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A new definition of the diffusion coefficient for use in reactor physics calculations is evaluated in this thesis. It is based on naturally flux-limited diffusion theory (FDT), sometimes referred to as Levermore-Pomraning diffusion theory. Another diffusion coefficient more loosely based on FDT is also evaluated in this thesis. Flux-limited diffusion theory adheres to the physical principle of flux-limiting, which is that the magnitude of neutron current is not allowed to exceed the scalar flux. Because the diffusion coefficients currently used in the nuclear industry are not flux-limited they may violate this principle in regions of large spatial gradients, and because they encompass other assumptions, they are only accurate when used in the types of calculations for which they were intended. The evaluations were performed using fine-mesh diffusion theory. They are in one spatial dimension and in 47, 4, and 2 energy groups, and were compared against a transport theory benchmark using equivalent energy structures and spatial discretization. The results show that the flux-limited diffusion coefficient (FD) outperforms the standard diffusion coefficient in calculations of single assemblies with vacuum boundaries, according to flux- and eigenvalue-errors. In single assemblies with reflective boundary conditions, the FD yielded smaller improvements, and tended to improve only the fast-group results. The results also computationally confirm that the FD adheres to flux-limiting, while the standard diffusion coefficient does not.

<p>Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. In this thesis, the most important existing reactivity determination methods have been evaluated experimentally in the subcritical YALINA-experiments in Belarus. It is concluded that the existing methods are sufficient for calibration purposes, but not for reactivity monitoring during regular operation of an accelerator-driven system. Conditions for successful utilization of the various methods are presented, based on the experimental experience.</p>

This thesis documents significant contributions to the quantification of input and modelling uncertainties in the simulation of nuclear power plants. This work is intended to support the simulations that are performed to demonstrate the safety of nuclear power plants in general, and in CANDU reactors specifically. The work presented in this thesis extends the methodologies for uncertainty propagation established internationally to CANDU plants and pioneers the integration of these tools with important plant features in CANDUs, such as online fueling. This thesis documents a series of simulation studies performed to quantify the impact of uncertainties (primarily nuclear data uncertainties), on simulations of CANDU stations and light water reactors (LWRs). The novel part of this work includes quantifying the role of operational feedbacks such as online refuelling and reactor control systems, and important modelling uncertainties, on CANDU simulations. To achieve this objective, this thesis examines 4 important areas as documented in journal papers. To demonstrate understanding of the tools developed for the UAM-LWR benchmark and to support the ongoing international effort, select studies from the UAM-LWR benchmark study exercises were performed and published in the first journal paper. Time-dependent PWR neutronics exercises, considering both nuclear data and manufacturing uncertainties, were completed. This work found that the relative importance of nuclear data uncertainties and manufacturing uncertainties depended on whether the parameter of interest was “local”, such as pin power factors, or “global”, such as homogenized assembly properties. The second publication in this thesis documents the adaption of the tools from the first paper to consider CANDU specific features, such as spatial control systems and online refuelling. This paper demonstrated the significant effect that consistent feedback from fuelling operations has on reducing the total uncertainty in core level simulations of CANDU plants. The tools developed for this work were used to support downstream studies by generating extensive sets of realistic initial conditions for many different possible nuclear datasets. The next publications utilized the tools developed above and then extends the methods to include operational aspects of CANDUs in the assessments for the first time. In the third paper these methods were then used to demonstrate the tools’ capabilities to simulate an operational transient (a power maneuver from 100% full power to 59% full power) in a CANDU station and compared the resultant prediction and uncertainties to measure plant responses. A further study, on the role of nuclear data and initial burnup distribution uncertainty on a CANDU plant’s response to perturbations to liquid zone controller levels, was also performed to examine the effect of the commonly used “superposition principle” utilized in industry to make safety analysis of CANDU’s various fueling states more tractable. In both cases the role of nuclear data uncertainties was generally found to be similar in magnitude to the role of uncertainty in the core initial conditions. The results of this work support the continued safe operation of CANDU nuclear generating stations in Canada by quantifying the role of select uncertainties on safety simulation outputs, informing future BEPU analysis for CANDU plants and demonstrating the exceptional flexibility of the CANDU reactor design. This is reflected in one of the major conclusions of these works, which demonstrates that the natural feedbacks in CANDU operation help to minimize the effect of uncertainties in the outcome of many safety analysis. / Thesis / Candidate in Philosophy

Recent work by Yasseri and Rahnema has introduced a consistent spatial homogenization (CSH) method completely in transport theory. The CSH method can very accurately reproduce the heterogeneous flux shape and eigenvalue of a reactor, but at high computational cost. Other recent works for homogenization in diffusion or quasi-diffusion theory are accurate for problems with low heterogeneity, such as PWRs, but are not proven for more heterogeneous reactors such as BWRs or GCRs. To address these issues, a consistent hybrid diffusion-transport spatial homogenization (CHSH) method is developed as an extension of the CSH method that uses conventional flux weighted homogenized cross sections to calculate the heterogeneous solution. The whole-core homogenized transport calculation step of the CSH method has been replaced with a whole- core homogenized diffusion calculation. A whole-core diffusion calculation is a reasonable replacement for transport because the homogenization procedure tends to smear out transport effects at the core level. The CHSH solution procedure is to solve a core-level homogenized diffusion equation with the auxiliary source term and then to apply an on-the-fly transport-based re-homogenization at the assembly level to correct the homogenized and auxiliary cross sections. The method has been derived in general geometry with continuous energy, and it is implemented and tested in fine group, 1-D slab geometry on controlled and uncontrolled BWR and HTTR benchmark problems. The method converges to within 2% mean relative error for all four configurations tested and has computational efficiency 2 to 4 times faster than the reference calculation.

Made available in DSpace on 2014-10-09T12:54:34Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:09:26Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:54:34Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:09:26Z (GMT). No. of bitstreams: 0 / Este trabalho apresenta uma solução analítica obtida pelo método de expansão para cinética espacial usando o modelo de difusão e considerando meios homogêneos multiplicativos subcríticos acionados por fonte externa. Em particular, partindo de modelos mais simples e aumentando a complexidade do sistema, resultados foram obtidos para diferentes tipos de transientes. Inicialmente, uma solução analítica foi obtida considerando um grupo de energia sem nêutrons atrasados, em seguida considerou-se um sistema de um grupo de energia e uma família de precursores. A solução para o caso G grupos de energia e R famílias de precursores em forma fechada é obtida, apesar do fato que não possa ser resolvido analiticamente, uma vez que não existe forma explícita para os autovalores e métodos numéricos devem ser utilizados para resolver tal problema. Para ilustrar a solução geral um problema de multigrupo (três grupos de energia) dependente do tempo sem precursores é apresentada e os resultados numéricos obtidos usando um código de diferenças finitas são comparados com os resultados exatos para diferentes tipos de transientes. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP

Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. In this thesis, the most important existing reactivity determination methods have been evaluated experimentally in the subcritical YALINA-experiments in Belarus. It is concluded that the existing methods are sufficient for calibration purposes, but not for reactivity monitoring during regular operation of an accelerator-driven system. Conditions for successful utilization of the various methods are presented, based on the experimental experience. / QC 20101115

Neste trabalho foi realizado uma comparação entre três tipos de alvos (UAl2-Al, U-Ni cilíndrico e U-Ni placa) para a produção de 99Mo por fissão do 235U. Para isso foram desenvolvidas análises neutrônicas e termo-hidráulicas. Também foram realizados experimentos para se validar as metodologias de cálculo termo-hidráulica e neutrônica utilizadas neste trabalho. Para os cálculos neutrônicos foram utilizados os programas NJOY99.0, AMPX-II e HAMMERTECHNION, para geração das seções de choque, e os programas SCALE 6.0 e CITATION para os cálculos tridimensionais dos núcleos, queima do combustível e produção de 99Mo. Para os cálculos termo-hidráulicos foram utilizados os programas MTRCRIEAR1 e ANSYS CFX para calcular as variáveis térmicas e hidráulicas dos dispositivos de irradiação e compará-las a limites e critérios de projeto estabelecidos. Primeiro foram realizadas análises neutrônicas e termo-hidráulicas para o reator IEA-R1 com os alvos de UAl2-Al (10 miniplacas). As análises demonstraram que a atividade total obtida para o 99Mo nas miniplacas não atende à demanda dos hospitais brasileiros (450 Ci/semana) e que nenhum limite de projeto termo-hidráulico é ultrapassado. Em seguida foram realizados os mesmos cálculos para os três tipos de alvo no Reator Multipropósito Brasileiro (RMB). As análises neutrônicas demonstraram que os três alvos podem atender à demanda dos hospitais brasileiros. As análises termo-hidráulicos demonstram que será necessário uma velocidade mínima no dispositivo de irradiação de 7 m/s para o UAl2, de 8 m/s para o alvo de U-Ni cilíndrico e de 9 m/s para o alvo de U-Ni placa para que nenhum limite de projeto seja ultrapassado. Foram realizados experimentos em uma bancada de aferição de vazão para se validar a metodologia de cálculo termo-hidráulico. Os experimentos realizados para se validar os cálculos neutrônicos foram feitos no reator IPEN/MB-01. Todos os experimentos foram simulados com as metodologias acima descritas e os resultados comparados entre si. Os resultados das simulações apresentaram boa concordância com os resultados experimentais. / In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl2-Al, U-Ni cylindrical and U-Ni plate) used for the production of 99Mo by fission of 235U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of 99Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl2-Al (10 miniplates). Analyses have shown that the total activity obtained for 99Mo on the miniplates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl2-Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental results.

Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort. / UOIT