This project aims to investigate and analyze crucial aspects related to the deployment of Small Modular Reactors (SMRs) in Sweden. SMRs have risen as a competitor to large-scale nuclear power plants with cost-competitiveness and cogeneration purposes in focus. Alongside the advantages of SMRs, it is important to address the non-proliferation and safeguards aspects of the deployment of SMRs in Sweden. The first stage of this project aimed to predict the electricity consumption in Sweden in 2045. Based on scenarios made by researchers and authorities, a prediction was made that the electricity consumption in Sweden will be approximately 240 TWh in 2045. Subsequently, a study was conducted to assess the suitability of different SMR concepts for deployment in Sweden. The results were that the GE Hitachi BWRX-300 and the Rolls Royce SMR are the concepts most suitable for electricity generation and the VTT LDR 50-SMR is suitable for district heating in Sweden. Based on the predicted electricity consumption and the predicted need for district heating, the SMRs were deployed in three different scenarios which yield the following results: In scenario i), a total of 15-23 SMRs were deployed to the existing nuclear sites in Sweden. (15 SMRs if BWRX-300 is chosen and 23 SMRs if the Rolls Royce SMR is chosen). In scenario ii), 16-24 SMRs in total were deployed to bidding area 1 (SE1), SE2, and SE4. In scenario iii), a total of 25 new reactors were deployed to eight different sites, close to where the electricity consumption was predicted to be large. The implications on material accounting and safeguards from the deployment in the scenarios were that the transportation and storage of spent nuclear fuel will rise by roughly a factor of two, depending on which scenario is chosen and the choice of SMR. There will be more sites for inspectors to visit. New harbors need to be built and licensed. New transportation infrastructure may need to be developed if scenario three is implemented. Furthermore, a new intermediate storage of spent nuclear fuel may be needed.

The high and ever growing demand for electricity coupled with environmental concerns and a worldwide desire to shed petroleum dependence, all point to a shift to utilization of renewable sources of energy. The under developed nature of truly renewable energy sources such as, wind and solar, along with their limitations on the areas of applicability and the energy output calls for a renaissance in nuclear energy. In this second nuclear era, deliberately small reactors are poised to play a major role with a number of Small Modular Reactors (SMRs) currently under development in the U.S. In this work, an SMR model of the Integral Pressurized Water Reactor (IPWR) type is created, analyzed and optimized to meet the publically available performance criteria of the mPower SMR from B&W. The Monte Carlo codes MCNP5/MCNPX are used to model the core. Fuel enrichment, core inventory, core size are all variables optimized to meet the set goals of core lifetime and fuel utilization (burnup). Vital core behavior characteristics such as delayed neutron fraction and reactivity coefficients are calculated and shown to be typical of larger PWR systems, which is necessary to ensure the inherent safety and to achieve rapid deployment of the reactor by leveraging the vast body of operational experience amassed with the larger commercial PWRs. Inherent safety of the model is analyzed with the results of an analytical single channel analysis showing promising behavior in terms of axial and radial fuel element temperature distributions, the critical heat flux, and the departure from nucleate boiling ratio. The new fleet of proposed SMRs is intended to have increased proliferation resistance (PR) compared to the existing fleet of operating commercial PWRs. To quantify this PR gain, a PR analysis is performed using the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR) code developed by the Nuclear Science and Security Policy Institute at Texas A&M University. The PRAETOR code uses multi-attribute utility analysis to combine 63 factors affecting the PR value of a facility into a single metric which is easily comparable. The analysis compared hypothetical spent fuel storage facilities for the SMR model spent fuel assembly and one for spent fuel from a Westinghouse AP1000. The results showed that from a fuel material standpoint, the SMR and AP1000 had effectively the same PR value. Unable to analyze security systems and methods employed at specific nuclear power plant sites, it is premature to conclude that the SMR plants will not indeed show increased PR as intended.

<p>Small modular reactors (SMRs) are expected as a suitable candidate to fulfill energy needs in the future. The regulation of the emergency planning zone (EPZ) has been a controversial issue. The possibility of smaller EPZs because of their small core size and passive safety functions is still under discussion. The major emergency responses to radiological incidents in the early phase are evacuation from the area and shelter-in-place within a building. Comparison between the dose incurred during evacuation and that with shelter-in-place is necessary to consider the proper protective actions. The effect of shelter-in-place from small modular reactor hypothetical accident was studied. The source term came from a long-term station blackout (LTSBO) and loss of cooling accident (LOCA), and the time change of air concentration and the ground deposition data through the atmospheric spread around the plant was calculated with Radiological Assessment System for Consequence Analysis (RASCAL), a software developed by United States Nuclear Regulatory Commission (NRC) to provide dose projection around the plant. Then general one-story and two-story houses were set up, and 6 wall materials were selected for calculating indoor doses. Cloudshine and groundshine were calculated with Monte Carlo methods. In addition, the conservation of mass, air flow model was established to evaluate the inhalation for sheltered cases. The shielding function of each house for each pathway was evaluated by comparing the indoor dose with outdoor dose. The projected dose for sheltered cases was much smaller than that for unsheltered cases. Even though the projected dose will not completely perish, it was quite effective to reduce radiation exposure and can be superior to evacuation. The result will be a basis for calculating the radiological dose for sheltered cases in case of nuclear emergency for SMRs, which will be valuable to have a more effective emergency planning.</p>

This is a report of a verification study of the multiphysics simulation framework MOOSE which was preformed at the company Seaborg Technologies. In the process of designing molten salt reactors there is a special need of making credible multiphysics simulations since the fuel is in motion. In this study the incompressible version of Navier-Stokes equations of finite volumes available in the Navier-Stokes module of the MOOSE framework is verified by modelling and simulations of fluid flow and heat transfer in two different systems with available benchmarks. The first system, a thin buoyancy driven molten sodium hydroxide test loop which is verified by a similar model made with the high fidelity CFD software STAR-CCM+ as benchmark. The second system, forced convection of air through a straight pipe with heated walls which is verified by comparisons with an analytical solution. The resulting velocity profiles from simulations of the first system corresponds well with the benchmark but certain conclusions can not be drawn from it since the the transient simulations stops to converge before reaching equilibrium. The results from simulations of the second system corresponds well with the analytical solution and no convergence issues arise. The conclusion from the results is that the incompressible version of Navier-Stokes equations of finite volumes available in the Navier-Stokes module of the MOOSE framework has potential to be used in multiphysics simulations of molten salt reactors but seemingly not in cases of buoyancy driven flows in thin geometries. Two proposals for further work is recommended. The first is that this implementation is applied in a context with forced fluid flow or a context with thicker fluid domain. The second proposal is that the other available abilities of MOOSE such as finite element method and/or the compressible version of the Navier-Stokes equations should be tested.

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

Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines, as the fuel used could not conceivably be as highly enriched. Nuclear-powered propulsion would allow ships to operate with low fuel costs, long refueling intervals, and minimal emissions; however, currently such systems remain largely confined to military vessels. This research project undertakes computational modeling of possible soluble-boron-free (SBF) reactor core designs for this application, with a view to informing design decisions in terms of choices of fuel composition, materials, core geometry and layout. Computational modeling using appropriate reactor physics (e.g. WIMS, MONK, Serpent and PANTHER), thermal-hydraulics etc. codes (e.g. COBRA-EN) is used for this project. With an emphasis on reactor physics, this study investigates possible fuel assembly and core designs for civil marine propulsion applications. In particular, it explores the feasibility of using uranium/thorium-rich fuel in a compact, long-life reactor and seek optimal choices and designs of the fuel composition, reactivity control, assembly geometry, and core loading in order to meet the operational needs of a marine propulsion reactor. In this reactor physics and 3D coupled neutronics/thermal-hydraulics study, we attempt to design a civil marine reactor core that fulfills the objective of providing at least 15 effective full-power-years (EFPY) life at 333 MWth. In order to unleash the benefit of thorium in a long life core, the micro-heterogeneous ThO2-UO2 duplex fuel is well-positioned to be utilized in our proposed civil marine core. Unfortunately, A limited number of studies of duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life small modular rector (SMR) core. Therefore, we assumed micro-heterogeneous ThO2-UO2 duplex fuel for our proposed marine core in order to explore its capability. For the proposed civil marine propulsion core design, this study uses 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel. To provide a basis for comparison we also evaluate the performance of homogeneously mixed 15% U-235 enriched all-UO2 fuel. This research also attempts to design a high power density core with 14 EFPY while satisfying the neutronic and thermal-hydraulics safety constraints. A core with an average power density of 100 MW/m3 has been successfully designed while obtaining a core life of 14 years. The average core power density for this core is increased by ∼50% compared to the reference core design (63 MW/m3 and is equivalent to Sizewell B PWR (101.6 MW/m3 which means capital costs could be significantly reduced and the economic attractiveness of the marine core commensurately improved. In addition, similar to the standard SMR core, a reference core with a power density of 63 MW/m3 has been successfully designed while obtaining a core life of ∼16 years. One of the most important points that can be drawn from these studies is that a duplex fuel lattice needs less burnable absorber than uranium-only fuel to achieve the same poison performance. The higher initial reactivity suppression and relatively smaller reactivity swing of the duplex can make the task of reactivity control through BP design in a thorium-rich core easier. It is also apparent that control rods have greater worth in a duplex core, reducing the control material requirements and thus potentially the cost of the rods. This research also analyzed the feasibility of using thorium-based duplex fuel in different cases and environments to observe whether this fuel consistently exhibit superior performance compared to the UO2 core in both the assembly and whole-core levels. The duplex fuel/core consistently exhibits superior performance in consideration of all the neutronic and TH constraints specified. It can therefore be concluded from this study that the superior performance of the thorium-based micro-heterogeneous ThO2-UO2 duplex fuel provides enhanced confidence that this fuel can be reliably used in high power density and long-life SBF marine propulsion core systems, offering neutronic advantages compared to the all-UO2 fuel. Last, but not least, considering all these factors, duplex fuel can potentially open the avenue for low-enriched uranium (LEU) SBF cores with different configurations. Motivated by growing environmental concerns and anticipated economic pressures, the overall goal of this study is to examine the technological feasibility of expanding the use of nuclear propulsion to civilian maritime shipping and to identify and propose promising candidate core designs.

Cílem práce bylo vytvořit základní koncept integrovaného výměníku tepla pro solí chlazený reaktor vyvíjený společností Terrestrial Energy s využitím programu Promex. První kapitola se zabývá historií a současnou situací v oblasti výzkumu malých modulárních reaktorů chlazených fluoridovými solemi. Ve druhé kapitole jsou popsány vlastnosti fluoridových solí a konstrukčních materiálů. Poslední kapitola se zabývá simulací tepelného výměníku pomocí programu Promex, validací modelu, transformací protiproudého výměníku na výměník s U trubkami a vizualizací výměníku s použitím CAD Invetoru.