燃耗截面基本库对输运燃耗耦合计算的影响分析

基于蒙特卡罗中子输运程序和ORIGEN2点燃耗程序的蒙特卡罗输运燃耗耦合计算方法应用广泛。但现有评价库中子连续截面的核素个数远小于燃耗计算涉及到的核素数量,即通过输运计算得到的燃耗截面不足以完全替代燃耗计算的基本库。采用经过栅元验证的蒙特卡罗燃耗程序MCBMPI,对最新的VERA燃耗计算基准题进行验证计算,对比分析不同的燃耗截面基本库对输运燃耗计算的影响。分析结果表明:1)在实际应用中尽量不要采用典型热中子截面库,会带来较大偏差;2)在燃耗计算核素替换较多的情况下,对该基准题而言,选取典型压水堆基本库还是典型快堆基本库,对结果影响不大,二者keff偏差在8‰以内,燃耗末期235U偏差在4‰以内,135Xe偏差在5‰左右;3)建议选取与研究对象能谱相近的基本库。     

Development and validation of depletion code system IMPC Burnup for ADS

Depletion calculation is important for studying the transmutation efficiency of minor actinides and longlife fission products in accelerator-driven subcritical reactor system(ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEAADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover,the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.     

Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor–corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP–ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems’ k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.     

A perturbation theory for the non-linear system of neutron transport and burnup equations

A perturbation theory for use in nuclear reactor burnup analysis is derived. An important characteristic function is defined, and the related adjoint equation is obtained simply by using the variational principle. The adjoint matrix operator is evaluated directly from its defining differential expression. Responses at the end of cycle due to changes in initial material inventory, in nuclear data, and in power demands can be calculated using the previously determined forward and adjoint solutions. The method has additional applications, notably in selecting beginning of cycle conditions so as to achieve a particular end of cycle condition.     

基于CAD技术的特征线中子输运计算程序开发

特征线方法(Method of Characteristics,MOC)能否应用于复杂几何关键在于能否将特征线方法与有效的几何处理方法结合起来。本文在菱形差分特征线理论基础上,基于FDS团队自主研发的核与辐射输运计算自动建模软件MCAM的几何处理引擎,研发了基于CAD技术的特征线中子输运计算程序,并利用相关基准例题对程序进行了数值验证,其结果与参考值吻合良好,表明本文方法和程序的可行性、正确性与可靠性。     

基于离散纵标输运计算方法的三维燃耗程序发展研究

为了精确描述和分析具有强烈各向异性中子注量率空间分布的反应堆燃耗过程,本文实现了三维SN 输运计算与燃耗计算的耦合,发展了相应的三维输运燃耗耦合计算程序.该程序系统采用接口程序自动耦合三维SN输运计算程序和同位素燃耗计算程序的方法实现对三维中子学计算模型的精细燃耗计算,获得燃料同位素成分、燃耗反应性、中子注量率空间分布等参数随燃耗时间的变化量.采用IAEA 基准校核例题对程序系统进行了校核,计算结果初步证明了所开发的三维燃耗程序系统的正确性.     

一维特征线法中子输运程序开发及验证

为探讨两维/一维综合法堆芯分析方法,本文基于特征线法研制了一维中子输运程序--PEACH-1D.不同于通常的平源近似特征线方法,PEACH-1D可对子区的中子源项作线性近似;程序运用指数函数插值表和渐近源外推技术来加速计算过程.相关数值结果表明,PEACH-1D具有很高的计算精度和效率,线性源近似的特征线法具备处理较粗网格的能力,值得推广.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Neutronics and burnup analysis of an accelerator-based tru-nitride fuel transmutation system with the ATRAS code

The neutronics and burnup analyses of an accelerator-based transmutation system with tungsten target and TRU-nitride fuel were performed with a newly developed code system named ATRAS (Accelerator-based Transmutation Reactor Analysis System). The ATRAS code is an integrated code system which can perform the hadronic cascade process above 20 MeV and neutron transport and core burnup process below 20 MeV with the spallation neutron source.

The specifications of the transmutation system are investigated. The core consists of the central spallation target region and the surrounding TRU-mononitride fuel region. The core is driven by protons at an energy of 1.0 GeV. This system was also proposed as a benchmark problem in the “OECD NEA/NSC Benchmark on Physics aspects of Different Transmutation Concepts”.

According to the calculation results by the ATRAS code, higher power density and transmutation rate were achieved with nitride fuel, and the neutron spectrum was slightly harder than that of the metallic fuel system. The burnup calculation for thermal power 800 MW was also performed with the ATRAS code. It is shown that about 300 kg of TRU are transmuted annually.     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。     

A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

Utilization of Mixed Uranium–Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse ‘AVVER-1000LEUandMOXAssemblyComputationalBenchmark’ and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group ‘JEFF31GX’ cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k_∞). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium–Gadolinium (UGD) pin, fission rate distributions in UGD, UO₂ and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.     

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample.     

Improvements and validation of the transient analysis code MOREL for molten salt reactors

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.     

Development of three-dimensional reactor analysis code system for accelerator-driven system,ADS3D and its application with subcriticality adjustment mechanism

In order to perform the parametric survey for an accelerator-driven system (ADS) core with the subcriticality adjustment mechanism, a new calculation code system, ADS3D, was developed on MARBLE which is a comprehensive and versatile framework for reactor analysis. The application of ADS3D was also demonstrated on the neutronics design of ADS operated by control rod (CR) movement. Through the neutronics calculation, it was shown that the maximum proton beam current was decreased from 20.5 to 11.6 mA due to the switch from beam-operated to CR-operated core.     

基于蒙特卡罗方法与指数欧拉法耦合的反应堆三维燃耗计算程序的研发

燃耗计算在反应堆设计、分析研究中起着重要作用.一维、二维耦合燃耗程序因其几何限制难以满足先进反应堆精细设计分析的要求.本文研发了基于蒙特卡罗方法与指数欧拉法耦合的三维燃耗程序.程序采用编写耦合MCNP与FISPACT的接口程序的方法,结合了MCNP处理复杂几何能力强,FISPACT计算核素全面、能谱多样的特点,实现了考...     

Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries.     

Preparation and verification of mixed high-energy neutron crosssection library for ADS

An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.     

Analytical solution for the Feynman-Alpha formula for ADS with pulsed neutron sources

The theory of Feynman-alpha measurements is elaborated for the case of a “stochastically pulsed” subcritical system. The corresponding physical situation is when a pulsed neutron source is used, and no synchronisation between the start of the measurement time gate and the pulsing is made. This is the case in the European Community supported research project MUSE.

The solution to the Feynman-alpha formula was obtained for such a case through complex function techniques in an analytical form by Laplace transform and residue calculus. The final expression is a smoothly regular function with a simple periodic modulation. It consists of a Feynman-curve corresponding to a stationary source, plus an infinite sum of periodic sine functions squared. The series converges as 1/n⁶ with the summation index n, thus in practice two or three terms are sufficient for a high accuracy quantitative result. This few-term representation amounts to a compact closed form analysis solution. Such a solution is well suitable for use in the determination of the subcritical reactivity from measurements, in contrast to the case of deterministic pulsing (measurement start synchronized with pulsing), where no simple solution is available, and where no explicit relationship between the continuous and pulsed forms of the Feynman-alpha exists.