This work deals with questions of spent fuel storage casks thermal and physical properties investigation. Foundations of mathematics which are necessary for describing field of temperature are included. The work itself contains calculation methods which are split into two parts. The first one deals with simplified analytic solution and the second part solves the whole problem by the numerical computation.

NR 20140805

Oxidative dissolution of std. UO2 and UO2 doped with Cr2O3 and Al2O3, i.e. ADOPT, induced by H2O2 and γ radiation has been the main focus in this licentiate thesis. The catalytic decomposition of H2O2 on oxides like Gd2O3, HfO2, CeO2, Fe2O3 and CuO were also investigated. A kinetic study was performed by determining first and second order rate constants together with Arrhenius parameters for the decomposition of H2O2. The reactivity of H2O2 towards the oxides mentioned was observed to differ significantly despite their similarities. In the mechanistic study, the yields and dynamics of the formation of the intermediate hydroxyl radical from the decomposition of H2O2 was determined for the oxides and found to differ considerably. A turnover point could be found for most of oxides studied, i.e. an increase in the rate of hydroxyl radical scavenging after a specific amount of consumed H2O2. The reactivity of the std. UO2 and ADOPT towards H2O2 was similar to what was observed for other UO2-based materials in previous studies. The oxidative dissolution in radiation experiments showed a slight but significant difference. This was attributed to a difference in exposed surface area instead of an effect of doping. The difference in oxidative dissolution yield was too small to be significant which supports the previous conclusion. Leaching experiments using spent nuclear fuel were also performed on the two types of fuel showing the same behavior as the unirradiated pellets, i.e., a slightly lower 238U release from ADOPT. The difference was attributed to difference in exposed surface area. The release of fission products with low UO2 solubility displayed a higher release from ADOPT which was attributed to a difference in matrix solubility. Cs was released to a larger extent from std. UO2. This is attributed to the larger grain size of ADOPT, extending the diffusion distance. The release of lanthanides and actinides was slightly higher for the conventional UO2, nevertheless the difference was relatively small. / <p>QC 20140527</p>

Denna licensiatavhandling behandlar oxidativ upplösning av UO2. Upplösning av UO2 studeras huvudsakligen då UO2-matrisen hos använt kärnbränsle förväntas fungera som en barriär mot frigörande av radionuklider i ett framtida djupförvar. Lösligheten av U(IV) är mycket låg under i djupförvaret rådande förhållanden emedan U(VI) har betydligt högre löslighet. Oxidation av UO2-matrisen kommer därför att påverka dess löslighet och därmed dess funktion som barriär. I denna avhandling studeras den relativa effektiviteten av en- och två-elektronoxidanter för upplösning av UO2. Vid låga oxidantkoncentrationer är utbytet för upplösningen för en-elektronoxidanter signifikant lägre än för två-elektronoxidanter. För en-elektronoxidanter ökar dock utbytet med ökande oxidanthalt, vilket kan förklaras av den ökade sannolikheten för två konsekutiva en-elektronoxidationer av samma reaktionssite och den ökade möjligheten till disproportionering. Radikaler och molekylära radiolysprodukters relativa inverkan på oxidativ upplösning av UO2 studeras också i denna avhandling genom mätning av mängden upplöst U(VI) i γ-bestrålade system som dominerades av olika oxidanter. Dessa studier visade att upplösningshastigheten av UO2 kan uppskattas från oxidantkoncentrationer framtagna genom simuleringar av radiolys i motsvarande homogena system och hastighetskonstanterna för ytreaktionerna. Simuleringarna visar att de molekylära oxidanterna kommer vara de viktigaste oxidanterna i alla system i denna studie vid långa bestrålningstider (&gt;10 timmar). Vid liknande simuleringar av α-bestrålade system fanns att vid förhållanden relevanta för ett djupförvar för använt kärnbränsle, är det endast de molekylära oxidanterna (i huvudsak H2O2) som är av betydelse för upplösningen av bränslematrisen. Då använt kärnbränsle innehåller en mängd radionuklider som utsätter UO2-matrisen för kontinuerlig bestrålning, är det av vikt att undersöka hur bestrålning påverkar reaktiviteten av UO2. Bestrålningseffekten på reaktionen mellan UO2 och MnO4- studerades. Dessa försök visade att bestrålning av UO2 vid doser &gt;40 kGy leder till att reaktiviteten ökar upp till 1.3 gånger reaktiviteten av obestrålad UO2. Den ökade reaktiviteten kvarstår efter bestrålningen och effekten kan därför möjligen tillskrivas permanenta förändringar i materialet. Vid uppskattning av reaktiviteten hos använt kärnbränsle måste hänsyn tas till denna effekt då bränslet redan efter ett par dagar i reaktor blivit utsatt för doser &gt;40 kGy. Det har tidigare föreslagits att hastigheten för en heterogen västka/fast-fas reaktion är beroende av partikelstorleken hos det fasta materialet, vilket har studerats för UO2-partiklar i denna avhandling. Experimentellt bestämda kinetiska parametrar jämförs med de föreslagna ekvationerna för fyra storleksfraktioner av UO2-pulver och en UO2-pellet. Studien visade partikelstorleksberoendet av andra ordningens hastighetskonstant och aktiveringsenergin för oxidation av UO2 med MnO4- beskrivs relativt väl av de föreslagna ekvationerna. / The general subject of this thesis is oxidative dissolution of UO2. The dissolution of UO2 is mainly investigated because of the importance of the UO2 matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO2-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one- and two electron oxidants in dissolving UO2 is studied. The oxidative dissolution yield of UO2 was found to differ between one- and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation. Furthermore, the relative impact of radical and molecular radiolysis products on oxidative dissolution of UO2 is investigated. Experiments were performed where the amount of dissolved U(VI) was measured in γ-irradiated systems dominated by different oxidants. We have found that the UO2 dissolution rate in systems exposed to γ-irradiation can be estimated from oxidant concentrations derived from simulations of radiolysis in the corresponding homogeneous systems and rate constants for the surface reactions. These simulations show that for all systems studied in this work, the molecular oxidants will be the most important oxidants for long irradiation times (&gt;10 hours). Similar simulations of α-irradiated systems show that in systems relevant for a deep repository for spent nuclear fuel, only the molecular oxidants (mainly H2O2) are of importance for the dissolution of the fuel matrix. The effect on UO2 reactivity by irradiation of the material is of importance when predicting the spent fuel dissolution rate since the fuel, due to its content of radionuclides, is exposed to continuous self-irradiation. The effect of irradiation on the reaction between solid UO2 and MnO4- in aqueous solutions was studied. It was found that irradiation of UO2 at doses &gt;40 kGy increases the reactivity of the material up to ~1.3 times the reactivity of unirradiated UO2. The increased reactivity remains after the irradiation and can possibly be attributed to permanent changes in the material. This issue must be taken into account when predicting the reactivity of spent nuclear fuel since the fuel is exposed to doses &gt;40 kGy after only a few days in the reactor. It has earlier been suggested that the rate of a heterogeneous liquid-solid reaction depends on the size of the solid particles. This was investigated for UO2 particles in this thesis. Experimental kinetic parameters are compared to the previously proposed equations for UO2 powder of four size fractions and a UO2 pellet. We have found that the particle size dependence of the second order rate constant and activation energy for oxidation of UO2 by MnO4- is described quite well by the proposed equations. / QC 20101123

It is accepted worldwide that the best final solution for spent nuclear fuel is to bury it in deep geological repositories. Despite the physical and chemical barriers that are supposed to isolate the nuclear waste for at least 100.000 years, some uncertainty factors may cause underground water to get in contact with the nuclear waste. Due to radioactivity and oxidation under air, dissolution experiments using UO2 pellets are difficult and frequently lead to incoherent results. Therefore, to enable a detailed study of the influence of microstructure and surface properties on the stability of spent nuclear fuel over time, it is necessary to produce analogues that closely resemble nuclear fuel in terms of crystallography and microstructure. At the same time, in-depth understanding of dissolution phenomena is crucial to geological processes such as dissolution precipitation creep and solvent mediated phase transformations. My thesis is based in two manuscripts. Paper I reports the microstructures obtained after sintering CaF2 powders at temperatures up to 1240°C. Pellets with microstructure, density and pore structure similar to that of UO2 spent nuclear fuel pellets were obtained in the temperature range between 900°C and 1000°C. Paper II reports how differences of surface chemistry and crystal symmetry, characteristics of each surface orientation, affect the topography of CaF2 pellets described in paper I during dissolution. I propose that every orientation of the fluorite structure can be decomposed in the three reference surfaces {100}, {110} and {111}. The {111} is the most stable surface with a dissolution rate of the top surface of 1,13x10-9 mol.m-2.s-1, and {112} the less stable surface with a dissolution rate 34 times faster that {111}. Surfaces that expose both Ca and F atoms in the same plan dissolve faster, possibly because the calcium is more susceptible to be solvated. The faster dissolving surfaces are replaced by the more stable {111} and {100} surfaces which causes the development of roughness on the top surface and stabilizes the surface on high energy sites; i.e. pores or grain boundaries. The main consequences of these observations are i) the increase of the total surface area; ii) the decrease of the overall surface energy. I present a dissolution model for surfaces of crystal with different surface energies. The main conclusions are: a) dissolution rates calculated from surface area are over estimated to the real dissolution rate; b) dissolution rates are faster at the beginning of dissolution and tend to diminish with time until a minimum value is reached.

The kinetics for radiation induced dissolution of spent nuclear fuel is a key issue in the safety assessment of a future deep repository. Spent nuclear fuel mainly consists of UO2 and therefore the release of radionuclides (fission products and actinides) is assumed to be governed by the oxidation and subsequent dissolution of the UO2 matrix. The process is influenced by the dose rate in the surrounding groundwater (a function of fuel age and burn up) and on the groundwater composition. In this licentiate thesis the effects of HCO3- (a strong complexing agent for UO22+) and ionic strength on the kinetics of UO2 oxidation and dissolution of oxidized UO2 have been studied experimentally. The experiments were performed using aqueous UO2 particle suspensions where the oxidant concentration was monitored as a function of reaction time. These reaction systems frequently display first order kinetics. Second order rate constants were obtained by varying the solid UO2 surface area to solution volume ratio and plotting the resulting pseudo first order rate constants against the surface area to solution volume ratio. The oxidants used were H2O2 (the most important oxidant under deep repository conditions), MnO4- and IrCl62-. The kinetics was studied as a function of HCO3- concentration and ionic strength (using NaCl and Na2SO4 as electrolytes). The rate constant for the reaction between H2O2 and UO2 was found to increase linearly with the HCO3- concentration in the range 0-1 mM. Above 1 mM the rate constant is independent of the HCO3- concentration. The HCO3- concentration independent rate constant is interpreted as being the true rate constant for oxidation of UO2 by H2O2 [(4.4 ± 0.3) x 10-6 m min-1] while the HCO3- concentration dependent rate constant is used to estimate the rate constant for HCO3- facilitated dissolution of UO22+ (oxidized UO2) [(8.8 ± 0.5) x 10-3 m min-1]. From experiments performed in suspensions free from HCO3- the rate constant for dissolution of UO22+ was also determined [(7 ± 1) x 10-8 mol m-2 s-1]. These rate constants are of significant importance for simulation of spent nuclear fuel dissolution. The rate constant for the oxidation of UO2 by H2O2 (the HCO3- concentration independent rate constant) was found to be independent of ionic strength. However, the rate constant for dissolution of oxidized UO2 displayed ionic strength dependence, namely it increases with increasing ionic strength. The HCO3- concentration and ionic strength dependence for the anionic oxidants is more complex since also the electron transfer process is expected to be ionic strength dependent. Furthermore, the kinetics for the anionic oxidants is more pH sensitive. For both MnO4- and IrCl62- the rate constant for the reaction with UO2 was found to be diffusion controlled at higher HCO3- concentrations (~0.2 M). Both oxidants also displayed ionic strength dependence even though the HCO3- independent reaction could not be studied exclusively. Based on changes in reaction order from first to zeroth order kinetics (which occurs when the UO2 surface is completely oxidized) in HCO3- deficient systems the oxidation site density of the UO2 powder was determined. H2O2 and IrCl62- were used in these experiments giving similar results [(2.1 ± 0.1) x 10-4 and (2.7 ± 0.5) x 10-4 mol m-2, respectively]. / QC 20101116

Chloride-induced stress corrosion cracking (CISCC) has been identified as the main degradation mechanism for spent nuclear fuel dry storage canisters. This type of induced cracking is complex and depends on several factors, such as material composition, exposure temperature, relative humidity, applied tensile stress, and atmospheric salt concentration. An accelerated experiment was designed to simulate marine environments in a controlled fogging chamber to examine 304 and 304L stainless steel U-bend and welded U-bend samples. The samples were exposed to chloride rich and humid fogging in a corrosion chamber at 35℃ continuously for 4 weeks, 8 weeks, and 12 weeks. The same experiment was repeated at 50℃ for 4 weeks, 8 weeks, and 14 weeks to study the sensitivity of CISCC to temperature changes. A qualitative evaluation of optical micrographs from a 3D Surface Profiler was performed for 16 corroded samples and compared with 2 reference samples. Cracking was observed on 12 out of 16 samples exposed to 35℃ and 50℃ for durations ranging from 8 to 14 weeks. Likely cracking observations were noted on 4 out of 16 samples. A quantitative statistical analysis was also performed using surface profile depth (valley) data from corroded and reference samples. The quantitative analysis examined the effect of temperature, welding, exposure duration, and material composition. The quantitative results were compared with the qualitative results and literature published in CISCC. / Master of Science / Most nuclear power plants are currently using dry storage canisters (DSCs) which are made of a concrete vault and a stainless steel canister that houses the spent nuclear fuel (SNF) assemblies. Multiple conditions must be present simultaneously for chloride-induced stress corrosion cracking (CISCC) to develop, such as the presence of a susceptible alloy, high relative humidity, high temperature, high atmospheric salt concentrations, and applied tensile stresses. DSCs are typically made from 300-series austenitic stainless steels which are susceptible to this type of corrosion during long-term storage near marine environments. Therefore, understanding of the factors leading to CISCC is critically important for proper management and mitigation and to estimate the service life of DSCs for the safe long-term storage of SNF. An accelerated experiment was designed to examine the effects of temperature, exposure duration, and welding on pitting and cracking for 304 and 304L U-bend samples. The experimental results concluded that stainless-steel grades 304 and 304L are susceptible to CISCC when exposed for 8 weeks or longer to fogging at temperatures between 35℃ and 50℃, 95% relative humidity, and 5% salt concentration. This study also concluded that increasing exposure duration from 8 to 12 weeks or the temperature from 35℃ to 50℃ had no significant effect on the acceleration of CISCC. Also, unwelded samples were deemed more susceptible to CISCC than welded samples and the susceptibility of 304 and 304L grades were relatively similar.

This thesis presents a comparative analysis of surrogate models for the dissolution of spent nuclear fuel, with a focus on the use of deep learning techniques. The study explores the accuracy and efficiency of different machine learning methods in predicting the dissolution behavior of nuclear waste, and compares them to traditional modeling approaches. The results show that deep learning models can achieve high accuracy in predicting the dissolution rate, while also being computationally efficient. The study also discusses the potential applications of surrogate modeling in the field of nuclear waste management, including the optimization of waste disposal strategies and the design of more effective containment systems. Overall, this research highlights the importance of surrogate modeling in improving our understanding of nuclear waste behavior and developing more sustainable waste management practices.

Nuclear energy has gained crucial importance since it has a minor impact on climate change and greenhouse gas releases; additionally, the other energy sources are insufficient to reach the world's energy needs without nuclear energy. Another sign that the Generation IV International Forum (Kelly, Gen IV International Forum: A decade of progress through international cooperation, 2014) has pointed out is to utilize uranium resources to the maximum and recycle spent nuclear fuel through burn-up in the Generation IV reactor designs, one of which is the molten salt reactor (MSR). Therefore, the MSR can use the spent nuclear fuel as a fresh fuel when the actinides recycle. That reprocessing of spent fuel could be one of the opportunities to contribute to future nuclear energy goals. This study aims to develop a modified pyroprocessing method to prepare molten salt fuels for MSR from spent oxide nuclear fuel that was burned in light water reactors (LWRs). The process diagram illustrated as (1) spent fuel treatment, (2) chopping and voloxidation of spent oxide fuel, (3) oxide reduction of spent fuel, and then depending on the fuel structure and composition for the MSR, it continues by one or two of the following; – electrorefining, – chlorination, and – fluorination. The subject of this study focused on oxide reduction in two categories: chemical reduction and electrochemical reduction. The system designs have been optimized in calcium salts since they have high calcium metal and calcium oxide solubility. The significant results indicated that both methods would substantially reduce the solid uranium dioxide pellet. The chemical reduction will reduce the total solid pellet at 850oC in the composition of 55.73mol%CaCl2-12.37mol%CaF2-26.58mol%Ca-5.32mol%UO2 over 12 hours. The total reduction in the electrochemical test is seen at 850oC during 12 hours with a salt composition of 79mol%CaCl2-17mol%CaF2-4mol%CaO. These oxide reduction mechanisms are convenient ways to reprocess spent oxide fuel from LWRs to utilize in the MSR. Additionally, the reduced fuel is also applicable to using other next-generation reactors. The prospect of this research is the explicit comparison between chemical and electrochemical methods in calcium salts. / M.S. / Nuclear energy is a crucial energy production to meet the world’s future energy needs. The 6 (six) next-generation reactor design has been determined based on their sustainability, economic, and peaceful application for the world. One of those designs is molten salt reactors (MSRs) which have more attention due to their fuel choice. Most MSRs use the reprocessed fuel from current reactors or the fuel with the breeder blanket that creates more fuel while the reactor operates. This study aims to provide a diagram showing the various steps involved in the preparation of molten salt fuel from spent oxide fuel, which is a mainly utilized form of fuel in current and previous operations. The flowsheet’s first step is the treatment of spent fuel that releases most of the decay heat. The second step is that spent fuel chopping and voloxidation, which meets the requirements of removing gas products and cladding material from used fuel. Afterward, the spent oxide fuel reduces into its metal form chemically or electrochemically in oxide reduction. Then, the molten salt fuel could be fabricated in n one or two more steps from reduced metals: electrorefining, chlorination, or fluorination. Chlorination and fluorination pass through the specific gas components to convert the metal forms into salt. Electrorefining could be applied to arrange the composition of the reduced metal, and this stage is strongly dependent on the MSR designs; it may get eliminated due to its unnecessity. The oxide mechanisms mentioned above were examined under different design conditions to acquire a total reduction of the fuel pellet in calcium salts. The chemical reduction and electroreduction experiments have shown the reduced whole pellet at 850oC with two different salt mixtures. The design impacts of the reduction mechanism were discussed extensively between chemical and electrochemical reductions to identify the benefits and limitations.

In this work, the antineutrino spectrum as specifically generated by nuclear reactors is studied. The topics covered include corrections and higher-order effects in reactor antineutrino experiments, one of which is covered in Ref. [1] and another contributes to Ref. [2]. In addition, a practical application, antineutrino safeguards for nuclear reactors, as summarized in Ref. [3,4] and Ref. [5], is explored to determine its viability and limits. The work will focus heavily on theory, simulation, and statistical analyses to explain the corrections, their origins, and their sizes, as well as the applications of the antineutrino signal from nuclear reactors. Chapter [1] serves as an introduction to neutrinos. Their origin is briefly covered, along with neutrino properties and some experimental highlights. The next chapter, Chapter [2], will specifically cover antineutrinos as generated in nuclear reactors. In this chapter, the production and detection methods of reactor neutrinos are introduced as well as a discussion of the theories behind determining the antineutrino spectrum. The mathematical formulation of neutrino oscillation will also be introduced and explained. The first half of this work focuses on two corrections to the reactor antineutrino spectrum. These corrections are generated from two specific sources and are thus named the spent nuclear fuel contribution and the non-linear correction for their respective sources. Chapter [3] contains a discussion of the spent fuel contribution. This correction arises from spent nuclear fuel near the reactor site and involves a detailed application of spent fuel to current reactor antineutrino experiments. Chapter [4] will focus on the non-linear correction, which is caused by neutron-captures within the nuclear reactor environment. Its quantification and impact on future antineutrino experiments are discussed. The research projects presented in the second half, Chapter [5], focus on neutrino applications, specifically reactor monitoring. Chapter [5] is a comprehensive examination of the use of antineutrinos as a reactor safeguards mechanism. This chapter will include the theory behind safeguards, the statistical derivation of power and plutonium measurements, the details of reactor simulations, and the future outlook for non-proliferation through antineutrino monitoring. / Ph. D.