Methods for determining the assumed uranium-plutonium isotopic composition of the reprocessing plant solution system

Abstract

This invention relates to a method for determining the assumed uranium-plutonium isotope composition of a reprocessing plant solution system. Based on the reprocessing target, the method determines the component parameters and irradiation history parameters of the target, and then determines a conservative combination of nuclide composition calculation parameters for criticality safety analysis. Nuclide composition calculations are performed for different control rod insertion states, and a nuclide composition database is established. Combining the control rod insertion range in the irradiation history of different reactor types' spent fuel assemblies and the conservative axial burnout distribution of the target components, the method determines the average burnout limit for spent fuel assemblies with different initial enrichment levels that satisfies the uranium-plutonium isotope composition assumption. After comparing and adjusting the iterative analysis with the range and distribution of average burnout of spent fuel assemblies with different initial enrichment levels, the method ensures that, under the determined uranium-plutonium isotope composition assumption and its corresponding average burnout limit for spent fuel assemblies, the reprocessing plant can handle the vast majority of spent fuel assemblies, thereby maximizing the economic efficiency of criticality safety design.

Description

TECHNICAL FIELD

[0001] The present application relates to the field of nuclear fuel reprocessing, in particular to a method for determining the proportion of actual spent fuel assemblies that need to be processed in a reprocessing plant. BACKGROUND

[0002] The burnup credit method, as a method for nuclear criticality safety analysis, is different from the method of assuming spent fuel as new fuel. The burnup credit method takes into account the overall decrease in reactivity due to the irradiation of fuel in the reactor core and the increase in cooling time, and excavates a certain calculation margin, thereby improving the economy of nuclear equipment or nuclear facilities designed by using the method.

[0003] In view of the criticality safety design requirements of the spent fuel reprocessing plant for processing high initial enrichment and high discharge burnup assemblies, the application of the burnup credit method in the spent fuel reprocessing plant is studied, so that the large-scale spent fuel reprocessing plant can make full use of the reactivity reduction of the spent fuel due to burnup, and improve the processing capacity of the key process flow of the spent fuel reprocessing plant. The burnup credit method is gradually applied to the nuclear criticality safety design of the spent fuel reprocessing plant.

[0004] The spent fuel reprocessing plant applies the burnup credit method, and needs to determine a uranium-plutonium isotope composition assumption in advance for the criticality safety design of the solution system related equipment. For the uranium-plutonium isotope composition assumption, the corresponding average burnup limit value of the spent fuel assembly is specified, so that the reprocessing plant can process most of the spent fuel assemblies, and the specified limit value is conservative enough. On the premise of meeting the limit value requirement, the solution formed after the spent fuel assemblies with different irradiation histories are dissolved has a reactivity smaller than that of the solution system under the uranium-plutonium isotope composition assumption, so that the criticality safety design is conservative, and the characteristics of the fissile nuclides consumed with burnup in the spent fuel assembly are utilized as much as possible to improve the economy of the criticality safety design. However, the prior art cannot determine the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant. SUMMARY

[0005] The purpose of the present application is to provide a method for determining the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant, which meets the need of determining the uranium-plutonium isotope composition assumption of the solution system in advance when the reprocessing plant applies the burnup credit method for criticality safety design.

[0006] The present embodiment provides a method for determining the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant, which includes the following steps:

[0007] S1, determining the processing object of the reprocessing plant, and determining the preset proportion target of the spent fuel assemblies that need to be processed;

[0008] S2, for the determined processing object of the reprocessing plant, carry out spent fuel assembly burnup depth investigation and prediction analysis, determine the range and distribution of average burnup of spent fuel assemblies with different initial enrichment;

[0009] S3, determine the parameters of the spent fuel assembly of the processing object, the irradiation history parameters, and further determine the conservative combination of the calculation parameters of the nuclide composition for critical safety analysis;

[0010] S4, using the conservative combination of the calculation parameters of the nuclide composition for critical safety analysis, carrying out nuclide composition calculation under different control rod insertion states, and establishing a nuclide composition database;

[0011] S5, given the uranium-plutonium isotope composition assumption, determine the average burnup limit value of the spent fuel assembly with different initial enrichment that meets the uranium-plutonium isotope composition assumption;

[0012] S6, find the average burnup limit value of the spent fuel assembly with different initial enrichment that meets the uranium-plutonium isotope composition assumption from the nuclide composition database, and judge whether the proportion of the actual spent fuel assembly corresponding to the average burnup limit value is the preset proportion target of the spent fuel assembly to be processed;

[0013] S7, if the proportion of the actual spent fuel assembly corresponding to the average burnup limit value is the preset proportion target of the spent fuel assembly to be processed, then determine the proportion of the actual spent fuel assembly to be processed by the reprocessing plant; otherwise, adjust the conservatism of the uranium-plutonium isotope composition assumption, and after the adjustment, return to execute steps S5 and S6 until the proportion of the actual spent fuel assembly corresponding to the average burnup limit value is the preset proportion target of the spent fuel assembly to be processed by the reprocessing plant.

[0014] Preferably, in step S1, the determination of the processing object of the reprocessing plant includes any one or more of the type, initial enrichment range, and defined after-discharge cooling time of the spent fuel assembly planned to be processed by the reprocessing plant.

[0015] Preferably, in step S2, the range and distribution of average burnup of spent fuel assemblies with different initial enrichment are for spent fuel assemblies generated by nuclear power plants from which the reprocessing plant processes spent fuel assemblies, and for spent fuel assemblies to be generated according to a prospective fuel management scheme.

[0016] Preferably, in step S3, the conservative combination of the calculation parameters of the nuclide composition for critical safety analysis refers to a conservative analysis method based on the burnup credit, which is used to determine the conservative combination of the relevant parameters of the irradiation history for the spent fuel assembly predicted to be processed by the reprocessing plant, under the fuel management scheme that has been adopted or will be adopted by the nuclear power plant, with the reaction conservatism for critical safety analysis as the measure.

[0017] Preferably, in the step S3, the conservative combination of nuclide composition calculation parameters for criticality safety analysis includes at least two of the following: spent fuel assembly power, fuel temperature, coolant density, cooling temperature, soluble boron concentration, burnable poison type, and burnable poison amount.

[0018] Different conservative combinations of nuclide composition calculation parameters for criticality safety analysis can be established for different nuclear power plant types.

[0019] Preferably, in the step S4, the nuclide composition database includes the nuclide composition of spent fuel within a limited range based on different post-discharge cooling times, for any one or several of the following parameters: nuclear power plant type, initial enrichment of spent fuel assembly, control rod insertion state, and burnup point.

[0020] Preferably, in the step S4, the database of spent fuel nuclide composition within a limited range based on different post-discharge cooling times considers different post-discharge cooling times for each burnup point; and the control rod insertion state is: no control rod insertion, control rod insertion, gray rod control rod insertion, or black rod control rod insertion, according to the control rod control strategy of the nuclear power plant type.

[0021] Preferably, in the step S5, the determination of the average burnup limit of spent fuel assembly of different initial enrichments satisfying the uranium-plutonium isotope composition assumption refers to the determination of the average burnup limit of spent fuel assembly of different initial enrichments based on the uranium isotope composition assumption and the plutonium isotope composition assumption, respectively, and the selection of the smaller value as the average burnup limit of spent fuel assembly of different initial enrichments satisfying the uranium-plutonium isotope composition assumption.

[0022] Preferably, the step S3 further includes the following steps: determining the control rod insertion range in the irradiation history that needs to be considered for different types of spent fuel assemblies by processing the parameters of the spent fuel assembly and the irradiation history, and determining the axial burnup distribution of the spent fuel assembly to be processed.

[0023] In the step S5, when determining the average burnup limit of spent fuel assembly of different initial enrichments, the nuclide composition data in the control rod insertion range of different axial heights of the spent fuel assembly are selected based on the determined control rod insertion range in the irradiation history that needs to be considered for different types of spent fuel assemblies, and the nuclide composition calculation results under the corresponding control rod insertion state are selected.

[0024] The burnup value of each axial segment is determined by using the axial burnup distribution of the spent fuel assembly to be processed, and the nuclide composition calculation results under the corresponding control rod insertion state are selected, and then the average uranium-plutonium isotope composition of the spent fuel assembly is obtained by weighting according to the fuel volume ratio of each axial segment, and the average burnup limit of spent fuel assembly of different initial enrichments is determined by comparing with the uranium-plutonium isotope composition assumption.

[0025] and / or, the average burnup limit of the spent fuel assembly corresponding to different initial enrichment is determined respectively for different sources of nuclear power plant reactor types;

[0026] and / or, the average burnup limit of the spent fuel assembly corresponding to different initial enrichment is determined respectively for different post-discharge cooling times.

[0027] Preferably, in the step S3, the axial burnup distribution of the spent fuel assembly to be treated is conservatively adjusted, and the axial burnup distribution is determined respectively for different burnup depths of the spent fuel assembly. Different conservative axial burnup distributions are established for different sources of nuclear power plant reactor types.

[0028] Preferably, in the step S7, the specific method for conservatively adjusting the uranium-plutonium isotope composition assumption is: if the proportion of the corresponding actual spent fuel assembly is higher than the preset proportion target of the spent fuel assembly to be treated, the uranium-plutonium isotope composition assumption is conservatively adjusted to be reduced; if the proportion of the corresponding actual spent fuel assembly is lower than the preset proportion target of the spent fuel assembly to be treated, the uranium-plutonium isotope composition assumption is conservatively adjusted to be increased.

[0029] Preferably, in the step S7, the conservatively increasing of the uranium-plutonium isotope composition assumption means adjusting the uranium-plutonium isotope composition assumption to increase the reactivity of the solution system, in which case the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the uranium-plutonium isotope composition assumption is reduced.

[0030] The conservatively reducing of the uranium-plutonium isotope composition assumption means adjusting the uranium-plutonium isotope composition assumption to reduce the reactivity of the solution system, in which case the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the uranium-plutonium isotope composition assumption is increased.

[0031] Preferably, in the step S7, the finally determined uranium-plutonium isotope composition assumption is applicable to different sources of nuclear power plant reactor types, spent fuel assemblies within the limited range of the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the uranium-plutonium isotope composition assumption, and different post-discharge cooling times within the limited range of the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the uranium-plutonium isotope composition assumption.

[0032] For the finally determined uranium-plutonium isotope composition assumption, different average burnup limits of the spent fuel assembly corresponding to different initial enrichment are set for different sources of nuclear power plant reactor types, or different average burnup limits of the spent fuel assembly corresponding to different initial enrichment are set for different post-discharge cooling times.

[0033] The present application aims at the need of determining the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant when the reprocessing plant applies the burnup credit system for critical safety design, and provides a specific determination method and analysis process.

[0034] The present application provides a determination method and analysis process of the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant when the reprocessing plant applies the burnup credit system for critical safety design, considers various conservative factors required in the application process of the burnup credit system, and the determined uranium-plutonium isotope composition assumption and the corresponding spent fuel assembly average burnup limit value are conservative and reliable, which can improve the economy of the critical safety design as much as possible under the premise that the reprocessing plant can process most of the spent fuel assemblies. BRIEF DESCRIPTION OF DRAWINGS

[0035] Figure 1 The method for determining the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant in Example 3.

[0036] Figure 2 The example graph of the envelope curve of the conservative axial burnup distribution of the spent fuel assembly determined in Example 3.

[0037] Figure 3 The example graph of the average burnup limit value of the spent fuel assembly with different initial enrichments determined in Example 3. DETAILED DESCRIPTION

[0038] In order to make the purpose, technical scheme and advantages of the present application clearer, the present application is further described in detail below in combination with the drawings and examples. It should be understood that the specific examples described herein are only used to explain the present application, and are not used to limit the present application.

[0039] Example 1

[0040] The present example provides a method for determining the uranium-plutonium isotope composition assumption of the solution system of the reprocessing plant, which comprises the following steps:

[0041] S1, determining the processing object of the reprocessing plant, and determining the preset proportion target of the spent fuel assembly to be processed;

[0042] S2, for the determined processing object of the reprocessing plant, carrying out spent fuel assembly burnup depth research and prediction analysis, and determining the range and distribution of the average burnup of the spent fuel assembly with different initial enrichments;

[0043] S3, determining the spent fuel assembly parameters and irradiation history parameters of the processing object, and further determining the conservative combination of the calculation parameters of the critical safety analysis nuclide composition;

[0044] S4, adopt the conservative combination of the nuclide composition calculation parameters for criticality safety analysis, carry out nuclide composition calculation of different control rod insertion states, and establish a nuclide composition database;

[0045] S5, determine the average burnup limit values of the spent fuel assemblies of different initial enrichment degrees that meet the uranium-plutonium isotope composition assumption, given the uranium-plutonium isotope composition assumption;

[0046] S6, find the average burnup limit values of the spent fuel assemblies of different initial enrichment degrees that meet the uranium-plutonium isotope composition assumption from the nuclide composition database, and judge whether the proportion of the actual spent fuel assemblies corresponding thereto is the preset proportion target of the spent fuel assemblies that need to be treated;

[0047] S7, if the proportion of the actual spent fuel assemblies corresponding thereto is the preset proportion target of the spent fuel assemblies that need to be treated, determine the proportion of the actual spent fuel assemblies that need to be treated by the reprocessing plant; otherwise, conservatively adjust the uranium-plutonium isotope composition assumption, and after the conservative adjustment, return to execute steps S5 and S6 until the proportion of the actual spent fuel assemblies corresponding thereto is the preset proportion target of the spent fuel assemblies that need to be treated by the reprocessing plant.

[0048] The beneficial effects of the embodiment are as follows: The embodiment provides a method for determining the uranium-plutonium isotope composition assumption that can be used in the criticality safety design of the solution system of a reprocessing plant based on the burnup credit system, considers various conservative factors required in the application process of the burnup credit system, and the determined uranium-plutonium isotope composition assumption and the corresponding average burnup limit values of the spent fuel assemblies are conservative and reliable. Under the premise that the reprocessing plant can treat most of the spent fuel assemblies, the economy of the criticality safety design can be improved as much as possible.

[0049] Embodiment 2

[0050] The embodiment provides a method for determining a uranium-plutonium isotope composition assumption of a solution system of a reprocessing plant, which comprises the following steps:

[0051] S1, determine the processing object of the reprocessing plant and determine the preset proportion target of the spent fuel assemblies that need to be treated;

[0052] S2, carry out spent fuel assembly burnup depth research and prediction analysis for the determined processing object of the reprocessing plant, and determine the range and distribution of the average burnup of the spent fuel assemblies of different initial enrichment degrees;

[0053] S3, determine the parameters of the spent fuel assemblies of the processing object and the irradiation history parameters, and further determine the conservative combination of the nuclide composition calculation parameters for criticality safety analysis;

[0054] S4, adopt the conservative combination of the nuclide composition calculation parameters for criticality safety analysis, carry out nuclide composition calculation of different control rod insertion states, and establish a nuclide composition database;

[0055] S5, determining the average burnup limit of the spent fuel assemblies with different initial enrichment satisfying the uranium-plutonium isotope composition assumption;

[0056] S6, finding the average burnup limit of the spent fuel assemblies with different initial enrichment satisfying the uranium-plutonium isotope composition assumption from the nuclide composition database, and judging whether the proportion of the corresponding actual spent fuel assemblies is the preset proportion target of the spent fuel assemblies to be treated;

[0057] S7, if the proportion of the corresponding actual spent fuel assemblies is the preset proportion target of the spent fuel assemblies to be treated, determining the proportion of the actual spent fuel assemblies to be treated in the reprocessing plant; otherwise, conservatively adjusting the uranium-plutonium isotope composition assumption, and returning to execute steps S5 and S6 until the proportion of the corresponding actual spent fuel assemblies is the preset proportion target of the spent fuel assemblies to be treated in the reprocessing plant.

[0058] Preferably, in the step S1, the determination of the treatment object of the reprocessing plant includes any one or several of the type, the initial enrichment range, and the limited after-discharge cooling time of the spent fuel assemblies planned to be treated by the reprocessing plant.

[0059] Preferably, in the step S2, the range and distribution of the average burnup of the spent fuel assemblies with different initial enrichments are for the spent fuel assemblies generated by the nuclear power plant from which the spent fuel assemblies to be treated by the reprocessing plant are derived and the spent fuel assemblies to be generated according to the prospective fuel management scheme.

[0060] Preferably, in the step S3, the conservative combination of the nuclide composition calculation parameters for criticality safety analysis refers to a conservative analysis method based on the burnup credit system for the spent fuel assemblies derived from the nuclear power plant predicted to be treated by the reprocessing plant and having adopted or to be adopted fuel management scheme, to determine the conservative combination of the related parameters of the irradiation history by taking the reactivity conservation for criticality safety analysis as the measurement.

[0061] Preferably, in the step S3, the conservative combination of the nuclide composition calculation parameters for criticality safety analysis includes at least two of the spent fuel assembly power, the fuel temperature, the coolant density, the cooling temperature, the soluble boron concentration, the type of burnable poison, and the number of burnable poisons. Different conservative combinations of the nuclide composition calculation parameters for criticality safety analysis can be established for different source nuclear power plant types.

[0062] Preferably, in the step S4, the nuclide composition database includes the nuclide composition of the spent fuel assemblies within the limited range based on different after-discharge cooling times under any one or several parameters of the nuclear power plant type, the initial enrichment of the spent fuel assembly, the control rod insertion state, and the burnup point.

[0063] Preferably, the step S4 is based on the database of the spent fuel nuclide composition within the limited range of the different post-irradiation cooling time, and each burnup point considers different post-irradiation cooling time; according to the control rod control strategy of the nuclear power plant type, the control rod insertion state is: no control rod insertion, control rod insertion, or gray rod control rod insertion, black rod control rod insertion.

[0064] Preferably, in the step S5, the determination of the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments under the assumption of uranium-plutonium isotopic composition refers to the determination of the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments under the assumption of uranium isotopic composition and the assumption of plutonium isotopic composition, respectively, and the smaller value is selected as the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments under the assumption of uranium-plutonium isotopic composition.

[0065] Preferably, the step S3 further comprises the following steps: determining the control rod insertion range in the irradiation history that needs to be considered for different types of spent fuel assemblies by processing the spent fuel assembly parameters and the irradiation history parameters, and determining the conservative axial burnup distribution of the spent fuel assembly of the processing object.

[0066] In the step S5 of determining the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments, according to the determined control rod insertion range in the irradiation history that needs to be considered for different types of spent fuel assemblies, the nuclide composition data in the control rod insertion range at different axial heights of the spent fuel assembly is selected to use the nuclide composition calculation result under the corresponding control rod insertion state.

[0067] The average uranium-plutonium isotopic composition of the spent fuel assembly is obtained by weighting the fuel volume ratio of each axial segment using the nuclide composition calculation result under the corresponding control rod insertion state, and then the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments is determined by comparing the uranium-plutonium isotopic composition assumption.

[0068] And / or, the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments is determined for different sources of nuclear power plant types.

[0069] And / or, the average burnup limit value of the spent fuel assembly corresponding to different initial enrichments is determined for different post-irradiation cooling times.

[0070] Preferably, in the step S3, the conservative axial burnup distribution of the processing object spent fuel assembly is determined for different burnup depths of the spent fuel assembly, and different conservative axial burnup distributions are established for different sources of nuclear power plant types.

[0071] The parameter combination selected in the nuclide composition calculation is relatively conservative for the critical safety design. The control rod insertion state and the conservative axial burnup distribution of the spent fuel assembly to be treated are considered when determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the assumption of the uranium-plutonium isotope composition, which can ensure the conservatism and reliability of the critical safety design of the reprocessing plant.

[0072] Preferably, the specific method for conservatively adjusting the uranium-plutonium isotope composition assumption in step S7 is: if the proportion of the corresponding actual spent fuel assembly is higher than the preset proportion target of the spent fuel assembly to be treated, the uranium-plutonium isotope composition assumption is conservatively adjusted to be reduced; if the proportion of the corresponding actual spent fuel assembly is lower than the preset proportion target of the spent fuel assembly to be treated, the uranium-plutonium isotope composition assumption is conservatively adjusted to be increased.

[0073] Preferably, in step S7, the uranium-plutonium isotope composition assumption conservatively adjusted to be increased means adjusting the uranium-plutonium isotope composition assumption to increase the reactivity of the solution system, in which case the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the assumption of the uranium-plutonium isotope composition is reduced.

[0074] The uranium-plutonium isotope composition assumption conservatively adjusted to be reduced means adjusting the uranium-plutonium isotope composition assumption to decrease the reactivity of the solution system, in which case the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the assumption of the uranium-plutonium isotope composition is increased.

[0075] Preferably, in step S7, the finally determined uranium-plutonium isotope composition assumption is applicable to different source nuclear power plant types, spent fuel assemblies within the limited range of the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the assumption of the uranium-plutonium isotope composition, and different cooling times after discharge within the limited range of the average burnup limit of the spent fuel assembly corresponding to different initial enrichment under the assumption of the uranium-plutonium isotope composition.

[0076] For the finally determined uranium-plutonium isotope composition assumption, different average burnup limits of the spent fuel assembly corresponding to different initial enrichment are set for different source nuclear power plant types, or different average burnup limits of the spent fuel assembly corresponding to different initial enrichment are set for different cooling times after discharge.

[0077] The embodiment relates to a method and an analysis process for determining a uranium-plutonium isotope composition assumption of a solution system of a reprocessing plant, determining component parameters, irradiation history parameters of a processing object of the reprocessing plant, and further determining a conservative combination of calculation parameters of a nuclide composition for critical safety analysis, carrying out nuclide composition calculation under different control rod insertion states, establishing a nuclide composition database, and combining a control rod insertion range in an irradiation history which needs to be considered for different reactor type spent fuel components, and a conservative axial burnup distribution of the processing object, so that different initial enrichment degree spent fuel component average burnup limit values which meet a certain uranium-plutonium isotope composition assumption can be determined, and iteration analysis is adjusted after comparison with the range and distribution of the average burnup of the spent fuel components with different initial enrichment degrees, so that the reprocessing plant can just process most of the spent fuel components under the determined uranium-plutonium isotope composition assumption and the corresponding spent fuel component average burnup limit value.

[0078] The embodiment has the following beneficial effects: the embodiment provides a method and an analysis process for determining a uranium-plutonium isotope composition assumption which can be used in critical safety design of a solution system of a reprocessing plant based on a burnup credit system, various conservative factors which need to be considered in the application process of the burnup credit system are considered, the determined uranium-plutonium isotope composition assumption and the corresponding spent fuel component average burnup limit value are conservative and reliable, and the economy of the critical safety design can be improved as much as possible under the premise that most of the spent fuel components can be processed by the reprocessing plant.

[0079] Embodiment 3

[0080] Specifically, the following actual process for determining a uranium-plutonium isotope composition assumption and the corresponding spent fuel component average burnup limit value in the research and design of a certain reprocessing plant is used to illustrate the application of the embodiment.

[0081] Figure 1 A flowchart of the method for determining a uranium-plutonium isotope composition assumption of a solution system of a reprocessing plant is given, which can provide an image understanding of the content of the invention.

[0082] The embodiment provides a method for determining a uranium-plutonium isotope composition assumption of a solution system of a reprocessing plant, which comprises the following steps:

[0083] Firstly, the processing object of the reprocessing plant is determined: in the embodiment, the reprocessing plant processes AFA-3G type spent fuel components, the possible sources of which are CNP600 (121 reactor cores), M310 (157 reactor cores) and ACP1000 (177 reactor cores), the maximum initial enrichment degree is not more than 4.45%, and the minimum cooling time after discharge is not less than 5 years. And for the spent fuel components from these sources, the preset proportion target of the acceptable spent fuel components which can be processed is 95%, that is, 95% of the spent fuel components generated by these reactor cores can be processed by the spent fuel reprocessing plant.

[0084] Further, the burnup depth of the spent fuel assemblies is researched and predicted to determine the range and distribution of the average burnup of the spent fuel assemblies with different initial enrichments: In this embodiment, the burnup distribution of the discharged spent fuel assemblies is counted based on the design values of the 18-month fuel management of the three types of reactors. The discharge burnup depth of the spent fuel assemblies with an initial enrichment of 4.45% under the balanced cycle is greater than BU (GWd / tU) (the discharge burnup depth of the spent fuel assemblies with other initial enrichments is not shown, and the discharge burnup depth of the spent fuel assemblies produced under the transition cycle and other fuel management schemes is less than BU (GWd / tU)).

[0085] Further, the parameters of the spent fuel assemblies and the irradiation history parameters of the processing objects are determined to further determine the conservative combination of the nuclide composition calculation parameters for criticality safety analysis: In this embodiment, the coolant outlet temperature and the corresponding density of the reactor core are used for different types of reactors; the average boron concentration during the cycle life is used for the soluble boron concentration; the PYREX rod or the WABA rod used in the burnup range of the first cycle is considered; the decrease in the fuel loading of the Gd rod is considered, and the spectral hardening effect of the Gd rod as a combustible absorber is not considered; the maximum power density is used for the power density; the fuel pellet temperature corresponding to the maximum power density is used for the fuel pellet temperature; and the conservative combination of the nuclide composition calculation parameters for criticality safety analysis is determined.

[0086] Further, the control rod insertion range in the irradiation history of the spent fuel assemblies of different types of reactors is determined: In this embodiment, for the three types of reactors considered, the control rods are only inserted in a small top section during the core burnup process, so the nuclide composition analysis for criticality analysis only needs to insert the control rods in a small top section, and the specific range is determined according to the conditions of each type of reactor.

[0087] Further, the axial burnup distribution of the spent fuel assembly is determined: in this embodiment, for the three types of reactors considered, factors affecting the axial burnup distribution of the spent fuel assembly, such as the control rod insertion depth and the residence time during the operation of the reactor core, are considered, and the applicable axial burnup envelope of the spent fuel assembly is generated. The basic idea is that, from the theoretical calculation or measured burnup information of the spent fuel assembly, the spent fuel assemblies with an average discharge burnup in a certain burnup range are selected, the axial burnup distribution of each spent fuel assembly is normalized, if the minimum normalized burnup of the same axial segment is less than 1, the minimum normalized burnup of the axial segment of each spent fuel assembly is taken as the normalized burnup of the axial segment; if the minimum normalized burnup of the same axial segment is greater than 1, the maximum normalized burnup of the axial segment of each spent fuel assembly is taken as the normalized burnup of the axial segment; finally, the normalized burnup of the segments with a maximum normalized burnup greater than 1 is re-normalized, thereby obtaining the axial burnup envelope curve in a certain burnup range. A set of axial burnup envelope curves that have been proven to be conservative can also be used.

[0088] Figure 2 An example of a determined and conservative envelope curve of the axial burnup distribution of the spent fuel assembly is given.

[0089] In other embodiments, considering that the axial burnup distribution of the spent fuel assembly varies greatly for different reactor types, different conservative axial burnup distributions are established for different sources of nuclear power plant reactor types.

[0090] Further, a conservative combination of nuclide composition calculation parameters for criticality safety analysis is used to carry out nuclide composition calculation for different control rod insertion states, and a nuclide composition database is established: in this embodiment, for the control rod control strategies of the three types of reactors considered, the nuclide composition calculation for different control rod insertion states refers to the nuclide composition calculation in the two states of no control rod insertion and control rod insertion. In this embodiment, the established nuclide composition database is for spent fuel assemblies with initial enrichments of 1.8%, 2.4%, 3.1%, 3.7%, 3.9%, 4.45%, etc. For each initial enrichment of the spent fuel assembly, the nuclide composition calculation burnup point starts from 1000 MWd / tU, and the nuclide composition calculation results are output every 1000 MWd / tU until 72000 MWd / tU, and for each burnup point, two cooling times of 5 years and 8 years are considered.

[0091] In other embodiments, the spent fuel assembly is derived from a nuclear power plant reactor type that uses the movement of a control rod bank to complete the reactor operation regulation control strategy, and the nuclide composition calculation for different control rod insertion states refers to the nuclide composition calculation in the two states of inserting a gray rod control rod and inserting a black rod control rod.

[0092] Further, given an initial uranium-plutonium isotopic composition assumption, with conservative spent fuel assembly axial burn-up distribution, considering control rod insertion range, for the case of 5 years of post-irradiation cooling time, determine the different initial enrichment spent fuel assembly average burn-up limit values that meet the uranium-plutonium isotopic composition assumption.

[0093] Further, determine whether the different initial enrichment spent fuel assembly average burn-up limit values that meet the uranium-plutonium isotopic composition assumption can just meet the preset proportion target of acceptable spent fuel assemblies that can be handled: if there are many that cannot be handled, increase the conservatism of the uranium-plutonium isotopic composition assumption; if the limit value is too small, decrease the conservatism of the uranium-plutonium isotopic composition assumption.

[0094] Further, re-determine the different initial enrichment spent fuel assembly average burn-up limit values that meet the uranium-plutonium isotopic composition assumption until the determined uranium-plutonium isotopic composition assumption and its corresponding spent fuel assembly average burn-up limit value can just meet the preset proportion target of acceptable spent fuel assemblies that can be handled.

[0095] In this embodiment, the finally determined uranium-plutonium isotopic composition assumption is: 235 U remaining enrichment is W u235 , 240 Pu content is W Pu240 For example, for a spent fuel assembly with an initial enrichment of 4.45%, the corresponding spent fuel assembly average burn-up limit value is about (BU-1) (GWd / tU), which is slightly less than the minimum discharge burn-up depth BU (GWd / tU) of the spent fuel assembly with an initial enrichment of 4.45% under the 18-month fuel management scheme. Therefore, the reprocessing plant can handle the expected spent fuel assemblies with an initial enrichment of 4.45%, and only a small number of spent fuel assemblies produced under transition cycles and other fuel management schemes do not meet the limit requirement.

[0096] In this embodiment, the spent fuel assembly average burn-up limit value (BU-1) (GWd / tU) for the case of 5 years of post-irradiation cooling time can envelope the case of longer cooling time, and the minimum post-irradiation cooling time of the handling object limit range is not less than 5 years. In this embodiment, different spent fuel assembly average burn-up limit values are not set for different cooling times and for different sources of nuclear power plant types, i.e., the spent fuel assembly average burn-up limit value is applicable to different sources of nuclear power plant types, and the spent fuel assembly cooling time is greater than or equal to 5 years.

[0097] Figure 3An example of the average burnup limit of the spent fuel assembly with different initial enrichment determined in the embodiment is given. The limit determined in the embodiment considers various conservative factors in the application of the burnup credit system and is conservative enough to make the reactivity of the solution system formed by dissolving the spent fuel assemblies with different irradiation histories and under the actual uranium-plutonium isotope composition less than the reactivity of the solution system under the assumed uranium-plutonium isotope composition.

[0098] In some other embodiments, different average burnup limits of the spent fuel assembly with different initial enrichment corresponding to different assumed uranium-plutonium isotope compositions are provided for different nuclear power plant types.

[0099] In some other embodiments, different average burnup limits of the spent fuel assembly with different initial enrichment corresponding to different after-cooling time are provided.

[0100] The beneficial effects of the embodiment are as follows: The embodiment provides a method and a flow for determining the assumed uranium-plutonium isotope composition for the critical safety design of the solution system of the reprocessing plant based on the burnup credit system, considers various conservative factors required in the application of the burnup credit system, and the determined assumed uranium-plutonium isotope composition and the corresponding average burnup limit of the spent fuel assembly are conservative and reliable, which can improve the economy of the critical safety design as much as possible under the premise that the reprocessing plant can process most of the spent fuel assemblies.

[0101] Obviously, those skilled in the art can make various modifications and variations to the present application without departing from the spirit and scope of the present application. Thus, if these modifications and variations of the present application belong to the scope of the claims of the present application and their equivalents, the present application also intends to include these modifications and variations.

Claims

1. A method of determining the assumed isotopic composition of uranium and plutonium in a reprocessing plant solution system, characterized in that, The method comprises the following steps: S1, determining the processing object of the reprocessing plant, and determining the preset proportion target of the spent fuel assembly to be processed; S2, for the determined processing object of the reprocessing plant, carrying out spent fuel assembly burnup depth research and prediction analysis, and determining the range and distribution of the average burnup of spent fuel assemblies with different initial enrichments; S3, determining the spent fuel assembly parameters and irradiation history parameters of the processing object, and then determining the conservative combination of the nuclide composition calculation parameters for critical safety analysis; S4, using the conservative combination of the nuclide composition calculation parameters for critical safety analysis, carrying out nuclide composition calculation under different control rod insertion states, and establishing a nuclide composition database; S5, given the uranium-plutonium isotope composition assumption, determining the average burnup limit value of the spent fuel assembly with different initial enrichments that meets the uranium-plutonium isotope composition assumption; S6, finding out the average burnup limit value of the spent fuel assembly with different initial enrichments that meets the uranium-plutonium isotope composition assumption from the nuclide composition database, and judging whether the proportion of the actual spent fuel assembly corresponding thereto is the preset proportion target of the spent fuel assembly to be processed; S7, if the proportion of the actual spent fuel assembly corresponding thereto is the preset proportion target of the spent fuel assembly to be processed, the proportion of the actual spent fuel assembly to be processed by the reprocessing plant is determined; otherwise, the uranium-plutonium isotope composition assumption is conservatively adjusted, and after the conservative adjustment, steps S5 and S6 are executed until the proportion of the actual spent fuel assembly corresponding thereto is the preset proportion target of the spent fuel assembly to be processed by the reprocessing plant.

2. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 1, characterized in that, In the step S1, the determination of the processing object of the reprocessing plant includes any one or several of the type, initial enrichment range, and defined after-discharge cooling time of the spent fuel assembly to be processed by the reprocessing plant.

3. The method for determining a postulated uranium-plutonium isotope composition of a reprocessing plant solution system as claimed in claim 1, wherein, In the step S2, the range and distribution of the average burnup of the spent fuel assembly with different initial enrichments are for the spent fuel assembly generated by the nuclear power plant from which the spent fuel assembly to be processed by the reprocessing plant is sourced and not processed, and the spent fuel assembly to be generated according to the prospective fuel management scheme.

4. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 1, wherein, In the step S3, the conservative combination of the nuclide composition calculation parameters for critical safety analysis refers to the conservative combination of the related parameters of the irradiation history of the spent fuel assembly predicted to be sourced from the nuclear power plant that has adopted or will adopt the fuel management scheme, based on the conservative analysis method of the burnup credit, with the reaction conservation for critical safety analysis as the measure.

5. The method of determining a postulated Pu isotope composition of a reprocessing plant solution system as claimed in claim 1 or 4, characterized in that, In the step S3, the conservative combination of the nuclide composition calculation parameters for critical safety analysis includes at least two of the spent fuel assembly power, fuel temperature, coolant density, cooling temperature, soluble boron concentration, and burnable poison type and quantity.

6. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 1, wherein, In the step S4, the nuclide composition database includes the nuclide composition of the spent fuel assembly within the defined range based on different after-discharge cooling times for any one or several parameters of the nuclear power plant type, initial enrichment of the spent fuel assembly, control rod insertion state, and burnup point.

7. The method of determining a postulated uranium-plutonium isotope composition of a reprocessing plant solution system as claimed in claim 6 wherein, The step S4 includes the following steps: determining the control rod insertion range in the irradiation history of the spent fuel assembly of different reactor types according to the spent fuel assembly parameters and irradiation history parameters of the processing object; and determining the conservative axial burnup distribution of the spent fuel assembly of the processing object.

8. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 1, wherein, The step S5 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium isotope composition assumption and the plutonium isotope composition assumption; and selecting the smaller value as the average burnup limit of the spent fuel assembly corresponding to different initial enrichments satisfying the uranium-plutonium isotope composition assumption.

9. The method of determining a postulated Pu isotope composition of a reprocessing plant solution system as claimed in claim 1 or 8, characterized in that, The step S3 includes the following steps: determining the control rod insertion range in the irradiation history of the spent fuel assembly of different reactor types according to the spent fuel assembly parameters and irradiation history parameters of the processing object; and determining the conservative axial burnup distribution of the spent fuel assembly of the processing object. In the step S5, when determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments, the nuclide composition data in the control rod insertion range at different axial heights of the spent fuel assembly are selected according to the determined control rod insertion range in the irradiation history of the spent fuel assembly of different reactor types. The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed. The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed. The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed.

10. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 9, wherein, The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed.

11. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 1, wherein, The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed.

12. The method of determining a postulated uranium-plutonium isotope composition of a reprocessing plant solution system as claimed in claim 11 wherein, The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed. The step S7 includes the following steps: determining the average burnup limit of the spent fuel assembly corresponding to different initial enrichments according to the uranium-plutonium isotope composition assumption; and adjusting the uranium-plutonium isotope composition assumption according to the proportion of the actual spent fuel assembly corresponding to the uranium-plutonium isotope composition assumption to the preset proportion target of the spent fuel assembly to be processed. The uranium-plutonium isotope composition assumption is conservatively adjusted to make the solution system reactivity smaller, and in this case, the average burnup limit of the spent fuel assembly with different initial enrichment satisfying the uranium-plutonium isotope composition assumption is increased.

13. The method of determining a postulated uranium-plutonium isotope composition of a solution system of a reprocessing plant as claimed in claim 1, wherein, In the step S7, the finally determined uranium-plutonium isotope composition assumption is applicable to nuclear power plant types of different sources, spent fuel assemblies within the limited range of the average burnup limit of the spent fuel assembly with different initial enrichment satisfying the uranium-plutonium isotope composition assumption, and different cooling times after discharge satisfying the average burnup limit of the spent fuel assembly with different initial enrichment satisfying the uranium-plutonium isotope composition assumption.

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