船用反应堆堆芯精细功率重构计算

针对现有船用反应堆安全分析仿真软件不能计算堆芯精细功率分布这一缺陷,开展了堆芯径向和轴向精细功率分布重构计算和分析。采用三次样条插值法对轴向精细功率进行重构计算,采用双线性和双三次插值法对径向精细功率进行重构计算,并与采用细网差分的专业物理程序的计算结果进行比较。结果表明,本工作精细功率重构计算简单、可靠,有较高的精度,对船用反应堆安全运行分析和监督管理具有重要的参考价值。     

Whole Core Calculations of Power Reactors by Use of Monte Carlo Method

Abstract

Whole core calculations have been performed for a commercial size PWR and a prototype LMFBR by using vectorized Monte Carlo codes. Geometries of cores were precisely represented in a pin by pin model. The calculated parameters were k _eff, control rod worth, power distribution and so on. Both multigroup and continuous energy models were used and the accuracy of multigroup approximation was evaluated through the comparison of both results. One million neutron histories were tracked to considerably reduce variances. It was demonstrated that the high speed vectorized codes could calculate k _eff, assembly power and some reactivity worths within practical computation time. For pin power and small reactivity worth calculations, the order of 10 million histories would be necessary. It would be difficult for the conventional scalar code to solve such large scale problems while the present codes consumed computation time less than 30 min for a PWR and 1 hour for an LMFBR. Required number of histories to achieve target design accuracy were estimated for those neutronic parameters.     

轻水堆燃料组件计算程序包TPFAP

TPFAP是一个同时适用于PWR和BWR的穿透几率法燃料组件燃耗计算程序包。它首先利用碰撞几率方法在库能群结构下完成三区或四区圆环几何的栅元输运计算。载钆燃料棒或硼棒可燃毒物栅元的有效吸收截面由微燃耗程序CMB产生,两维穿透几率法组件计算是在(x,y)几何下进行。基模计算用来考虑中子泄漏修正。根据反应率等效,计算组件等效扩散参数。在每一燃料棒和可燃毒物棒进行燃耗计算,TPFAP给出每一燃耗步的组件和栅元少群截面、功率分布,提供核设计和安全分析所需参数。     

Neutron-kinetic code validation against measurements in the Moscow V-1000 zero-power facility

Measurements carried out in an original-size VVER-1000 mock-up (V-1000 facility, Kurchatov Institute, Moscow) were used for the validation of three-dimensional neutron-kinetic codes, designed for VVER safety calculations. The significant neutron flux tilt measured in the V-1000 core, which is caused only by radial-reflector asymmetries, was successfully modeled. A good agreement between calculated and measured steady-state powers has been achieved, for relative assembly powers and inner-assembly pin power distributions. Calculated effective multiplication factors exceed unity in all cases. The time behaviour of local powers, measured during two transients that were initiated by control rod moving in a slightly super-critical core, has been well simulated by the neutron-kinetic codes.     

Pin Power Reconstruction Methods of the Few-Group BWR Core Simulator NEREUS

An improved pin power reconstruction method has been incorporated in the few-group nodal BWR core simulator NEREUS, which is based on the analytic polynomial nodal method. With the analytic polynomial nodal method, accurate node surface fluxes are available, which are used later to reconstruct pin powers. The intranodal homogeneous thermal flux is corrected using a semi-empirical proportional relation between surface transition components of the homogeneous and heterogeneous fluxes. This correction method is effective for BWR calculations, especially for controlled assemblies or mixed loading of off-set assemblies. A unified model accounting for effects of spectral histories, caused by spectral interactions between fuel assemblies and the control blade insertion, was also developed for intranodal burnup correction. The change in pin powers due to the control blade history can be predicted well, without laborious assembly depletion calculations with control blade insertion. R-factors used in critical power ratio calculations are also reconstructed from the pin powers. The NEREUS pin power reconstruction method was verified against heterogeneous multi-assembly depletion calculations.     

Results on the DECD/NEA C5G7-MOX benchmark obtainedwith the discrete ordinates code dort

The well-known multigroup Discrete Ordinates code DORT has been employed to perform calculations on the DECD/NEA C5G7-MOX benchmark. The participants were required to supply the effective multiplication factor as well as the pin power distribution of the problem specification. We show our submitted results to be well consistent with the reference solution, which was produced by Monte-Carlo calculations. Furthermore, we point out some improvements being made on the computational procedure, which give rise to an additional gain in accuracy.     

Accuracy evaluation of pin exposure calculations in current LWR core design codes

The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at the Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

The shut-down of the Barsebäck 1 BWR: A unique opportunity to measure the power distribution in nuclear fuel rods

Reactor poolside measurements of gamma radiation specific for the fission product ¹⁴⁰La (1596 keV) have been used for an experimental determination of axial power distributions in 55 nuclear fuel rods irradiated in the Barsebäck 1 BWR nuclear power plant. The measurements take advantage of the unique situation of a very short last reactor cycle of only three months due to the out-phasing of the reactor unit at November 30 1999. ¹⁴⁰La whose decay is controlled by the mother nuclide ¹⁴⁰Ba with the half-life 12.75 days reflects an average power distribution, representative for the latest weeks of core operation (in this case basically during November 1999). The measured intensities have been transformed into a 25 nodal representation to allow a precise and direct comparison with the corresponding calculated power distribution. The 55 rods were selected from two different fuel assemblies with average burn-ups of 1.9 and 9.7 MWd/kgU, respectively (that is one fresh bundle and one slightly more than one cycle bundle). The stability and the linearity of the measurement system were evaluated. The linearity was checked using the two-source method. The stability was checked by recurrent measurements on a reference fuel rod. The results have been used in the validation of the pin power reconstruction model of Westinghouse 3D core simulator POLCA-7. The deviation between measured and calculated ¹⁴⁰Ba concentration (expressed as radial error) is typically a few percent on rod level. Results indicate that also Gd-rods are properly modelled over a broad range of conditions. It is indicated that predictions for fuel rods in their first month of operation are less accurate than for the rest of the rods.     

The irradiation test of inert-matrix fuel in comparison to uranium plutonium mixed oxide fuel at the halden reactor

Conservative modelling for pin layout shows that the relatively low thermal conductivity of Inert-Matrix Fuel (IMF) causes higher temperatures and therefore higher fission gas release than in uranium plutonium mixed oxide (MOX). According to neutronic calculations, performance differences will also arise from different evolutions of the respective radial power and burnup distributions. Modelling of these effects as well as a 10% greater production of Xe in the thermal spectrum of the Halden reactor is well within the capabilities of appropriate codes. Some of the data and models used for the pre-calculations are preliminary and will be revised after the first experimental data have become available.     

Statistical evaluation of pin gaps in fuel clusters for fugen

The computer program STGAP has been developed to estimate pin gaps in a fuel assembly for FUGEN. The program optionally computes the probable distribution of the pin gap between any adjacent pair of fuel pins, either at a desired location in an assembly or longitudinally averaged over the total effective length of a pin, based on the measured manufacturing and assembling tolerances in geometrical dimensions and mechanical properties of all the independent elements composing a fuel assembly. It also correlates the computed fuel gap distribution with the minimum critical heat flux ratio in the corresponding local subchannel. Sample calculations were performed for the probable distributions of the pin gaps between pairs of adjacent fuel pins in the outermost layer of a FUGEN fuel assembly using the program and satisfactory agreement was obtained with the corresponding measured distributions.     

Analytical Study on Detailed Void Distributions Inside BWR Fuel Bundle under Turbine Trip Event Considering Time-Dependent Pin Power Distributions

A one-way coupling system between the plant simulator TRAC/BF1-ENTRÉE and the subchannel code with the improved cross flow model, NASCA, has been developed. Based on a scenario of turbine trip tests in the Peach Bottom Unit 2, the wide and rapid reactivity insertion transient induced by the system pressure rise was calculated. The pin power distribution in hot bundles was re-constructed considering heterogeneity of the fuel bundle. When the neighboring control blade is withdrawn, NASCA predicted that the steady-state bundle exit void distribution was nearly flat with regardless of the pin power distribution. However, void distributions in the middle and lower bundle regions became complicated depending on the pin power and the two-phase flow regime in each subchannel. The pin power distribution rapidly changed according to traveling of control blades. However, influence in the void distribution was delayed and damped due to the fuel heat conduction. The detailed void distribution under transient events has an impact on the location of dryout. It has been shown that the coupling calculation including the subchannel code is useful in understanding transition of the detailed void distribution depending on the pin power and the two-phase flow regime.     

In-core fuel management practice in HANARO

The HANARO (High-flux Advanced Neutron Application Research reactOr) is a newly created research reactor. Its initial criticality was achieved on February 8, 1995. In its design stage, the HANAFMS, which is the nuclear analysis and fuel management code system for the HANARO, was verified using the results from the MCNP4A full core model because there was no similar research reactor in the world. It was needed to verify the HANAFMS with reactor physics experiments, which were performed during the reactor commissioning and power operation. The calculated results for the criticality, power distribution and control absorber rod (CAR) worth were compared with the measured ones. In the criticality calculation, the clean and depleted cores were applied and in the comparison of power distribution, the gamma scanning data of the fuel assemblies were used. The CAR worth was calculated following the measurement positions and then compared with the measurements. The calculated results for verifying the HANAFMS are in good agreement with the measured ones.     

快堆燃料组件内精细功率分布的计算

堆芯热通道因子是堆芯热工设计及安全分析的一项重要参数,确定热通道因子需用中子学计算给出较准确的燃料组件内元件棒功率分布。在三维六角形几何节块扩散理论基础上,使用多项式重构的方法计算节块内中子通量密度分布和功率密度分布。针对快堆六角形燃料组件的特点,用小六角形积分的方法计算组件内元件棒功率,得到组件内各元件棒功率分布。在NAS程序基础上,编制了元件棒功率分布计算模块NAS PIN。通过与蒙特卡罗程序的校验可发现,二者计算结果符合较好,计算精度可满足工程设计的需要。     

Investigation of axial power gradients near a control rod tip

Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations.     

Three-dimensional pin power reconstruction for the axially heterogeneous region in BWR

A three-dimensional pin power reconstruction method was proposed and verified by applying to the axially heterogeneous region problem of the BWR core calculation. Because the production assembly calculation has been carried out by two-dimensional deterministic calculation methods like current coupling collision probability or the method of characteristics, it has been difficult to predict the detailed three-dimensional heterogeneous pin power distribution of the axially heterogeneous region. Consequently, only radial intranodal homogeneous power distributions have been considered, and axial intranodal homogeneous power distributions have not been considered in the estimation of linear-heat-generation-ratio at common BWR core calculations.     

焊接残余应力有限元分析技术研究

以岭澳核电站控制棒驱动机构耐压壳Ω环焊接修复为例,应用ANSYS有限元生死单元技术模拟焊接流程,计算出焊接后残余应力的分布,绘制出残余应力分布曲线,并与美国WSI公司的计算结果进行对比分析.结果表明,本课题的计算结果与美国焊接公司(WSI公司)一致.因此,焊接残余应力有限元分析技术可以用于反应堆耐压壳焊接修复评价.     

SIMS characterisation of actinide isotopes in irradiated nuclear fuel

Numerical codes are used to calculate the inventory of plutonium and minor actinides produced during irradiation in industrial power reactors. Secondary ion mass spectrometry (SIMS) is a useful method for validating theses inventory calculations especially with regards to isotopic characterisation. Isotopic ratios measured by SIMS on the radius of an UO₂ irradiated fuel pellet are presented and compared with calculated values in this paper. The SIMS results tally well with the calculations, as both show an increase in the actinide concentration on the pellet periphery due to the rim effect.     

Integral Data Test of HENDL1.0/MG with Neutronics Shielding Experiments(Ⅱ)

The multi-group working nuclear data library HENDL1.0/MG is numerically tested with a series of existent benchmark spherical shell experiments (Si, Cr, Fe, Cu, Zr and Nb) by calculations using the multi-functional neutronics code VisualBUS. The ratio of calculated/measured neutron leakage rates and the neutron leakage spectra are presented in tabular and figural forms.The results from the calculations with the code ANISN and IAEA data library FENDL2.0/MGwere also included for comparison, where the origination of the data used is different from that of HENDL1.0/MG.