Thesis (Ph. D.)--Oregon State University, 2009. / Printout. Includes bibliographical references (leaves 331-360). Also available on the World Wide Web.

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Using the fuel burnup as objective function, it is desired to determine the initial distribution of the fuel in a reactor in order to obtain the maximum energy possible, for which, without changing a fixed initial fuel mass, the results for different initial fuel and control poison configurations are analyzed and the corresponding running times compared. One-dimensional, two energy-group theory is applied to a reflected cylindrical reactor using U-235 as fuel and light water as moderator and reflector. Fissions in both fast and thermal groups are considered. The reactor is divided into several annular regions, and the constant flux approximation in each depletion step is then used to solve the fuel and fission-product poisons differential equations in each region. The computer code OPTIME was developed to determine the time variation of core properties during the fuel cycle. At each depletion step, OPTIME calls ODMUG, [12] a criticality search program, from which the spatially-averaged neutron fluxes and control poison cross sections are obtained. A uniform initial fuel distribution was chosen as a benchmark and the results for several different fuel configurations were analyzed. Two different initial control poison distributions were investigated for each fuel configuration: a uniform and a fuel proportional distribution. Using an iterative approach fuel was moved from the low burnup regions toward the high burnup regions; reactor running times were in this way increased from 9000 to 11,500 hours in the fuel proportional control poison distribution case and from 9000 to 11,000 hours in the uniform control poison distribution case. Beyond this point not only did the running time not increase, but no criticality was reached. / Ph. D.

Increased conversion ratios and burnup can be achieved by mechanically changing the fuel-to-water volume ratio of a reactor over the core lifetime. As the fuel-to-water ratio decreases, the neutron spectrum softens, thereby increasing core reactivity. Proposed mechanical spectral shift reactors utilize this concept. RFD-1, a 1-dimensional, 4-group code was developed to compute fuel burnup cycles for spectral shift reactors. The code calculates burnup for a triangular core lattice having a beginning fuel to water ratio as high as 1.30. Core shutdown occurs at a fuel to water ratio of 0.50. The microscopic cross sections were obtained through use of the VIM code and tabulated for use in RFD-1 as a function of fuel to water ratio and burnup time. The fission product group cross sections were developed using the VIM and TOAFEW codes. The flexibility of RFD-1 allows the user to study a wide variety of possible core configurations. Results of RFD-1 show that increased conversion and burnup, using lower initial enrichments than that of standard Pressurized Water Reactors, result for mechanical spectral shift designs. The next step is to study specific spectral shift designs in greater detail. The RFD-1 code could be improved primarily through refinements in its cross section data tables. / Master of Science

Improving computer technology and the desire to more accurately model the heterogeneity of the nuclear reactor environment have made the use of Monte Carlo depletion codes more attractive in recent years, and feasible (if not practical) even for 3-D depletion simulation. However, in this case statistical uncertainty is combined with error propagating through the calculation from previous steps. In an effort to understand this error propagation, four test problems were developed to test error propagation in the fuel assembly and core domains. Three test cases modeled and tracked individual fuel pins in four 17x17 PWR fuel assemblies. A fourth problem modeled a well-characterized 330MWe nuclear reactor core. By changing the code's initial random number seed, the data produced by a series of 19 replica runs of each test case was used to investigate the true and apparent variance in k-eff, pin powers, and number densities of several isotopes. While this study does not intend to develop a predictive model for error propagation, it is hoped that its results can help to identify some common regularities in the behavior of uncertainty in several key parameters.

Nuclear nonproliferation research and forensics have a need for improved software solutions, particularly in the estimates of the transmutation of nuclear fuel during burnup and depletion. At the same time, parallel computers have become effectively sized to enable full core simulations using highly-detailed 3d mesh models. In this work, the capability for modeling 3d reactor models is researched with PENBURN, a burnup/depletion code that couples to the PENTRAN Parallel Sn Transport Solver and also to the Monte Carlo solver MCNP5 using the multigroup option. This research is computationally focused, but will also compare a subset of results of experimental Pressurized Water Reactor (PWR) burnup spectroscopy data available with a designated BR3 PWR burnup benchmark. Also, this research will analyze large-scale Cartesian mesh models that can be feasibly modeled for 3d burnup, as well as investigate the improvement of finite differencing schemes used in parallel discrete ordinates transport with PENTRAN, in order to optimize runtimes for full core transport simulation, and provide comparative results with Monte Carlo simulations. Also, the research will consider improvements to software that will be parallelized, further improving large model simulation using hybrid OpenMP-MPI. The core simulations that form the basis of this research, utilizing discrete ordinates methods and Monte Carlo methods to drive time and space dependent isotopic reactor production using the PENBURN code, will provide more accurate detail of fuel compositions that can benefit nuclear safety, fuel management, non-proliferation, and safeguards applications.

La nécessité de simulations précises d’un réacteur nucléaire et spécialement dans des cas de cœurs et de configurations de combustible complexes, a imposé un usage accru de Codes Neutroniques Stochastiques (CNS). De plus, une demande a émergé pour des CNS à capacité inhérente d’estimation en continu de la variation de la composition isotopique du cœur ainsi qu’à couplage thermo-hydraulique optimisé. Des capacités supplémentaires sont exigées pour ces codes au vu de leur utilisation pour l’étude de nouveaux concepts de réacteur comme les Réacteurs Conduits par Accélérateur (RCA). Plus précisément, le réacteur hybride comprenant un réacteur nucléaire conventionnel et un accélérateur, nécessite l’analyse des deux composantes (réacteur – accélérateur) par un outil capable de couvrir le spectre énergétique neutronique extrêmement étendu qui caractérise ce système hybride.Ce travail présente les principales caractéristiques et capacités du nouveau CNS ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) développé en collaboration du NCSR Demokritos (Grèce) avec CNRS/IDRIS et UPMC (France) et couvrant autant que possible les exigences exposées ci-dessus. ANET est basé sur la version ouverte du code PHE GEANT3.21 et est destiné à effectuer des analyses de cœurs de réacteurs conventionnels de génération II et III ainsi que des RCA. ANET est construit avec la capacité inhérentea) d’effectuer des calculs d’évolution du combustibleb) de simuler le processus de spallation dans le cas des RCAtout en tenant compte de la thermo-hydraulique du système.La version actuelle d’ANET utilise les trois estimateurs standard Monte Carlo pour le calcul du facteur de multiplication neutronique effectif (keff), soit l’estimateur de collision, celui d’absorption et celui de longueur de trace. Pour ce qui est du calcul du débit de fluence neutronique et des taux de réaction, les estimateurs de collision et de longueur de trace sont implémentés dans ANET suivant la procédure standard Monte Carlo. Pour ce qui concerne les calculs d’évolution (par exemple la consommation du combustible), une approche purement stochastique est implémentée dans ANET. A noter que la procédure usuelle consiste à coupler le code neutronique stochastique avec un code déterministe qui calcule la consommation du combustible. Pour les besoins d’analyse des RCA, le module INCL/ABLA a été incorporé dans ANET de façon à ce que le processus de spallation soit simulé par le code. La capacité d’ANET de simuler des configurations classiques a été démontrée en utilisant des résultats de mesures et des simulations de vérification effectuées en utilisant d’autres codes bien établis, ainsi qu’il est montré par la suite.Des données provenant de plusieurs installations et des analyses de problèmes-type internationaux ont été utilisés pour vérifier et valider les capacités d’ANET.Pour conclure, les résultats obtenus lors des comparaisons avec des mesures ou avec des simulations effectuées en utilisant d’autres codes neutroniques stochastiques ou déterministes, montrent qu’ANET possède la capacité de calculer correctement d’importants paramètres de systèmes critiques ou sous-critiques. Par ailleurs, l’application préliminaire d’ANET à des problèmes dépendant du temps fournit des résultats encourageants. ANET produit des estimations de consommation de combustible raisonnables, compte tenu que des incertitudes dans ce domaine sont souvent de l’ordre de 20% ou plus. Finalement, les performances du code dans le cas de KUCA montrent qu’ANET peut analyser des RCA de façon satisfaisante. / The necessity for precise simulations of a nuclear reactor especially in case of complex core and fuel configurations has imposed the increasing use of Monte Carlo (MC) neutronics codes. Besides, a demand of additional stochastic codes’ inherent capabilities has emerged regarding mainly the simulation of the temporal variations in the core isotopic composition as well as the incorporation of the T-H feedback. In addition to the above, the design of innovative nuclear reactor concepts, such as the Accelerator Driven System (ADSs), imposed extra requirements of simulation capabilities. More specifically, the combination of an accelerator and a nuclear reactor in the ADS requires the simulation of both subsystems for an integrated system analysis. Therefore a need arises for more advanced simulation tools, able to cover the broad neutrons energy spectrum involved in these systems.This work presents the main features and capabilities of the new MC neutronics code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback), being developed in NCSR Demokritos (Greece) in cooperation with CNRS/IDRIS and UPMC (France) and intending to meet as effectively as possible the above described modelling requirements. ANET is based on the open-source version of the HEP code GEANT3.21 and is targeting to the creation of an enhanced computational tool in the field of reactor analysis, capable of simulating both GEN II/III reactors and ADSs. ANET is structured with inherent capability of (a) performing burnup calculations and (b) simulating the spallation process in the ADS analysis, while taking T-H feedback into account.The current ANET version utilizes the three standard Monte Carlo estimators for the neutron multiplication factor (keff) calculation, i.e. the collision estimator, the absorption estimator and the track-length estimator. Regarding the simulation of neutron fluence and reaction rates, the collision and the track-length estimators are implemented in ANET following the standard Monte Carlo procedure. For the burnup calculations ANET attempts to apply a pure Monte Carlo approach, adopting the typical procedure followed in stochastic codes. With respect to code improvements for the ADS analysis, so far ANET has incorporated the INCL/ABLA code so that the spallation process can be inherently simulated. The ANET reliability in typical computations was tested using observational data and parallel simulations by different codes as described in the following chapters.Various installations and international benchmarks were considered suitable for the verification and validation of all the previously mentioned features incorporated in the new code ANET. The obtained results are compared with experimental data from the nuclear infrastructures and with computations performed by well-established stochastic or deterministic neutronics codes and show satisfactory agreement with both measurements and independent computations, verifying thus ANET’s ability to successfully simulate important parameters of critical and subcritical systems. Also, the preliminary ANET application for dynamic analysis is encouraging since it indicates the code capability to inherently provide a reasonable prediction for the core inventory evolution taking into account the uncertainties of the order of 20% and even higher that are traditionally expected in core inventory evolution calculations. Lastly, the code performance in the KUCA case was found satisfactory demonstrating thus inherent capability of analyzing ADSs.