Development of Improved Chord-length Sampling Method and Its Application in Monte Carlo Simulation of Dispersion Fuel

Dispersion fuel is widely used in high-temperature gas-cooled reactor (HTGR), accident tolerant fuel, experimental research reactor, naval nuclear power plant and so on. The chord-length sampling (CLS) method can simplify the geometry modeling of dispersion fuel, which can improve the efficiency. However, traditional CLS can only handle the packing of single particle, and has large error when the packing fraction is high. Aiming to solve these two problems, the improve CLS method was developed in reactor Monte Carlo code RMC, and applied to the fully ceramic micro-encapsulated fuel pin case and HTGR fuel pebble with mixed fuel and poison particles. Results show that the proposed method can handle mixed particles with multiple types, and preserve the accuracy of packing fraction, which provide precise and high efficiency for the critical and burnup calculations.     

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the “HTR-N” project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides.These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the “HTR-N” project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel.These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes.     

改进弦长抽样方法开发及在弥散燃料蒙特卡罗模拟的应用

弥散型燃料广泛应用于高温气冷堆、事故容忍燃料、实验研究堆及核动力舰船等,是重要的燃料类型之一。弦长抽样(CLS)方法可简化弥散燃料几何建模,提高计算效率,然而传统CLS方法只能描述单种颗粒的填充,同时在高体积填充率时误差较大。针对CLS方法的两大问题,本文在自主化堆用蒙特卡罗程序RMC中开发了改进CLS方法,并应用于全陶瓷微胶囊封装燃料棒算例及含毒物颗粒的高温堆燃料球算例。计算结果表明,改进CLS方法可解决多种颗粒混合填充的问题,并且可保证体积填充率的准确性,为弥散燃料的临界及燃耗计算提供了高效、精确的方法。     

Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters — the temperature and burnup — on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 172–179, September, 2007.     

HTGR nuclear fuel based on plutonium oxide microspheres

The results of investigations performed at the conceptual design stage of the development of rod-shaped fuel elements based on spherical plutonium dioxide particles with protective coatings (fuel microelements) for a modular HTGR (GT-MGR type) with a gas turbine are presented. The basic requirements for fuel elements and their components, ensuring that the elements are used effectively in a reactor for weapons plutonium utilization, are formulated. Technological-material-engineering investigations have been performed on UO₂ and PuO₂−x fuel cores. The basic parameters of the technological processes have been determined, and the required setups have been developed and put into operation. Certain physical and mechanical properties of fuel elements and their components without irradiation have been investigated, 5 figures, 3 references. Scientific and Industrial Association “Luch”. Translated from Atomnaya énergiya, Vol. 88, NO. 1, pp. 35–38, January, 2000.     

Irradiation behavior of HTGR coated particle fuel at abnormally high temperature

Irradiation behavior of high temperature gas-cooled reactor (HTGR) coated particles under temperature transient conditions was investigated in accordance with a design-base accident scenario for HTTR, a 30 MWth HTGR under construction at JAERI. One of the scenarios predicts that the fuel temperature of the block-type fuel elements rises to abnormally high temperature by blocking a coolant channel with some foreign substance. For simulating this scenario the fuel compacts incorporating the coated particles were irradiated at normal temperature in three capsules, followed by temperature transient up to a maximum of approximately 2000°C. The post-irradiation examinations, including surface inspection, metrology, ceramography and a measurement of coated particle failure were applied to the fuel compacts to investigate the thermal-transient effect on the fuel integrity. Integrity of the fuel compact was also assessed by an estimation of tangential stress introduced into the compact by the temperature transient.     

Fuel elements based on spherical fuel pellets with a protective coating for enhanced-safety reactors

The principal investigations performed at the Scientific and Industrial Association Luch on the development of fuel elements based on spherical pellets of nuclear fuel with protective ceramic and metallic coatings for HTGR. VVéR, and other types of reactors are reviewed. The main solutions concerning the construction and technical and materials-science aspects of the fabrication of fuel micropellets, fuel microelements, and fuel elements of different modifications (spherical, rod-shaped) are examined. The characteristics of fuel elements and their components at the fabrication and preliminary and reactor testing stages are presented. It is shown that because of additional protective barriers the fuel elements which have been developed effectively confine the fission products and ensure safety. The directions of possible practical utilization of the fuel elements developed for improved enhanced-safety reactors are described. 7 figures, 6 tables, 42 references. Scientific and Industrial Association Luch, Translated from Atomnaya énergiya, Vol. 87, No. 6, pp. 451–462, December, 1999     

Modeling of fuel performance and metallic fission product release behavior during HTTR normal operating conditions

The computer codes PANAMA and FRESCO developed at the Research Center Jülich have been used for the prediction of fuel performance and fission product release behavior during the normal operation of the Japanese High-Temperature Engineering Test Reactor, HTTR. Basis for the calculations was the so-called ‘Standard HTTR Operation Plan’ with a nominal operation time of 660 efpd including a 110 efpd period with enhanced fuel temperatures. Fuel performance model calculations with the PANAMA code have shown that for the temperature distribution given, only a small additional failure fraction is expected. The diffusive release of metallic fission products from the fuel occurs mainly from the central core layers with the maximum temperatures whereas there is little contribution from the upper layer. Silver most easily escapes the fuel. The release data for strontium and cesium also reveal a significant fraction to originate from still intact particles. The comparison with the calculations obtained with the JAERI models has shown a good agreement for the release from the coated particles.     

Dynamic response analysis for fuel assemblies under earthquake and LOCA accidents

The fuel assembly is the main component in 300 MWe PWRs, located in an environment of high temperature, high pressure, irradiation and erosion of the reactor core. Whether or not it can keep functioning during the normal operation of plants, especially under severe accidents such as the earthquake and LOCA, will affect directly the safety of the NPPs. In this paper, by using the general-use computer code ANSYS, a horizontal modal analysis model and a non-linear impact analysis model as well as a vertical non-linear impact analysis model for the fuel assemblies are constructed. The results of the modal analysis are compared with that obtained from tests and other computer codes. Relevant parameters in the non-linear impact calculating models are selected and adjusted by comparing with the shaking table test data of the fuel assemblies. Based on the comparisons and adjustments, the dynamic response calculations of the 1×13 horizontal impact models for Qinshan and CHASNUPP under the earthquake (SSE) and LOCA conditions are performed and corresponding results are given. The response calculations of the vertical impact model for three kinds of postulated LOCA cases are also performed and relevant calculated results are provided.     

Fabrication of the First-Loading Fuel of the High Temperature Engineering Test Reactor

The High Temperature Engineering Test Reactor (HTTR). which is the first high temperature gas-cooled reactor (HTGR) in Japan, attained its first criticality in November 1998. The fabrication of the first-loading fuel started June 1995 and in December 1997, 150 fuel assemblies were completely formed. A total of 66,780 fuel compacts, corresponding to 4,770 fuel rods, were successfully produced through the fuel kernel, coated fuel particle and fuel compact processes. Fabrication technology for the fuel was established through a lot of research and development activities and fabrication experiences of irradiation samples. As-fabricated fuel compacts contained almost no through-coatings failed particles and few SiC-defective particles. Average through-coatings and SiC defective fractions were as low as 2 × 10^–6 and 8 × 10^–6, respectively. This paper describes (1) characteristics of as-fabricated fuel, (2) the experiences obtained from the first mass-production and (3) prediction of irradiation performance of the fuel in the HTTR.     

Irradiation swelling, creep and thermal-stress analysis of LWR fuel elements, computer code isune-2

The irradiation swelling, creep, and thermal-stress analysis of light-water reactor (LWR) oxide (UO₂) fuel elements is analysed. The analysis is based on the basic physical and mathematical assumptions and the experimental data of the fuel and cladding (or canning) materials. In the analysis, the nuclear, physical, metallurgical, and thermo-mechanical properties of the fuel and cladding materials under irradiation environment are examined carefully. The objectives of the paper are mainly (1) to formulate and carry out the irradiation swelling, irradiation creep, and thermal-stress analysis of fuel elements for LWR power reactors, and (2) to develop a computer code which will facilitate the computations for fuel element design, safety analysis, and economic optimization of the power reactors. In a general procedure of the analysis, the irradiation swelling, irradiation creep, temperature distribution, etc. in the fuel and cladding of the oxide fuel elements during the reactor in operation are studied. Some theoretical models and empirical relations (on the basis of accepted experimental data) for irradiation swelling and creep in the fuel and irradiation creep in cladding materials are postulated and developed. Some analytical and empirical relations (based on test results) for heat generation and temperature distribution in the fuel during fuel restructuring are derived. The fuel restructure is, in general, divided into the central void, columnar grain, equiaxed grain, and unaffected grain zones (or regions) after a sufficiently long period for the fuel elements to be irradiated (or operated). From these relations derived for irradiation swelling, irradiation creep, and temperature distribution in the fuel and cladding, together with the well-known strain-stress, incompressibility, compatibility, and stress equilibrium equations, the irradiation swelling, creep, and thermal-stress analysis for the LWR fuel elements can be carried out.From the analytical results obtained, a computer code, ISUNE-2 (which is in the sequence of computer code ISUNE-1 and -1A developed and used previously for liquid-metal fast breeder reactor fuel element design and safety and economic analysis), can be developed. With some reliable experimental data (measured during fuel elements in operation) as input, the computer code may predict various cases of LWR (oxide or carbide) fuel elements in operation. The general scope and resulting contribution of this paper is to provide a realistic analysis and a reliable operating LWR fuel element code for use by nuclear power utilities to predict the fuel element behavior in power reactors. The fuel element design, safety analysis, and economic optimization depend largely on the fuel element behavior in the power reactors.     

Numerical simulation of three-dimensional thermal-hydraulic behavior for HTTR (High Temperature Engineering Test Reactor)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are now in progress in order to verify the inherent safety features and to improve safety designing and analysis technologies for future HTGR (high temperature gas-cooled reactor). Coolant flow reduction test is one of the safety demonstration tests for the purpose of demonstration of inherent HTGR safety features in the case that coolant flow is reduced by tripping of helium gas circulators. If reactor core element temperature and core internal structure temperature during abnormal events are estimated by numerical simulation with high-accuracy, developed numerical simulation method can be applied to future HTGR designing efficiently. In the present research, three-dimensional in- and ex-vessel thermal-hydraulic calculations for the HTTR are performed with a commercially available thermal-hydraulic analysis code “STAR-CD^®” with finite volume method. The calculations are performed for normal operation and coolant flow reduction tests of the HTTR. Then calculated temperatures are compared with measured ones obtained in normal operation and coolant flow reduction test. As the result, calculated temperatures are good agreement with measured ones in normal operation and coolant flow reduction test.     

Verification of the BONUS module integrated into the SOCRAT code

The results of a verification of the BONUS method as regards predictions of the time dependence of the mass and activity of fission products produced in thermal reactors are presented. Different standard regimes of fuel irradiation in VVER-1000 are examined, and the results calculated using the autonomous version of the BONUS code and as a module integrated into SOCRAT code are compared with the results obtained using other codes, including precision program complexes. On the whole, the calculation shows good agreement between BONUS and alternative codes; the standard deviation variance is 6–13.5% for fission product mass and activity, which is comparable to the discrepancies between different variants of the alternative calculations themselves.     

Computational investigation of the temperature and geometric characteristics of VVÉR fuel in the Kozlodui nuclear power plant (Bulgaria) under normal operating conditions

A method of performing stationary thermomechanical calculations of VVéR-440 and-1000 fuel elements, using the TRANSURANUS computer code to obtain the dependence of the temperature and radius of the fuel elements on the lineal power ensity and burnup, is described. These dependences are intended for use in neutron-physical calculations of the VVéR reactor at the Kozlodui nuclear power plant in stationary and transient regimes. The results obtained with this computer program are compared with calculations performed using the certified TOPRA-s code. The comparison shows reasonable agreement between the results of calculations of the fuel temperature. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 336–342, November, 2006.     

Taking account of the effect of the curvature of VVÉR-1000 fuel assemblies on the power of the fuel elements

A comparative analysis is made of the deterministic and statistical methods of taking into account the effect of the curvature of VVéR-1000 fuel assemblies on the power of fuel elements. The fuel-element distribution of the energy release in the core for any random distribution of the gaps between the fuel assemblies is simulated, using the MEX code, on the basis of precise calculations (MCU code) and design calculations (BIPR-7 and PERMAK codes). The Monte Carlo method (Zazor code) was used to model the nominal density distribution of gaps in the core for different degrees of curvature of the fuel assemblies. It is shown that the power gain, obtained for the fuel elements by the probabilistic-statistical method, due to the curvature of the fuel assemblies is smaller and makes it possible to substantiate core safety with large perturbations, in contrast to the deterministic “maximum gaps near the most-energy stressed fuel element” method. 5 figures, 1 table, 3 references. Special Design Office “Gidropress.” Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 210–213, September, 1999.     

U_3Si_2Al弥散型燃料的辐照肿胀研究

介绍了U3Si2 Al弥散型燃料的辐照肿胀机理。将弥散型燃料的芯体视为连续基体中的微型燃料元件 ,应用裂变气体的行为机理描述燃料相中的气泡形成过程。研究结果表明 :燃料相的肿胀引起燃料颗粒和金属基体之间的力学相互作用 ,金属基体能抑制燃料颗粒的辐照肿胀。在一定辐照条件下 ,本模型对燃料元件辐照肿胀的预测值与测量值相符     

Swelling behavior of SM reactor fuel elements under irradiation conditions with elevated parameters

A comprehensive investigation of SM reactor fuel elements with 20% higher nuclear fuel load after reactor tests with elevated parameter values is performed. The radial swelling behavior of fuel elements and their kernels as a function of the fission products concentration, fission density, heat-flux density, and test temperature is presented. The swelling behavior of fuel particles in the transfer cross section of a fuel element kernel is examined. The dependences found make it possible to evaluate quantitatively the swelling of fuel elements for different values of the thermophysical parameters and choose safe values for the parameters taking account of the maximum swelling of the fuel elements and the technical possibilities. Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 158–162, March, 2009.     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

Release Behavior of Fission Gas from Coated Fuel Particles under Irradiation

TRISO coated fuel particles for HTGR were irradiated by two sweep gas capsules in order to study the release behavior of the fission gas and try to predict the failure fraction of the particles on the basis of the measurement. For verification of the predicted failure fraction, post irradiation examination was conducted, and failure fraction in a visual inspection and acid leaching fraction were measured. Agreement between the predicted failure fraction and the acid leaching fraction was good for these samples except one. From the release behavior from the intact particles, in-pile diffusion coefficients of Kr in LTI-PyC were estimated and expressed as D=(2.9–6.0)×10⁴exp(-2.55×10°/RT) (cm²/s), where R ids the gas constant (=8.314 J/K) and T the absolute temperature. It was recognized that the release from failed particles was controlled by diffusion at 1,600°C and that from intact particles, predominantly by recoil at 1,400°C.