京都大学 / 0048 / 新制・課程博士 / 博士(生命科学) / 甲第18626号 / 生博第317号 / 新制||生||42(附属図書館) / 31526 / 京都大学大学院生命科学研究科統合生命科学専攻 / (主査)教授 上村 匡, 教授 西田 栄介, 教授 James Hejna / 学位規則第4条第1項該当 / Doctor of Philosophy in Life Sciences / Kyoto University / DGAM

京都大学 / 0048 / 新制・課程博士 / 博士(理学) / 甲第18498号 / 理博第4013号 / 新制||理||1578(附属図書館) / 31384 / 京都大学大学院理学研究科生物科学専攻 / (主査)准教授 半谷 吾郎, 教授 湯本 貴和, 教授 高田 昌彦 / 学位規則第4条第1項該当 / Doctor of Science / Kyoto University / DFAM

The yields of xenon and krypton from the neutron- induced fission of Am241 and Am242 have been measured with a mass spectrometer. This was accomplished by irradiating samples of Am241 for different lengths of time so that the effect of the growth of highly fissionable Am242 could be determined. These studies reveal that both the degree of fine structure in the mass yield curve and the fission-product charge distribution are dependent on the energy of the incident neutrons. This has not been previously observed for any fissioning nuclide. These studies also reveal effects of the 50-neutron shell and of the neutron-proton ratio of the fissioning nuclide on the mass yield curve. / Thesis / Doctor of Philosophy (PhD)

No description available.

The Precision Reactor Oscillation and SPECTrum (PROSPECT) experiment was designed to probe short baseline oscillations of electron antineutrinos in search of eV-scale sterile neutrinos and precisely measure the 235U reactor antineutrino spectrum from the High Flux Isotope Reactor (HFIR) at Oak Ridge national Laboratory (ORNL). The PROSPECT antineutrino detector (AD) provided excellent background rejection due to its segmented design and use of 6Li-loaded liquid scintillator for a neutron capture target. By tracking the neutron capture lifetime from cosmogenic neutrons and a 252Cf neutron source, we suspect the 6Li content of our scintillator changed over time. We look at this evolution and uncertainty in the PROSPECT oscillation and spectrum analyses. Additionally, the 252Cf source data taken with the PROSPECT AD for detector calibrations are used to make a new measurement on the neutron multiplicity probability distribution emitted during spontaneous fissions, with an average multiplicity of 3.81 ± 0.05 neutrons per fission. / Physics

There is a significant need for 3-D steady-state and transient neutron transport formulations and codes that yield accurate, high-fidelity solutions with reasonable computing resources and time. These tools are essential for modeling innovative nuclear systems, such as next-generation reactor designs. The existing methods generally compromise heavily between accuracy and affordability in terms of computation times. In this dissertation, novel algorithms for simulation of reactor transient conditions have been developed and implemented into the RAPID code system. In addition, extensive computational verification and experimental validation of RAPID's steady-state and transient algorithms was performed, and a novel virtual reality system (VRS) web-application was developed for the RAPID code system. The new algorithms, collectively referred to as tRAPID, are based on the Transient Fission Matrix (TFM) methodology. By decoupling the kinetic neutron transport problem into two different stages (an accurate pre-calculation to generate a database and an on-line solution of linear partial differential equations) the method ensures the preservation of the highest level of accuracy while also allowing for high-fidelity modeling and simulation of nuclear reactor kinetics in a short time with minimal computing resources. The tRAPID algorithms have been computationally verified using several computational benchmarks and experimentally validated using the JSI TRIGA Mark-II reactor. In order to develop these algorithms, first the steady-state capabilities of RAPID have been successfully benchmarked against the GBC-32 spent fuel cask system, also highlighting issues with the standard eigenvalue Monte Carlo calculations that our code is capable of overcoming. A novel methodology for accounting for the movement of control rods in the JSI TRIGA reactor has been developed. This methodology, referred to as FM-CRd, is capable of determining the neutron flux distribution changes due to the presence of control rod in real-time. The FM-CRd method has been validated with successfully using the JSI TRIGA reactor. The time-dependent, kinetic capabilities of the new tRAPID algorithm have been implemented based on the Transient Fission Matrix (TFM) method. tRAPID has been verified and validated using the Flattop-Pu benchmark and reference calculations and measurements using the JSI TRIGA reactor. In addition to the main tRAPID algorithms development and benchmarking efforts, a new web-application for the RAPID Code System for input preparation and interactive output visualization was developed. VRS-RAPID greatly enhances the usability, intuitiveness, and outreach possibilities of the RAPID Code System. / Doctor of Philosophy / The simulation of the behavior of nuclear systems (such as power reactors) relies on the development of innovative software that allows for calculating nuclear-relevant quantities in support of the design, operation, and safety of said systems. Traditional codes are often very complex and need to rely on approximations and/or require a very large amount of time to perform even a single calculation. The RAPID Code System is based on a methodology that allows for pre-calculation of a database that can later be used to simulate nuclear systems in real-time while achieving high levels of accuracy. For this dissertation, several new algorithms for simulation of equilibrium and transient conditions of nuclear systems have been developed for the RAPID Code System. In particular, the main features and additions are the ability of simulating the insertion of control rods (devices that are used to control the fission chain reaction) in nuclear reactors and the ability of analyzing the kinetics of nuclear systems. This latter feature, in particular, is extremely important and difficult to simulate, as it involves the fast variation in time of the nuclear quantities under analysis. Finally, a Virtual Reality System (VRS) is embedded with RAPID for easy utilization of the code through a web-application. All these new algorithms and tools have been benchmarked and validated, against reference high-fidelity computational predictions and experimental data. This dissertation demonstrates RAPID's ability of achieving accurate high quality solutions in a rather short time.

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.

Fuel burnup calculations provide material concentrations and intrinsic neutron and gamma source strengths as a function of irradiation and cooling time. Detailed, full-core 3D burnup calculations are critical for nuclear fuel management studies, including core design and spent fuel storage safety and safeguards analysis. For core design, specifically during refueling, full- core pin-wise, axially-dependent burnup distributions are necessary to determine assembly positioning to efficiently utilize fuel resources. In spent fuel storage criticality safety analysis, detailed burnup distributions enable best-estimate analysis which allows for more effective utilization of storage space. Additionally, detailed knowledge of neutron and gamma source distributions provide the ability to ensure nuclear material safeguards. The need for accurate and efficient burnup calculations has become more urgent for the simulation of advanced reactors and monitoring and safeguards of spent fuel pools. To this end, the Virginia Tech Transport Theory Group (VT3G) has been working on advanced computational tools for accurate modeling and simulation of nuclear systems in real-time. These tools are based on the Multi-stage Response-function Transport (MRT) methodology. For monitoring and safety evaluation of spent fuel pools and casks, the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system has been developed. This dissertation presents a novel methodology and algorithm for performing 3D fuel bur- nup calculations, referred to as bRAPID- Burnup with RAPID . bRAPID utilizes the existing RAPID code system for accurate calculation of 3D fission source distributions as the trans- port calculation tool to drive the 3D burnup calculation. bRAPID is capable of accurately and efficiently calculating assembly-wise axially-dependent fission source and burnup dis- tributions, and irradiated-fuel properties including material compositions, neutron source, gamma source, spontaneous fission source, and activities. bRAPID performs 3D burnup calculations in a fraction of the time required by state-of-the-art methodologies because it utilizes a pre-calculated database of response functions. The bRAPID database pre-calculation procedure, and its automation, is presented. The ex- isting RAPID code is then benchmarked against the MCNP and Serpent Monte Carlo codes for a spent fuel pool and the U.S. Naval Academy Subcritical Reactor facility. RAPID is shown to accurately calculate eigenvalue, subcritical multiplication, and 3D fission source dis- tributions. Finally, bRAPID is compared to traditional, state-of-the art Serpent Monte Carlo burnup calculations and its performance will be evaluated. It is important to note that the automated pre-calculation proceedure is required for evaluating the performance of bRAPID. Additionally, benchmarking of the RAPID code is necessary to understand RAPID's ability to solve problems with variable burnups distributions and to asses its accuracy. / Ph. D. / In a nuclear reactor, the energy released from a fission reaction, the splitting of an atomic nucleus into smaller parts, is harnessed to generate electricity. Nuclear reactors rely on fuel, typically comprised of uranium oxide (UO₂). While the reactor is operating and the fuel is being used, or “burned”, for power production it is undergoing numerous nuclear reactions, including fission, and radioactive decays which alter the material composition. Knowing the time evolution of fuel as it is burned in the reactor, i.e., concentration of isotopes and sources of radiation, is critical. Nuclear reactor designers and operators use this information to optimize power production and perform safety analysis of used nuclear fuel. By performing fuel burnup calculations, material concentrations and radiation source strengths are obtained as a function of time in an operating nuclear reactor. Using traditional computational techniques, these calculations are extremely time consuming and, for certain problems, can be difficult to obtain an accurate solution. Ideally, a reactor designer would like to know the three-dimensional (3D) distribution of material compositions and sources; however this level of detail would require an excessive amount of calculation time, therefore simplified models and assumptions are used. For the design of the new generation of nuclear reactors, and monitoring and safeguards analysis, this level of detail will be required in lieu of the availability of experimental facilities which do not currently exist. This dissertation presents a novel methodology and algorithm for performing accurate 3D fuel burnup calculations in real-time, referred to as bRAPID (Burnup with RAPID). bRAPID utilizes an existing nuclear software, RAPID (Real-time Analysis for Particle transport and In-situ Detection), developed in the Virginia Tech Transport Theory Group (VT3G), which has been shown to accurately solve time-independent nuclear calculations in significantly less time than traditional approaches. bRAPID is capable of accurately calculating 3D material and source distributions as a function of time in an operating nuclear reactor, and requires significantly less time and computational resources than traditional approaches. To ensure that bRAPID is relatively easy to use, a number of automated routines have been developed and are presented. RAPID is benchmarked against the traditional code systems MCNP (Monte Carlo N-Particle) and Serpent, both of which are widely used in the nuclear community, for a spent fuel storage pool and the U.S. Naval Academy subcritical nuclear reactor facility. RAPID is shown to accurately calculate system parameters (eigenvalue and subcritical multiplication factor) and 3D fission source distributions. Finally, bRAPID is compared to the traditional burnup approach, using the Serpent code system. bRAPID is shown to accurately calculate system parameters and 3D material and source distributions in significantly less time than the traditional approach.

Most cellular proteins retain a stable concentration as cells grow and divide, but there are exceptions. Some cell cycle regulators change in concentration with cell size. In fission yeast, Cdc13 (cyclin B), an important activator of the core cell cycle kinase Cdc2 (CDK1), increases in concentration as cells grow. It has been proposed that the concentration of such cell cycle regulators serves as a proxy for cell size and makes cell cycle progression dependent on cell size, thereby contributing to cell size homeostasis. The underlying mechanisms for the size-dependent scaling of these cell cycle regulators are poorly understood. Here, I show that Cdc13 protein concentration, but not mRNA concentration, increases with cell size. Furthermore, only the nuclear, but not the cytoplasmic, fraction of Cdc13 increases in concentration as cell size increases. Computational modeling along with half-life measurements suggests that stabilization of Cdc13 in the nucleus plays an important role in establishing this pattern. Taken together, my results suggest that Cdc13 scales with time, and therefore only indirectly—not directly—with cell size. This leaves open the possibility that Cdc13 contributes to cell size homeostasis, but in a different way than originally proposed. / Master of Science / Cells maintain their size very efficiently, but how they manage to do so is not well characterized. It has been suggested that cells sense their size by the size-dependent concentration changes of cell cycle proteins. I have investigated how cyclin B may serve as such a proxy for cell size in fission yeast. My data suggest that fission yeast cyclin B indirectly scales with cell size through an unknown time-based mechanism.

Theoretical models are used in support of the I2S-LWR (Integral Inherently Safe LWR) project for a direct comparison of fuel swelling and fission gas bubble formation between U₃Si₂ and UO₂ fuels. Uranium silicide is evaluated using a model developed by Dr. J. Rest with the fuel in a amorphous state. The uranium dioxide is examined with two separate models developed using a number of papers. One model calculates the swelling behavior with a fixed grain radius while the second incorporates grain growth into the model. Uranium silicide rapidly becomes amorphous under irradiation. The different mechanisms controlling the swelling of the fuels are introduced including the knee point caused by the amorphous state for the U₃Si₂. The outputs of each model are used to compare the fuels.