• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• Applied Chemistry for Engineering
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• v.24 no.3
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• pp.227-230
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• 2013

The phase change of $(U_{1-x}Nd_x)_3O_8$ powder produced by oxidation of Nd-doped $UO_2$ pellet at $500^{\circ}C$ was investigated by high temperature oxidation heat treatment at $900{\sim}1500^{\circ}C$ under an air atmosphere. The XRD analysis results showed that the formation of $(U_{1-y}Nd_y)O_{2+z}$ phase and $U_3O_8$ phase from metastable $(U,Nd)_3O_8$ phase initiated at a temperature of $1000^{\circ}C$. The relative integrated intensity of $(U_{1-y}Nd_y)O_{2+z}$ phase to $U_3O_8$ phase increased with increasing of the oxidation temperature from 1100 to $1500^{\circ}C$. And also, it was found from the SEM observation that the particle size of $(U_{1-y}Nd_y)O_{2+z}$ phase increased with increasing of the oxidation temperature. However, electrone probe X-ray microanalyzer (EPMA) analysis results showed that Nd contents in $(U_{1-y}Nd_y)O_{2+z}$ phase decreased with increasing of the oxidation temperature. This behavior on the ground of XRD, SEM, and EPMA analysis data could be interpreted in terms of the transportation of U ions from $U_3O_8$ phase into $(U_{1-y}Nd_y)O_{2+z}$ phase through the interface of two phases during high temperature oxidation.

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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• pp.407-408
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• 2011

Natural boron consists of two stable isotopes 10B and 11B with natural abundance of 18.8 atom percent of 10B and 81.2 atom percent of 11B. The thermal neutron absorption cross-section for 10B and 11B are 3837 barn and 0.005 barn respectively. 10B enriched specific compounds are used for control rods and as a reactor coolant additives. In this work 2 methods for boron enrichment were analysed: 1) Gas irradiation in static conditions. Dissociation occurs due to multiphoton absorption by specific isotopes in appropriately tuned laser field. IR shifted laser pulses are usually used in combination with increasing the laser intensity also improves selectivity up to some degree. In order to prevent recombination of dissociated molecules BCl3 is mixed with H2S 2) SILARC method. Advantages of this method: a) Gas cooling is helpful to split and shrink boron isotopes absorption bands. In order to achieve better selectivity BCl3 gas has to be substantially rarefied (~0.01%-5%) in mixture with carrier gas. b) Laser intensity is lower than in the first method. Some preliminary calculations of dissociation and recombination with carrier gas molecules energetics for both methods will be demonstrated Boron separation in SILARC method can be represented as multistage process: 1) Mixture of BCl3 with carrier gas is putted in reservoir 2) Gas overcooling due to expansion through Laval nozzle 3) IR multiphoton absorption by gas irradiated by specifically tuned laser field with subsequent gradual gas condensation in outlet chamber It is planned to develop software which includes these stages. This software will rely on the following available software based on quantum molecular dynamics in external quantized field: 1) WavePacket: Each particle is treated semiclassicaly based on Wigner transform method 2) Turbomole: It is based on local density methods like density of functional methods (DFT) and its improvement- coupled clusters approach (CC) to take into account quantum correlation. These models will be used to extract information concerning kinetic coefficients, and their dependence on applied external field. Information on radiative corrections to equation of state induced by laser field which take into account possible phase transition (or crossover?) can be also revealed. This mixed phase equation of state with quantum corrections will be further used in hydrodynamical simulations. Moreover results of these hydrodynamical simulations can be compared with results of CFD calculations. The first reasonable question to ask before starting the CFD simulations is whether turbulent effects are significant or not, and how to model turbulence? The questions of laser beam parameters and outlet chamber geometry which are most optimal to make all gas volume irradiated is also discussed. Relationship between enrichment factor and stagnation pressure and temperature based on experimental data is also reported.

• Journal of Adhesion and Interface
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• v.19 no.3
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• pp.113-117
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• 2018

Boron carbide is lower in hardness than diamond or boron nitride but has a hardness of more than 30 GPa and is used for manufacturing tank armors and ammo shells due to its high hardness. It is also used as a neutron absorber due to its ability to absorb neutrons, which is increasing its use in nuclear power projects. Neutrons have no interaction with electrons and are known to pass through the material without interactions. Along with boron carbide, the atoms with high interaction with neutrons are hydrogen, and high hydrogen concentration polyesters and epoxy polymers including boron are used as materials for manufacturing products for nuclear power generation waste. In this paper, the surface of boron carbide is treated with iron oxide and tungsten to improve interaction between modified boron carbide and epoxy polymer. XRD and XPS were used to confirm that iron oxide and tungsten are well attached on the surface of boron carbide, respectively. The mechanical strength of the surface treated boron carbide was measured by a universal testing machine (UTM) and the dynamic characteristics of the cured product were observed by using a dynamic analyzer (DMA).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.243-260
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• 2018

A literature review on the effects of high temperature and radiation on radiation shielding concrete in Spent Fuel Dry Storage is presented in this study with a focus on concrete degradation. The general threshold is $95^{\circ}C$ for preventing long-term degradation from high temperature, and it is suggested that the temperature gradient should be less than $60^{\circ}C$ to avoid crack generation in concrete structures. The amount of damage depends on the characteristics of the concrete mixture, and increases with the temperature and exposure time. The tensile strength of concrete is more susceptible than the compressive strength to degradation due to high temperature. Nuclear heating from radiation can be neglected under an incident energy flux density of $10^{10}MeV{\cdot}cm^{-2}{\cdot}s^{-1}$. Neutron radiation of >$10^{19}n{\cdot}cm^{-2}$ or an integrated dose of gamma radiation exceeding $10^{10}$ rads can cause a reduction in the compressive and tensile strengths and the elastic moduli. When concrete is highly irradiated, changes in the mechanical properties are primarily caused by variation in water content resulting from high temperature, volume expansion, and crack generation. It is necessary to fully utilize previous research for effective technology development and licensing of a Korean dry storage system. This study can serve as important baseline data for developing domestic technology with regard to concrete casks of an SF (Spent Fuel) dry storage system.