This report describes computer studies which were done to determine the effect on burnup of modifying the heavy water reflector in a 600 MWe CANDU-PHW reactor. It is shown that the burnup penalty increases rapidly as the reflector thickness is reduced. The burnup penalty is significantly lower for mixed reflectors in which some of the heavy water in the outer region of the reflector is replaced by graphite, an organic liquid, or light water, while maintaining the original reflector thickness. / Thesis / Master of Engineering (MEngr)

No description available.

No description available.

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.

This thesis presents a detailed analysis of the bRAPID (burnup for RAPID – Real Time Analysis for Particle transport and In-situ Detection) code system, and the implementation and validation of two new algorithms for improved burnup simulation. bRAPID is a fuel burnup algorithm capable of performing full core 3D assembly-wise burnup calculations in real time, through its use of the RAPID Fission Matrix methodology. A study into the effect of time step resolution on isotopic composition in Monte Carlo burnup calculations is presented to provide recommendations for time step scheme development in bRAPID. Two novel algorithms are implemented into bRAPID, which address: i) the generation of time-dependent correction factors for the fission density distribution in boundary nuclear fuel assemblies within a reactor core; ii) the calculation of pin-wise burnup distributions and isotopic concentrations. Time step resolution analysis shows that a variable time step scheme, developed to accurately characterize important isotope evolution, can be used to optimize burnup calculations and minimize computation time. The two new algorithms have been benchmarked against the Monte Carlo code system Serpent. Results indicate that the time-dependent boundary correction algorithm improves fission density distribution calculations by including a more detailed representation of boundary physics. The pin-wise burnup algorithm expands bRAPID capabilities to provide material composition data at the pin level, with accuracy comparable to the reference calculation. In addition, wall-clock time analyses show that burnup calculations performed using bRAPID with these novel algorithms require a fraction of the time of Serpent. / Master of Science / Fuel burnup modeling is an important aspect of nuclear reactor design that provides information about the energy extracted (called burnup) and isotopes created or used in the fuel of a reactor over time. A reactor core is a collection of fuel assemblies, and assemblies are simply a bundle of fuel pins, which contain nuclear fuel such as Uranium. The desire for accurate and fast computer codes to calculate fuel burnup rises each year as engineers working in reactor core design seek to arrange fuel assemblies in an optimal pattern to extract the most energy. State of the art burnup codes exist, however they have certain limitations due to their underlying methodologies. To satisfy this need, the bRAPID algorithm was developed by the Virginia Tech Transport Theory Group (VT³G). bRAPID is a new methodology capable of performing full core fuel burnup calculations in real time. bRAPID is able to calculate the criticality and burnup distribution of a reactor orders of magnitude faster than comparable algorithms, while addressing many of the shortcomings seen in other burnup codes. In this thesis, studies of standard burnup codes are conducted in order to aid in bRAPID analysis: first in the form of a detailed study of the reference Monte Carlo model used in this thesis, and secondly in an investigation of the effect of time step selection – or the time intervals used in burnup calculations – on isotope concentration. Both of these studies are conducted using the benchmark code system, Serpent, with the latter study providing useful insight that can be used for bRAPID database development. This thesis then presents two new algorithms for bRAPID that expand its capability and improve performance. First, an algorithm to more accurately simulate the boundary regions of the core – called the time dependent boundary correction algorithm – is presented and benchmarked. Next, an algorithm to expand bRAPID capability from assembly-wise to pin-wise burnup calculations is implemented and tested. These two algorithms are benchmarked against the Serpent Monte Carlo based burnup code.

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

En règle générale, les études de sûreté-criticité concernant les combustibles usés stockés, transportés ou retraités sont très conservatives et considèrent ce combustible comme neuf donc le plus réactif possible. Le " Crédit Burn-up " (CBU) est la prise en compte de l'antiréactivité du combustible irradié par rapport au combustible neuf. Une méthodologie CBU rigoureuse, développée par le CEA en collaboration avec AREVA-NC a récemment été validée et réévaluée pour les combustibles REP-UOx. Cependant, 22 réacteurs sur les 58 que compte la France utilisent également du combustible MOx. De plus en plus d'assemblages MOx irradiés doivent donc être entreposés et transportés, ce qui conduit les industriels à s'intéresser à la prise en compte du CBU pour ces applications, dans le but de pouvoir gagner des marges en terme de dimensionnement des installations. Des publications récentes et les travaux du Groupe de Travail Français sur le CBU ont souligné l'importance de la prise en compte des 15 produits de fission stables et non volatiles les plus absorbants qui sont à l'origine de la moitié de l'antiréactivité totale apportée dans les combustibles REP-MOx. C'est pourquoi, dans le but de garantir la sous-criticité de la configuration étudiée suivant les dispositions règlementaires relatives à la sûreté des installations, les biais de calcul affectant leur bilan-matière et leur effet individuel en réactivité doivent également être pris en considération dans les études de sûreté-criticité s'appuyant sur des calculs de criticité. Dans ce contexte, une revue bibliographique exhaustive a permis d'identifier les particularités des combustibles REP-MOx et une démarche rigoureuse a été suivie afin de proposer une méthodologie CBU adaptée à ces combustibles validée et physiquement représentative, permettant de prendre en compte les produits de fission et permettant d'évaluer les biais liés au bilan-matière et à l'antiréactivité des isotopes considérés. Cette démarche s'est articulée autour des études suivantes : * détermination de facteurs correctifs isotopiques permettant de garantir le conservatisme du calcul de criticité sur la base de la qualification du formulaire d'évolution DARWIN-2.3 pour les applications REP-MOx et d'une analyse des données nucléaires des produits de fission métalliques afin de déterminer l'impact des incertitudes associées sur le calcul de leur bilan matière ; * évaluation de l'antiréactivité individuelle des produits de fission sur la base des résultats d'interprétation des expériences d'oscillation des programmes CBU et MAESTRO, réalisés dans le réacteur expérimental MINERVE à Cadarache, avec le formulaire dédié PIMS développé au SPRC/LEPh avec mise à jour des schémas de calcul pour la criticité ; * élaboration de matrices de covariances réalistes associées à la capture de deux des principaux produits de fission du CBU REP-MOx : 149Sm et le 103Rh associées à l'évaluation JEFF-3.1.1 ; * détermination des biais et incertitudes " a posteriori " dus aux données nucléaires des actinides et produits de fission considérés pour deux applications industrielles (piscine d'entreposage et château de transport) par une étude de transposition réalisée avec l'outil RIB, développé au SPRC/LECy, qui a bénéficié à cette occasion de développements spécifiques et de mises à jour des données utilisées (importation des données de covariance issues de la bibliothèque COMAC V0 associée à JEFF-3.1.1 pour les isotopes 235,238U, 238,239,240,241,242Pu, 241Am et 155Gd et prise en compte des corrélations inter-réactions pour un même isotope). * évaluation de la méthodologie proposée pour deux applications industrielles (piscine d'entreposage et château de transport), démonstration de son intérêt et de sa robustesse.

O objetivo deste trabalho é verificar a validade e a acurácia dos resultados fornecidos pelos programas computacionais FRAPCON-3.4a e FRAPTRAN-1.4, utilizados no processo de simulação do comportamento de varetas combustíveis de reatores a água leve pressurizada PWR (Pressurized Water Reactor), sob situações operacionais de regimes permanente e transiente, em condições de alta queima (high burnup). Para realizar a verificação, foi utilizada a base de dados FUMEX-III, que fornece dados relativos a experimentos realizados com diversos tipos de combustíveis nucleares, submetidos a diversas condições operacionais. Através dos resultados obtidos nas simulações computacionais com os programas FRAPCON-3.4a e FRAPTRAN-1.4 e da sua comparação com os dados experimentais da base FUMEX-III, foi possível constatar que os programas empregados possuem um boa capacidade de predizer o comportamento operacional de varetas combustíveis de PWR em regime permanente a altas queimas e sob condição de transiente inicializado por reatividade (Reactivity Initiated Accident RIA). / The objective of this work is to verify the validity and accuracy of the results provided by the computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods in steady-state and transient operational conditions at high burnup. To perform the verification, the database FUMEX-III was used to provide data on experiments with different nuclear fuel types, under various operating conditions. Through the comparison of the computational simulation results of the programs FRAPCON-3.4a e FRAPTRAN-1.4 with the experimental data of the database FUMEX III, it was found that the computer programs used have good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under Reactivity Initiated Accident (RIA).

The Monte Carlo method is, practically, the only approach capable of giving detail insight into complex neutron transport problems. In reactor physics, the method has been used mainly for determining the keff in criticality calculations. In the last decade, the continuously growing computer performance has allowed to apply the Monte Carlo method also on simple burnup simulations of nuclear systems. Nevertheless, due to its extensive computational demands the Monte Carlo method is still not used as commonly as deterministic methods. One of the reasons for the large computational demands of Monte Carlo criticality calculations is the necessity to carry out a number of inactive cycles to converge the fission source. This thesis presents a new concept of fission matrix based Monte Carlo criticality calculations where inactive cycles are not required. It is shown that the fission matrix is not sensitive to the errors in the fission source, and can be thus calculated by a Monte Carlo calculation without inactive cycles. All required results, including keff, are then derived via the final fission matrix. The confidence interval for the estimated keff can be conservatively derived from the variance in the fission matrix. This was confirmed by numerical test calculations of Whitesides's ``keff of the world problem'' model where other Monte Carlo methods fail to estimate the confidence interval correctly unless a large number of inactive cycles is simulated.   Another problem is that the existing Monte Carlo criticality codes are not well shaped for parallel computations; they cannot fully utilise the processing power of modern multi-processor computers and computer clusters. This thesis presents a new parallel computing scheme for Monte Carlo criticality calculations based on the fission matrix. The fission matrix is combined over a number of independent parallel simulations, and the final results are derived by means of the fission matrix. This scheme allows for a practically ideal parallel scaling since no communication among the parallel simulations is required, and no inactive cycles need to be simulated.   When the Monte Carlo criticality calculations are sufficiently fast, they will be more commonly applied on complex reactor physics problems, like non-linear steady-state calculations and fuel cycle calculations. This thesis develops an efficient method that introduces thermal-hydraulic and other feedbacks into the numerical model of a power reactor, allowing to carry out a non-linear Monte Carlo analysis of the reactor with steady-state core conditions. The thesis also shows that the major existing Monte Carlo burnup codes use unstable algorithms for coupling the neutronic and burnup calculations; therefore, they cannot be used for fuel cycle calculations. Nevertheless, stable coupling algorithms are known and can be implemented into the future Monte Carlo burnup codes. / QC 20100709

The swelling mechanisms of U3Si2 under neutron irradiation in reactor conditions are not unequivocally known. The limited experimental evidence that is available suggests that the main driver of the swelling in this material would be fission gases accumulation at crystalline grain boundaries. The steps that lead to the accumulation of fission gases at these locations are multiple and complex. However, gradually, the gaseous fission products migrate by diffusion. Upon reaching a grain boundary, which acts as a trap, the gaseous fission products start to accumulate, thus leading to formation of bubbles and hence to swelling. Therefore, a quantitative model of swelling requires the incorporation of phenomena that increase the presence of grain boundaries and decrease grain sizes, thus creating sites for bubble formation and growth. It is assumed that grain boundary formation results from the conversion of stored energy from accumulated dislocations into energy for the formation of new grain boundaries.This thesis attempts to develop a quantitative model for grain subdivision in U3Si2 based on the above mentioned phenomena to verify the presence of this mechanism and to use in conjunction with swelling codes to evaluate the total swelling of the pellet in the reactor during its lifetime.