Calculation Method for Nuclear Density of Poison in External Fuel Circulation of Pebble Bed High-Temperature Gas-Cooled Reactor

Abstract

The present invention discloses a method for calculating the nuclear density of poisons in fuel spheres circulating outside the reactor of a pebble-bed high-temperature gas-cooled reactor. First, the distribution of multiple neutron fluxes within the reactor is obtained; the burnup chain of important poisons is simplified, and the burnup equation of the simplified burnup chain is established and solved to obtain the equilibrium concentration of the batch that will return to the reactor core in all areas of the bottom layer of the pebble bed at the current stage, i.e., the initial fuel sphere poison concentration; a burnup equation is established after the fuel spheres are discharged from the reactor, and the nuclear density of a batch of poisons discharged from a flow channel, after staying outside the reactor for a period of time before returning to the reactor core, is derived and calculated; until the change in the nuclear density of poisons in each flow channel that will return to the reactor core is calculated; the nuclear density of the same batch of poisons in different flow channels that have decayed outside the reactor for a period of time is calculated as a volume-weighted average based on the area ratio of each flow channel; and finally, the nuclear density of the poisons in the batch that returns to the reactor core after decaying outside the reactor and being evenly mixed is obtained. The method of the present invention makes the poison concentration of the fuel spheres that have returned to the reactor core after circulating outside the reactor more reasonable.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear reactor physics calculation, and in particular to a method for calculating the nucleon density of poisons in the external circulation fuel spheres of a pebble bed high-temperature gas-cooled reactor. Background Art

[0002] The pebble-bed high-temperature gas-cooled reactor (HTGR) uses spherical fuel elements approximately 6 cm in diameter, with approximately 420,000 fuel pellets loaded into the core. The fuel pellets are constructed from a graphite matrix with TRISO fuel particles dispersed throughout. The core pebble bed is constructed from a graphite reflector layer, with a cylindrical upper structure and a funnel-shaped lower structure connected to a discharge pipe. Fuel pellets are continuously loaded into the core from the top of the reactor, while spent fuel pellets are continuously discharged from a discharge pipe at the bottom of the core. After discharge, the discharged fuel pellets are measured for burnup, and those that have not reached the predetermined burnup depth are returned to the reactor for reuse. Fuel pellets undergo an average of more than 10 cycles within the reactor. After discharge, the fuel pellets undergo the aforementioned process before returning to the core. During this time, the poisons within the fuel pellets decay, causing changes in the nuclear density. Failure to account for this effect would introduce errors in subsequent full-core diffusion calculations, necessitating corrections for the poison concentrations upon the return of the fuel pellets to the core.

[0003] The existing technique uses a volume-weighted average of the density of poison nuclei from the same batch of different flow channels at the bottom layer of the core, which need to be returned to the core, based on the radial cross-sectional area of ​​each flow channel. This technique accounts for the mixing effect after the fuel spheres are discharged, but does not consider the decay of the poison within the fuel. This introduces a certain error in the homogenized cross-section of the topmost region after discharge, thus affecting the accuracy of the diffusion calculation. Summary of the Invention

[0004] To address the aforementioned problems in the prior art, the present invention provides a method for calculating the nuclear density of poisons in the recirculating fuel spheres of a pebble-bed high-temperature gas-cooled reactor. This method takes into account the decay of poisons in the recirculating fuel spheres. By solving a simplified burnup equation, the decayed poison nuclear density is used to replace the poison concentration in the fuel spheres at the time of discharge from the core. This method results in a more reasonable poison concentration after the recirculated fuel spheres return to the core.

[0005] In order to achieve the above objectives, the present invention adopts the following technical solutions to be implemented:

[0006] A method for calculating the nuclear density of poisons in the fuel spheres of a pebble-bed high-temperature gas-cooled reactor comprises the following steps:

[0007] Step 1: Use the high-temperature gas-cooled reactor core physics calculation program to calculate the multi-group neutron flux distribution in the reactor;

[0008] Step 2: Simplify the burnup chain of important poisons, establish and solve the burnup equation of the simplified burnup chain to obtain the equilibrium concentration of poisons in each batch of fuel balls in the bottom area of ​​each flow channel at a certain discharge stage;

[0009]

[0010] Where:

[0011] N I,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 135 I equilibrium concentration

[0012] N Xe,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 135 Xe equilibrium concentration

[0013] N Pm,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 149 Pm equilibrium concentration

[0014] N Sm,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 149 Sm equilibrium concentration

[0015] γ I,h ——The hth energy group 135 The fission yield of I

[0016] γ Xe,h ——The hth energy group 135 Xe fission yield

[0017] γ Pm,h ——The hth energy group 149 Pm fission yield

[0018] Σ f,h,j,i ——The macroscopic fission cross section of the hth energy group in the i-th batch in the lowest region of the j-th flow channel

[0019] ——Neutron flux density of the hth energy group in the i-th batch of balls in the lowest region of the j-th flow channel

[0020] λ I —— 135 The decay constant of I

[0021] λ Xe —— 135 Xe decay constant

[0022] λ Pm —— 149 Decay constant of Pm

[0023] λ Sm —— 149 Sm decay constant

[0024] —— 135 Microscopic absorption cross section of the h-th energy group of Xe —— 149 The microscopic absorption cross section of the h-th energy group of Sm;

[0025] Step 3: Repeat step 2 until the equilibrium concentration of the fuel ball poisons in all areas of the bottom layer of the pebble bed that will return to the core is obtained. This concentration is the initial concentration of the fuel ball poisons in each batch in each bottom layer when they are unloaded from the core.

[0026] Step 4: Establish the burnup equation after the fuel balls are unloaded from the reactor, and derive and calculate the nuclear density of a batch of fuel ball poisons unloaded from a flow channel after staying outside the reactor for a period of time before returning to the core;

[0027]

[0028]

[0029] Where:

[0030] N I,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 135 I Nuclear density

[0031] N Xe,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 135 Xe nucleon density

[0032] N Pm,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 149 Pm nucleon density

[0033] N Sm,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 149 Sm nucleon density;

[0034] Step 5: Repeat step 4 until the density of poison nuclei in each fuel ball of each flow channel that will return to the core is obtained;

[0035] Step 6: The density of poison nuclei of the same batch of different flow channels that have decayed outside the reactor for a period of time is calculated as a volume-weighted average based on the area ratio of each flow channel;

[0036]

[0037] Where:

[0038] N x,i (t)——nuclear density of the i-th batch of x poison nuclides before they finally return to the core

[0039] N x,i,j (t)——Nuclear density of x poison nuclide in batch i of j flow channel

[0040] S j ——the radial cross-sectional area of ​​the jth flow channel;

[0041] Until the density of poison nuclei is obtained for each batch that will return to the core after decaying outside the core and being evenly mixed.

[0042] Compared with existing technologies, the present invention has the following advantages: It takes into account the decay of poisons in the recirculating fuel spheres. By solving a simplified burnup equation, the decayed poison nucleus density is used to replace the poison concentration in the fuel spheres at the time of discharge from the core. This results in a more reasonable poison concentration after the recirculated fuel spheres return to the core. BRIEF DESCRIPTION OF THE DRAWINGS

[0043] Figure 1 Simplified burnup chain for xenon.

[0044] Figure 2 Simplify the burnup chain for samarium.

[0045] Figure 3 This is the overall flow chart of the calculation method for the nuclear density of poisons in the external fuel circulation of a pebble bed high-temperature gas-cooled reactor. DETAILED DESCRIPTION

[0046] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments:

[0047] The present invention provides a method for calculating the nuclear density of poisons in fuel spheres circulating outside a pebble bed high-temperature gas-cooled reactor. The method comprises the following steps: first, using a high-temperature gas-cooled reactor core physics calculation program to calculate the distribution of multiple neutron fluxes within the reactor; simplifying the burnup chain of important poisons, establishing and solving the burnup equation of the simplified burnup chain to obtain the equilibrium concentration of poisons in each batch of fuel spheres in the bottom region of each flow channel at a certain discharge stage; and finally obtaining the equilibrium concentration of poisons in batches of fuel spheres that will return to the core in all regions of the bottom region of the pebble bed. This concentration is the initial concentration of poisons in each batch of fuel spheres in each region when discharged from the core; establishing a burnup equation after the fuel spheres are discharged from the reactor, and deriving and obtaining the nuclear density of poisons in a batch of fuel spheres discharged from a flow channel after staying outside the reactor for a period of time before returning to the core; and finally obtaining the change in the nuclear density of poisons in each batch of fuel spheres that will return to the core in each flow channel; and calculating the volume-weighted average of the nuclear densities of poisons in the same batch of fuel spheres in different flow channels that have decayed outside the reactor for a period of time based on the area ratio of each flow channel. Finally, the nuclear density of poisonous substances in the fuel balls returned to the core is obtained after decaying outside the core and being evenly mixed.

[0048] The following describes the application of the calculation of the nuclear density of poisons in the extra-core circulating fuel balls by taking the transition core from the full reactor loading of 4.2% enrichment fuel balls to the full reactor loading of 8.5% enrichment fuel balls as an example. Figure 3 As shown, the specific steps include:

[0049] Step 1: Use the high-temperature gas-cooled reactor core physics calculation program NECP-Panda to calculate the multi-group neutron flux distribution in the reactor at the current stage;

[0050] Step 2: Simplify the burnup chains of important poisons such as iodine 135, xenon 135, promethium 149, and samarium 149. Figure 1 The figure shows a simplified xenon burnup chain. The production of xenon 135 consists of two parts: the decay of iodine 135 and the fission of nuclear fuel. Its disappearance is through its own decay and neutron capture reaction, as shown in Figure 2 The figure shows a simplified samarium burnup chain. The source of samarium 149 is the beta - The decay is due to the neutron capture reaction of the fuel itself. The equilibrium concentration of the poison in each batch of fuel balls in the bottom area of ​​each flow channel at the current pouring stage is obtained by solving the simplified burnup equation of the burnup chain.

[0051]

[0052] Where:

[0053] N I,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 135 I equilibrium concentration

[0054] N Xe,i,j(∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 135 Xe equilibrium concentration

[0055] N Pm,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 149 Pm equilibrium concentration

[0056] N Sm,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 149 Sm equilibrium concentration

[0057] γ I,h ——The hth energy group 135 The fission yield of I

[0058] γ Xe,h ——The hth energy group 135 Xe fission yield

[0059] γ Pm,h ——The hth energy group 149 Pm fission yield

[0060] Σ f,h,j,i ——The macroscopic fission cross section of the hth energy group in the i-th batch in the lowest region of the j-th flow channel

[0061] ——Neutron flux density of the hth energy group in the i-th batch of balls in the lowest region of the j-th flow channel

[0062] λ I —— 135 The decay constant of I

[0063] λ Xe —— 135 Xe decay constant

[0064] λ Pm —— 149 Decay constant of Pm

[0065] λ Sm —— 149 Sm decay constant

[0066] —— 135 Microscopic absorption cross section of the h-th energy group of Xe —— 149 The microscopic absorption cross section of the h-th energy group of Sm;

[0067] Step 3: Repeat step 2 until the equilibrium concentration of the fuel ball poisons in all areas of the bottom layer of the pebble bed that will return to the core is obtained. This concentration is the initial concentration of the fuel ball poisons in each batch in each bottom layer when they are unloaded from the core.

[0068] Step 4: Establish the burnup equation after the fuel balls are unloaded outside the reactor, derive and calculate the nuclear density of a batch of fuel ball poisons unloaded from a flow channel that stays outside the reactor for a period of time before returning to the core.

[0069]

[0070]

[0071] Where:

[0072] N I,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 135 I Nuclear density

[0073] N Xe,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 135 Xe nucleon density

[0074] N Pm,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 149 Pm nucleon density

[0075] N Sm,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 149 Sm nucleon density;

[0076] Step 5: Repeat step 4 until the density of poison nuclei in each fuel ball of each flow channel that will return to the core is obtained;

[0077] Step 6: The density of poison nuclei of the same batch in different flow channels that have decayed outside the reactor for a period of time is calculated as a volume-weighted average based on the area ratio of each flow channel.

[0078]

[0079] Where:

[0080] N x,i (t)——nuclear density of the i-th batch of x poison nuclides before they finally return to the core

[0081] N x,i,j (t)——Nuclear density of x poison nuclide in batch i of j flow channel

[0082] S j ——the radial cross-sectional area of ​​the jth flow channel;

[0083] Until the density of poison nuclei is obtained for each batch that will return to the core after decaying outside the core and being evenly mixed.

[0084] The ingenuity of the method of the present invention lies in that it approximately assumes that the fuel balls in the bottom area of ​​the core are in a shutdown state after being unloaded from the core, and the neutron flux density is zero. Solving the burnup equation obtains the density of poison nuclides after decay outside the core, effectively taking into account the decay effect of the poison nuclides in the fuel balls participating in the outside-core circulation.

Claims

1. A method for calculating the nuclear density of poisons in the external fuel circulation of a pebble-bed high-temperature gas-cooled reactor, characterized by: The following steps are involved: Step 1: Use the high-temperature gas-cooled reactor core physics calculation program to calculate the multi-group neutron flux distribution in the reactor; Step 2: Simplify the burnup chain of important poisons, establish and solve the burnup equation of the simplified burnup chain to obtain the equilibrium concentration of poisons in each batch of fuel balls in the bottom area of ​​each flow channel at a certain discharge stage; Where: N I,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 135 I equilibrium concentration N Xe,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 135 Xe equilibrium concentration N Pm,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 149 Pm equilibrium concentration N Sm,i,j (∞)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel 149 Sm equilibrium concentration γ I,h ——The hth energy group 135 The fission yield of I γ Xe,h ——The hth energy group 135 Xe fission yield γ Pm,h ——The hth energy group 149 Pm fission yield Σ f,h,j,i ——The macroscopic fission cross section of the hth energy group in the i-th batch in the lowest region of the j-th flow channel ——Neutron flux density of the hth energy group in the i-th batch of balls in the lowest region of the j-th flow channel λ I —— 135 The decay constant of I λ Xe —— 135 Xe decay constant λ Pm —— 149 Decay constant of Pm λ Sm —— 149 Sm decay constant —— 135 Microscopic absorption cross section of the h-th energy group of Xe —— 149 The microscopic absorption cross section of the h-th energy group of Sm; Step 3: Repeat step 2 until the equilibrium concentration of the fuel ball poisons in all areas of the bottom layer of the pebble bed that will return to the core is obtained. This concentration is the initial concentration of the fuel ball poisons in each batch in each bottom layer when they are unloaded from the core. Step 4: Establish the burnup equation after the fuel balls are unloaded from the reactor, and derive and calculate the nuclear density of a batch of fuel ball poisons unloaded from a flow channel after staying outside the reactor for a period of time before returning to the core; Where: N I,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 135 I Nuclear density N Xe,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 135 Xe nucleon density N Pm,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 149 Pm nucleon density N Sm,i,j (t)——the number of balls in the i-th batch in the lowest area of ​​the j-th flow channel that stay outside the reactor for a period of time t before returning to the core 149 Sm nucleon density; Step 5: Repeat step 4 until the density of poison nuclei in each fuel ball of each flow channel and each batch that will return to the core is obtained; Step 6: The density of poison nuclei of the same batch of different flow channels that have decayed outside the reactor for a period of time is calculated as a volume-weighted average based on the area ratio of each flow channel; Where: N x,i (t)——Nuclear density N of the i-th batch of x poison nuclides before they finally return to the core x,i,j (t)——Nuclear density of x poison nuclide in batch i of j flow channel S j ——the radial cross-sectional area of ​​the jth flow channel; The final result is the density of poison nuclei after the batches of poisons that will be returned to the core decay outside the reactor and are evenly mixed.

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