CN105810254B - Lack the Restart Method and system in group cross-section library for making reactor nuclear design - Google Patents
Lack the Restart Method and system in group cross-section library for making reactor nuclear design Download PDFInfo
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- CN105810254B CN105810254B CN201410840816.5A CN201410840816A CN105810254B CN 105810254 B CN105810254 B CN 105810254B CN 201410840816 A CN201410840816 A CN 201410840816A CN 105810254 B CN105810254 B CN 105810254B
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- 238000000034 method Methods 0.000 title claims abstract description 53
- 238000013461 design Methods 0.000 title claims abstract description 27
- 239000000446 fuel Substances 0.000 claims abstract description 149
- ZOXJGFHDIHLPTG-UHFFFAOYSA-N Boron Chemical compound [B] ZOXJGFHDIHLPTG-UHFFFAOYSA-N 0.000 claims description 27
- 229910052796 boron Inorganic materials 0.000 claims description 27
- 239000002574 poison Substances 0.000 claims description 22
- 231100000614 poison Toxicity 0.000 claims description 22
- 230000008859 change Effects 0.000 claims description 19
- 230000000704 physical effect Effects 0.000 claims description 6
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 6
- 230000008569 process Effects 0.000 claims description 5
- 238000003780 insertion Methods 0.000 claims description 3
- 230000037431 insertion Effects 0.000 claims description 3
- 230000032258 transport Effects 0.000 description 22
- 238000004364 calculation method Methods 0.000 description 13
- 238000012545 processing Methods 0.000 description 8
- 238000002485 combustion reaction Methods 0.000 description 7
- 230000000694 effects Effects 0.000 description 7
- 238000004891 communication Methods 0.000 description 4
- 238000004519 manufacturing process Methods 0.000 description 4
- 238000003860 storage Methods 0.000 description 4
- 239000003795 chemical substances by application Substances 0.000 description 3
- 238000010586 diagram Methods 0.000 description 3
- 239000000463 material Substances 0.000 description 3
- 238000004458 analytical method Methods 0.000 description 2
- 241000208340 Araliaceae Species 0.000 description 1
- 235000005035 Panax pseudoginseng ssp. pseudoginseng Nutrition 0.000 description 1
- 235000003140 Panax quinquefolius Nutrition 0.000 description 1
- 230000000712 assembly Effects 0.000 description 1
- 238000000429 assembly Methods 0.000 description 1
- 230000008901 benefit Effects 0.000 description 1
- 229910002056 binary alloy Inorganic materials 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 238000005520 cutting process Methods 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 238000011156 evaluation Methods 0.000 description 1
- 238000010304 firing Methods 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 235000008434 ginseng Nutrition 0.000 description 1
- 238000000265 homogenisation Methods 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 238000004321 preservation Methods 0.000 description 1
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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Abstract
Present disclose provides a kind of Restart Methods and system for lacking group cross-section library for making reactor nuclear design.This method comprises: (a) obtains the burn-up level of fuel assembly；(b) variation of one or more parameters of the fuel assembly under the burn-up level is obtained；(c) parameter pretreatment is executed for the variation of one or more of parameters；And it is (d) pretreated based on the parameter as a result, neutron transport calculating is executed, to obtain few group cross-section.The system includes: burn-up level acquiring unit, for obtaining the burn-up level of fuel assembly；Fuel parameter acquiring unit, for obtaining the variation of one or more parameters of the fuel assembly under the burn-up level；Parameter pretreatment unit executes parameter pretreatment for the variation for one or more of parameters；And few group cross-section acquiring unit, it is calculated for pretreated based on the parameter as a result, executing neutron transport, to obtain few group cross-section.
Description
Technical field
This disclosure relates to which reactor nuclear design field, more particularly relates to production reactor nuclear design and lacks group cross-section library
Restart Method and system.
Background technique
In the use process of nuclear reactor, after the burn-up level of fuel assembly reaches a certain level, need to its into
Row replacement.However, due to the fuel assembly that there are a large amount of (for example, 157) in reactor, and due to the difference of burn-up level,
These fuel assemblies will not be replaced all simultaneously.Therefore, in refuelling component, inevitably by new fuel assembly
The fuel assembly of burn-up levels different from having reached is used together.At this time, it may be necessary to according to the relevant parameter (example of each fuel assembly
Such as, burn-up level, average enrichment, burnable poison rod parameter etc.) design the layout of fuel assembly.In addition, in order to react
Guarantee that the fuel assembly of reactor is in the critical state of safety and stability during stack operation, it is also desirable to core be carried out to reactor and set
Meter, to prevent unstable emergent power, overheat, spend situations such as slowing down.Therefore, in the application of nuclear reactor, nuclear design system
It is all essential.
Currently, the wider nuclear design system of use scope all uses two-step method in the world, i.e., obtained first using component programs
Few group cross-section data of (such as: power, fuel temperature, moderator temperature, concentration of soluble boron etc.) component under different operating conditions, and
Few group cross-section library is formed, then reactor core calculation procedure carries out core physics performance evaluation using few group cross-section library.
In few group cross-section data of fuel assembly under component programs obtain different operating conditions, each nuclear design system in the world
It is all developed according to respective few group cross-section library production demand and restarts calculation method.For example, the APA core of US Westinghouse company is set
Meter systems, CASMO-SIMULATE nuclear design system of Studsvik company etc..But since different IPs designing system lacks group cross-section
The difference of library Consideration, there is certain differences for the function of restarting in component programs.With in nuclear design system
Few group cross-section library needs the understanding of Consideration constantly to deepen, and the continuous of the requirement for nuclear design program computational accuracy mentions
Height, on the basis of existing technology, more comprehensively, deep consideration influence few group cross-section under computation module difference operating condition because
Element, and realized on restarting function, it becomes more and more important.
Summary of the invention
To solve the above-mentioned problems, it provides and group cross-section is lacked for making reactor nuclear design according to the embodiment of the present disclosure
The Restart Method and system in library.
A kind of lack restarting for group cross-section library for making reactor nuclear design in a first aspect, providing according to the disclosure
Method.This method comprises: (a) obtains the burn-up level of fuel assembly；(b) fuel assembly is obtained under the burn-up level
One or more parameters variation；(c) parameter pretreatment is executed for the variation of one or more of parameters；And
(d) pretreated based on the parameter to be calculated as a result, executing neutron transport, to obtain few group cross-section.
In some embodiments, one or more of parameters include at least one of the following: power, concentration of soluble boron, combustion
Material temperature degree, moderator temperature and control rod are inserted into/propose reactor core active region.
In some embodiments, if one or more of parameters include power, the parameter pretreatment includes: to obtain
Take relevant information of the fuel assembly under the burn-up level；Variation based on the relevant information and the power is come really
The variation of the fuel temperature, moderator temperature and moderator-density of the fixed fuel assembly；And based on the variation come again
Obtain nucleic microscopic cross and nucleon density.
In some embodiments, if one or more of parameters include fuel temperature, the parameter pretreatment packet
It includes: obtaining relevant information of the fuel assembly under the burn-up level；Based on the relevant information and the fuel temperature
Variation come determine the fuel assembly microscopic cross variation；And microcosmic section of nucleic is reacquired based on the variation
Face and nucleon density.
In some embodiments, if one or more of parameters include moderator temperature, the parameter pretreatment
It include: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the relevant information and the moderator
The variation for changing to determine the moderator-density of the fuel assembly of temperature；And nucleic is reacquired based on the variation
Microscopic cross and nucleon density.
In some embodiments, if one or more of parameters include that control rod is inserted into/proposes reactor core active region,
The parameter pretreatment includes: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the related letter
Breath and the control rod are inserted into/propose the variation of reactor core active region to determine the variation of the geometry of the fuel assembly；And base
Nucleic microscopic cross and nucleon density are reacquired in the variation.
In some embodiments, the relevant information includes at least nucleon of the fuel assembly under the burn-up level
Density.
In some embodiments, step (d) further include: pretreated based on the parameter as a result, executing neutron transport meter
It calculates and burnup calculates, to obtain few group cross-section under the burn-up level and under subsequent burn-up level.
In some embodiments, one or more if to obtain few group cross-section under the subsequent burn-up level
A parameter can also include that burnable poison rod group proposes reactor core active region.
In some embodiments, if one or more of parameters include concentration of soluble boron, the parameter pretreatment
It include: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the relevant information and the soluble boron
The variation for changing to determine the nucleic relevant information of the fuel assembly of concentration；And core is reacquired based on the variation
Plain microscopic cross and nucleon density.
In some embodiments, the nucleic relevant information includes at least information relevant to B-10 and B-11.
In some embodiments, if one or more of parameters include that burnable poison rod group proposes reactor core active region,
Then the parameter pretreatment includes: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the correlation
Information and the burnable poison rod group propose the variation of reactor core active region to determine the variation of the geometry of the fuel assembly；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
In some embodiments, described that the combustion is determined based on the variation of the relevant information and the moderator temperature
The step of expecting the variation of the moderator-density of component includes: the variation based on the moderator temperature, passes through interpolation thermal-hydraulic
Physical property table obtains the variation of moderator-density.
In some embodiments, it before the neutron transport of step (d) calculates, also executes resonance and calculates.
In some embodiments, reactor is presurized water reactor.
According to the second aspect of the disclosure, provides and a kind of lack restarting for group cross-section library for making reactor nuclear design
System.The system includes: burn-up level acquiring unit, for obtaining the burn-up level of fuel assembly；Fuel parameter acquiring unit,
For obtaining the variation of one or more parameters of the fuel assembly under the burn-up level；Parameter pretreatment unit is used
It is pre-processed in the variation for one or more of parameters to execute parameter；And few group cross-section acquiring unit, for being based on
The parameter is pretreated to be calculated as a result, executing neutron transport, to obtain few group cross-section.
In some embodiments, one or more of parameters include at least one of the following: power, concentration of soluble boron, combustion
Material temperature degree, moderator temperature and control rod are inserted into/propose reactor core active region.
In some embodiments, if one or more of parameters include power, the parameter pretreatment includes: to obtain
Take relevant information of the fuel assembly under the burn-up level；Variation based on the relevant information and the power is come really
The variation of the fuel temperature, moderator temperature and moderator-density of the fixed fuel assembly；And based on the variation come again
Obtain nucleic microscopic cross and nucleon density.
In some embodiments, if one or more of parameters include fuel temperature, the parameter pretreatment packet
It includes: obtaining relevant information of the fuel assembly under the burn-up level；Based on the relevant information and the fuel temperature
Variation come determine the fuel assembly microscopic cross variation；And microcosmic section of nucleic is reacquired based on the variation
Face and nucleon density.
In some embodiments, if one or more of parameters include moderator temperature, the parameter pretreatment
It include: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the relevant information and the moderator
The variation for changing to determine the moderator-density of the fuel assembly of temperature；And nucleic is reacquired based on the variation
Microscopic cross and nucleon density.
In some embodiments, if one or more of parameters include that control rod is inserted into/proposes reactor core active region,
The parameter pretreatment includes: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the related letter
Breath and the control rod are inserted into/propose the variation of reactor core active region to determine the variation of the geometry of the fuel assembly；And base
Nucleic microscopic cross and nucleon density are reacquired in the variation.
In some embodiments, the relevant information includes at least nucleon of the fuel assembly under the burn-up level
Density.
In some embodiments, few group cross-section acquiring unit is also used to: pretreated based on the parameter as a result, holding
Row neutron transport calculating and burnup calculate, to obtain few group cross-section under the burn-up level and under subsequent burn-up level.
In some embodiments, one or more if to obtain few group cross-section under the subsequent burn-up level
A parameter can also include: that burnable poison rod group proposes reactor core active region.
In some embodiments, if one or more of parameters include concentration of soluble boron, the parameter pretreatment
It include: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the relevant information and the soluble boron
The variation for changing to determine the nucleic relevant information of the fuel assembly of concentration；And core is reacquired based on the variation
Plain microscopic cross and nucleon density.
In some embodiments, the nucleic relevant information includes at least information relevant to B-10 and B-11.
In some embodiments, if one or more of parameters include that burnable poison rod group proposes reactor core active region,
Then the parameter pretreatment includes: the relevant information for obtaining the fuel assembly under the burn-up level；Based on the correlation
Information and the burnable poison rod group propose the variation of reactor core active region to determine the variation of the geometry of the fuel assembly；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
In some embodiments, described that the combustion is determined based on the variation of the relevant information and the moderator temperature
The step of expecting the variation of the moderator-density of component includes: the variation based on the moderator temperature, passes through interpolation thermal-hydraulic
Physical property table obtains the variation of moderator-density.
In some embodiments, few group cross-section acquiring unit also executes the meter that resonates before son transports calculating in commission
It calculates.
In some embodiments, reactor is presurized water reactor.
By using disclosed method and system, not only consider that power itself changes to few group in changed power calculating
The influence that parameter calculates also calculates fuel temperature caused by the changed power of fuel assembly by fuel effective temperature computing module
The variation of degree, moderator temperature, and the slowing down due to caused by moderator temperature change is obtained by interpolation thermal-hydraulic physical property table
The variation of agent density.In addition, not only considering that moderator temperature itself calculates Few group parameter in the calculating of moderator temperature change
Influence, it is also contemplated that the variation of moderator-density caused by the moderator temperature change of fuel assembly.That is the disclosure
Not only allow for direct effect caused by factor changes, it is also contemplated that second order effect caused by the factor changes.To consider certain
The influence that a factor generates is more comprehensively, accurately.In addition, the disclosure considers fuel stack in branch calculates in terms of effect
Power, fuel temperature, moderator temperature, concentration of soluble boron, the control rod of part are inserted into/propose the shadow of the factors such as reactor core active region
It rings, in afterflame consumption calculates, it is contemplated that the power of fuel assembly, fuel temperature, moderator temperature, concentration of soluble boron, control rod
Insertion/proposition reactor core active region, burnable poison rod propose the influence of the factors history effects such as reactor core active region, and Consideration can
Fully consider that fuel assembly in reactor core lacks the influence of group cross-section various aspects.It, can also basis meanwhile when making few group cross-section library
Other core physics analyze the needs of program, arbitrarily select above-mentioned Consideration, easy to use, flexible.
Detailed description of the invention
By illustrating preferred embodiment of the present disclosure with reference to the accompanying drawing, above and other purpose, the spy of the disclosure will be made
Advantage of seeking peace is clearer, in which:
Fig. 1 is the component programs for lacking group cross-section library for making reactor nuclear design shown according to the embodiment of the present disclosure
The example flow diagram of Restart Method.
Fig. 2 is to show the block diagram of the example system for executing method shown in Fig. 1 according to the embodiment of the present disclosure.
Specific embodiment
Preferred embodiment of the present disclosure is described in detail with reference to the accompanying drawings, is omitted in the course of the description for this
It is unnecessary details and function for open, to prevent understanding of this disclosure from causing to obscure.
Reactor core nuclear design will be substantially introduced first.The target of Nuclear design is mainly to during reactor operation
Reactor core neutron behavior is accurately predicted.What usual neutron gauges were carried out based on Nuclear Data at last, in the Nuclear Data, for
Various nucleic are stored with details relevant to reaction probability (that is, the section) of neutron and target nucleus.Cutting in original Nuclear Data
Face has very delicate energy resolved degree, which is enough to reproduce extremely complex neutron behavior.For fission reactor
For used heavy isotope, it there may come a time when to reach hundreds of thousands for accurately reproducing the number of energy lattice point of original section.
Theoretically, most accurate result can be provided to the direct utilization in this section indicated with sophisticated energy.So
And the calculation amount that this neutron transport calculates is (for example, 20 ten thousand to 30 in the whole three-dimensional reactor core geometry of presurized water reactor (PWR)
In the case where ten thousand energy groups) it the use of current most advanced, most powerful computer is also to be difficult to complete within the reasonable time.When
So, if using magnanimity parallel super computer, it is final it is also possible that realizing.But consider from cost-benefit angle,
It will not usually do so.
Therefore, for this respect the considerations of, it will usually approximation be carried out to it by the way of approximate solution.In order to reduce energy
The quantity of group has used multigroup (for example, more than 25 energy groups), in multigroup, to the energy in the section in particular energy section
Dependence is averaged (homogenization).Then, by several approximating steps will the number of group gradually decrease at least group (for example, 2
~4 energy groups).
It is thus possible to by above-mentioned approximate way using few group cross-section library come the middle sub-line in approximate representation (prediction) reactor core
For, and group cross-section library can be lacked based on this and carry out subsequent required nuclear design.For example, required and each group when fuel assembly is replaced
The related assembly layout of part burn-up level, thermal hydraulic analysis, safety analysis etc..
In some embodiments, Nuclear design calculating can be divided into two steps.The first step, component programs are according to fuel assembly
The various operating conditions (for example, power, boron concentration, fuel temperature, moderator temperature etc.) faced in reactor core, calculate under different operating conditions
Component physical characteristic parameter.Few group that component characteristic parameter combination under various difference operating conditions just forms the component together cuts
Face library.Second step, reactor core program can be lacked in the library of group cross-section according to component in the actual condition interpolation in reactor core obtains respective sets
Part characterisitic parameter.
In the first step, component programs calculate first is fired under given a certain operating condition (commonly referred to as base operation condition)
Consumption calculates, and then under base operation condition on a certain burnup point, changes a certain duty parameter (for example, power, boron concentration, fuel temperature
Degree, moderator temperature etc.), the physical characteristic parameter for carrying out the burnup point calculates the spy of (branch's calculating) and/or subsequent burnup point
Property parameter calculate (afterflame consumption calculate).Change on the basis of base operation condition some or multiple duty parameters (for example, power,
Boron concentration, fuel temperature, moderator temperature etc.) the calculation method of component physical characteristic parameter be called Restart Method.
In addition, in component programs calculating, it first can be by the information preservations such as nucleon density, power under base operation condition to two
In binary data file, the binary file then is read when the branch introduced in detail below calculates or afterflame consumption calculates
Information changes certain parameter on the basis of this document information.Therefore, which also referred to as restarts binary system
File.
Hereinafter, the disclosure is described in detail for the disclosure is applied to the scene of pressurized water reactor.But this
Open to be not limited thereto, the disclosure also can be applied to other types of reactor or other kinds of reactor fuel group
Part, as long as these reactors and/or fuel assembly need to carry out similar nuclear design work.
It generally can wrap according to the Restart Method that the production reactor nuclear design of the embodiment of the present disclosure lacks group cross-section library
Include branch's calculation method and/or afterflame consumption calculation method.
Branch's calculation method
Branch's calculation method refers to when fuel assembly changes a certain parameter under a certain burn-up level, according to disclosure reality
Apply example branch's computing function can be re-started under the burn-up level parameter pretreatment, then carry out resonance calculating and/or
Neutron transport calculates, to obtain few group cross-section.Specifically, which can carry out for the variation of following various parameters
Parameter pretreatment, resonance calculate and/or neutron transport calculates etc..
(1) changed power branch calculates:
Restart Method is required to meet the needs that simulated fuel assembly changes power under a certain burn-up level.Work as function
When rate changes, the related letter such as the nucleon density of the component under corresponding burn-up level can be read in binary file from restarting
Breath, and the ginseng such as fuel temperature, moderator temperature, moderator-density caused by changed power is calculated by thermal-hydraulic computing module
Several variations, and the processing of relevant parameter is re-started according to the variation of above-mentioned parameter.Then, carry out resonance calculating and/or
Neutron-transport equation solves, to obtain few group cross-section.The process only carries out under a certain burn-up level, but without burnup meter
It calculates.
(2) concentration of soluble boron changes branch's calculation method:
Restart Method is required to meet simulated fuel assembly in the needs of a certain burn-up level variation concentration of soluble boron.
It, can that nucleon of the component under corresponding burn-up level is read in binary file be close from restarting when concentration of soluble boron variation
The relevant informations such as degree, and the nucleic such as B-10, B-11 relevant information is handled again.Then, carry out resonance calculating and/or in
Sub- transport equation calculates, and obtains few group cross-section, but calculate without burnup.
(3) fuel temperature variation branch calculates:
Restart Method is required to meet the needs that simulated fuel assembly changes fuel temperature under a certain burn-up level.
It, can be from restarting the nucleon density for reading the component in binary file under corresponding burn-up level when fuel temperature variation
Etc. relevant informations, and consider the variation of microscopic cross caused by the variation of fuel temperature.Then, re-start resonance calculate and/
Or neutron transport calculates, and obtains few group cross-section.The process only carries out under a certain burn-up level, but calculates without burnup.
(4) moderator temperature change branch calculates:
Restart Method is required to meet the need that simulated fuel assembly changes moderator temperature under a certain burn-up level
It wants.It, can be from restarting the core for reading the component in binary file under corresponding burn-up level when moderator temperature change
The relevant informations such as sub- density, and consider the variation of moderator-density caused by the variation of moderator temperature, according to above-mentioned parameter
Variation re-starts parameter processing.Then, it carries out resonance calculating and/or neutron-transport equation calculates, obtain few group cross-section.The mistake
Journey only carries out under a certain burn-up level, but calculates without burnup.
(5) control rod is inserted into/proposes the calculating of reactor core active region branch:
Restart Method be required to meet simulated fuel assembly control rod under a certain burn-up level be inserted into/propose reactor core
Branch calculate needs.When control rod is inserted into/propose reactor core when, can read the component in file from restarting and accordingly firing
The relevant informations such as the nucleon density under depth are consumed, and several how relevant informations are handled again.Then, resonance calculating is carried out
And/or neutron-transport equation calculates, and obtains few group cross-section.The process only carries out under a certain burn-up level, but without burnup
It calculates.
Afterflame consumes calculation method
Afterflame consumption calculation method refers to when fuel assembly changes a certain parameter under a certain burn-up level, according to the disclosure
The afterflame consumption computing function of embodiment can re-start parameter pretreatment under the burn-up level, then carry out resonance calculating
And/or neutron transport calculating and burnup calculate, to obtain few group cross-section under the burn-up level and subsequent burn-up level.Specifically
Ground, it is defeated that afterflame consumption computing function can carry out parameter pretreatment, resonance calculating and/or neutron for the variation of following various parameters
Fortune calculating etc..
(1) changed power afterflame consumes calculation method:
Restart Method is required to meet the subsequent burnup that simulated fuel assembly changes power under a certain burn-up level
The needs of calculating.When changed power, the component can be read in file in the nucleon density of corresponding burn-up level from restarting
Etc. relevant informations, and by thermal-hydraulic computing module calculate changed power caused by fuel temperature, moderator temperature, moderator
The variation of the parameters such as density re-starts parameter processing according to the variation of above-mentioned parameter.Then, carry out resonance calculating and/or in
Sub- transport equation solves, and obtains few group cross-section, and carry out burnup calculating, obtains few group cross-section under subsequent burnup point.
(2) variation concentration of soluble boron afterflame consumption calculates:
Restart Method is required to meet after simulated fuel assembly changes concentration of soluble boron under a certain burn-up level
The needs that afterflame consumption calculates.When concentration of soluble boron variation, the component can be read in file in corresponding burnup depth from restarting
The relevant informations such as the nucleon density under degree, and the nucleic such as B-10, B-11 relevant information is handled again.Then, it is total to
Vibration calculates and/or neutron-transport equation calculates and burnup calculates.
(3) become fuel temperature afterflame consumption to calculate:
Restart Method is required to meet simulated fuel assembly changes the subsequent of fuel temperature under a certain burn-up level
The needs that burnup calculates.When fuel temperature variation, the component can be read in binary file in corresponding burnup from restarting
The relevant informations such as the nucleon density under depth, and consider the variation of microscopic cross caused by the variation of fuel temperature.Then, it carries out
Resonance calculating and/or neutron transport calculating and burnup calculate.
(4) slack-off agent temperature afterflame consumption calculates:
Restart Method is required to meet after simulated fuel assembly changes moderator temperature under a certain burn-up level
The needs that afterflame consumption calculates.When moderator temperature change, the component can be read in binary file corresponding from restarting
The relevant informations such as the nucleon density under burn-up level, and consider the variation of moderator-density caused by moderator temperature, according to upper
The variation for stating parameter re-starts parameter processing.Then, resonance calculating and/or neutron transport calculating are carried out and burnup calculates.
(5) control rod is inserted into/proposes reactor core afterflame consumption and calculates:
Restart Method be required to meet simulated fuel assembly control rod under a certain burn-up level be inserted into/propose reactor core
Afterflame consumption calculate needs.When control rod is inserted into/propose reactor core when, the component can be read in file corresponding from restarting
The relevant informations such as the nucleon density under burn-up level, and several how relevant informations are handled again.Then, resonance meter is carried out
Calculation and/or neutron transport calculating and burnup calculate.
(6) burnable poison rod group proposes that reactor core afterflame consumption calculates
Restart Method is required to meet simulated fuel assembly burnable poison rod group proposition heap under a certain burn-up level
The needs that the afterflame consumption of core calculates.When burnable poison rod group proposes reactor core, the component can be read in file exist from restarting
The relevant informations such as the nucleon density under corresponding burn-up level, and several how relevant informations are handled again.Then, it is total to
Vibration calculating and/or neutron transport calculating and burnup calculate.In fact, due to just there is flammable poison usually only in first circulation
Object, and following cycle need to only propose burnable poison rod without being inserted into burnable poison rod.So in restarting calculating usually
Pertain only to the proposition to burnable poison rod.
By using the Restart Method according to the embodiment of the present disclosure, group cross-section can be lacked according to Nuclear design programming system
The production demand in library, to calculate few group cross-section parameter of the fuel assembly under different operating conditions.
In the embodiments of the present disclosure, not only consider that power itself changes to Few group parameter calculating in changed power calculating
It influences, fuel temperature caused by the changed power of fuel assembly, moderator is also calculated by fuel effective temperature computing module
The variation of temperature, and pass through the change of interpolation thermal-hydraulic physical property table acquisition moderator-density due to caused by moderator temperature change
Change.The influence that moderator temperature itself calculates Few group parameter is not only considered in the calculating of moderator temperature change, it is also contemplated that
The variation of moderator-density caused by the moderator temperature change of fuel assembly.That is the disclosure not only allows for factor change
Direct effect caused by changing, it is also contemplated that second order effect caused by the factor changes.To consider the influence of some factor generation
More comprehensively, accurate.
In addition, in the embodiments of the present disclosure, considered in above-mentioned branch calculates the power of fuel assembly, fuel temperature,
Moderator temperature, concentration of soluble boron, control rod are inserted into/propose the influence of the factors such as reactor core active region, in afterflame consumption calculates, examine
Considered fuel assembly power, fuel temperature, moderator temperature, concentration of soluble boron, control rod be inserted into/propose reactor core active region,
Burnable poison rod proposes therefore the influence of the factors history effects such as reactor core active region can fully take into account fuel stack in reactor core
Part lacks the influence of group cross-section various aspects.Meanwhile when making few group cross-section library, program can also be analyzed according to other core physics
Needs, arbitrarily select above-mentioned Consideration, be easy to use, flexibly.
Fig. 1 be show executed in the system 10 shown in Fig. 2 according to the embodiment of the present disclosure for making reactor
Nuclear design lacks the flow chart of the Restart Method 100 in group cross-section library.As shown in Figure 1, method 100 may include step S110,
S120, S130 and S140.According to the disclosure, some steps of method 100 can be individually performed or combine execution, and can be simultaneously
Row executes or sequence executes, it is not limited to concrete operations sequence shown in FIG. 1.
Fig. 2 is to show to be opened again according to the example for lacking group cross-section library for making reactor nuclear design of the embodiment of the present disclosure
The block diagram of dynamic system 10.As shown in Fig. 2, restarting system 10 may include: burn-up level acquiring unit 12, fuel parameter acquisition
Unit 14, parameter pretreatment unit 16 and few group cross-section acquiring unit 18.
Burn-up level acquiring unit 12 can be used for obtaining the burn-up level of fuel assembly.Burn-up level acquiring unit 12 can
It, can be the central processing unit (CPU) of system 10, digital signal processor (DSP), microprocessor, microcontroller etc.
With the communications portion (for example, radio receiving-transmitting unit, Ethernet card, xDSL modem etc.) and/or storage unit with system 10
Divide (for example, hard disk, CD-ROM, RAM, SD card etc.) to match, is obtained by communications portion by other hardware (for example, sensor)
The burn-up level detected, or burn-up level that is precalculated or being got by other modules is read from storage section.
Fuel parameter acquiring unit 14 can be used for obtaining fuel assembly in the combustion determined by burn-up level acquiring unit 12
Consume the variation of one or more parameters under depth.Fuel parameter acquiring unit 14 can be the central processing unit of system 10
(CPU), digital signal processor (DSP), microprocessor, microcontroller etc., can be with the communications portion (example of system 10
Such as, radio receiving-transmitting unit, Ethernet card, xDSL modem etc.) and/or storage section (for example, hard disk, CD-ROM, RAM,
SD card etc.) it matches, the one or more parameters detected by other hardware (for example, sensor) are obtained by communications portion
Variation, or the variation of the one or more parameters got by other modules is read from storage section.
Parameter pretreatment unit 16 can be used for executing parameter pretreatment for the variation of one or more parameters.Parameter
Pretreatment unit 16 can be the central processing unit (CPU) of system 10, digital signal processor (DSP), microprocessor, micro-control
Device processed etc. can be directed to the variation of the one or more parameters got by fuel parameter acquiring unit 14 as described above,
And in view of various secondary variations caused by relevant parameter variation, Lai Jinhang branch is calculated and/or afterflame consumption calculates.
Few group cross-section acquiring unit 18 can be used for pretreated based on parameter as a result, neutron transport calculating is executed, to obtain
Group cross-section must be lacked.Few group cross-section acquiring unit 18 can be the central processing unit (CPU) of system 10, digital signal processor
(DSP), microprocessor, microcontroller etc. can be pre-processed according to the various parameters obtained by parameter pretreatment unit 16
As a result, being calculated to execute neutron transport, to finally obtain required few group cross-section.
Below with reference to Fig. 1 and Fig. 2, to according to the embodiment of the present disclosure execute on system 10 for making reactor
Nuclear design lacks the Restart Method 100 in group cross-section library and system 10 is described in detail.
Method 100 starts from step S110, in step s 110, can by the burn-up level acquiring unit 12 of system 10
Obtain the burn-up level of fuel assembly.
In step S120, fuel assembly can be obtained by the fuel parameter acquiring unit 14 of system 10 under burn-up level
One or more parameters variation.
In step S130, can be held by the parameter pretreatment unit 16 of system 10 for the variation of one or more parameters
The pretreatment of row parameter.
In step S140, it is pretreated as a result, holding parameter can be based on by system 10 and few group cross-section acquiring unit 18
Row neutron transport calculates, to obtain few group cross-section.
In some embodiments, one or more parameters may include at least one of following: power, concentration of soluble boron, combustion
Material temperature degree, moderator temperature and control rod are inserted into/propose reactor core active region.
In some embodiments, if one or more parameters include power, parameter pretreatment may include: to obtain combustion
Expect relevant information of the component under burn-up level；The fuel temperature of fuel assembly is determined based on the variation of relevant information and power
The variation of degree, moderator temperature and moderator-density；And nucleic microscopic cross is reacquired based on the variation and nucleon is close
Degree.
In some embodiments, if one or more parameters include fuel temperature, parameter pretreatment may include: to obtain
Take relevant information of the fuel assembly under burn-up level；Fuel assembly is determined based on the variation of relevant information and fuel temperature
The variation of microscopic cross；And nucleic microscopic cross and nucleon density are reacquired based on the variation.
In some embodiments, if one or more parameters include moderator temperature, parameter pretreatment may include:
Obtain relevant information of the fuel assembly under burn-up level；Fuel stack is determined based on the variation of relevant information and moderator temperature
The variation of the moderator-density of part；And nucleic microscopic cross and nucleon density are reacquired based on the variation.
In some embodiments, if one or more parameters include that control rod is inserted into/proposes reactor core active region, parameter
Pretreatment may include: the relevant information for obtaining fuel assembly under burn-up level；It is inserted into/mentions based on relevant information and control rod
The variation for changing to determine the geometry of fuel assembly of reactor core active region out；And based on the variation to reacquire nucleic microcosmic
Section and nucleon density.
In some embodiments, relevant information at least may include nucleon density of the fuel assembly under burn-up level.
In some embodiments, step S140 can also include: pretreated by parameter as a result, executing based on neutron transport
It calculates and burnup calculates, to obtain few group cross-section under the burn-up level and under subsequent burn-up level.
In some embodiments, if to obtain few group cross-section under subsequent burn-up level, one or more parameters are also
It may include: that burnable poison rod group proposes reactor core active region.
In some embodiments, if one or more parameters include concentration of soluble boron, parameter pretreatment may include:
Obtain relevant information of the fuel assembly under burn-up level；Fuel stack is determined based on the variation of relevant information and concentration of soluble boron
The variation of the nucleic relevant information of part；And nucleic microscopic cross and nucleon density are reacquired based on the variation.
In some embodiments, nucleic relevant information at least may include information relevant to B-10 and B-11.
In some embodiments, if one or more parameters include that burnable poison rod group proposes reactor core active region, join
Number pretreatment may include: the relevant information for obtaining fuel assembly under burn-up level；Based on relevant information and burnable poison rod
Group proposes the variation of reactor core active region to determine the variation of the geometry of fuel assembly；And nucleic is reacquired based on the variation
Microscopic cross and nucleon density.
In some embodiments, determine that the moderator of fuel assembly is close based on the variation of relevant information and moderator temperature
The step of variation of degree may include: the variation based on moderator temperature, obtain slowing down by interpolation thermal-hydraulic physical property table
The variation of agent density.
In some embodiments, before the neutron transport of step S140 calculates, resonance can also be performed and calculate.
In some embodiments, reactor can be presurized water reactor.
So far preferred embodiment is had been combined the disclosure is described.It should be understood that those skilled in the art are not
In the case where being detached from spirit and scope of the present disclosure, various other changes, replacement and addition can be carried out.Therefore, the disclosure
Range be not limited to above-mentioned specific embodiment, and should be defined by the appended claims.
Claims (16)
1. a kind of Restart Method for lacking group cross-section library for making reactor nuclear design, comprising:
(a) burn-up level of fuel assembly is obtained；
(b) variation of one or more parameters of the fuel assembly under the burn-up level is obtained；
(c) parameter pretreatment is executed for the variation of one or more of parameters；And
(d) pretreated based on the parameter to be calculated as a result, executing neutron transport, to obtain few group cross-section.
2. according to the method described in claim 1, wherein, one or more of parameters include at least one of the following: power, can
Molten boron concentration, fuel temperature, moderator temperature and control rod are inserted into/propose reactor core active region.
3. according to the method described in claim 2, wherein, if one or more of parameters include power, the parameter
Pretreatment includes:
Obtain relevant information of the fuel assembly under the burn-up level；
Determined based on the variation of the relevant information and the power fuel temperature of the fuel assembly, moderator temperature and
The variation of moderator-density；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
4. according to the method described in claim 2, wherein, if one or more of parameters include concentration of soluble boron, institute
Stating parameter pretreatment includes:
Obtain relevant information of the fuel assembly under the burn-up level；
The nucleic relevant information of the fuel assembly is determined based on the variation of the relevant information and the concentration of soluble boron
Variation；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
5. described if one or more of parameters include fuel temperature according to the method described in claim 2, wherein
Parameter pre-processes
Obtain relevant information of the fuel assembly under the burn-up level；
The variation of the microscopic cross of the fuel assembly is determined based on the variation of the relevant information and the fuel temperature；With
And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
6. according to the method described in claim 2, wherein, if one or more of parameters include moderator temperature, institute
Stating parameter pretreatment includes:
Obtain relevant information of the fuel assembly under the burn-up level；
The change of the moderator-density of the fuel assembly is determined based on the variation of the relevant information and the moderator temperature
Change；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
7. according to the method described in claim 2, wherein, if one or more of parameters include control rod insertion/proposition
Reactor core active region, then the parameter pretreatment include:
Obtain relevant information of the fuel assembly under the burn-up level；
The variation of reactor core active region is inserted into/proposed based on the relevant information and the control rod to determine the fuel assembly
The variation of geometry；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
8. the method according to any one of claim 3~7, wherein the relevant information includes at least the fuel stack
Nucleon density of the part under the burn-up level.
9. according to the method described in claim 1, wherein, step (d) further include:
It is pretreated based on the parameter to be calculated as a result, executing neutron transport calculating and burnup, to obtain under the burn-up level
And few group cross-section under subsequent burn-up level.
10. according to the method described in claim 9, wherein, if to obtain few group cross-section under the subsequent burn-up level,
One or more of parameters include at least one of the following: power, concentration of soluble boron, fuel temperature, moderator temperature, control rod
Insertion/proposition reactor core active region and burnable poison rod group propose reactor core active region.
11. according to the method described in claim 4, wherein, the nucleic relevant information includes at least related to B-10 and B-11
Information.
12. according to the method described in claim 9, wherein, if one or more of parameters include that burnable poison rod group mentions
Reactor core active region out, then the parameter pretreatment include:
Obtain relevant information of the fuel assembly under the burn-up level；
The variation of reactor core active region is proposed based on the relevant information and the burnable poison rod group to determine the fuel assembly
Geometry variation；And
Nucleic microscopic cross and nucleon density are reacquired based on the variation.
13. according to the method described in claim 6, wherein, the change based on the relevant information and the moderator temperature
Change come the step of determining the variation of the moderator-density of the fuel assembly and includes:
Based on the variation of the moderator temperature, the variation of moderator-density is obtained by interpolation thermal-hydraulic physical property table.
14. according to the method described in claim 1, wherein, reactor is presurized water reactor.
15. according to the method described in claim 1, wherein, before the neutron transport of step (d) calculates, also executing resonance meter
It calculates.
16. a kind of restart system for make that reactor nuclear design lacks group cross-section library, comprising:
Burn-up level acquiring unit, for obtaining the burn-up level of fuel assembly；
Fuel parameter acquiring unit, for obtaining the change of one or more parameters of the fuel assembly under the burn-up level
Change；
Parameter pretreatment unit executes parameter pretreatment for the variation for one or more of parameters；And
Few group cross-section acquiring unit is calculated for pretreated based on the parameter as a result, executing neutron transport, to obtain few group
Section.
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| CN108062989B (en) * | 2016-11-09 | 2019-07-09 | 国家电投集团科学技术研究院有限公司 | Nuclear power station design and operation supports method and system |
| CN107122564B (en) * | 2017-05-12 | 2020-01-24 | 西安交通大学 | Parameterization method for calculating few group constants of pin-by-pin of pressurized water reactor |
| CN110457802B (en) * | 2019-07-31 | 2021-08-20 | 上海交通大学 | Precision optimization implementation method for SFCOMPO fuel consumption experiment benchmark question check simulation |
| CN113868591B (en) * | 2021-09-28 | 2022-12-09 | 西安交通大学 | Method for obtaining high-precision total reaction cross section of unresolved resonance region |
-
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Non-Patent Citations (3)
| Title |
|---|
| COSINE软件包组件程序中基于MOC方法的输运模块开发与初步验证;李硕等;《原子能科学技术》;20131231;第47卷;第515-519页 |
| 分层数据格式在COSINE物理子系统中的应用;王苏等;《原子能科学技术》;20131231;第47卷;第721-724页 |
| 微观燃耗模型在COSINE软件包堆芯程序中的研究与应用;胡啸宇等;《原子能科学技术》;20131231;第47卷;第503-506页 |
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