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The fission product releases from spherical fuel elements used in modern high temperature gas cooled reactors are one of the first source terms used in describing the safety of planned nuclear plants during normal and accident conditions. The verification and validation of the model and code used to predict the gaseous fission product behaviour and release from spherical fuel elements for the PBMR were documented in this dissertation. The PBMR is the latest design in high temperature gas cooled reactor technology utilizing spherical fuel elements based on the LEU TRISO-coated particle design. Fission products can be divided into relatively short-lived noble gas and halogens, and relatively long-lived metallic fission and activation products. Each group is described by its own models and sets of transport parameters. The noble gases and halogen fission product releases from the fuel elements are direct indications of fuel performance and are modelled by the Booth equation. The fission product release legacy code NOBLEG for noble gases and halogens was developed previously to calculate this diffusion model for high temperature reactors. The model and code are verified and validated for use in PBMR design and analyses under normal operating conditions. The history of irradiation experiments conducted on coated fuel particles and spherical fuel elements was investigated, and the most suitable irradiation tests with their post irradiation investigations were identified for the purpose of validation of the model and code. The model used to determine gaseous fission product behaviour and release from spherical fuel elements is described in detail. The application of this model in the code is verified mathematically with the Booth model, and by inspection of the source code. The thermohydraulic model used by NOBLEG to calculate fuel temperatures is verified with code to code comparisons with the core neutronics code VSOP. The irradiation tests HFR-K5 and -K6 were selected to validate the gaseous fission product code NOBLEG. An investigation was done into the development of NOBLEG to calculate gaseous fission product release under oxidizing conditions caused by water ingress events. New relationships were derived from water vapour injection tests done during the irradiation experiment HFR-K6, that allows NOBLEG to estimate the increase in gaseous fission product release under oxidizing conditions. A new model was proposed to explain peculiarities observed during the water injection tests. / Prof. P.P. Coetzee

No description available.

Fluoride salt-cooled high-temperature reactor (FHR) is a safer and potentially less expensive alternative to light water reactor due to the low pressure of primary system, passive decay heat cooling system, chemically inert coolant salt, and high-temperature power cycle. However, one challenge presented by this reactor is that fission products may leak into the primary system from its TRISO particle fuel during normal operation. Consequently, the circulating fission products within the primary coolant would be a potential radioactive source. On the other hand, the containment material of the molten salt such as nickel-based alloys may be corroded, and its species may stay in the salt. Thus, the installment of the primary coolant clean-up system and the study on the contaminant species' chemistry and separation are necessarily needed. Electrochemical separation technique has been proposed as the online coolant clean-up method for FHR for removing some impurities from the salt such as lanthanides and corrosion products. The present research focuses on the electrochemical separations of fission products and corrosion products in molten FLiNaK salt (46.5LiF-11.5NaF-42KF mol%) which is the surrogate of the primary coolant candidate FLiBe (67LiF-33BeF2, mol%) for FHR. The main objective is to investigate the electrochemical behaviors of fission products and corrosion products in molten FLiNaK salt to achieve its separations, and provide fundamental properties to instruct the conditions needed to be applied for a desired electrochemical separation. La and Ce are two main elements concerned in this study since they are major lanthanide fission products. Electrochemical behavior of LaF3 in molten FLiNaK salt was studied on both W and Mo inert working electrodes. Although the standard reduction potential of La (III) is more cathodic than that of the primary salt melt constituents K (I) and Na (I), the electrochemical separation of La from molten FLiNaK salt was achieved by merely using inert working electrode because of the formed LaF63- when KF or NaF exists as the salt constituents. Fundamental properties of La in molten FLiNaK salt were also studied at various situations by electroanalytical methods including cyclic voltammetry (CV), chronopotentiometry (CP), and potentiodynamic polarization scan (PS). Ce is another fission product to be separated out from molten FLiNaK salt. Both inert (W) and reactive working electrodes (Cu and Ni) were utilized to realize the extraction of Ce. The electrochemical behaviors of Ce observed on inert W electrode are similar to the ones obtained in FLiNaK-LaF3 system. Reactive electrodes Cu and Ni were used to precede the electrochemical deposition potential of Ce by forming intermetallic compounds. It turned out only Ni electrode was feasible for preceding the deposition potential and the intermetallic compound was identified as CeNi5. The dissolution of chromium metal in the form of chromium fluoride into molten FLiNaK salt is the main concern of alloy corrosion in FHR. To understand the alloy corrosion and removal of the corrosion products from the FHR salt coolant, the electrochemical behavior and fundamental properties of Cr in molten FLiNaK salt were investigated in the present study as well. A new analysis method for the Cr two-step electrochemical reaction in the salt was developed. The method can be applied to other two-step reactions as well. Liquid bismuth was proposed to be the extraction media for liquid/liquid multistage separation of fission products in molten salt reactor. It also can be used as the cathode to extract the fission product of which the electrodeposition potential is close to or more negative than that of the main constituents of molten salt. Activity and activity coefficient are essential factors for assessing the extraction behavior and viability of bismuth in separating fission products. Hence, in the present study, the activity and activity coefficient of fission products and alkali metals (Li and K) at different concentrations and temperatures were determined by experiment and simulation methods respectively. To conduct the parametric study for the electrochemical reaction process and predict fundamental properties, an electrochemical model including single-step reversible, irreversible, and quasi-reversible reactions, multiple-reaction, and two-step consecutive charge transfer reaction was developed based on MOOSE. Although the model was not applied to analyze the experimental data in the present study, this model provides an efficient and easy way to understand the effect of various parameters on electrochemical reaction process. The present study supplied a comprehensive study on the electrochemical separation of fission products and corrosion products in molten FLiNaK salt and will contribute greatly to the development of FHR. / Doctor of Philosophy / There is a significant increased demand for the generation of electricity with the fast development of modern society and economy. For well over 100 years, the dominant energy sources for producing electricity in the industrialized world are fossil fuels, notably coal, oil, and natural gas. The generation of electricity from fossil fuels is a major and growing contributor to the emission of greenhouse gases that contribute significantly to global warming. As clean and efficient energy, the nuclear power source has been an attractive alternative to traditional fossil fuels. The fluoride salt cooled high temperature reactor (FHR) is a promising Generation-IV advanced nuclear reactor. FHR is a salt-cooled reactor in which the core contains a solid fuel and liquid salt coolant. It combines attractive attributes from previously developed reactors and has the advantages of, for example, low-pressure operation, high temperature power cycle, and passive decay heat rejection. However, the primary salt coolant can unavoidably acquire fission products from the fuel particles and corrosion products from structural material corrosion. Therefore, it is necessary to have a primary coolant clean-up system installed in the FHR to mitigate the contamination and ensure the continued operation of the reactor. Electrochemical separation technique has been proposed as the online coolant clean-up method for FHR. Electrochemical separation can be typically done in a three-electrode cell system (working, counter, and reference electrodes). Through applying a proper electrical potential or a current, the target metal ions in the molten salt will be deposited on the working electrode. In that way, the contaminants, including fission products and corrosion products, can be taken out with a working electrode from the molten salt coolant. In this study, the fundamental behaviors of separation of La, Ce (represent lanthanide fission products) and Cr (represents corrosion products) in FLINAK were investigated. To achieve their separations, the present dissertation provided a comprehensive study about the electrochemical behaviors of La, Ce, and Cr species in molten FLiNaK salt at various situations, and relevant fundamental properties for guiding the conditions needed to be applied for the desired electrochemical separation. Considering the use of liquid bismuth as the extraction media for liquid/liquid separation and the electrode for electrochemical separation of fission products the fundamental properties of fission products and alkali metals in liquid bismuth are also determined in the present study to evaluate the separation behavior and viability. Finally, an electrochemical model for understanding the electrochemical process in the FHR salt coolant clean-up was developed. Overall, the work performed in this study will contribute greatly to facilitate the FHR development.

In the event of a nuclear power plant accident, fission products could be released into the atmosphere potentially affecting the health of local citizens. In order to triage the possibly large number of people impacted, a detection device is needed that can acquire data quickly and that is sensitive to internal contamination. The portal monitor TPM-903B was investigated for use in the event of a fission product release. A list of fission products released from a Pressurized Water Reactor (PWR) was generated and separated into two groups--Group 1 (gamma- and beta-emitting fission products) and Group 2 (strictly beta-emitting fission products.) Group one fission products were used in the previously validated Monte Carlo N-Particle Transport Code (MCNP) model of the portal monitor. Two MIRD anthropomorphic phantom types were implemented in the MCNP model--the Adipose Male and Child phantoms. Dose and Risk Calculation software (DCAL) provided inhalation biokinetic data that were applied to the output of the MCNP modeling to determine the radionuclide concentrations in each organ as a function of time. For each phantom type, these data were used to determine the total body counts associated with each individual gamma-emitting fission product. Corresponding adult and child dose coefficients were implemented to determine the total body counts per 250 mSv. A weighted sum of all of the isotopes involved was performed. The ratio of dose associated with gamma-emitting fission products to the total of all fission products was determined based on corresponding dose coefficients and relative abundance. This ratio was used to project the total body counts corresponding to 250mSv for the entire fission product release inhalation--including all types of radiation. The developed procedure sheets will be used by first response personnel in the event of a fission product release.

O objetivo dessa dissertação foi analisar a influência de possíveis reações de fusão nuclear quando elementos leves, definidos como alvos são adicionados ao combustível nuclear. A energia cinética dos produtos de fissão, na concepção desse estudo, é utilizada para vencer a repulsão elétrica entre os núcleos. Para esse trabalho foram considerados a adição de hidrogênio, deutério, lítio e berílio em fração mássica de 0,30. A escolha desses elementos foi realizada através da análise da energia de ligação, do número atômico e das seções de choque experimentais de fusão. A avaliação da fusão nuclear foi realizada via principio da incerteza, pois não há um modelo que pudesse cobrir um grande número de reações possíveis. A probabilidade de reação foi aproximada como uma função da probabilidade de um produto de fissão encontrar determinado elemento leve. Adicionalmente foram calculados os tempos de freamento eletrônico para os produtos de fissão com isso verificou-se que a escala de freamento não representa limitação para a fusão. A principal alteração, devido à adição dos alvos, foi na distribuição dos tempos de meia vida dos produtos de fissão, na direção de maiores tempos. A alteração nessa grandeza indica que é possível reduzir a atividade média do rejeito nuclear, em especial o de alto nível de radioatividade, tornando sua manipulação mais segura. Os alvos de hidrogênio e lítio apresentaram os melhores resultados para o aumento do tempo de meia vida médio dos produtos de fissão, conseguindo altera-la em média duas ordens de grandeza. Outras alterações, que não foram alvo desse estudo, são esperadas devido à inserção desses alvos, como por exemplo, na cinética. / The objective of this dissertation was to analyze the influence of possible nuclear fusion reactions when light elements, defined as targets are added to the nuclear fuel. The kinetic energy of the fission products, in the conception of this study, is used to overcome the electrical repulsion between the nuclei. For this work were considered the addition of hydrogen, deuterium, lithium and beryllium in mass fraction of 0.30. The choice of these elements was performed by analyzing the binding energy, the atomic number and experimental fusion cross sections. The assessment of fusion was performed via the uncertainty principle, since there is no model that could cover a large number of possible reactions. The probability of reaction was approximated as a function of the probability of fission product find a particular light element. Additionally times were calculated for the electronic stopping fission products where was found that scale of stopping does not impose limits on fusion. The main change due to the addition of the targets was the distribution of halflives of the fission products, towards longer times. The change in this magnitude indicates that it is possible to reduce the average activity of the nuclear waste, in particular the high level of radioactivity, making handling safer. The targets of hydrogen and lithium showed the best results for increasing the half-life of the average fission products, managing change a on the average of two orders of magnitude. Other modifications, which were not discussed in this study, are expected due to the insertion of these targets, e.g. kinetics.