Fast reactor core deformation due to stainless steel swelling and thermal bowing with irradiation enhanced and thermal creep

In order to establish a technique which will properly account for the LMFBR core deformation under various reactor operating conditions, two separate computer programs have been developed, and coupled together. The first of the two, “FEMCRP”, is a finite element creep analysis code for a long, straight, hexagonal tube. The second program owes its origin to BOX-V developed by ANL, but was later modified considerably by the authors to include the effect of stainless steel swelling. The computer program, HICODEM, is the cross between the two. It can analyze the equilibrium configuration of the core consisting of fuel subassemblies of various burnup histories, with corresponding degree of swelling and creep. The code can assume a mechanical clamping on the periphery of the core during the power operation, and after a desired length of time, it can bring the entire core to the refueling temperature and unclamp the constraint, the resultant equilibrium core configuration being successively obtained.     

A CFD simulation process for fast reactor fuel assemblies

A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-? and SST (Menter) k-ω were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.     

Online embedded integrated scanning system for identifying failed fuel sub assemblies in fast reactor

The physical integrity of the fuel in fast reactor is of utmost concern for the healthiness of the reactor and operating people. Hence details of the failed fuel location in the core shall be determined at the earliest, to minimize reactor down time and radiation exposure. In the present reactor under construction, i.e., 500 MWe Prototype Fast Breeder Reactor (PFBR), a system for failed fuel identification, was proposed. The system follows a novel scheme to locate the failed fuel using failed fuel location module along with necessary instrumentation and control. This paper details out the scheme followed.     

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Abstract

Preliminary studies of used fuel generated in the US Department of Energy's Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg^–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.     

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.     

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。     

Core and fuel cycle for an advanced sodium-cooled fast reactor

The development of BN-1200 is based on the greatest possible use of tested and scientifically validated and developed technical solutions implemented in BN-350, -600, and the BN-800 design as well as new technical solutions that increase facility cost-effectiveness and safety. The BN-1200 design must permit the reactor to operate with different cores, including with denser fuel. The main fuel variant considered is oxide fuel and for the nearest term nitride fuel, for which the production technology involves the same steps as the oxide technology. The main approaches for choosing the parameters of the BN-1200 core as well as the results of computational studies are presented.     

Modelling the dynamical behaviour of nuclear reactor fuel assemblies

The computation of the motion of fuel rods in a nuclear reactor core is generally difficult because the number of elements is very large. This paper is devoted to an homogenization process likening the fluid-tube-spacer system to an equivalent homogeneous medium for which the constitutive laws are derived. It is shown how non-linearity (non-linear spacers, shocks) can be taken into account. Some numerical computation schemes are evoked.     

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.     

SAMURA—A fast computer code for reactor fuel pin performance modelling and analysis

A very fast integral numerical computer code for the modelling of transient and steady-state thermal and mechanical behaviour of Zircaloy-clad UO₂ fuel pins in water reactors has been developed. The computational technique which determines the stress and deformation state of the fuel pin is based upon an extremely efficient finite difference scheme, i.e. the non-linear terms in the constitutive equations which produce a non-linear system of equations have been linearised using a Taylor expansion technique coupled with a very sophisticated error minimization algorithm and then solved with great accuracy. An improved numerical method has also been developed for the fast and efficient solution of the transient heat conduction equation. In this way a very stable and economical one-dimensional code (with appropriate provisions made for its conversion to a quasi two-dimensional code) has been obtained. The physical processes included are thermo-elastic deformation, thermal and irradiation creep, plasticity, fission gas swelling and release, formation of cracks in the fuel, hot pressing, densification, pore migration and dish or central void filling. Here the mathematical basis of SAMURA is presented along with some preliminary calculations and benchmarkings. It is concluded that SAMURA is quite fast indeed, converges to accurate results and within the margins of the error criterion chosen has very reasonable computer demands. It is also stable under all conditions tested.     

On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO₂ fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.     

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.