A generalized software capability has been developed for the pin-wise loading optimization of light water reactor (LWR) fuel lattices with the enhanced flexibility of control variables that characterize heterogeneous or blended target pins loaded with non-standard compositions, such as minor actinides (MAs). Furthermore, this study has developed the software coupling to evaluate the performance of optimized lattices outside their reflective boundary conditions and within the realistic three-dimensional core-wide environment of a LWR. The illustration of the methodologies and software tools developed helps provide a deeper understanding of the behavior of optimized lattices within a full core environment. The practical applications include the evaluation of the recycling (destruction) of “undesirable” minor actinides from spent nuclear fuel such as Am-241 in a thermal reactor environment, as well as the timely study of planting Np-237 (blended NpO2 + UO2) targets in the guide tubes of typical commercial pressurized water reactor (PWR) bundles for the production of Pu-238, a highly “desirable” radioisotope used as a heat source in radioisotope thermoelectric generators (RTGs). Both of these applications creatively stretch the potential utility of existing commercial nuclear reactors into areas historically reserved to research or hypothetical next-generation facilities. In an optimization sense, control variables include the loadings and placements of materials; U-235, burnable absorbers, and MAs (Am-241 or Np-237), while the objective functions are either the destruction (minimization) of Am-241 or the production (maximization) of Pu-238. The constraints include the standard reactivity and thermal operational margins of a commercial nuclear reactor. Aspects of the optimization, lattice-to-core coupling, and tools herein developed were tested in a concurrent study (Galloway, 2010) in which heterogeneous lattices developed by this study were coupled to three-dimensional boiling water reactor (BWR) core simulations and showed incineration rates of Am-241 targets of around 90%. This study focused primarily upon PWR demonstrations, whereby a benchmarked reference equilibrium core was used as a test bed for MA-spiked lattices and was shown to satisfy standard PWR reactivity and thermal operational margins while exhibiting consistently high destruction rates of Am-241 and Np to Pu conversion rates of approximately 30% for the production of Pu-238.

At Oregon State University the Multi-Application Small Light Water Reactor (MASLWR) integral effects testing facility is being prepared for safety analysis matrix testing in support of the NuScale Power Inc. (NSP) design certification progress. The facility will be used to simulate design basis accident performance of the reactor's safety systems. The design includes an initially evacuated, high pressure capable containment system simulated by a 5 meter tall pressure vessel. The convection-condensation process that occurs during use of the Emergency Core Cooling System has been characterized during two experimental continuous blowdown events. Experimental data has been used to calculate an average heat transfer coefficient for the containment system. The capability of the containment system has been analytically proven to be a conservative estimate of the full scale reactor system. / Graduation date: 2013

Within the presented thesis there were collected data of effluents of 14C chemical forms from ventilation stacks of the Unit 1, the Unit 2 and the Auxiliary Building of the Temelin NPP for the period from 2001 to 2006. These data are compared to power of both reactors and concentration of ammonium ions in coolant of the primary circuit of the Unit 1 and Unit 2.

In this work, which deals with problematics of releases from nuclear power plants, I tried to make an overview of various types of nuclear power plants and radioizotopes which are released through liquid and gasseous effluents. As a part of this comparison evaluation of czech and world nuclear power plants gaseous and liquid releases was made. Introductory part of this work contains information about different types of nuclear power plants and radioizotopes, which are produced in their reactors. It continues with today{\crq}s legislative and information about releasing levels and methods and systems used for measurement of radioactive gaseous and liquid effluents from nuclear power plants. Second part of this work describes the aim of this work and hypothesis. Third part explains the methods, which were used for completing of this work. Fourth part contains results in the form of tables and graphs. Fifth part represents discussion of the results. Last part is a summarization of the results.

Um dos principais mecanismos de falha que causam riscos de fratura a reatores de água pressurizada é a corrosão sob tensão de ligas metálicas em água do circuito primário (CSTAP). É causada por uma combinação das tensões de tração, meio ambiente em temperatura e microestruturas metalúrgicas susceptíveis. Ela pode ocorrer, dentre outros locais, nos bocais do mecanismo de acionamento das barras de controle. Essa fratura pode causar acidentes que comprometem a segurança nuclear através do bloqueio das barras de controle e vazamentos de água do circuito primário reduzindo a confiabilidade e a vida útil do reator. O objetivo desta Tese de Doutorado é o estudo de modelos e uma proposta de modelagem para fraturas por corrosão sob tensão em liga 75Ni15Cr9Fe (liga 600), em água de circuito primário de reator de água pressurizada nesses bocais. São superpostos modelos eletroquímicos e de mecânica da fratura e validados com dados obtidos em experimentos e na literatura. Na parte experimental foram utilizados resultados obtidos pelo CDTN no equipamento recém-instalado de ensaio por taxa de deformação lenta. Na literatura está proposto um diagrama que exprime a condição termodinâmica de ocorrerem diversos modos de CSTAP na liga 600: partiu-se de diagramas de potencial x pH (diagramas de Pourbaix), para a liga 600 imersa em água primária à alta temperatura (3000C a 3500C). Sobre ele, determinaram-se os submodos de corrosão, a partir de dados experimentais. Em seguida acrescentou-se uma dimensão adicional ao diagrama, correlacionando uma variável a que se denominou fração de resistência à corrosão sob tensão. No entanto, é possível acrescentar-se outras variáveis que exprimem a cinética de iniciação e/ou crescimento de trinca, provenientes de outras modelagens de CSTAP. A contribuição original deste trabalho se insere nessa fase: partindo-se de uma condição de ensaio de potencial versus pH, foram iniciadas as modelagens de um modelo empírico-comparativo, um semi-empírico-probabilístico, um de tempo de iniciação e um de taxa de deformação, a partir dos ensaios experimentais e superpostas a essa condição. Esses exprimem respectivamente a susceptibilidade à CSTAP, o tempo de falha, e nos dois últimos o tempo de iniciação de falha por corrosão sob tensão. Os resultados foram comparados com os da literatura e se mostraram coerentes. Através desse trabalho, obteve-se uma metodologia de modelagem a partir de dados experimentais. / One of the main failure mechanisms that cause risks to pressurized water reactors is the primary water stress corrosion cracking (PWSCC) occurring in alloys. It can occurs, besides another places, at the control reactor displacement mechanism nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause accidents that reduce nuclear safety by blocking the rods displacement and may cause leakage of primary water, reducing the reactors life. In this work it is proposed a study of the existing models and a modeling proposal to primary water stress corrosion cracking in these nozzles in a nickelbased Alloy 600. It is been superposed electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained at CDTN-Brazilian Nuclear Technology Development Center, in a recent installed slow strain rate testing equipment. In the literature it is found a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC submodes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Alloy 600 in high temperature primary water (3000C till 3500C). Over it, were located the PWSCC submodes, using experimental data. It was added a third parameter called stress corrosion strength fraction. However, it is possible to superpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. Here is the proposition of the original contribution of this work: from an original experimental condition of potencial versus pH, it was superposed, an empiric-comparative, a semi-empiric-probabilistic, an initiation time, and a strain rate damage models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. It was modeling from our experimental data. The results were compared with the literature and it showed to be coherent. From this work was obtained a modeling methodology from experimental data.

Der Reaktordynamikcodes DYN3D wurde in der neu entwickelten Mehrgruppen-Version DYN3D-MG für die Anwendung auf wassergekühlte Reaktoren alternativ zu industriellen DWR und SWR ertüch-tigt. Es wurde die Anwendbarkeit für den graphitmoderierten Druckröhrenreaktor EGP-6 (KKW Bilibi-no), eine Konzeptstudie eines fortgeschrittenen Siedewasserreaktors mit schnellem Neutronenspekt-rum (RMWR) und das Reaktorkonzept RUTA-70 zur Wärmeversorgung nachgewiesen. Beim RUTA-Reaktor geht es vor allem um die Modellierung des Naturumlaufs des Kühlmittels bei niedrigen Sys-temdrücken. Zur Validierung wurden Experimente zu flashing-induzierten Naturumlaufinstabilitäten an der Versuchsanlage CIRCUS der TU Delft mit RELAP5 nachgerechnet. Für die Anwendung von DYN3D auf die alternativen Reaktorkonzepte wurden Modellerweiterungen und Anpassungen vorgenommen, u.a. Modifikationen in den Wärmeleitungs- und -übergangsmodellen. Vergleichsrechnungen mit dem stationären russischen Feingitter-Diffusionscode ACADEM ergänzen die Verifikationsdatenbasis von DYN3D-MG. Zur Validierung wurden zwei reak-tordynamische Experimente am Reaktor EGP-6 nachgerechnet. Für Reaktoren EGP-6, RMWR und RUTA wurden verschiedene Transienten mit Ausfahren von Re-gelstäben mit und ohne Reaktorschnellabschaltung gerechnet. Weiterhin wurden Analysen für den ATWS-Störfall \"Abschalten aller Hauptkühlmittelpumpen bei Vollleistung\" für den RUTA-Reaktor mit den gekoppelten Programmkomplexen DYN3D/ATHLET und DYN3D/RELAP5 durchgeführt. Der Reaktor geht in einen sicheren Zustand mit reduzierter Leistung bei Naturumlauf des Kühlmittels über. Die Ergebnisse von Analysen zum unkontrollierten Ausfahren einer Regelgruppe für den RMWR lassen dagegen eine belastbare Schlussfolgerung bezüglich der Beherrschbarkeit des Aus-fahrens einer Regelgruppe nicht zu. Abschließend wurde der Nutzen der Programmertüchtigung von DYN3D für die Anwendung auf GenIV -Konzepte und LWR mit hohem Konversionsfaktor bewertet.

Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial description of the reactor core. The TH code uses few, typically 5 to 20 TH channels, which represent the core. The NK code uses explicit node for each fuel assembly. Therefore, a spatial mapping of coarse grid TH and fine grid NK domain is necessary. However, improper mappings may result in loss of valuable information, thus causing inaccurate prediction of safety parameters. The purpose of this thesis is to study the sensitivity of spatial coupling (channel refinement and spatial mapping) and develop recommendations for NK-TH mapping in simulation of safety transients – Control Rod Drop, Turbine Trip, Feedwater Transient combined with stability performance (minimum pump speed of recirculation pumps). The research methodology consists of spatial coupling convergence study, as increasing number of TH channels and different mapping approach the reference case. The reference case consists of one TH channel per one fuel assembly. The comparison of results has been done under steady-state and transient conditions. Obtained results and conclusions are presented in this licentiate thesis.

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.

<p>Severe accidents in a Light Water Reactor (LWR) have been a subject of the research for the last three decades. The research in this area aims to further understanding of the inherent physical phenomena and reduce the uncertainties surrounding their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors. The research is also focusing on evaluation of the proposed accident management schemes for mitigating the consequences of such accidents.</p><p>During a hypothetical severe accident, whatever the scenario, there is likelihood that the core material will be relocated and accumulated in the lower plenum in the form of a debris bed or a melt pool. Physical phenomena involved in a severe accident progression are complex. The interactions of core debris or melt with the reactor structures depend very much on the debris bed or melt pool thermal hydraulics. That is why predictions of heat transfer during melt pool formation in the reactor lower head are important for the safety assessment.</p><p>The main purpose of the present study is to advance a method for describing turbulent natural convection heat transfer of a melt pool, and to develop a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of Core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident.</p><p>Given the insights gained from Computational Fluid Dynamics (CFD) simulations, a physics-based model and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a Boiling Water Reactor (BWR) during the late phase of an in-vessel core melt progression. A model is developed for the core debris bed heat up and formation of a melt pool in the lower head of the reactor vessel, and implemented in a commercial CFD code. To describe the natural convection heat transfer inside the volumetrically decay-heated melt pool, we advanced the Effective Convectivity Conductivity Model (ECCM), which was previously developed and implemented in the MVITA code. In the present study, natural convection heat transfer is accounted for by only the Effective Convectivity Model (ECM). The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables simulations of heat transfer of a high Rayleigh melt pool in 3D large dimension geometry.</p><p>In order to describe the phase-change heat transfer associated with core debris, a temperature-based enthalpy formulation is employed in the ECM (the phase-change ECM or so called the PECM). The PECM is capable to represent possible convection heat transfer in a mushy zone. The simple approach of the PECM method allows implementing different models of the fluid velocity in a mushy zone for a non-eutectic mixture. The developed models are validated by a dual approach, i.e., against the existing experimental data and the CFD simulation results.</p><p>The ECM and PECM methods are applied to predict thermal loads to the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heat up and melting in the BWR lower plenum. Applying the ECM and PECM to simulations of reactor-scale melt pool heat transfer, the results of the ECM and PECM calculations show an apparent effectiveness of the developed methods that enables simulations of long term accident transients. It is also found that during severe accident progression, the cooling by water flowing inside the CRGTs plays a very important role in reducing the thermal load on the reactor vessel wall. The results of the CFD, ECM and PECM simulations suggest a potential of the CRGT cooling as an effective mitigative measure during a severe accident progression.</p>

Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents.  In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment.  The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis.   The CFD method, on the one hand, is indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR) lower plenum. It is found that during the accident progression, the CRGT cooling plays a very important role in reducing the thermal loads on the reactor vessel wall. Results of the ECM and PECM simulations suggest a high potential of the CRGT cooling to be an effective measure for severe accident management in BWRs. / <p>QC 20100812</p>