Subchannel analysis of supercritical light water-cooled fast reactor assembly

Subchannel analyses have been carried out for supercritical water-cooled fast reactor fuel assembly. Peak cladding surface temperature difference arising from subchannel heterogeneities have been calculated by using the improved subchannel analysis code STARS and was evaluated to be about 18.5 °C. Several suggestions have been also made for reducing the PCST difference arising from channel heterogeneity. Influences of local power peaking on deflection of cladding surface temperature are explained with pin power distribution taken from core depletion calculation in this paper. Maximum cladding surface temperature at nominal condition is evaluated to be 645.3 °C over the cycle. Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Maximum statistical design uncertainty of PCST is calculated to be 31 °C and is in a good agreement with that from RTDP method. Influence of downward flow in seed region on system sensitivity is investigated by improved Monte-Carlo thermal design procedure. Limiting thermal condition of MCST is 681 °C (650 °C of nominal + 31 °C) within 95/95 limit for SWFR.     

Oxygen redistribution in fast reactor oxide fuel

Out-of-pile tests were carried out in order to investigate the oxygen redistribution in uranium-plutonium mixed oxides exposed to a thermal gradient. In hypostoichiometric oxide fuel the oxygen migrates towards the low temperature region of the pellet and in hyperstoichiometric fuel the oxygen migrates in the opposite direction. The oxygen transport is explained on the basis of solid-state thermal diffusion and occurs via vacancies and interstitials. It has been shown that the heats of oxygen transport are a function of plutonium and uranium valencies for hypo- and hyperstoichiometric oxides, respectively. The experimental results allowed to construct a practical example in which oxygen profiles in fuel pins were calculated as a function of initial stoichiometry and burnup.     

Restructure behavior analysis in fast breeder reactor MOX fuel by X-ray CT

Both high- and low-density MOX fuel pellets of uranium and plutonium oxides were irradiated in the experimental fast reactor JOYO. After irradiation, these fuel pellets were examined by X-ray CT and their irradiation behavior was evaluated for formation of the central void. In particular, the central void size and temperature of fuel pellets at the beginning and end of irradiation were analyzed. The central voids in the low-density fuel pellets were bigger than those of the high-density fuel pellets at the same linear heating rate (LHR), and the threshold LHR and temperature at which the central voids were formed were lower than those of the high-density fuel pellets. It was understood from these results that the irradiation behaviors of high- and low-density fuel pellets were different.     

Experimental techniques for liquid metal cooled fast breeder reactor fuel assembly thermal/hydraulic tests

Sensors and methods of experimental measurement being employed in fast breeder reactor fuel assembly tests are reviewed. Such tests are being carried out in sodium, water and air environments. In sodium tests direct measurement of bundle performance parameters such as temperature, flow, pressure, boiling inception, and void fraction are being performed. Development of improved instrumentation is needed for reliable fast-response, high-temperature pressure detection and small, more readily interpretable, void detectors. Water and air environment tests are being undertaken to measure parameters used in models which predict design behavior in sodium. Parameters being measured are subchannel average velocity, local axial and transverse velocities, wall shear stress, salt and other tracer concentrations, and turbulence parameters. Adequate techniques exist for measurement of each of these parameters.     

Optimizing fuel assembly placement in a research reactor

Institute for Nuclear Research, Academy of Sciences of the Ukrainian SSR. Translated from Atomnaya Energiya, Vol. 70, No. 4, pp. 257–259, April, 1991.     

Molten fuel dispersal in hypothetical fast reactor accidents

Some aspects of molten fuel dispersal in hypothetical fast reactor accidents are considered, ranging from the two-phase flow fluid equation forms appropriate to modelling molten fuel dispersal to an analytical self-similar solution as a function of space and time for the dispersal of a two-phase molten fuel/fission gas mixture by fission gas pressures. The analytical solution provides both scaling laws for fuel dispersal velocities as a function of gas content and a solution with which to check code results. A discussion of the COMCYL program for molten fuel dispersal in a sodium volded channel is used to illustrate the types of problems that need to be tackled in a molten fuel dispersal program and typical results obtained from the application of COMCYL to hypothetical loss-of-coolant flow accidents are presented.     

On the failure of fast reactor fuel pins

The prediction of the timing and position of fuel pin failures is an important task in the modelling of fast reactor fuel behaviour. The range of processes that can provoke failure of fast reactor fuel pins in normal operating conditions and during hypothetical accidents is reviewed. Some of the mechanisms of failure are examined in more detail and the effect of hot spots and local stress concentrations is discussed. A review of failure criteria used in fast reactor fuel pin codes is given elsewhere, but the difficulties in applying various types of criteria are examined. Some discussion is also given on probabilistic approaches. Recommendations are given for a future approach to the problem of failure prediction, resolving the dilemma between inadequate empirical criteria and over-complex physically based approaches.     

Transient thermal-hydraulic analysis for reactor cores

A new computational method is presented for a transient, thermal-hydraulic, multichannel analysis. The method is developed based on the concept of artificial compressibility to preserve the elliptic character of the reactor core flow in order to satisfy the realistic pressure boundary conditions, and to account for the discontinuities of the emprical correlations simulating the flow resistances. The computer code (RETSAC) developed by implementing the method presented in this paper can be categorized as a fourth generation multichannel computer code. This new computer code has been compared with the widely used marching techniques, such as COBRA IIIC (the third generation). The numerical results clearly indicate the situations in which the marching technique may or may not be appropriate. Furthermore, the RETSAC computer code can calculate various normal or off-normal reactor core flows which the third generation codes could not handle without a substantial increase of computer time.     

Partial flow blockage analysis of the hottest fuel assembly in SNCLFR-100 reactor core

In this paper, we perform an unprotected partial flow blockage analysis of the hottest fuel assembly in the core of the SNCLFR-100 reactor, a 100 MW_(th) modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China. The flow blockage shall cause a degradation of the heat transfer between the fuel assembly and the coolant potentially,which can eventually result in the clad fusion. An analysis of core blockage accidents in a single assembly is of great significance for LFR. Such scenarios are investigated by using the best estimation code RELAP5. Reactivity feedback and axial power profile are considered. The crosssectional fraction of blockage, axial position of blockage,and blockage-developing time are discussed. The cladding material failure shall be the biggest challenge and shall be a considerable threat for integrity of the fuel assembly if the cross-sectional fraction of blockage is over 94%. The blockage-developing time only affects the accident progress. The consequence will be more serious if the axial position of a sudden blockage is closer to the core outlet.The method of analysis procedure can also be applied to analyze similar transient behaviors of other fuel-type reactors.     

A perspective on fast reactor fuel cycle in India

The growing energy needs of India can be fulfilled only by judicious mix of all the fuel resources. It is possible to achieve energy security and sustainability through the introduction of fast reactors in an expeditious manner and closing the fuel cycle. This approach is inevitable in view of the limited uranium resources in India. The Fast Breeder Test Reactor (FBTR) built by India uses mixed carbide as fuel and the 500 MW(e) Fast Breeder Reactor Project (PFBR), to be operational in 2010, will use mixed oxide as fuel. It has also been decided that fast reactors beyond 2020, with enhanced safety features and having better economy, will use metallic fuel. Having successfully operated FBTR with carbide fuels, we need to develop the fuel cycles for both the mixed oxide fuel in the near future and the metallic fuel expeditiously. The progress achieved so far and the plans for implementation are discussed in this paper.     

Cesium-uranium-oxygen chemistry in uranium-plutonium oxide fast reactor fuel pins

The Cs-U-O system has been reinvestigated in light of recently reported thermodynamic data for Cs₂U₄O₁₂ and recent phase data showing the existence of a new Cs-U-O compound with an $\text{O}\text{(U + Cs)}$ atom ratio less than that of Cs₂UO₄. Our experiments have confirmed the existence of a new phase, allowed the formula Cs₂UO_3.56 to be assigned, and generated thermodynamic data for this new compound. This new phase exists only at oxygen potentials that are too negative to be encountered in uranium-plutonium oxide fast reactor fuel pins. The compound Cs₂UO₄ appears to be the most likely one to be formed, with the formation occurring at the fuel-blanket interface.     

Plutonium redistribution in fast reactor mixed oxide fuel pins

Some redistribution effects of uranium and plutonium, caused by thermal diffusion and evaporation-condensation processes in mixed oxide fuels, are discussed by means of autoradiographs of sections of fuel pins irradiated in the fast flux of the RAPSODIE reactor. The change in the stoichiometric state as a function of burnup and the radial distribution of oxygen are described and their influence on the redistribution processes is discussed. A model and suitable data are given to calculate redistribution effects on the basis of thermal diffusion in fast reactor fuels. In fuel pins with power ratings of 500 W/cm and 600 W/cm the enrichment of plutonium around the central cavity produces an increase in the central temperature of about 100°C and 250° C, respectively.     

Comparing VVR-M reactor fuel assembly heat technological capabilities

Translated from Atomnaya Énergiya, Vol. 67, No. 2, pp. 97–100, August, 1989.     

虚拟装配技术及其在中国快堆技术中的应用展望

阐述虚拟装配技术的主要研究内容和现状,指出该技术的关键内容并分别进行分析.针对中国快堆技术发展中的机械设备装配现状,分析虚拟装配技术在快堆技术中应用的重要意义,提出了该技术在快堆技术应用中的发展目标.     

小型模块化熔盐快堆燃料管理初步分析

由于燃料随熔盐流动的特性以及可以进行在线添料与处理的特点,液态燃料熔盐堆的燃耗分析与燃料管理和传统固态燃料反应堆有很大不同,需要针对液态燃料熔盐堆的特点重新开发燃耗分析与管理程序。本文针对液态燃料熔盐堆的熔盐流动特性以及在线添料与处理功能,基于MCNP5和ORIGEN2.1燃耗耦合程序,开发了适用于液态燃料熔盐堆的燃料管理程序,并应用于一种小型模块化熔盐快堆的燃料管理和分析,对比分析了5种不同运行方案以及分批在线添料情况下,运行30年期间keff的变化情况及重要核素的演化情况。计算结果表明,采用不断调整添料率的连续在线添料运行方案和固定批量添料的运行方案,都可以让小型模块化熔盐快堆维持运行在一个较小的keff波动范围之内。开发的燃料管理程序适用于液态燃料熔盐堆的研究,同时可以为液态燃料熔盐堆的设计及燃耗管理和分析提供有价值的参考。     

Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

Fast reactors based on thorium fuel have enhanced inherent safety. Fluoride salt performs well as a coolant in high-temperature nuclear systems. In this paper,we present a reference core for a large fluoride-salt-cooled solid-fuel fast reactor(LSFR) using thorium–uranium fuel cycle. Neutronics physics of the LSFR reference core is investigated with 2D and 3D in-core fuel management strategy. The design parameters analyzed include the fuel volume fraction, power density level and continuous removal of fission products with 3D fuel shuffling that obtains better equilibrium core performance than 2D shuffling. A self-sustained core is achieved for all cases,and the core of 60% fuel volume fraction at 50 MW/m~3 power density is of the best breeding performance(average breeding ratio 1.134). The LSFR core based on thorium fuel is advantageous in its high discharge burn-up of 20–30% fissions per initial heavy metal atom, small reactivity swing over the whole lifetime(to simplify the reactivity control system), the negative reactivity temperature coefficient(intrinsically safe for all cases) and accepted cladding peak radiation damage. The LSFR reactor is a good alternative option for the deployment of a self-sustained thorium-based nuclear system.     

Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

A synchrotron radiation source called TURKAY is proposed as a sub-project of the Turkish Accelerator Center Project. The storage ring of TURKAY is a low emittance synchrotron and the radiation ranges between 0.01 and 60 keV can be generated from the insertion devices and bending magnets placed on it. The injector system of the facility will mainly consist of a 150 MeV linac and full energy booster. In this study, we present design considerations and beam dynamics studies of the pre-injector linac and booster ring for TURKAY.