CN103365825B - A kind of reactor core active region cooling agent absolute neutron flux spectrum computational methods - Google Patents

Abstract

The present invention relates to the computational methods of pile neutron Flux Spectrum, be specifically related to a kind of reactor core active region cooling agent absolute neutron flux spectrum computational methods.The method uses the method that reactor core neutron transport program and Monte-Carlo code combine, reactor core calculation procedure is used to calculate the average flux absolute value of reactor core, calculated by Monte-Carlo code simulation and provide the Neutron flux distribution of self-defined energy group in cooling agent and the flux ratio of reactor core and cooling agent, finally can realize calculating the absolute neutron flux spectrum of reactor core active region cooling agent difference energy group distribution, improve tradition to use the value of average neutron flux to calculate reactor to be correlated with the computational methods of source item, improve computational accuracy.

Description

A kind of reactor core active region cooling agent absolute neutron flux spectrum computational methods

Technical field

The present invention relates to the computational methods of pile neutron Flux Spectrum, be specifically related to a kind of reactor core active region cooling agent absolute Neutron flux spectrum computational methods.

Background technology

Nuclear reactor is the device of a kind of energy released by Neutron chain reaction reaction, is that one is the hugest and multiple Miscellaneous system, in order to allow the operation of reactor safety economy, generally requires the distribution of neutron flux in knowing reaction heap.Reactor The distribution of neutron flux operationally is affected by factors different in heap, and the distribution therefore calculating its time and space is also Complex, in order to adapt to different demands, in reactor, the distribution of zones of different neutron flux can use different meters Calculation method obtains, and the computational accuracy that distinct methods obtains and the consumption to the time are different.In commonly used reactor core Son transports calculation procedure and can be given under different reactor core state faster, the absolute value of the neutron flux in reactor, this value one Mixed average flux beaten by Ban Shi reactor fuel district and cooling agent, and often only gives two groups or the distribution of few group, counts at some Calculate the Neutron flux distribution such as group few when source item calculates and be insufficient for required precision.Such as, in accurate estimation reactor core Product tritium amount, need to calculate the amount of 10B and neutron reaction in cooling agent, and the cross-section of 10B be in different Neng Qun districts Territory is different, very big in some reaction cross-section, Low Energy Region, and high energy region is relatively small, if therefore can provide the energy group relatively segmented The result that when neutron flux spectrum in region produces tritium amount for estimating, just neutron flux spectrum than few group obtains is the most accurate.

Summary of the invention

It is an object of the invention to provide a kind of reactor core active region cooling agent absolute neutron flux spectrum computational methods, the method will Reactor core neutron transport program and Monte-Carlo code combine, it is possible to obtain higher computational accuracy.

Technical scheme is as follows: a kind of reactor core active region cooling agent absolute neutron flux spectrum computational methods, the party Method uses reactor core neutron diffusion program to calculate the average neutron flux absolute value of reactor core, uses Monte-Carlo code simulation to calculate The Neutron flux distribution of self-defined energy group in the neutron flux ratio of cooling agent and reactor core and cooling agent, COMPREHENSIVE CALCULATING goes out reactor core and lives Property district cooling agent difference energy group distribution absolute neutron energy spectrum.

Further, reactor core active region as above cooling agent absolute neutron flux spectrum computational methods, wherein, concrete meter Calculation process is as follows:

Fuel assembly is divided into fuel pellet, involucrum, gap and four regions of cooling agent, uses Monte-Carlo code system The meter each region of fuel assembly is with the Neutron flux distribution of energy variation, the relative neutron flux of cooling agent region each energy group Being calculated by Monte-Carlo code, total relative neutron flux in this region is calculated by formula (4), and four regions are total Average relative neutron flux is calculated by the volume weighted in each region according to formula (5)；Formula (4) and formula (5) are obtained Result and by the average neutron flux absolute value Φ of the reactor core calculated reactor core of neutron diffusion program^absSubstitute into formula (2) In, can be cooled the absolute value of neutron flux in agent；The share shared by neutron flux of cooling agent respectively energy group is by formula (3) calculate, then, by the absolute neutron flux of energy group each in the available cooling agent of formula (1)

${\mathrm{\Φ}}_{g,H2O}^{\mathrm{abs}}={\mathrm{\Φ}}_{H2O}^{\mathrm{abs}}\·f_g---\left(1\right)$

${\mathrm{\Φ}}_{H2O}^{\mathrm{abs}}={\mathrm{\Φ}}^{\mathrm{abs}}\·\frac{{\mathrm{\Φ}}_{H2O}^{\mathrm{rel}}}{\stackrel{\‾}{{\mathrm{\Φ}}^{\mathrm{rel}}}}---\left(2\right)$

$f_g={\mathrm{\Φ}}_{g,H2O}^{\mathrm{rel}}/{\mathrm{\Φ}}_{H2O}^{\mathrm{rel}}---\left(3\right)$

${\mathrm{\Φ}}_{H2O}^{\mathrm{rel}}=\underset{g=1}{\overset{G}{\mathrm{\Σ}}}{\mathrm{\Φ}}_{g,H2O}^{\mathrm{rel}}---\left(4\right)$

$\stackrel{\‾}{{\mathrm{\Φ}}^{\mathrm{rel}}}=\underset{m=1}{\overset{M}{\mathrm{\Σ}}}{\mathrm{\Φ}}_m^{\mathrm{rel}}{V}_m/\underset{m=1}{\overset{M}{\mathrm{\Σ}}}{V}_m---\left(5\right)$

In formula:

The absolute neutron flux of each energy group in cooling agent；

Neutron flux absolute value in cooling agent；

f_g, the share shared by neutron flux of each energy group in cooling agent；

Φ^abs, the average neutron flux absolute value of reactor core, calculate for reactor core neutron diffusion program and get；

Relative neutron flux in cooling agent in lattice cell；

The relative neutron flux that lattice cell is average；

The relative neutron flux of the difference energy group in cooling agent, is calculated by Monte-Carlo code；

G represents the quantity of energy group, and according to difference and the requirement of subsequent applications of object to be analyzed, can arbitrarily divide can group Interval；

The relative neutron flux of the lattice cell of zones of different；

V_m, the volume of the lattice cell in zones of different；

M represents the region quantity of division.

Beneficial effects of the present invention is as follows: the present invention uses reactor core neutron transport program and Monte-Carlo code to combine Method, uses reactor core calculation procedure to calculate the average flux absolute value of reactor core, Monte-Carlo code simulation calculating provide cooling In agent and the flux ratio of reactor core and cooling agent self-defined can the Neutron flux distribution of group, finally can realize calculating reactor core active region The absolute neutron flux spectrum of cooling agent difference energy group distribution, improves tradition and uses the value of average neutron flux to calculate reactor phase Close the computational methods of source item, improve computational accuracy.

Accompanying drawing explanation

Fig. 1 is ACP1000 unit balanced recycle reactor core burnup profile schematic diagram (beginning of life)；

Fig. 2 is ACP1000 unit balanced recycle longevity 47 groups of Neutron flux distribution figures of interim cooling agent；

Fig. 3 is ACP1000 unit balanced recycle longevity 16 groups of Neutron flux distribution figures of interim cooling agent.

Detailed description of the invention

With embodiment, the present invention is described in detail below in conjunction with the accompanying drawings.

As a example by ACP1000 unit is composed by the cooling agent absolute neutron flux carrying out being provided when source item calculates.This unit Heap in-core have 177 assemblies, first carry out the calculating of reactor core program, by the result of calculation of reactor core calculation procedure understand certain balance Two groups of neutron flux absolute values in circulatory life-time, now boron concentration, coolant density and temperature, fuel temperature and one in reactor core The conditions such as the nucleic density determining assembly under burnup are also knowable.Monte-Carlo code Simulation Core is used to carry out three-dimensional computations, Calculating can use the reactor core of simulation whole 177 assemblies composition to calculate, it is possible to uses the fuel assembly model more simplified Calculate, as a example by being used herein as fuel assembly model.

Assembly average burn-up distribution map that this equilibrium core longevity is interim is as it is shown in figure 1, by the core of all component volume weighted Element density is as the nucleic density of component model, and nucleic kind considers actinides and part fission product, and actinides is examined Consider following nucleic:²³⁴U、²³⁵U、²³⁶U、²³⁸U、²³⁸Pu、²³⁹Pu、²⁴⁰Pu、²⁴¹Pu、²⁴²Pu, part fission product considers:²⁴¹Am、^242mAm、²⁴³Am、²⁴²Cm、²⁴³Cm、²⁴⁴Cm、¹³⁵I、¹³⁵Xe、¹⁴⁷Pm、¹⁴⁸Pm、¹⁴⁹Pm、¹⁴⁹Sm.In order to make component model more Good Simulation Core, employs the ginsengs such as the running status provided in reactor core program coolant density described above, temperature, boron concentration Number, assembly is divided into fuel pellet, involucrum, gap and cooling agent are divided into four regions, the modeling figure of program as shown in Figure 2, Use the Neutron flux distribution covering the card program statistics above-mentioned each district of fuel assembly with energy variation, obtain in cooling agent with energy quantitative change The relative Neutron flux distribution changed.The absolute neutron flux of each energy group in estimation cooling agentTime, by covering card program statistics The relative Neutron flux distribution of regional difference energy groupAbsolute with the reactor core average flux that reactor core calculation procedure is given Value Φ^abs, the absolute flux in each energy group's Main Coolant can be extrapolated according to equation below.

${\mathrm{\Φ}}_{g,H2O}^{\mathrm{abs}}={\mathrm{\Φ}}_{H2O}^{\mathrm{abs}}\·f_g---\left(1\right)$

${\mathrm{\Φ}}_{H2O}^{\mathrm{abs}}={\mathrm{\Φ}}^{\mathrm{abs}}\·\frac{{\mathrm{\Φ}}_{H2O}^{\mathrm{rel}}}{\stackrel{\‾}{{\mathrm{\Φ}}^{\mathrm{rel}}}}---\left(2\right)$

$f_g={\mathrm{\Φ}}_{g,H2O}^{\mathrm{rel}}/{\mathrm{\Φ}}_{H2O}^{\mathrm{rel}}---\left(3\right)$

${\mathrm{\Φ}}_{H2O}^{\mathrm{rel}}=\underset{g=1}{\overset{G}{\mathrm{\Σ}}}{\mathrm{\Φ}}_{g,H2O}^{\mathrm{rel}}---\left(4\right)$

$\stackrel{\‾}{{\mathrm{\Φ}}^{\mathrm{rel}}}=\underset{m=1}{\overset{M}{\mathrm{\Σ}}}{\mathrm{\Φ}}_m^{\mathrm{rel}}{V}_m/\underset{m=1}{\overset{M}{\mathrm{\Σ}}}{V}_m---\left(5\right)$

In formula:

The absolute neutron flux of each energy group in cooling agent；

Neutron flux absolute value in cooling agent；

f_g, the share shared by neutron flux of each energy group in cooling agent；

Φ^abs, the average neutron flux absolute value of reactor core, calculate for reactor core neutron diffusion program and get；

Relative neutron flux in cooling agent in lattice cell；

The relative neutron flux that lattice cell is average；

The relative neutron flux of the difference energy group in cooling agent, is calculated by Monte-Carlo code；

G represents the quantity of energy group, and according to difference and the requirement of subsequent applications of object to be analyzed, can arbitrarily divide can group Interval；

The relative neutron flux of the lattice cell of zones of different；

V_m, the volume of the lattice cell in zones of different；

M represents the region quantity of division.

Concrete calculating process is as follows: fuel assembly is divided into fuel pellet, involucrum, gap and four districts of cooling agent Territory, uses the Monte-Carlo code statistics each region of fuel assembly with the Neutron flux distribution of energy variation, each energy in cooling agent region The relative neutron flux of groupBeing calculated by Monte-Carlo code, total relative neutron flux in this region is by formula (4) Calculate, the total average relative neutron flux in four regions according to formula (5) by the volume weighted calculating in each region；Will Result that formula (4) and formula (5) obtain and by the average neutron flux of the reactor core calculated reactor core of neutron diffusion program Absolute value Φ^absSubstitute in formula (2), can be cooled the absolute value of neutron flux in agent；The neutron of cooling agent respectively energy group Share shared by flux is calculated by formula (3), then, by the absolute neutron of energy group each in the available cooling agent of formula (1) Flux

Above-mentioned reactor core neutron diffusion program, Monte-Carlo code are techniques known.

In covering card program, user self-defined group can be distributed the statistical counting carrying out neutron flux, in order to adapt to source The different demands that item calculates, it is possible to use different can carry out counting statistics in group interval, if Fig. 2 is by using the present invention to be calculated In ACP1000 unit cooling agent, the distribution map of 47 groups of neutron flux, can be used for¹⁴The estimation of C generation amount, Fig. 3 is for using the present invention The distribution map of calculated 16 groups of neutron flux, can be used for producing the estimation of tritium amount.

Obviously, those skilled in the art can carry out various change and the modification essence without deviating from the present invention to the present invention God and scope.So, if these amendments and modification to the present invention belong to the model of the claims in the present invention and equivalent technology thereof Within enclosing, then the present invention is also intended to comprise these change and modification.

Claims (1)

1. reactor core active region cooling agent absolute neutron flux spectrum computational methods, it is characterised in that: the method uses in reactor core Sub-diffusion process calculates the average neutron flux absolute value of reactor core, uses Monte-Carlo code simulation to calculate cooling agent and reactor core Neutron flux ratio and cooling agent in self-defined can the Neutron flux distribution of group, COMPREHENSIVE CALCULATING goes out reactor core active region cooling agent not With the absolute neutron energy spectrum of energy group's distribution, concrete calculating process is as follows:

Fuel assembly is divided into fuel pellet, involucrum, gap and four regions of cooling agent, uses Monte-Carlo code statistics combustion The material each region of assembly is with the Neutron flux distribution of energy variation, the relative neutron flux of cooling agent region each energy groupBy covering Special Caro program is calculated, and total relative neutron flux in this region is calculated by formula (4), total average in four regions Neutron flux relatively is calculated by the volume weighted in each region according to formula (5)；The knot that formula (4) and formula (5) are obtained Fruit and by the average neutron flux absolute value Φ of the reactor core calculated reactor core of neutron diffusion program^absSubstitute in formula (2), Can be cooled the absolute value of neutron flux in agent；The share shared by neutron flux of cooling agent respectively energy group is by formula (3) Calculate, then, by the absolute neutron flux of energy group each in the available cooling agent of formula (1)

${\mathrm{\Φ}}_{g,H2O}^{abs}={\mathrm{\Φ}}_{H2O}^{abs}\·f_g---\left(1\right)$

${\mathrm{\Φ}}_{H2O}^{abs}={\mathrm{\Φ}}^{abs}\·\frac{{\mathrm{\Φ}}_{H2O}^{rel}}{\stackrel{\‾}{{\mathrm{\Φ}}^{rel}}}---\left(2\right)$

$f_g={\mathrm{\Φ}}_{g,H2O}^{rel}/{\mathrm{\Φ}}_{H2O}^{rel}---\left(3\right)$

${\mathrm{\Φ}}_{H2O}^{rel}=\underset{g=1}{\overset{G}{\Σ}}{\mathrm{\Φ}}_{g,H2O}^{rel}---\left(4\right)$

$\stackrel{\‾}{{\mathrm{\Φ}}^{rel}}=\underset{m=1}{\overset{M}{\Σ}}{\mathrm{\Φ}}_m^{rel}{V}_m/\underset{m=1}{\overset{M}{\Σ}}{V}_m---\left(5\right)$

In formula:

The absolute neutron flux of each energy group in cooling agent；

Neutron flux absolute value in cooling agent；

f_g, the share shared by neutron flux of each energy group in cooling agent；

Φ^abs, the average neutron flux absolute value of reactor core, calculate for reactor core neutron diffusion program and get；

Relative neutron flux in cooling agent in lattice cell；

The relative neutron flux that lattice cell is average；

The relative neutron flux of the difference energy group in cooling agent, is calculated by Monte-Carlo code；

G represents the quantity of energy group, according to difference and the requirement of subsequent applications of object to be analyzed, can arbitrarily divide between energy group Every；

The relative neutron flux of the lattice cell of zones of different；

V_m, the volume of the lattice cell in zones of different；

M represents the region quantity of division.