reactor designs, such as sodium-cooled fast reactors and gas-cooled reactors, no longer rely on water as a coolant. Some water-cooled reactor designs have also proposed boric acid-free designs. Reactivity control in these advanced reactors can no longer rely on soluble boron, so control rods remain the primary control method.

[0003] Since the energy spectrum of advanced reactors is usually hard, the common control rod absorber material is mainly B4C. B4C is similar to boric acid and both rely on 10 B has a larger neutron absorption cross section. 10 After absorbing neutrons, B will transform into a 11 B, therefore, the reactivity control ability of the absorbent material will vary with 10 During the life of a reactor, the content of boric acid 10 While boron consumption can be compensated through boration in the chemical volume system, control rod absorber material consumption cannot be compensated during operation. Therefore, assessing control rod absorber material consumption during operation, determining its value, and, based on this, deciding whether to replace the control rods is crucial for the safe operation of the reactor.

[0004] On the other hand, control rods are often divided into different control rod groups according to the control strategy, and their movement logic is also different. When evaluating the control rod fuel consumption, it is also necessary to distinguish them according to their movement logic.

[0005] Currently, various units and research institutions have noticed the issue of control rod burnup when designing advanced nuclear reactors. However, there is no clear technical approach or relevant calculation method to evaluate control rod burnup. Therefore, it is very necessary to propose and develop a method suitable for evaluating control rod burnup in advanced nuclear reactors. Summary of the Invention

[0006] In order to overcome the problems of the prior art, the present invention aims to provide a control rod burnup evaluation method for advanced nuclear reactors. The control rod burnup evaluation method has the characteristics of processing each absorber region on the control rod. 10The function of B nuclear density changing with fuel consumption is taken into account. At the same time, factors such as the reduction in control rod value caused by control rod fuel consumption and the change in the length of the material area involved in fuel consumption calculation caused by the movement of the control rod position are comprehensively considered. The calculation results are more in line with the actual situation.

[0007] In order to achieve the above objectives, the present invention adopts the following technical solutions to be implemented:

[0008] A control rod burnup assessment method for an advanced nuclear reactor comprises the following steps:

[0009] Step 1: First, based on the core control rod movement logic, the control rods in the core are divided into a fine burnup assessment rod group and a direct burnup assessment rod group;

[0010] Step 2: For the fine burnup assessment rod group, in each burnup step, based on the results of the steady-state neutron transport calculation, the average neutron flux density of each segment of the control rod assembly is obtained by the following formula:

[0011]

[0012] Where:

[0013] ——the average neutron flux density of the control rod energy group ig in the i-th burnup step

[0014] ——Neutron flux density of the nth control rod assembly segment

[0015] ——nth b - Neutron flux density of one control rod assembly segment, n b -1 is the number of the bottom node of the control rod

[0016] ——nth t + Neutron flux density of 1 control rod assembly segment, n t +1 is the number of the top node of the control rod

[0017] n——Number of each segment of the control rod assembly

[0018] h——Height of each segment of the control rod assembly

[0019] h t ——Height of the top segment of the control rod assembly

[0020] h b ——Height of the bottom segment of the control rod assembly

[0021] l is the total length of the control rods

[0022] Step 3: For the fine burnup evaluation rod group, after obtaining the average neutron flux density, solve the following absorber 10 B's burnup equation, and the calculated control rod 10 The B nucleon density is fed back into the neutron transport calculation for the next burnup step;

[0023]

[0024] Where:

[0025] —— 10 B nucleon density / cm -3

[0026] N0—— 10 B initial nucleon density / cm -3

[0027] —— 10 Microscopic absorption cross section of B / cm 2

[0028] ——Neutron flux density / cm -2 ·s -1

[0029] t——time / s

[0030] Step 4: For the direct burnup assessment rod group, according to the results of the steady-state neutron transport calculation, the average neutron flux density of each segment of the control rod assembly is obtained by formula (1);

[0031] Step 5: For the direct burnup assessment rod group, only the average neutron flux density of each segment is accumulated to calculate the total neutron flux.

[0032] Step 6: For the direct burnup assessment rod group, after all burnup steps are calculated, use the total neutron flux during the lifetime to calculate 10 The burnup equation (2) of B is used to obtain the material composition of the control rod group after burnup;

[0033] Step 7: The core burnup calculation is completed.

[0034] Compared with the prior art, the present invention has the following outstanding advantages:

[0035] 1. Ability to calculate each absorber area on the control rod in each burn-up step 10 B nuclear density changes with burnup function, which can output the value of each node in each burnup step. 10 B nuclear density;

[0036] 2. The calculation accuracy is higher by comprehensively considering factors such as the reduction in control rod value caused by control rod burnup and the change in the length of the material area involved in the burnup calculation caused by the movement of the control rod position. BRIEF DESCRIPTION OF THE DRAWINGS

[0037] Figure 1 Flowchart of the control rod burnup assessment method applied to advanced nuclear reactors according to the present invention.

[0038] Figure 2 This is the core layout diagram of the Super Phoenix fast breeder reactor.

[0039] Figure 3 This is a diagram of the grouping of compensation rods in the right third of the Super Phoenix reactor core. DETAILED DESCRIPTION

[0040] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments.

[0041] The specific implementation steps are as follows Figure 1 As shown. First, the program processes the input card to obtain the information of each control rod group. According to the control rod movement logic of the core, the control rods in the core are divided into a fine burnup assessment rod group and a direct burnup assessment rod group. Then, the burnup calculation and analysis of the core life are performed. In each burnup step, the average neutron flux density of each segment of the control rod assembly is obtained based on the results of the steady-state neutron transport calculation. Among the control rods currently used in fast neutron reactors, 10 B is the most important neutron absorbing nuclide, so the 10 B nuclear density is used as a measure of control rod burnup, and the control rod burnup depth is 10 B is the reduction of nuclear density. Then, for the direct burnup assessment rod group, only the average neutron flux density of each block is accumulated to calculate the total neutron flux. For the fine burnup assessment rod group, after obtaining the average neutron flux density, the absorber is solved immediately. 10 B's burnup equation, and the calculated control rod 10 The B nucleon density is fed back to the neutron transport calculation of the next burnup step. When all burnup step calculations are completed, the total neutron flux during the lifetime is used to calculate the direct burnup assessment rod group. 10 The burnup equation of B is used to obtain the material composition of the control rod after the burnup. The control rod burnup evaluation strategy has the ability to process each absorber area on the control rod. 10 The function of changing the nuclear density with the burnup of B, while taking into account factors such as the reduction in the value of the control rods due to the burnup of the control rods and the change in the length of the material area involved in the burnup calculation due to the movement of the control rods, makes the calculation results more consistent with the actual situation. The following uses the Super Phoenix fast neutron breeder reactor problem to describe the specific implementation steps:

[0042] Step 1: Figure 2 This is the core layout of the Super Phoenix fast breeder reactor. The right third of the core is cut off for calculation. The compensating rod assembly is divided into 7 groups as the control rods in the core, such as Figure 3 As shown in the figure. First, according to the core control rod movement logic, the control rods in the core are divided into a fine burnup assessment rod group and a direct burnup assessment rod group. In order to compare the difference in the calculation results of the fine burnup assessment rod group and the direct burnup assessment rod group, the seven control rod groups are calculated as fine burnup assessment rod groups and direct burnup assessment rod groups respectively, and the burnup step is set to 3 steps.

[0043] Step 2: For the fine burnup assessment rod group, in each burnup step, based on the results of the steady-state neutron transport calculation, the average neutron flux density of each segment of the seven control rod assemblies is obtained by the following formula:

[0044]

[0045] Where:

[0046] ——the average neutron flux density of the control rod energy group ig in the i-th burnup step

[0047] ——Neutron flux density of the nth control rod assembly segment

[0048] ——nth b - Neutron flux density of one control rod assembly segment, n b -1 is the number of the bottom node of the control rod

[0049] ——nth t + Neutron flux density of 1 control rod assembly segment, n t +1 is the number of the top node of the control rod

[0050] n——Number of each segment of the control rod assembly

[0051] h——Height of each segment of the control rod assembly

[0052] h t ——Height of the top segment of the control rod assembly

[0053] h b ——Height of the bottom segment of the control rod assembly

[0054] l is the total length of the control rods

[0055] Step 3: For the fine burnup evaluation rod group, after obtaining the average neutron flux density of the seven control rod assemblies, the following absorber is solved for each control rod assembly: 10B's burnup equation, and the calculated absorber in the control rod 10 The B nucleon density is fed back into the neutron transport calculation for the next burnup step. Table 1 shows the detailed burnup assessment rod group of the Super Phoenix fast breeder reactor at various burnup points. 10 B nuclear density;

[0056]

[0057] Where:

[0058] —— 10 B nucleon density / cm -3

[0059] N0—— 10 B initial nucleon density / cm -3

[0060] —— 10 Microscopic absorption cross section of B / cm 2

[0061] ——Neutron flux density / cm -2 ·s -1

[0062] t——time / s

[0063] Table 1 Super Phoenix fast breeder reactor fine burnup assessment rod set 10 BNuclear density

[0064]

[0065] Step 4: For the direct burnup assessment rod group, based on the results of the steady-state neutron transport calculation, the average neutron flux density of each segment of the seven control rod assemblies is obtained by equation (1);

[0066] Step 5: For the direct burnup assessment rod group, only the average neutron flux density of each node is accumulated to calculate the total neutron flux density of each direct burnup assessment rod group;

[0067] Step 6: For the direct burnup assessment rod group, after all three burnup steps are completed, use the total neutron flux during the lifetime to calculate 10 The burnup equation of B is (2), which is used to obtain the burnup of this group of control rods. 10 B nuclear density, Table 2 is the direct burnup assessment rod group of the Super Phoenix fast neutron breeder reactor 10 B nuclear density;

[0068] Table 2 Super Phoenix fast breeder reactor direct burnup assessment rod set 10BNuclear density

[0069]

[0070] Step 7: The core burnup calculation is completed.

[0071] To verify the effectiveness of the present invention, the calculation results of the Super Phoenix fast breeder reactor using the present invention as the fine burnup assessment rod group and the traditional method using the Super Phoenix fast breeder reactor using the seven control rod groups as the direct burnup assessment rod group are compared. Table 3 shows the calculation results of the various control rod groups for the Super Phoenix fast breeder reactor using the present invention and the traditional method. 10 B comparison of loss.

[0072] Table 3 Comparison of calculation results between the present invention and the traditional method

[0073]

[0074] As can be seen from Table 3, the present invention can not only calculate the absorber area of ​​each control rod in each burnup step, but also calculate the absorber area of ​​each control rod in each burnup step. 10 B nuclear density changes with the burnup, and can also output each node in each burnup step 10 B nuclear density. It also comprehensively considers factors such as the reduction in control rod value due to control rod burnup and the change in the length of the material area involved in the burnup calculation due to control rod position movement. This results in higher calculation accuracy and plays a relatively important role in overall core design and core calculations.

Claims

1. A control rod burnup assessment method for an advanced nuclear reactor, characterized by: The steps include: Step 1: First, based on the core control rod movement logic, the control rods in the core are divided into a fine burnup assessment rod group and a direct burnup assessment rod group; Step 2: For the fine burnup assessment rod group, in each burnup step, based on the results of the steady-state neutron transport calculation, the average neutron flux density of each segment of the control rod assembly is obtained by the following formula: Where: ——the average neutron flux density of the control rod energy group ig in the i-th burnup step ——Neutron flux density of the nth control rod assembly segment ——nth b - Neutron flux density of one control rod assembly segment, n b -1 is the number of the bottom node of the control rod ——nth t + Neutron flux density of 1 control rod assembly segment, n t +1 is the number of the top node of the control rod n——Number of each segment of the control rod assembly h——Height of each segment of the control rod assembly h t ——Height of the top segment of the control rod assembly h b ——Height of the bottom segment of the control rod assembly l is the total length of the control rods Step 3: For the fine burnup evaluation rod group, after obtaining the average neutron flux density, solve the following absorber 10 B's burnup equation, and the calculated absorber in the control rod 10 The B nucleon density is fed back into the neutron transport calculation for the next burnup step; Where: —— 10 B nucleon density / cm -3 N0—— 10 B initial nucleon density / cm -3 —— 10 Microscopic absorption cross section of B / cm 2 ——Neutron flux density / cm -2 ·s -1 t——time / s Step 4: For the direct burnup assessment rod group, according to the results of the steady-state neutron transport calculation, the average neutron flux density of each segment of the control rod assembly is obtained by formula (1); Step 5: For the direct burnup assessment rod group, only the average neutron flux density of each segment is accumulated to calculate the total neutron flux density; Step 6: For the direct burnup assessment rod group, after all burnup steps are calculated, use the total neutron flux during the lifetime to calculate the absorber 10 The burnup equation of B (2) is used to obtain the absorber after the burnup of this group of control rods 10 B nuclear density; Step 7: The core burnup calculation is completed.

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