CN114003856A - Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor - Google Patents

Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor Download PDF

Info

Publication number
CN114003856A
CN114003856A CN202111296918.1A CN202111296918A CN114003856A CN 114003856 A CN114003856 A CN 114003856A CN 202111296918 A CN202111296918 A CN 202111296918A CN 114003856 A CN114003856 A CN 114003856A
Authority
CN
China
Prior art keywords
calculation
energy
neutron
reactor
nuclide
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Granted
Application number
CN202111296918.1A
Other languages
Chinese (zh)
Other versions
CN114003856B (en
Inventor
王成龙
温永江
张大林
秋穗正
苏光辉
田文喜
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Xian Jiaotong University
Original Assignee
Xian Jiaotong University
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Xian Jiaotong University filed Critical Xian Jiaotong University
Priority to CN202111296918.1A priority Critical patent/CN114003856B/en
Publication of CN114003856A publication Critical patent/CN114003856A/en
Application granted granted Critical
Publication of CN114003856B publication Critical patent/CN114003856B/en
Active legal-status Critical Current
Anticipated expiration legal-status Critical

Links

Images

Classifications

    • G—PHYSICS
    • G06—COMPUTING; CALCULATING OR COUNTING
    • G06F—ELECTRIC DIGITAL DATA PROCESSING
    • G06F17/00—Digital computing or data processing equipment or methods, specially adapted for specific functions
    • G06F17/10—Complex mathematical operations
    • G06F17/11—Complex mathematical operations for solving equations, e.g. nonlinear equations, general mathematical optimization problems

Landscapes

  • Engineering & Computer Science (AREA)
  • Physics & Mathematics (AREA)
  • Mathematical Physics (AREA)
  • General Physics & Mathematics (AREA)
  • Pure & Applied Mathematics (AREA)
  • Mathematical Optimization (AREA)
  • Mathematical Analysis (AREA)
  • Computational Mathematics (AREA)
  • Data Mining & Analysis (AREA)
  • Theoretical Computer Science (AREA)
  • Operations Research (AREA)
  • Algebra (AREA)
  • Databases & Information Systems (AREA)
  • Software Systems (AREA)
  • General Engineering & Computer Science (AREA)
  • Monitoring And Testing Of Nuclear Reactors (AREA)

Abstract

一种核热推进反应堆停堆状态外环境辐射场计算方法，步骤如下：1、确定核热推进反应堆几何结构，运行史，燃料分布，屏蔽材料与结构参数；2、使用SCALE软件的TRITON控制模块计算核热推进反应堆运行产物的放射性强度及分布；3、使用SCALE软件的ORIGEN‑S控制模块计算运行产物的中子能谱与光子能谱；4、使用SCALE软件的多维点核分析程序QADS计算核热推进反应堆外环境辐射；5、增加计算区域数量，跳至步骤2进行新的计算，比较两次计算结果，若计算结果变化不在可接受范围内，则增加计算区域继续计算；若计算结果变化在可接受范围内，则不再增加计算区域数量，进一步完成辐射场的计算。本发明的方法可以得到精确的计算结果。

Figure 202111296918

A method for calculating the external environment radiation field in the shutdown state of a nuclear thermal propulsion reactor, the steps are as follows: 1. Determine the nuclear thermal propulsion reactor geometry, operation history, fuel distribution, shielding materials and structural parameters; 2. Use the TRITON control module of SCALE software Calculate the radioactive intensity and distribution of the operating products of the nuclear thermal propulsion reactor; 3. Use the ORIGEN‑S control module of the SCALE software to calculate the neutron energy spectrum and photon energy spectrum of the operating products; 4. Use the multi-dimensional point nuclear analysis program QADS of the SCALE software to calculate External environmental radiation of nuclear thermal propulsion reactor; 5. Increase the number of calculation areas, skip to step 2 to perform a new calculation, compare the two calculation results, if the calculation results change is not within the acceptable range, increase the calculation area and continue the calculation; if the calculation results If the change is within the acceptable range, the number of calculation areas will not be increased, and the calculation of the radiation field will be further completed. The method of the present invention can obtain accurate calculation results.

Figure 202111296918

Description

Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor
Technical Field
The invention relates to the field of nuclear thermal propulsion reactors, in particular to a method for calculating an external environment radiation field in a shutdown state of a nuclear thermal propulsion reactor.
Background
The nuclear power propulsion has the great advantages of ultra-long endurance, strong maneuverability, high concealment and the like, and attracts the sight of the whole world. However, the nuclear thermal propulsion reactor can generate a large amount of radioactive substances during operation and after shutdown, and release a large amount of radiation, and meanwhile, due to economic considerations, the nuclear thermal propulsion reactor generally has no perfect radiation shielding device, so that a large amount of radiation pollution can be caused to the external environment, and the use and the storage of the nuclear thermal propulsion reactor are greatly influenced.
Disclosure of Invention
In order to overcome the above problems, an object of the present invention is to provide a method for calculating an external environment radiation field in a shutdown state of a nuclear thermal propulsion reactor, which considers the influence of uneven distribution of radioactive materials and performs calculation by gradually increasing the calculation area, so as to obtain the radiation field intensity distribution of the external environment in the shutdown state of the nuclear thermal propulsion reactor. The invention provides theoretical suggestion and guidance for the calculation of the environmental radiation in the shutdown state of the nuclear thermal propulsion reactor, and provides a method for designing the nuclear thermal propulsion reactor with high efficiency and safety.
In order to achieve the purpose, the invention adopts the following technical scheme:
a method for calculating an ambient radiation field in a shutdown state of a nuclear thermal propulsion reactor comprises the following steps:
step 1: determining the geometric structure, the operation history, the fuel distribution, the shielding material and the structural parameters of the nuclear thermal propulsion reactor;
step 2: TRITON is a multipurpose control module in SCALE software, couples a transport calculation program KENO and a burnup calculation program ORIGEN-S, and solves a neutron transport equation (1) and a burnup equation (2):
neutron transport equation:
Figure BDA0003334681220000021
the burnup equation:
Figure BDA0003334681220000022
in the formula
Omega-unit vector of direction of motion
Phi (r, omega, E) -neutron angular fluence rate/m at position r, with omega direction of motion and E energy-2·s-1
Phi (r, omega ', E') -neutron angular fluence rate/m with energy E 'at position r, motion direction omega', and energy E-2·s-1
r-spatial position/m
Σt-probability of neutron collision
ΣsNeutron moderation probability
Σf-probability of fission
E-energy/J
X (E) -neutron fission spectrum
V-neutron velocity/m.s-1
f (r, E '→ E, Ω' → Ω) — the energy changes from E 'to E, and the neutron whose direction of motion changes from Ω' to Ω accounts for the proportion of all other angles and colliding neutrons of energy
NiAtomic density/atom.cm of nuclide i-3
NjAtomic density/atom.cm of nuclide j-3
NkAtomic density of nuclide k/atom · cm-3
λiDecay constant of nuclide i
λjDecay constant of nuclide j
σiSpectral average neutron absorption cross section of nuclide i/b
σkSpectral average neutron absorption cross section/b of nuclide k
Figure BDA0003334681220000031
-space and energy average neutron flux/m-2·s-1
lij-branching ratio/% of radioactive decay of other nuclides
fik-branch ratio/% of the other nuclide k absorbing the neutron-producing nuclide i
Setting material parameters by using a multipurpose control module TRITON, establishing a geometric model of the nuclear thermal propulsion reactor, setting the running time and the running power, and calculating the radioactivity of fission products and actinide products after the nuclear thermal propulsion reactor is finished running and the radioactivity of each fuel region with different uranium concentrations; in the results output by the multipurpose control module TRITON, each fuel area with uranium concentration only outputs the radioactivity results of one fission product and one actinide product, the area which only outputs the radioactivity results of one fission product and one actinide product is regarded as a calculation area, and the radioactive substances and the fuel uranium concentration in the calculation area are regarded as uniform distribution; because each calculation region only outputs the radioactivity results of one fission product and actinide product when the multipurpose control module TRITON outputs the results, when a more detailed radioactive substance distribution result of a certain calculation region needs to be known, the region needs to be split into a plurality of calculation regions, the number of the calculation regions is increased, and the specific steps of splitting the calculation regions and increasing the number of the calculation regions are as follows:
1): when the multipurpose control module TRITON is used for setting material parameters, material numbers are added to materials in a calculation area needing to be split, namely a plurality of material numbers are created for one material at the same time;
2): when a multipurpose control module TRITON is used for establishing a geometric model of the nuclear thermal propulsion reactor, geometric modeling is carried out on a calculation region needing to be split again, the calculation region needing to be split is not regarded as a geometric body any more, but is split into a plurality of small geometric bodies for modeling; meanwhile, when setting material parameters, although the material of each small geometric body is the same, the material number of each small geometric body cannot be set to be the same and is sequentially set to be the material number created in 1);
3): adding the material number created in 1) into the material number setting set by the output result of the multipurpose control module TRITON;
4): after the steps are completed, when the multipurpose control module TRITON completes the calculation, calculation results of more calculation areas are obtained from the calculation results;
and step 3: inputting the calculation results of the radioactivity of the TRITON fission products and the actinide products of the multipurpose control module into an ignition energy consumption calculation program ORIGEN-S to calculate photon energy spectrums and neutron energy spectrums of the nuclear thermal propulsion reactor full stack and the operation products in each calculation region; the gamma ray source intensity and energy spectrum calculated by the ignition loss calculation program ORIGEN-S include X-rays, gamma rays, bremsstrahlung, spontaneous fission gamma rays and photons generated by gamma rays accompanying the (α, n) reaction, the nuclear data is stored in the binary format library of the ignition loss calculation program ORIGEN-S, and the photon energy spectrum of the user-directly specified energy group structure is calculated by the formula (3):
Ig＝Ia(Ea/Eg) (3)
in the formula
Ia-actual photon intensity/photons s in gamma library-1
Ea-actual photon energy/MeV
Eg-energy group mean energy/MeV
Ig-photon intensity/photons.s of energy group-1
The neutron source intensity and energy spectrum calculated by the ignition depletion calculation program ORIGEN-S includes spontaneous fission (α, n) reactions and (β) reactions, where (β) reactions are not important in typical spent fuels; the (α, n) reaction then calculates the effect of the surrounding medium by:
Figure BDA0003334681220000051
in the formula
S (E) -Total stopping force of Medium/b
σi(E) Of a nuclear species i: (α, n) reaction section/b
E-neutron energy/MeV
Yi,kThe energy emitted by the nuclide k is EαNeutron yield per neutrons s of alpha particles of (2)-1
N-Total atomic Density/atom. cm-3
Ni-atomic density/atom.cm of target nuclide-3
And 4, step 4: establishing a nuclear thermal propulsion reactor geometric model by using a multi-dimensional point nuclear analysis program QADS (quality and intensity), inputting the distribution and the neutron energy spectrum of each calculation area, setting a calculation reference point, and calculating the radiation dose of the reference point; aiming at a typical cylindrical reactor, by axially dividing the reactor into small sections of cylinders, creating a new input card for each section, and then adjusting the radial radioactive substance distribution in each input card through weight statements, the precise description of the radioactive substance distribution is realized; the QADS adopts a point kernel integration algorithm, and the equation is as follows:
Figure BDA0003334681220000061
in the formula
Figure BDA0003334681220000062
-calculating the position/m of a point
Figure BDA0003334681220000063
-location of source/m in volume element v
V-volume of source/m3
Mu-total attenuation coefficient at energy E
Figure BDA0003334681220000064
-distance between source point and calculation point/m
K-flux-dose conversion factor
B-dose cumulative factor
Figure BDA0003334681220000065
Dose rate/rem.h-1
s-number of photons/photons s-1
E-photon energy/MeV
And 5: firstly, selecting some reference points, and only performing radiation dose calculation of the reference points before determining the number of calculation areas, wherein the selected reference points need to be uniformly distributed around a reactor and cannot be too far away from the reactor; after the first round of calculation is finished, increasing the number of calculation areas according to the method for dividing the calculation areas and increasing the number of calculation areas in the step 2, jumping to the step 2 to perform a new round of calculation, comparing the results of the previous and new rounds of calculation after the new round of calculation is finished, if the change of the calculation result is not within the acceptable range, continuing to increase the calculation areas, and recalculating until the change of the calculation result is within the acceptable range; if the change of the calculation result is within the acceptable range, the calculation region is selected sufficiently, the uneven distribution of the radioactive substance can be simulated, the calculation is carried out according to the calculation region setting of the new calculation, and the calculation of the whole radiation field is completed.
Compared with the prior art, the invention has the following outstanding characteristics:
(1) by increasing the number of calculation areas, a more accurate source item distribution result can be obtained, and the accurate calculation of the uneven distribution of the radioactive source can be realized; (2) the most economic and effective calculation region setting number can be searched, and the calculation efficiency and the calculation precision are improved; (3) the external environment radiation field caused by the operation products in the shutdown state of the nuclear thermal propulsion reactor can be accurately calculated.
Aiming at the existing problems, the invention provides a fine calculation method for the radiation of the external environment under the shutdown state of the nuclear thermal propulsion reactor, thereby providing a reference for the design and the safety of the nuclear thermal propulsion reactor.
Drawings
FIG. 1 is a flow chart of the method of the present invention.
Detailed Description
The invention is described in further detail below with reference to the following figures and detailed description:
the invention relates to a method for calculating an external environment radiation field in a shutdown state of a nuclear thermal propulsion reactor,
the nuclear thermal propulsion reactor has the characteristics of high energy density, small volume and light weight, and in order to realize miniaturization design and improve nuclear reactivity, the nuclear reactor necessarily adopts high-concentration uranium materials, and radioactive operation products are main radioactive source items during shutdown storage. The method comprises the steps of solving the radiation of the environment outside the reactor caused by radioactive operation products by using SCALE software, refining the calculated area by continuously increasing the number of the calculated areas in the reactor, obtaining the distribution condition of more refined operation products, and obtaining more accurate dose equivalent rate calculation results by inputting more accurate and detailed radiation sources when solving the dose equivalent rate; the method can calculate the external environment radiation caused by the operation products in the shutdown state of the nuclear thermal propulsion reactor, and can obtain an accurate dose equivalent rate calculation result in the calculation.
As shown in fig. 1, the method specifically includes the following steps:
step 1: determining the geometric structure, the operation history, the fuel distribution, the shielding material and the structural parameters of the nuclear thermal propulsion reactor;
step 2: TRITON is a multipurpose control module in SCALE software, couples a transport calculation program KENO and a burnup calculation program ORIGEN-S, and solves a neutron transport equation (1) and a burnup equation (2):
neutron transport equation:
Figure BDA0003334681220000081
the burnup equation:
Figure BDA0003334681220000082
in the formula
Omega-unit vector of direction of motion
Phi (r, omega, E) -neutron angular fluence rate/m at position r, with omega direction of motion and E energy-2·s-1
Phi (r, omega ', E') -neutron angular fluence rate/m with energy E 'at position r, motion direction omega', and energy E-2·s-1
r-spatial position/m
Σt-probability of neutron collision
ΣsNeutron moderation probability
Σf-probability of fission
E-energy/J
X (E) -neutron fission spectrum
V-neutron velocity/m.s-1
f (r, E '→ E, Ω' → Ω) — the energy changes from E 'to E, and the neutron whose direction of motion changes from Ω' to Ω accounts for the proportion of all other angles and colliding neutrons of energy
NiAtomic density/atom.cm of nuclide i-3
NjAtomic density/atom.cm of nuclide j-3
NkAtomic density of nuclide k/atom · cm-3
λiDecay constant of nuclide i
λjDecay constant of nuclide j
σiSpectral average neutron absorption cross section of nuclide i/b
σkSpectral average neutron absorption cross section/b of nuclide k
Figure BDA0003334681220000091
-space and energy average neutron flux/m-2·s-1
lij-branching ratio/% of radioactive decay of other nuclides
fikOther nuclide k-aspiratesBranch ratio/% of neutron-generating nuclide i
Setting material parameters by using a multipurpose control module TRITON, establishing a geometric model of the nuclear thermal propulsion reactor, setting the running time and the running power, and calculating the radioactivity of fission products and actinide products after the nuclear thermal propulsion reactor is finished running and the radioactivity of each fuel region with different uranium concentrations; in the results output by the multipurpose control module TRITON, each fuel area with uranium concentration only outputs the radioactivity results of one fission product and one actinide product, the area which only outputs the radioactivity results of one fission product and one actinide product is regarded as a calculation area, and the radioactive substances and the fuel uranium concentration in the calculation area are regarded as uniform distribution; because each calculation region only outputs the radioactivity results of one fission product and actinide product when the multipurpose control module TRITON outputs the results, when a more detailed radioactive substance distribution result of a certain calculation region needs to be known, the region needs to be split into a plurality of calculation regions, the number of the calculation regions is increased, and the specific steps of splitting the calculation regions and increasing the number of the calculation regions are as follows:
1): when the multipurpose control module TRITON is used for setting material parameters, material numbers are added to materials in a calculation area needing to be split, namely a plurality of material numbers are created for one material at the same time;
2): when a multipurpose control module TRITON is used for establishing a geometric model of the nuclear thermal propulsion reactor, geometric modeling is carried out on a calculation region needing to be split again, the calculation region needing to be split is not regarded as a geometric body any more, but is split into a plurality of small geometric bodies for modeling; meanwhile, when setting material parameters, although the material of each small geometric body is the same, the material number of each small geometric body cannot be set to be the same and is sequentially set to be the material number created in 1);
3): adding the material number created in 1) into the material number setting set by the output result of the multipurpose control module TRITON;
4): after the steps are completed, when the multipurpose control module TRITON completes the calculation, calculation results of more calculation areas are obtained from the calculation results;
and step 3: inputting the calculation results of the radioactivity of the TRITON fission products and the actinide products of the multipurpose control module into an ignition energy consumption calculation program ORIGEN-S to calculate photon energy spectrums and neutron energy spectrums of the nuclear thermal propulsion reactor full stack and the operation products in each calculation region; the gamma ray source intensity and energy spectrum calculated by the ignition loss calculation program ORIGEN-S include X-rays, gamma rays, bremsstrahlung, spontaneous fission gamma rays and photons generated by gamma rays accompanying the (α, n) reaction, the nuclear data is stored in the binary format library of the ignition loss calculation program ORIGEN-S, and the photon energy spectrum of the user-directly specified energy group structure is calculated by the formula (3):
Ig＝Ia(Ea/Eg) (3)
in the formula
Ia-actual photon intensity/photons s in gamma library-1
Ea-actual photon energy/MeV
Eg-energy group mean energy/MeV
Ig-photon intensity/photons.s of energy group-1
The neutron source intensity and energy spectrum calculated by the ignition depletion calculation program ORIGEN-S includes spontaneous fission (α, n) reactions and (β) reactions, where (β) reactions are not important in typical spent fuels; the (α, n) reaction then calculates the effect of the surrounding medium by:
Figure BDA0003334681220000111
in the formula
S (E) -Total stopping force of Medium/b
σi(E) -reaction section of (α, n) of nuclide i/b
E-neutron energy/MeV
Yi,kThe energy emitted by the nuclide k is EαNeutron yield per neutrons s of alpha particles of (2)-1
N-total atomsDensity/atom.cm-3
Ni-atomic density/atom.cm of target nuclide-3
And 4, step 4: establishing a nuclear thermal propulsion reactor geometric model by using a multi-dimensional point nuclear analysis program QADS (quality and intensity), inputting the distribution and the neutron energy spectrum of each calculation area, setting a calculation reference point, and calculating the radiation dose of the reference point; aiming at a typical cylindrical reactor, by axially dividing the reactor into small sections of cylinders, creating a new input card for each section, and then adjusting the radial radioactive substance distribution in each input card through weight statements, the precise description of the radioactive substance distribution is realized; the QADS adopts a point kernel integration algorithm, and the equation is as follows:
Figure BDA0003334681220000121
in the formula
Figure BDA0003334681220000122
-calculating the position/m of a point
Figure BDA0003334681220000123
-location of source/m in volume element v
V-volume of source/m3
Mu-total attenuation coefficient at energy E
Figure BDA0003334681220000124
-distance between source point and calculation point/m
K-flux-dose conversion factor
B-dose cumulative factor
Figure BDA0003334681220000125
Dose rate/rem.h-1
s-number of photons/photons s-1
E-photon energy/MeV
And 5: firstly, selecting some reference points, and only performing radiation dose calculation of the reference points before determining the number of calculation areas, wherein the selected reference points need to be uniformly distributed around a reactor and cannot be too far away from the reactor; after the first round of calculation is finished, increasing the number of calculation areas according to the method for dividing the calculation areas and increasing the number of calculation areas in the step 2, jumping to the step 2 to perform a new round of calculation, comparing the results of the previous and new rounds of calculation after the new round of calculation is finished, if the change of the calculation result is not within the acceptable range, continuing to increase the calculation areas, and recalculating until the change of the calculation result is within the acceptable range; if the change of the calculation result is within the acceptable range, the calculation region is selected sufficiently, the uneven distribution of the radioactive substance can be simulated, the calculation is carried out according to the calculation region setting of the new calculation, and the calculation of the whole radiation field is completed. The invention recognizes that the acceptable range can be selected to be 1% and this value can be increased or decreased depending on the circumstances.

Claims (1)

1.一种核热推进反应堆停堆状态外环境辐射场计算方法，其特征在于：包括以下步骤：1. a nuclear thermal propulsion reactor shutdown state external environment radiation field calculation method, is characterized in that: comprise the following steps: 步骤1：确定核热推进反应堆几何结构、运行史、燃料分布、屏蔽材料与结构参数；Step 1: Determine the nuclear thermal propulsion reactor geometry, operation history, fuel distribution, shielding materials and structural parameters; 步骤2：TRITON是SCALE软件中一个多用途控制模块，将输运计算程序KENO和燃耗计算程序ORIGEN-S进行耦合，求解中子输运方程(1)与燃耗方程(2)：Step 2: TRITON is a multi-purpose control module in the SCALE software. It couples the transport calculation program KENO and the burnup calculation program ORIGEN-S to solve the neutron transport equation (1) and the burnup equation (2): 中子输运方程：Neutron transport equation:
Figure FDA0003334681210000011
Figure FDA0003334681210000011
 燃耗方程：Fuel consumption equation:
Figure FDA0003334681210000012
Figure FDA0003334681210000012
 式中in the formula Ω——运动方向的单位矢量Ω - the unit vector of the direction of motion Φ(r,Ω,E)——位于位置r，运动方向为Ω，能量为E的中子角注量率/m-2·s-1 Φ(r,Ω,E)——the neutron fluence rate at position r, the direction of motion is Ω, and the energy is E/m -2 ·s -1 Φ(r,Ω',E')——位于位置r，运动方向为Ω’，能量为E'的中子角注量率/m-2·s-1 Φ(r,Ω',E')——the neutron fluence rate at position r, the direction of motion is Ω', and the energy is E'/m -2 ·s -1 r——空间位置/mr——spatial position/m Σt——中子发生碰撞概率Σ t — probability of neutron collision Σs——中子慢化概率Σ s — Neutron Moderation Probability Σf——裂变概率Σ f — fission probability E——能量/JE - energy/J χ(E)——中子裂变能谱χ(E)—neutron fission energy spectrum ν——中子速度/m·s-1 ν——neutron velocity/m s -1 f(r,E'→E,Ω'→Ω)——能量由E'变为E，运动方向由Ω'变为Ω的中子占所有其他角度和能量的发生碰撞的中子占比f(r,E'→E,Ω'→Ω)——The proportion of neutrons whose energy changes from E' to E, and the direction of motion changes from Ω' to Ω accounts for the proportion of colliding neutrons at all other angles and energies Ni——核素i的原子密度/atom·cm-3 Ni —— atomic density of nuclide i /atom·cm -3 Nj——核素j的原子密度/atom·cm-3 N j —— atomic density of nuclide j/atom·cm -3 Nk——核素k的原子密度/atom·cm-3 N k — atomic density of nuclide k/atom·cm -3 λi——核素i的衰变常数λ i ——the decay constant of nuclide i λj——核素j的衰变常数λ j ——the decay constant of nuclide j σi——核素i的谱平均中子吸收截面/bσ i —spectral average neutron absorption cross section of nuclide i/b σk——核素k的谱平均中子吸收截面/bσ k ——the spectrally averaged neutron absorption cross section of nuclide k/b
Figure FDA0003334681210000021
——空间和能量平均中子通量/m-2·s-1
Figure FDA0003334681210000021
——Space and energy average neutron flux/m -2 ·s -1 lij——其他核素放射性衰变的分支比/％l ij —— branching ratio/% of radioactive decay of other nuclides fik——其他核素k吸收中子产生核素i的分支比/％f ik — the branching ratio of other nuclides k to absorb neutrons to produce nuclides i/% 使用多用途控制模块TRITON设定材料参数，建立核热推进反应堆几何模型，设定运行时间、运行功率，计算出核热推进反应堆运行结束后裂变产物及锕系产物的放射性活度及每种铀浓度不同的燃料区域内的放射性活度；多用途控制模块TRITON输出结果中，每种铀浓度的燃料区域只输出一种裂变产物及锕系产物的放射性活度结果，将只输出一种裂变产物及锕系产物的放射性活度结果的区域视为是一个计算区域，计算区域内放射性物质和燃料铀浓度视作均匀分布；由于多用途控制模块TRITON在输出结果时，每个计算区域只输出一种裂变产物及锕系产物的放射性活度结果，当需要知道某个计算区域更详细的放射性物质分布结果时，则需要将该区域拆分成多个计算区域，增加计算区域数量，拆分计算区域增加计算区域数量的具体步骤如下：Use the multi-purpose control module TRITON to set the material parameters, establish the geometric model of the nuclear thermal propulsion reactor, set the operating time and operating power, and calculate the radioactivity of the fission products and actinide products and the radioactivity of each uranium after the nuclear thermal propulsion reactor is completed. The radioactivity in the fuel area with different concentrations; in the output results of the multi-purpose control module TRITON, the fuel area of each uranium concentration only outputs the radioactivity results of one fission product and actinide product, and only one fission product will be output. The area of the radioactive activity results of actinide and actinide products is regarded as a calculation area, and the concentration of radioactive substances and fuel uranium in the calculation area is regarded as a uniform distribution; since the multi-purpose control module TRITON outputs results, each calculation area only outputs one The radioactivity results of the fission products and actinide products of various fission products and actinide products. When you need to know the more detailed distribution results of radioactive substances in a calculation area, you need to split the area into multiple calculation areas, increase the number of calculation areas, and split the calculation area. The specific steps for increasing the number of calculation regions are as follows: 1)：在使用多用途控制模块TRITON设定材料参数时，对需要拆分的计算区域的材料增设材料编号，即一种材料同时创建多个材料编号；1): When using the multi-purpose control module TRITON to set the material parameters, add a material number to the material in the calculation area that needs to be split, that is, create multiple material numbers for one material at the same time; 2)：在使用多用途控制模块TRITON建立核热推进反应堆几何模型时，对需要拆分的计算区域重新进行几何建模，不再将需要拆分的计算区域视作一个几何体，而是将其拆分成数个小几何体进行建模；同时在设置材料参数时，虽然每个小几何体的材料相同，但是每个小几何体的材料编号不能设置为相同的，依次设置为1)中创建的材料编号；2): When using the multi-purpose control module TRITON to build the geometric model of the nuclear thermal propulsion reactor, the geometric modeling of the calculation area that needs to be split is re-modeled, and the calculation area that needs to be split is no longer regarded as a geometry, but Divide it into several small geometries for modeling; at the same time, when setting the material parameters, although the material of each small geometry is the same, the material number of each small geometry cannot be set to the same, and sequentially set to the material created in 1) Numbering; 3)：在多用途控制模块TRITON输出结果设置的物质编号设置中，加入1)中创建的材料编号；3): In the substance number setting of the multi-purpose control module TRITON output result setting, add the material number created in 1); 4)：完成上述步骤后，当多用途控制模块TRITON完成计算后，计算结果中就获得更多计算区域的计算结果；4): After completing the above steps, when the multi-purpose control module TRITON completes the calculation, the calculation results of more calculation areas are obtained in the calculation results; 步骤3：将多用途控制模块TRITON裂变产物及锕系产物的放射性活度计算结果输入点燃耗计算程序ORIGEN-S中计算核热推进反应堆全堆及各个计算区域内运行产物的光子能谱和中子能谱；点燃耗计算程序ORIGEN-S计算的伽马射线源强度和能谱包括X射线、伽马射线、韧致辐射、自发裂变伽马射线和伴随(α，n)反应的伽马射线产生的光子，核数据存储在点燃耗计算程序ORIGEN-S的二进制格式库中，通过式(3)计算出使用者直接指定的能量组结构的光子能谱：Step 3: Input the calculation results of the radioactivity of the multi-purpose control module TRITON fission products and actinide products into the ignition consumption calculation program ORIGEN-S to calculate the photon energy spectrum and neutralization of the entire nuclear thermal propulsion reactor and the operating products in each calculation area. Sub energy spectrum; gamma-ray source intensities and energy spectra calculated by the ignition consumption calculation program ORIGEN-S include X-rays, gamma rays, bremsstrahlung, spontaneous fission gamma rays and gamma rays accompanying (α, n) reactions The generated photons and nuclear data are stored in the binary format library of the ignition consumption calculation program ORIGEN-S, and the photon energy spectrum of the energy group structure directly specified by the user is calculated by formula (3): Ig＝Ia(Ea/Eg) (3)I g =I a (E a /E g ) (3) 式中in the formula Ia——伽马库中实际光子强度/photons·s-1 I a - the actual photon intensity in the gamma library/photons s -1 Ea——实际光子能量/MeVE a - actual photon energy/MeV Eg——能量组平均能量/MeVE g —— average energy of energy group/MeV Ig——能量组光子强度/photons·s-1 I g - energy group photon intensity/photons s -1 点燃耗计算程序ORIGEN-S计算的中子源强度和能谱包括自发裂变(α，n)反应和(β)反应，其中(β)反应在典型的乏燃料中并不重要；(α，n)反应则通过下式计算周围介质的影响：The neutron source intensities and energy spectra calculated by the ignition consumption calculation program ORIGEN-S include spontaneous fission (α,n) reactions and (β) reactions, where (β) reactions are not important in typical spent fuel; (α,n) ) reaction calculates the influence of the surrounding medium by the following formula:
Figure FDA0003334681210000041
Figure FDA0003334681210000041
 式中in the formula S(E)——介质总停止力/bS(E)——The total stopping force of the medium/b σi(E)——核素i的(α,n)反应截面/bσ i (E)——(α,n) reaction cross section of nuclide i/b E——中子能量/MeVE - neutron energy/MeV Yi,k——核素k发射的能量为Eα的α粒子的中子产额/neutrons·s-1 Y i,k ——the neutron yield of alpha particles with energy E α emitted by nuclide k/neutrons·s -1 N——总原子密度/atom·cm-3 N——Total atomic density/atom·cm -3 Ni——靶核素原子密度/atom·cm-3 Ni —— atomic density of target nuclide /atom·cm -3 步骤4：使用多维点核分析程序QADS建立核热推进反应堆几何模型，输入各计算区域的分布及中光子能谱，设定好计算参考点，计算出参考点的辐射剂量；针对典型的圆柱形反应堆，通过在轴向划分成小段圆柱形，每段都创建新的输入卡片，之后每个输入卡片中通过weight语句调整径向的放射性物质分布，如此实现对放射性物质分布的精确描述；多维点核分析程序QADS采用点核积分算法，其方程如下：Step 4: Use the multi-dimensional point nuclear analysis program QADS to establish the geometric model of the nuclear thermal propulsion reactor, input the distribution of each calculation area and the mesophoton energy spectrum, set the calculation reference point, and calculate the radiation dose at the reference point; for a typical cylindrical The reactor is divided into small cylindrical sections in the axial direction, and each section creates a new input card, and then adjusts the radial distribution of radioactive materials through the weight statement in each input card, so as to achieve an accurate description of the distribution of radioactive materials; multi-dimensional point The kernel analysis program QADS uses the point kernel integration algorithm, and its equation is as follows:
Figure FDA0003334681210000051
Figure FDA0003334681210000051
 式中in the formula
Figure FDA0003334681210000052
——计算点的位置/m
Figure FDA0003334681210000052
- Calculate the position of the point/m
Figure FDA0003334681210000053
——在体积元ν中的源的位置/m
Figure FDA0003334681210000053
- the position of the source in the volume element ν/m ν——源的体积/m3 ν - volume of source/m 3 μ——处于能量E的总衰减系数μ——the total attenuation coefficient at energy E
Figure FDA0003334681210000054
——源点和计算点之间的距离/m
Figure FDA0003334681210000054
- the distance between the source point and the calculation point/m K——通量-剂量转换系数K - flux-dose conversion factor B——剂量累计因子B - dose accumulation factor
Figure FDA0003334681210000055
——剂量率/rem·h-1
Figure FDA0003334681210000055
——Dose rate/rem·h -1 s——光子数量/photons·s-1 s——Number of photons/photons s -1 E——光子能量/MeVE - photon energy/MeV 步骤5：先选取一些参考点，在确定计算区域数量前只进行这些参考点的辐射剂量计算，选取的参考点需要在堆周围均匀分布且不能距离反应堆太远；完成第一轮计算后，按照步骤2中的拆分计算区域增加计算区域数量的方法增加计算区域数量，跳至步骤2进行新一轮计算，完成新一轮计算后比较上一轮和新一轮两次计算结果，若计算结果变化不在可接受范围内，则继续增加计算区域，重新计算，直至计算结果变化在可接受范围内；若计算结果变化在可接受范围内，说明计算区域选取足够多，能够模拟出放射性物质的不均匀分布，按照新一轮计算的计算区域设置进行计算，完成整个辐射场的计算。Step 5: Select some reference points first, and only calculate the radiation dose of these reference points before determining the number of calculation areas. The selected reference points should be evenly distributed around the reactor and should not be too far from the reactor; The method of splitting the calculation area in step 2 to increase the number of calculation areas increases the number of calculation areas, and skips to step 2 to perform a new round of calculation. After completing the new round of calculation, compare the calculation results of the previous round and the new round. If the change of the result is not within the acceptable range, continue to increase the calculation area and recalculate until the change of the calculation result is within the acceptable range; if the change of the calculation result is within the acceptable range, it means that the calculation area is selected enough to simulate the radioactive material. If the distribution is uneven, the calculation is performed according to the calculation area setting of the new round of calculation, and the calculation of the entire radiation field is completed.
CN202111296918.1A 2021-11-02 2021-11-02 Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor Active CN114003856B (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
CN202111296918.1A CN114003856B (en) 2021-11-02 2021-11-02 Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
CN202111296918.1A CN114003856B (en) 2021-11-02 2021-11-02 Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor

Publications (2)

Publication Number Publication Date
CN114003856A true CN114003856A (en) 2022-02-01
CN114003856B CN114003856B (en) 2022-08-05

Family

ID=79927152

Family Applications (1)

Application Number Title Priority Date Filing Date
CN202111296918.1A Active CN114003856B (en) 2021-11-02 2021-11-02 Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor

Country Status (1)

Country Link
CN (1) CN114003856B (en)

Cited By (2)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN114662041A (en) * 2022-02-28 2022-06-24 清华大学 Method for rapidly calculating core disk stock of short-life inert gas fission product
CN115270531A (en) * 2022-09-28 2022-11-01 国家电投集团科学技术研究院有限公司 Multi-radiation-source shielding calculation method and device, electronic equipment and storage medium

Citations (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN107122546A (en) * 2017-04-27 2017-09-01 西安交通大学 A kind of coupling of multiple physics method that presurized water reactor stable state is calculated
CN107194103A (en) * 2017-06-07 2017-09-22 哈尔滨工程大学 A kind of nuclear facilities model radiation field dosage emulation mode based on 3dsMax
CN107273582A (en) * 2017-05-23 2017-10-20 西安交通大学 A kind of computational methods for fast neutron reactor neutron transport burnup coupling analysis
CN109086506A (en) * 2018-07-20 2018-12-25 西安交通大学 A kind of burnup analysis calculation method suitable for liquid fuel molten salt reactor
CN109325282A (en) * 2018-09-13 2019-02-12 中国核动力研究设计院 A kind of reactor decommissioning three dimensional radiation field emulation mode and system
CN111414722A (en) * 2020-03-19 2020-07-14 西安交通大学 Simulation method for physical and thermal coupling of nuclear reactor core
CN112989651A (en) * 2021-02-06 2021-06-18 西安交通大学 Multi-physical field coupling method for reactor core
CN113494432A (en) * 2020-04-08 2021-10-12 中国科学院理化技术研究所 Nuclear heat thermoacoustic power generation system

Patent Citations (9)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN107122546A (en) * 2017-04-27 2017-09-01 西安交通大学 A kind of coupling of multiple physics method that presurized water reactor stable state is calculated
CN107122546B (en) * 2017-04-27 2020-06-26 西安交通大学 A Multiphysics Coupling Method for Steady-State Calculation of Pressurized Water Reactors
CN107273582A (en) * 2017-05-23 2017-10-20 西安交通大学 A kind of computational methods for fast neutron reactor neutron transport burnup coupling analysis
CN107194103A (en) * 2017-06-07 2017-09-22 哈尔滨工程大学 A kind of nuclear facilities model radiation field dosage emulation mode based on 3dsMax
CN109086506A (en) * 2018-07-20 2018-12-25 西安交通大学 A kind of burnup analysis calculation method suitable for liquid fuel molten salt reactor
CN109325282A (en) * 2018-09-13 2019-02-12 中国核动力研究设计院 A kind of reactor decommissioning three dimensional radiation field emulation mode and system
CN111414722A (en) * 2020-03-19 2020-07-14 西安交通大学 Simulation method for physical and thermal coupling of nuclear reactor core
CN113494432A (en) * 2020-04-08 2021-10-12 中国科学院理化技术研究所 Nuclear heat thermoacoustic power generation system
CN112989651A (en) * 2021-02-06 2021-06-18 西安交通大学 Multi-physical field coupling method for reactor core

Non-Patent Citations (3)

* Cited by examiner, † Cited by third party
Title
唐思邈等: "小型核电源传热及热电特性实验研究", 《核动力工程》 *
张磊等: "反应堆辐射屏蔽计算方法与程序概述", 《核电子学与探测技术》 *
段倩妮等: "余热排出系统中的热管设计及传热性能研究", 《原子能科学技术》 *

Cited By (4)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN114662041A (en) * 2022-02-28 2022-06-24 清华大学 Method for rapidly calculating core disk stock of short-life inert gas fission product
CN114662041B (en) * 2022-02-28 2024-10-18 清华大学 Method for rapidly calculating inventory of reactor core of short-service-life inert gas fission product
CN115270531A (en) * 2022-09-28 2022-11-01 国家电投集团科学技术研究院有限公司 Multi-radiation-source shielding calculation method and device, electronic equipment and storage medium
CN115270531B (en) * 2022-09-28 2022-12-16 国家电投集团科学技术研究院有限公司 Multi-radiation-source shielding calculation method and device, electronic equipment and storage medium

Also Published As

Publication number Publication date
CN114003856B (en) 2022-08-05

Similar Documents

Publication Publication Date Title
Wagner et al. Automated variance reduction of Monte Carlo shielding calculations using the discrete ordinates adjoint function
CN114003856B (en) Method for calculating environment radiation field in shutdown state of nuclear thermal propulsion reactor
He et al. Improved resonance calculation of fluoride salt-cooled high-temperature reactor based on subgroup method
Park et al. An improved DeCART library generation procedure with explicit resonance interference using continuous energy Monte Carlo calculation
Shultis et al. Radiation shielding
Rossi et al. Model design of a deuterium-deuterium neutron generator moderator and evaluation for delayed gamma-ray nondestructive assay for safeguards verification
Bourganel et al. Three-dimensional particle transport using Green’s functions in TRIPOLI-4 Monte Carlo code: application to PWR neutron fluence and ex-core response studies
Ziani et al. Eigenvalue sensitivity and nuclear data uncertainty analysis for the Moroccan TRIGA Mark II research reactor using SCALE6. 2 and MCNP6. 2
Rearden et al. Enhancements in continuous-energy Monte Carlo capabilities for SCALE 6.2
Ahmad et al. Evaluation of Reaction Cross Section of Radionuclide by Particles Induced Nuclear Reactions Using EXIFON Code
Sun et al. A study on the optimal position for the secondary neutron source in pressurized water reactors
Leconte et al. Feedback on 239Pu and 240Pu nuclear data and associated covariances through the CERES integral experiments
Xu et al. Development of the capability for multi-group shielding library generation in the nuclear data processing code NECP-Atlas
Koreshi Nuclear Engineering: Mathematical Modeling and Simulation
Wang et al. 3D SN and Monte Carlo calculations of the Utah TRIGA reactor core using PENTRAN and MCNP6
Yıldız An investigation of the effects of neutron energy-group structures and resonance treatment in a fusion–fission hybrid reactor fuelling with Tho2
Hernandez-Solis et al. Verification of the openmc homogenized myrrha-1.6 core model
Mumpower et al. Impact on Science and Technology
Chadwick et al. Reducing uncertainty in nuclear data
Ahmad et al. Plutonium build-up credits for a material test research reactor and influence of cross-section differences on actinide production
Wozniak From Michigan to Mars and Beyond
Timmons The Effects of a Power Iteration-Based K-Eigenvalue Solver for Various Subcritical Parameters and Calculations
Oprea et al. Mathematical modeling of neutron induced fission of 237Np nucleus
Talamo et al. Monte Carlo Analyses of the FOEHN Experiment
Cruz et al. A PHITS based-computational model of a TRIGA-fueled subcritical reactor for gamma dose mapping

Legal Events

Date Code Title Description
PB01 Publication
PB01 Publication
SE01 Entry into force of request for substantive examination
SE01 Entry into force of request for substantive examination
GR01 Patent grant
GR01 Patent grant