Search

Abstract: The MCNP code is a general Monte Carlo N-Particle Transport program that is widely used in health physics, medical physics and nuclear engineering for problems involving neutron, photon and [...]

Approximately one year before a single U[O.sub.2] fuel pellet is pressed and fresh fuel assemblies manufactured, design calculations are performed to effectively guarantee the performance of a light water reactor (LWR) core for an upcoming fuel cycle. Energy requirements must be fulfilled in conjunction with all other reactivity, thermal, and operational limits. Furthermore, several months prior to a reactor startup, the proposed core must be licensed by the regulating authority. Therefore, all built-in design conservatisms (i.e., target margins) established a priori must satisfy all safety, operational, and regulatory constraints, a posteriori. The magnitude of target margins directly impacts cycle energy efficiency, which is why this design cushion is often referred to as the 'cost of margin' because it ultimately affects the cost per generated kilowatt-hour by a LWR. This article illustrates the modern role of nuclear fuel management optimization in the LWR core reload design process, highlighting some of the history and recent advancements in the field, particularly, in the area of pin-by-pin optimization. Also, important limitations are highlighted to help define the new level of sophistication which the field must conquer for designers to ultimately be able to optimize LWRs 'one fuel pin at a time.' Index Terms--Nuclear fuel management, optimization.

AbstractThe goal of our research is to build upon the capability of RELAP5-3D to model molten lead systems. Molten lead has several potential uses in future advanced reactors, like the lead fast reactor or fusion reactors that utilize dual-coolant lead lithium blankets. This potential for use in future generations of reactors highlights the necessity of developing molten lead models to ensure that they can accurately predict thermohydraulic behavior. We have developed a RELAP5-3D model of the Lobo Lead Loop facility located at the University of New Mexico to verify the accuracy of RELAP5-3D via comparison to existing computational fluid dynamics results and analytical calculations.It was found that RELAP5-3D accurately calculated radiative heat transfer (within <1%) when compared to theoretical calculations. In addition, pressure drop calculations done in RELAP5-3D demonstrated reasonable agreement within 20 kPa, mostly within ~7% to 15%, when compared to the computational fluid dynamics model of the facility developed by the University of New Mexico, and captured the dependence of pressure drop on flow velocity accurately.Finally, a hypothetical loss-of-flow transient was imposed on the RELAP5-3D model to determine the feasibility of performing a similar experiment with the Lobo Lead Loop. It was found that such an experiment could be possible, as the RELAP5-3D model indicated that the temperatures of the fluid would not exceed the limiting temperatures of the structure (1658 K) nor the maximum temperature of the electromagnetic pump inlet (823 K). Although there are no experimental data to begin validation, the model will be readily available for future validation studies when the experimental data are generated, especially as the model continues to evolve over time. The results so far demonstrate a promising first step in the verification/validation of the RELAP5-3D model of the Lobo Lead Loop.The highlights from our research are as follows:1. The Lobo Lead Loop facility at the University of New Mexico is a good candidate for molten lead system code validation.2. The Lobo Lead Loop currently has extensive pressure drop results from a high-fidelity computational fluid dynamics model, which offers the opportunity for code-to-code verification of pressure drop in RELAP5-3D.3. A RELAP5-3D model of the Lobo Lead Loop has been developed to begin verification studies and to prepare for potential validation studies.4. Development of the model will continue throughout the future to prepare for potential validation studies. [ABSTRACT FROM AUTHOR]

This paper presents the development of a lattice physics-to-core simulator two-step procedure for the rapid analysis of the Advanced High Temperature Reactor (AHTR). Lattice physics, reflector, and control blade models were developed from which cross-section libraries could be generated for a nodal core simulator. Few-group structures for the core simulator were also generated to account for the neutronic characteristics of AHTR. After developing the AHTR two-step procedure, cross-section libraries were generated using the SERPENT continuous-energy Monte Carlo code. These libraries were then used in the core simulator NESTLE to perform full-core calculations, which were in turn benchmarked against reference SERPENT full-core models. Benchmarking results showed reasonable accuracy of the developed two-step procedure but revealed an inherent inadequacy in the one-dimensional radial reflector model and showed a likely need for a greater number of energy groups than were used in this study. [ABSTRACT FROM AUTHOR]

The proper calculation of core periphery discontinuity factors is important for accurate modeling when using an advanced nodal diffusion simulator. In cores with hexagonal assemblies, such as in the VVER-1000, most fuel assemblies share two faces with the radial reflector, and some even three faces. For this reason, use of a two-dimensional (2D) reflector model will more accurately capture the neutron physics near the core periphery. This article illustrates key points related to the use of 2D discontinuity factors in the reflector region, first by using an algorithm that applies the methodology proposed by Mittag et al. (2003) after correcting some minor typographical errors in the original publication, and then by employing the SCALE transport module NEWT to compute the appropriate quantities. Large and even negative discontinuity factors are an acceptable fact of this methodology when the diffusion approximation becomes invalid due to the problem’s localized features and the large flux gradients. [ABSTRACT FROM AUTHOR]

A query-based approach for adaptively retraining and restructuring a two-hidden-layer artificial neural network (ANN) has been developed for the speedy prediction of the fundamental mode eigenvalue of the neutron diffusion equation, a standard nuclear reactor core design calculation which normally requires the iterative solution of a large-scale system of nonlinear partial differential equations (PDE's). The approach developed focuses primarily upon the adaptive selection of training and cross-validation data and on artificial neural-network (ANN) architecture adjustments, with the objective of improving the accuracy and generalization properties of ANN-based neutron diffusion eigenvalue predictions. For illustration, the performance of a 'bare bones' feedforward multilayer perceptron (MLP) is upgraded through a variety of techniques; namely, nonrandom initial training set selection, adjoint function input weighting, teacher-student membership and equivalence queries for generation of appropriate training data, and a dynamic node architecture (DNA) implementation. The global methodology is flexible in that it can 'wrap around' any specific training algorithm selected for the static calculations (i.e., training iterations with a fixed training set and architecture). Finally, the improvements obtained are carefully contrasted against past works reported in the literature. Index Terms - ANN, DNA, eigenvalues, membership and equivalence queries, MLP, neutron diffusion.

The High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory utilizes a large annular beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron-absorbing and radiation-emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (³He and 6Li) buildup in the reflector regions and its effect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to ³He buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge and during postdischarge decay to assess the viability of transporting, storing, and ultimately disposing of the reflector regions currently stored in the spent-fuel pool. The post-irradiation curie inventories were used to determine whether, for disposal purposes, the reflector regions are discharged as transuranic (TRU) waste or become TRU waste during the decay period and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be TRU waste at discharge, and the other region was determined to become TRU waste <2 yr after being discharged due to irradiation of the initial uranium impurity content (0.0044 wt% uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only). [ABSTRACT FROM AUTHOR]

A numerical framework for modeling depletion and mass transport in liquid-fueled molten salt reactions is presented based on exponential time differencing. The solution method involves using the finite volume method to transform the system of partial differential equations (PDEs) into a much larger system of ordinary differential equations. The key part of this method involves solving for the exponential of a matrix. We explore six different algorithms to compute the exponential in a series of progression problems that explore physical transport phenomena in molten salt reactors. This framework shows good results for solving linear parabolic PDEs with each of the six matrix exponential algorithms. For large problems, the series solvers such as Padé and Taylor have large run times, which can be mitigated by using the Krylov subspace. [ABSTRACT FROM AUTHOR]

An analysis of the normal and shear stress distribution on the rake face of an orthogonal cutting tool is presented. A research-licensed version of DYNA3D finite element package was used to model the cutting process based upon mechanistic experiments that provided the empirical parameters of the Johnson–Cook constitutive model. The Johnson–Cook model uses strain, strain rate, and temperature to predict the material’s stress response. The computer model gave predictions for the normal and shear stresses on the rake lace that qualitatively agree with the experimentally determined stress distributions determined by other researchers. [Copyright &y& Elsevier]

This paper presents results obtained by reproducing a VVER-1000 lattice physics benchmark using the Serpent2 code. The benchmark data are related to the initial cycles of the Khmelnitskiy nuclear power plant. This study compares results obtained with the Serpent2 code against those calculations published in the benchmark document. Detailed descriptions are provided for two fuel assemblies; one without gadolinium in the fuel pins, and the other with gadolinium. A sensitivity study is performed utilizing different nuclear data libraries, different depletion sequences (including pin by pin burnup) and different number of radial regions within the fuel pins containing gadolinium. The infinite multiplication factor behavior and isotopic concentrations of major isotopes are compared and discussed. It is concluded that Serpent2 provided reasonable results in comparison with the other codes used in the benchmark. [ABSTRACT FROM AUTHOR]