This report investigates the sensitivity of the control rod depletion coefficients, Sg, to different input parameters and how this affects the accumulated 10B depletion, β. Currently the coefficients are generated with PHOENIX4, but the geometries can be more accurately simulated in McScram. McScram is used to calculate Control Rod Worth, which in turn is used to calculate Nuclear End Of Life, and Sg cannot be generated in the current version of McScram. Therefore, it is also analyzed whether the coefficients can be related to CRW and thus be studied indirectly through it. Simulations of the coefficients were done in PHOENIX4, simulations of CRW were done in both PHOENIX4 and McScram and simulations of β were done in POLCA7. All simulations were performed for a CR99 in a BWR reactor. The control rod coefficients were found to be sensitive to the enrichment of the fuel, void fraction of the water and the width of the gap, and these effects were also seen in the results of β. As a result, one of three steps could be taken. First, the parameter values should not be set arbitrarily, instead default values could be chosen such that Sg is calculated more accurately. Second, a set of tables of Sg could be generated for different parameter values so that β can be calculated with Sg from the current conditions, although this would mean that PHOENIX4 needs to be updated. Third, McScram can be updated to be able to calculate Sg directly. It has been concluded that Sg cannot be studied indirectly through CRW since the trends and the sensitivity to the different parameters were not consistent between Sg, CRW calculated with PHOENIX4 and CRW calculated with McScram, where PHOENIX4 was more sensitive than McScram. The results can instead be used to bench-mark the PHOENIX4 results.

The Master Thesis performs search concerning advanced small and medium power light-water reactors’ designs, including different possibilities to gain a license for their development and operation. The work covers the principal theory about the area of neutronics calculations, principal equations and simplifications. There are several different methods for solution of neutronics calculations. The thesis gives an overview of two principal groups of codes – deterministic methods and Monte Carlo method. The survey shows computational codes examples based on mentioned methods. The computational code PARCS is chosen for further study, which contained description of the input and output file, process of the model creation and conditions for neutronics calculation the of selected reactor design. Based on these facts, the transient calculation has been prepared within the thesis. Thee analyses are described – reactor emergency shutdown, reactor shutdown with stuck group of control and emergency shutdown rods and reactor shutdown with faulty reaction of emergency shutdown rods.

In this master thesis an input-model of a Nordic BWR power plant has been developed in APROS. The plant model contains key systems and major thermohydraulic components of the steam cycle, including I&C systems (i.e. power, pressure, level and flow controls). The plant model is primarily designed for balance of plant studies at discrete power levels. The input-model of the power plant focuses especially on the steam cycle which is crucial for analysing water and steam behaviour and its influence on the reactor power. At the current stage, the model primarily handles steady-state conditions of full-power operation, which has been the design point. It has also been shown that reduced-power operation can be simulated with a reasonable trendline of pressure and temperature progression over facility components. / Inom ramen för examensarbete har en indatafil (modell) av en nordisk kokvattenreaktor, BWR, utvecklats i simuleringsverktyget APROS. Anläggningsmodellen är främst utformad för att simulera diskreta effektnivåer och innehåller viktiga system och termohydrauliska komponenter som ingår i ångcykeln, inklusive instrumenterings- och kontrollutrustning (dvs. effekt-, tryck-, nivå- och flödesreglering). Fokus har lagts särskilt på att få till en bra representation av ångcykeln, vilket är avgörande för analys av vatten- och ångbeteendet och dess påverkan på reaktoreffekten. Modellen kan främst användas för simulering av jämviktstillstånd vid full effektdrift och till en viss grad även reducerad effektdrift.

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.