The demand for computational tools that are capable to reliably predict the behavior of a nuclear reactor core in a variety of static and dynamic conditions does inevitably require a proper qualification of these tools for the intended purposes. One of the qualification methods is the verification of the code in question. Hereby, the correct implementation of the applied model as well as its flawless implementation in the code are scrutinized. The present work concerns with benchmarking as a substantial part of the verification of the three-dimensional, multigroup neutron kinetics model employed in the transient code POLCA-T. The benchmarking is done by solving some specified and widely used space-time kinetics benchmark problems and comparing the results to those of other, established and well-proven spatial kinetics codes. It is shown that the obtained results are accurate and consistent with corresponding solutions of other codes. In addition, a sensitivity analysis is carried out with the objective to study the sensitivity of the POLCA-T neutronics to variations in different numerical options. It is demonstrated that the model is numerically stable and provide reproducible results for a wide range of various numerical settings. Thus, the model is shown to be rather insensitive to significant variations in input, for example. The other consequence of this analysis is that, depending on the treated transient, the computing costs can be reduced by, for instance, employing larger time-steps during the time-integration process or using a reduced number of iterations. Based on the outcome of this study, one can finally conclude that the POLCA-T neutron kinetics is modeled and implemented correctly and thus, the model is fully capable to perform the assigned tasks.

<p>The broad implementation of three-dimensional full-core modeling, with pin-resolved detail, for computational simulation and analysis of nuclear reactors highlights the importance of accuracy and efficiency in simulation codes for accurate and precise analysis. The primary objective of this dissertation is to develop a high-fidelity code capable of solving time-dependent neutron transport problems with 3D whole-core pin-resolved detail in nuclear reactor cores. Additionally, the dissertation explores the optimization of the code's parallelism to enhance its computational efficiency. To reduce the computational intensity associated with the direct 3D calculation of the neutron transport equation, a high-fidelity neutron transport code called PANDAS-MOC is developed using the 2D/1D approach. The 2D radial solution is obtained using the 2D Method of Characteristics (MOC), the axial 1D solution is determined through the Nodal Expansion Method (NEM), and then two solutions are coupled using transverse leakages to find the 3D solution. The convergence of the iterative scheme is accelerated using the multi-level coarse finite different mesh (ML-CMFD) technique. The code's validation and verification are carried out using the C5G7-TD benchmark exercises.</p> <p><br></p> <p>The significant and innovative aspect of this work involves parallelizing and optimizing the PANDAS-MOC code. Three parallel models are developed and evaluated based on the distributed memory and shared memory architecture: MPI parallel model (PMPI), Segment OpenMP threading hybrid model (SGP), and Whole-code OpenMP threading hybrid model (WCP). When computing the steady state of the C5G7 3D core with the same resources, the obtained speedup relationship between the three models is PMPI \(>\) WCP \(>\) SGP, whereas the WCP model only consumed 60\% of the memory of the PMPI model. Furthermore, the hybrid reduction in the ML-CMFD solver and the parallelism design of the MOC sweep are significant issues that decreased the speedup of WCP. Therefore, this study also addresses further optimizations of these two modules.</p> <p><br></p> <p>Concerning the MOC parallelism, two improvements are discussed: No-atomic schedule and Additional Axial Decomposition (AAD) parallelism. The No-atomic schedule evenly distributed the workload among threads and removes the \textit{omp atomic} clause from the code by predefining the MOC calculation sequence for each launched OpenMP thread while ensuring a thread-safe parallel environment. It can significantly reduce the calculation time and improve parallel efficiency. Furthermore, AAD divides the axial layers and OpenMP threads into multiple groups and restricts each thread to work on the layers designated to the same group. </p> <p>Meanwhile, Flag-Save-Update reduction is designed to increase the computational efficiency of the hybrid MPI/OpenMP reduction operations in the ML-CMFD module. It is accomplished by using the global arrays and status flags and establishing a tree configuration of all threads, and it includes no implicit and explicit barriers. In the case of the C5G7 3D core, the parallel efficiency of the MOC solver is about 0.872 when using 32 threads (=\#MPI \(\times\)\#OpenMP), and the Flag-Save-Update reduction yielded better speedup than the traditional hybrid MPI/OpenMP reduction, and its superiority is more obvious as more OpenMP threads are utilized. As a result, the WCP model outperforms the PMPI model for the overall steady-state calculation.</p> <p><br></p> <p>This research also investigates parallelizable preconditioners to accelerate the convergence of the generalized minimal residual method (GMRES) in the CMFD solver. Preconditioners such as Incomplete LU factorization (ILU), Symmetric Successive Over-relaxation (SOR), and Reduced Symmetric Successive Over-Relaxation (RSOR), are implemented in PANDAS-MOC. Except for RSOR, others are unsuitable for hybrid MPI/OpenMP parallel machines due to their inherent sequential nature and dependency on computation order. Their counterparts using the Red-Black ordering algorithm, namely RB-SOR, RB-RSOR, and RB-ILU, are formatted and examined on benchmark reactors such as TWIGL-2D, C5G7-2D, C5G7-3D, and their corresponding subplane models (TWIGL-2D(5S), C5G7-2D(5S), C5G7-3D(5S)), with relaxed convergence criteria (\(10^{-3}\)). Results show that all preconditioners significantly reduce the required number of iterations to converge the GMRES solutions, and RB-SOR is the best one for most reactors. In the case of C5G7-3D(5S), preconditioners exhibit similar sublinear speedup but demonstrate varying runtimes across all tests for both MG-GMRES and 1G-GMRES. However, the speedup results in 1G-GMRES are more than twice as high as those in MG-GMRES. RB-RSOR has an optimal efficiency of 0.6967 at (4,8), while RB-SOR and RB-ILU have optimal efficiencies of 0.6855 and 0.7275 at (32,1), respectively.</p>

<p dir="ltr">The research focuses on improving core simulation procedures in Boiling Water Reactors (BWRs) by leveraging machine learning techniques. Aimed at better fuel planning and enhanced safety, a machine learning model has been developed to predict errors in existing low-fidelity, diffusion-based core simulators. The machine learning models have demonstrated the capability to accurately and efficiently predict errors in core eigenvalue and power distribution in BWR Operations. This results in a significant improvement over conventional simulation methods in nuclear reactors without increasing computational complexity.</p>

Submitted by Mery Piedad Zamudio Igami (mery@ipen.br) on 2017-03-10T16:45:35Z No. of bitstreams: 0 / Made available in DSpace on 2017-03-10T16:45:35Z (GMT). No. of bitstreams: 0 / Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP) / O presente estudo foi realizado para verificar a possibilidade de redução do núcleo do reator IEA-R1 do IPEN/CNEN-SP. Cálculos neutrônicos foram desenvolvidos para um conjunto de novas configurações para que, a posteriori, a análise termo-hidráulica e de segurança pudessem ser realizadas. As novas configurações analisadas são menores por diversos motivos, como obter uma melhor utilização do combustível, melhor distribuição dos fluxos de nêutrons, dentre outros. Para que se possa atingir tais configurações, a densidade de Urânio no combustível deve ser aumentada. Neste estudo, combustíveis de U3Si2-Al com 4,8gU/cm3 foram testados e novos núcleos para o reator IEA-R1 foram propostos e discutidos. A análise neutrônica não impõe restrições aos núcleos estudados. A análise termohidráulica mostrou que as margens de segurança e os perfis de temperatura ao longo das placas combustíveis não excedem os limites de projeto. Os coeficientes de temperatura obtidos para os novos núcleos, no caso isotérmico, são todos negativos, conforme desejado. A queima mostrou que núcleos supercompactos não apresentam excesso de reatividade suficiente para o funcionamento dos mesmo, ao se utilizar combustíveis com 4,8gU/cm3. Um APR (Acidente de Perda de Refrigerante) foi simulado para os núcleos remanescentes. A ruptura da fronteira do primário se mostrou o acidente mais crítico, devido ao curto tempo para o esvaziamento completo da piscina do reator. As temperaturas atingidas após o descobrimento foram calculadas e não excedem aquelas cujos valores propiciam empolamento nas placas combustíveis (475 °! a 550 °!), uma vez que se obedeça os tempos de esvaziamento seguro da piscina para as novas configurações. / Tese (Doutorado em Tecnologia Nuclear ) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP / FAPESP: 11/17090-7

[ES] El análisis del decaimiento de combustible es fundamental para comprender los cambios a largo plazo en la composición del combustible del reactor debido al quemado del mismo. A medida que se consume el combustible, su composición isotópica cambia y eso afecta significativamente la vida útil operativa del reactor, su estabilidad y sus mecanismos de control. Para abordar estas complejidades, es crucial emplear un conjunto meticulosamente seleccionado de secciones eficaces y parámetros nucleares. Este enfoque no solo garantiza predicciones precisas del comportamiento del reactor tanto en condiciones estacionarias y transitorias, sino que también optimiza el ciclo del combustible y mejora el rendimiento global del reactor. Las librerías de secciones eficaces son la columna vertebral de cualquier código tridimensional utilizado en los cálculos del núcleo. Sin embargo, uno de los principales retos que plantea el cálculo del transporte de neutrones es la necesidad de que cada método empleado haga uso de secciones eficaces estructuradas con metodologías, formatos y contenidos distintos. Esta tesis lleva a cabo una exploración exhaustiva de la física de reactores, centrándose en estos problemas críticos. Su objetivo es desentrañar cómo se capturan y representan los fenómenos de los reactores mediante un análisis en profundidad de las librerías de secciones eficaces. Mediante la investigación de las fuentes de secciones eficaces y datos cinéticos, y la comprensión de los requisitos detallados para resolver diversos problemas, la tesis contribuye a avanzar en las evaluaciones de seguridad robustas y garantizar una representación precisa del comportamiento del reactor. Uno de los aspectos centrales de la tesis es la evaluación de la secuencia computacional CASMO-4/GenPMAXS/PARCS en el análisis de la operación de Reactores de Agua en Ebullición (BWR) con combustibles actuales. Esta evaluación implica una rigurosa verificación de las librerías de secciones eficaces a través de comparativas código a código, lo que garantiza consistencia y precisión en la predicción de la potencia radial y axial del reactor a lo largo del ciclo, mediante librerías de secciones eficaces colapsadas en dos grupos de energía. Además, se realiza un análisis de las predicciones del código nodal PARCS, que se compara con el simulador del núcleo de la planta, SIMULATE-3, utilizado como referencia en cada simulación. Adicionalmente, se incluye la validación de las librerías de secciones eficaces creadas y la comparación del modelo neutrónico del código de núcleo 3D PARCS con datos reales de planta utilizando el sistema detector In-Core Traveling Probe (TIP) con detectores gamma de alta resolución. La simulación de la respuesta del TIP es de importancia crucial para los simuladores del núcleo, ya que permite el uso fiable de las mediciones proporcionadas por este sistema para validar las predicciones y evaluar la precisión de las distribuciones de potencia radiales y axiales calculadas, contrastándolas con las tasas de reacción medidas por los instrumentos in-core. Este estudio emplea las mediciones del TIP para validar la capacidad del código PARCS en la modelización de diseños avanzados de combustible BWR y en el cálculo de distribuciones de potencia tridimensionales bajo condiciones operativas reales. La utilización de datos de planta no solo incrementa la fiabilidad de los modelos, sino que también refuerza el valor práctico de esta investigación dentro del campo de la física de los reactores nucleares. El impacto de las librerías de secciones eficaces en los análisis de seguridad se examina aplicándolas a los transitorios de cierre de la válvula de aislamiento de vapor principal (MSIVC) sin SCRAM (ATWS) mediante el código acoplado TRAC-BF1/PARCS. En un evento de MSIVC ATWS, las respuestas del núcleo se ven afectadas por la interacción entre la retroalimentación de reactividad debida al vacío, impulsada por el colapso del vacío, y la retroalimentación de reactivida / [CA] L'anàlisi del decaïment del combustible és fonamental per a comprendre els canvis a llarg termini en la composició del combustible del reactor deguts al seu cremat. A mesura que el combustible es consumeix, la seua composició isotòpica es modifica, la qual cosa afecta significativament la vida útil operativa del reactor, la seua estabilitat i els mecanismes de control associats. Per a abordar aquestes complexitats, resulta crucial emprar un conjunt meticulosament seleccionat de seccions eficaces i paràmetres nuclears. Aquest enfocament no sols garanteix prediccions precises sobre el comportament del reactor, tant en condicions estacionàries com transitòries, sinó que també optimitza el cicle del combustible i millora el rendiment global del reactor. Les llibreries de seccions eficaces constitueixen l'eix fonamental de qualsevol codi tridimensional utilitzat en els càlculs del nucli del reactor. No obstant això, un dels principals reptes que presenta el càlcul del transport de neutrons radica en la necessitat que cada mètode aplicat utilitze seccions eficaces estructurades conforme a diferents metodologies, formats i continguts. Aquesta tesi aborda una exploració exhaustiva de la física de reactors, centrant-se en aquestes qüestions crítiques. L'objectiu és desentranyar com es capten i es representen els fenòmens característics dels reactors mitjançant una anàlisi profunda de les llibreries de seccions eficaces. Mitjançant la investigació de les fonts de seccions eficaces i dels paràmetres cinètics, així com la comprensió detallada dels requisits necessaris per a resoldre diverses problemàtiques, aquest treball contribueix a avançar en la robustesa de les avaluacions de seguretat i a garantir una representació precisa del comportament del reactor. Un dels aspectes centrals d'aquesta investigació és l'avaluació de la seqüència computacional CASMO-4/GenPMAXS/PARCS en l'anàlisi de l'operació de reactors d'aigua en ebullició (BWR) amb combustibles contemporanis. Aquesta avaluació implica una rigorosa verificació de les llibreries de seccions eficaces mitjançant comparatives codi a codi, la qual cosa garanteix consistència i precisió en la predicció de la potència radial i axial del reactor al llarg del cicle, emprant llibreries de seccions eficaces col·lapsades en dos grups d'energia. A més, es realitza una anàlisi de les prediccions del codi nodal PARCS, comparant-les amb el simulador del nucli de la planta, SIMULATE-3, utilitzat com a referència en cada simulació. A més, s'inclou la validació de les llibreries de seccions eficaces generades i la comparació del model neutrònic tridimensional del codi PARCS amb dades reals de la planta, obtingudes a través del sistema detector In-Core Traveling Probe (TIP), equipat amb detectors gamma d'alta resolució. La simulació de la resposta del TIP és d'importància crucial per als simuladors del nucli, ja que permet l'ús fiable de les mesures proporcionades per aquest sistema per a validar les prediccions i avaluar la precisió de les distribucions de potència radials i axials calculades, contrastant-les amb les taxes de reacció mesurades pels instruments in-core. Aquest estudi empra les mesures del TIP amb l'objectiu de validar la capacitat del codi PARCS en la modelització de dissenys avançats de combustible BWR i en el càlcul de distribucions de potència tridimensionals sota condicions operatives reals. La utilització de dades de planta no sols augmenta la fiabilitat dels models, sinó que també reforça de manera significativa el valor pràctic d'aquesta investigació en l'àmbit de la física de reactors nuclears. L'impacte de les llibreries de seccions eficaces sobre els anàlisis de seguretat s'avalua a través de la seua aplicació en transitoris de tancament de la vàlvula principal d'aïllament de vapor (MSIVC) sense SCRAM (ATWS), emprant el codi acoblat TRAC-BF1/PARCS. En un escenari de MSIVC ATWS, la resposta del nucli es veu condicionada per la interacció entre la retroalimentació de reactiv / [EN] The analysis of fuel depletion is essential for understanding the long term changes in reactor fuel composition due to burnup. As fuel undergoes burnup, its isotopic composition alters, significantly influencing the reactor's operational life, stability, and control mechanisms. To address these complexities, the employment of a meticulously selected set of cross sections and nuclear parameters is crucial. This approach not only ensures accurate predictions of reactor behavior under both steady state and transient conditions but also optimizes the fuel cycle and enhances overall reactor performance. Cross section libraries form the backbone of any three dimensional code used in core calculations. However, a significant challenge in neutron transport calculations arises from the necessity for each method to utilize cross sections structured with varying methodologies, formats, and contents. This thesis undertakes a comprehensive exploration of reactor physics, focusing on these critical issues. It seeks to unravel how reactor phenomena are captured and represented through an in depth analysis of cross section libraries. By investigating the sources of cross sections and kinetic data, and understanding the detailed requirements for solving various problems, this work advances robust safety assessments and ensures an accurate representation of reactor behavior. A central focus of the research is the evaluation of the accuracy of the CASMO 4/GenPMAXS/PARCS computational sequence in analyzing modern Boiling Water Reactor (BWR) operations with current fuels. This entails rigorous verification of cross section libraries through code to code comparisons, ensuring consistency and accuracy in steady state performance parameters and two group energy cross sections. The predictions of the nodal code PARCS are meticulously assessed against the plant core simulator SIMULATE 3, which serves as a benchmark for each simulated case. Furthermore, the validation of the created cross section libraries is conducted through comparisons with real plant data utilizing the In Core Traveling Probe (TIP) system equipped with high resolution gamma detectors. Simulating the TIP response is a critical element for core simulators, enabling the reliable use of TIP measurements to validate predictions and assess the accuracy of calculated radial and axial power distributions by comparing them with measured in core instrument reaction rates. This study leverages TIP measurements to validate the capability of PARCS in modeling advanced BWR fuel designs and calculating 3D power distributions under actual reactor operating conditions. The utilization of real plant data not only enhances the reliability of the models but also significantly elevates the practical value of this research within the field of nuclear reactor physics. The impact of cross section libraries on safety analyses is further examined by applying them to Main Steam Isolation Valve Closure (MSIVC) transients without SCRAM (ATWS) through the coupled code TRAC BF1/PARCS. In an MSIVC ATWS event, core responses are influenced by the interplay between void reactivity feedback, driven by void collapse, and negative Doppler reactivity feedback. Consequently, the severity of the transient hinges on both system behavior and the accuracy of cross section libraries in predicting nuclear parameters. Given these considerations, the MSIVC ATWS scenario serves as an exemplary context for assessing the efficacy of cross section libraries in predicting the evolution of critical parameters under demanding transient conditions. This assessment enhances the modeling capabilities for such events and allows for the simulation of complex thermal hydraulic and feedback phenomena over extended durations. A significant contribution of this thesis is the identification of limitations within the NUREG/CR 7164 recommendations for modeling cross sections for BWR analysis. These recommendations fall short of encompassing the fu / Soler Martínez, MD. (2024). Impact of Nuclear Parameters Processing Techniques on BWR Dynamic Calculations [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/213212

PWR, fuel assembly, benchmark, burnup, lattice, SCALE, Polaris, validation, reactivity