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The isotope 232U is being considered for additive to uranium fuel for use as a tracer. The 232U decay chain has high energy gammas that can be used for tracer purposes. However, the presence and intensity of such gammas may increase effective dose rates to workers around such materials This study examines the dose rate from different uranium materials that varying amounts of 232U has been added to. Several materials and their respective storage geometries are modeled for particle transport calculations. For each material and 232U concentration, gamma source terms were generated using SCALE 6.2 ORIGEN. These source terms were then used in MCNP for each material and geometry. Upon analysis, it was determined that the baseline dose from enriched uranium dominates until the 232U concentration reaches about 10–100 parts per trillion (ppt), after which the dose rate increases linearly. • Uranium-232 is considered for use as a tracer in uranium fuels, but could lead to increased radiation dose. • Several stages of the nuclear fuel cycle are modeled with uranium-232 added. • Various concentrations of uranium-232 are examined. • Dose from uranium-232 becomes the primary contributor to dose between 10 and 100 parts per trillion. [ABSTRACT FROM AUTHOR]

• Exact 3D full core burnup calculation via stochastic codes is complicated and challenging problem. • Two step approach is currently widely used for full core neutronic calculation. • Microdepletion method can be implemented in the two step approach and it can evaluate burnup transmutation of the fuel in the core with limited calculation requirements. • Analysis of Plutonium quality after first year of operation of VVER-1000 nuclear reactor was conducted • Spent nuclear fuel contains around 500 kg of Plutonium after one year of operation. • The Plutonium amount with Pu-239 fraction higher than 80% is less than 8 Kg after irradiation process. Detailed three-dimensional (3D) spent nuclear fuel (SNF) isotopic composition analysis using a full-core 3D nodal simulator is not a widespread practice within production-level nuclear analysis fuel management codes. A 3D full-core pin-wise detailed isotopic analysis can be very time consuming, so more often than not, reduced geometries are used, either based on two-dimensional (2D) lattice-physics, single fuel pins, or point-reactor (0D) estimations. In fact, the typical macroscopic cross section based codes do not usually have the ability to evaluate the full detailed composition of SNF, but instead may track only a few isotopes in spatially homogenized regions. Accordingly, a microscopic cross section based depletion sequence was revisited within the otherwise macroscopic-based NESTLE code. To illustrate this capability, the plutonium content of SNF within a VVER-1000 benchmark was calculated and its non-proliferation risk was estimated. The total plutonium masses and 239Pu fractions were calculated on a 3D node-by-node basis using an established VVER-1000 NESTLE benchmark model. The total plutonium mass after the first cycle is estimated at 502.6 kg, and it increased to 523.4 kg after 90 days of decay. The mass of plutonium with a 239Pu content greater than 80% at the end of cycle is about 7.8 kg, while all remaining content shows to have a 239Pu fraction below 87%. The plutonium content with a 239Pu fraction greater than 80% after 20 days of the end of cycle is less than 7.9 kg. This article highlights a feature of the NESTLE code that can help make isotopic assessments radially and axially for fuel assemblies located in different regions of the core, thus accounting for the varying 3D conditions and flux distributions as a function of exposure. [ABSTRACT FROM AUTHOR]

• CASL VERA represents a new tool for fuel loading pattern optimization for SMR simulations. • The single fuel assembly calculations provided by the CASL VERA are in a good agreement against the Polaris or Serpent calculation. • The 3D full core calculations with water reflector are in a good agreement against Serpent calculation code. • The 3D full core calculations with stainless steel reflector are in a worse agreement with Serpent calculation code. The optimal stainless steel fraction to het reasonably good results against the Serpent code, was set up to 90%. • Stainless steel reflector extends core operation, moreover it balances the power generation in the SMR NuScale core. Three-dimensional (3D) full-core calculations are an integral part of fuel reload design for today's light water reactors (LWRs). The current approaches are typically based on the nodal diffusion codes that calculate criticality state points and power distributions as a function of burnup. The CASL VERA (Consortium for Advanced Simulation of Light Water Reactors, Virtual Environment for Reactor Applications) code represents one of the latest advancements in 3D full-core calculation analysis based on transport theory methods employing the Method of Characteristics (MOC) and coupled multi-physics. In this article, a publicly released version of the NuScale reactor core is analysed with the VERA code (version 3.9 and 4.1) and contrasted against some static Serpent and Polaris based simulations. The analysis shows an excellent agreement for the lattice-level calculations as well as with some of the 3D full-core models. However, larger deviations were found in cases with heavy reflector models, whereby the reflector composition was found to impact the differences between the VERA and Serpent results. In this analysis, it was determined that greater than 90% stainless steel content in the heavy reflector leads to higher deviations between the VERA results and the Serpent results. The burnup calculations showed that the presence of the heavy reflector extends the cycle length and also leads to a flatter power distribution in the core, which can generally be interpreted as more efficient. [ABSTRACT FROM AUTHOR]

• Benchmark study of a VVER 1000 core simulation of hot zero power (HZP) tests. • The Serpent-NESTLE combination is capable of providing high quality simulations. • The Serpent-NESTLE combination and MCNP, both, provided reasonable results. This article presents a benchmark study of a VVER 1000 core simulation of hot zero power (HZP) tests. Measured experimental data from the Temelin Unit 2 nuclear power plant's operation test during its first reactor startup were compared using MCNP as well as the combination of Serpent + NESTLE lattice-to-diffusion nodal simulation. Results from full-core 3D models included multiplication factor and moderator temperature coefficients (MTC) that were compared to experimental data, which revealed a very good agreement with criticality states during HZP tests, and a reasonable agreement for MTC. Therefore, it is concluded that the Serpent-NESTLE combination and MCNP, both, provided reasonable results for this benchmark. The Serpent-NESTLE code combination was used for the first time in this study and has shown to be capable of providing high quality simulations in an efficient fashion. To encourage future benchmarks by other researchers, this article makes a special effort to document the core design information and details that would facilitate such efforts. [ABSTRACT FROM AUTHOR]