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The UK nuclear innovation programme supported by the government includes preparation for future ABWR construction. The UK has significant expertise in building and operating gas-cooled nuclear reactors and some experience with PWRs, while there is limited knowledge in BWR technologies. Hence, an important aim of this work is to understand the discrepancies between codes to assess uncertainties in BWR lattice and depletion calculations, while identifying specific development demands to progress existing tools into extended applications. The objective of the study is to quantify the discrepancy between SCALE-6.2, CASMO5 and the UK WIMS-10A deterministic lattice code for BWR lattice and burnup modelling. Two models of BWR systems were considered for this new systematic comparison. They are a single BWR pin-cell with UO2 fuel only, and a 3 by 3 array of BWR UO2 fuel rods with gadolinia rod in the centre. Criticality over burnup was estimated for both models using these codes. Spectral indexes, number densities and neutron spectrum were compared for several burnup stages using SCALE-6.2 and WIMS-10A. The study showed that kinf obtained with CASMO5 was in a good agreement with the SCALE-6.2. A clear discrepancy in behaviour was observed between WIMS-10A and SCALE-6.2 as well as CASMO5. [ABSTRACT FROM AUTHOR]

• The burnup credit calculations for PBR spent fuel pebbles were performed using MCNP6. • The impact on the k eff as a function of water density in the storage cask was investigated. • Three single effects and four compound effects were studied. • The mechanisms of all the single and compound effects were researched in depth. This study attempts to investigate the influence due to the use of burnup credit in the criticality safety analysis for pebble-bed reactor (PBR) spent fuel pebbles. Recently, the development of PBR is very quick in China, thus the storage of spent fuel pebbles will become a significant issue in the foreseeable future. For the safe storage, the criticality safety analysis is of vital important. Furthermore, the utilization of burnup credit in the criticality safety analysis is essential due to the fact that it can result in a more compact design of storage system. In this study, all the calculations were performed using MCNP6 associated with the continuous energy neutron data library ENDF/B-VII. In addition, two geometrical models which represent the HTR-10 core and a proposed storage cask were adopted. The proposed storage cask model was established based on the real storage cask for HTR-10 spent fuel pebbles, but the capacity was increased in order to evaluate the impact of burnup credit. Moreover, the most conservative condition for this proposed cask occurs when the cask is full of water with a water density around 0.35 g/cm3. For burnup credit calculations, three operating parameters related to HTR-10 were investigated, including the usage of control rod, the fuel temperature, and the volume fraction of fuel pebbles in total pebbles. Additionally, these operating parameters were combined and classified as three single effects and four compound effects. The single and compound effects were defined as the influence on the effective multiplication factor (k eff) due to simultaneous variations of one and multiple operating parameters, respectively. It is worth noting that in most of the compound effects, the reactivity deviation (or the change of k eff , Δ k) resulting from the compound effect was not a summation of Δ k 's resulting from the associated single effects. This phenomena may affect the precise assessment to some extent. Finally, the mechanisms of both the single and compound effects were explored in depth from the analysis of the spectral distribution of fission. From this analysis, a harder spectral distribution of fission resulting from the single and compound effects corresponds a larger value of Δ k. [ABSTRACT FROM AUTHOR]

This study employs nuclide sensitivity coefficients to investigate the spent fuel reactivity of pressurized water reactors (PWRs) in terms of differences in predicted inventories under various depletion conditions. Considering a representative PWR spent fuel, the study began by selecting the top 20 nuclides of large macroscopic neutron absorption cross sections, which are expected to have a substantial effect on reactivity. The sensitivity coefficients (Δ k ∞ /Δ N i ) of each nuclide were first evaluated at various burnups and cooling times using MCNPX. These coefficients were used to facilitate a detailed analysis of the spent fuel reactivity and to identify which nuclides cause the effect in a quantitative manner. The approach was demonstrated by analyzing the effects on spent fuel reactivity of several depletion parameters, including the initial enrichment, specific power, operating history, boron concentration, fuel temperature, moderator temperature, control rod effect, and axial burnup profile. The results were generally in good agreement with those obtained from direct MCNPX criticality calculations. Two major advantages of this approach are: (1) the changes in spent fuel reactivity with various depletion parameters can be closely related to composition changes; and (2) it provides a quantitative estimate of the contribution of each nuclide to the reactivity effect based on the predicted inventory differences. [ABSTRACT FROM AUTHOR]