Search Results (6)

To quantify the uncertainties propagating from the fuel depletion calculation to the criticality calculation in the burnup credit system, this paper evaluates the effects of the nuclide concentration uncertainty on the criticality calculation based on Monte Carlo uncertainty sampling methods, and analyzes the assumption that the measured-to-calculated nuclide concentration ratio obeys a normal distribution with uncorrelation among isotopes in the Monte Carlo uncertainty sampling method by using the sensitivity and uncertainty analysis method and the Latin hypercube sampling method. The results indicated that the Monte Carlo uncertainty sampling method could effectively quantify the uncertainties with a calculation accuracy within 3%, and the criticality uncertainty calculation for the assumption that the measured-to-calculated concentration ratios obey normal distributions was more conservative than that of the samples according to their actual distributions. Thus, the assumption of a normal distribution is reasonable in the sampling process. Moreover, the uncertainty results of the criticality calculation considering the correlations among important isotopes presented a decrease of approximately 5% over those without the isotopic correlations. Therefore, introducing the correlations of significant isotopes could reduce the uncertainty of the criticality calculation for spent-nuclear-fuel storage systems. Full article

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO₂ fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies. Full article

This study performed criticality analysis for the GBC-68 storage cask loaded with boiling water reactor (BWR) spent fuel at the discharged burnups obtained from the full-core simulations. The analysis was conducted for: (1) different reloading scenarios; (2) target burnups; and (3) two fuel assembly types—GE14 and SVEA100—to estimate the impact each of the three factors has on the cask reactivity. The BWR spent fuel composition was estimated using the results of the nodal analysis for the advanced boiling water reactor (ABWR) core model developed in this study. The nodal calculations provided realistic operating data and axial burnup and coolant density profiles, for each fuel assembly in the reactor core. The estimated cask’s k_eff were compared with the fresh fuel and peak reactivity standards to identify the benefit of the burnup credit method applied to the BWR spent fuel at their potential discharge burnups. The analysis identified the significant cask criticality reduction from employing the burnup credit approach compared to the conventional fresh fuel approach. However, the criticality reduction was small compared to the peak reactivity approach, and could even disappear for low burnt fuel assemblies from non-optimal reloading patterns. In terms of cask manufacturing, the potential financial benefit from using the burnup credit approach was estimated to be USD 3.3 million per reactor cycle. Full article

Due to the challenges of spent fuel accumulation, the nuclear industry is exploring more cost-effective solutions for spent fuel management. The burnup-credit method, in application for storage and transport of the spent fuel, gained traction over recent decades since it can remove the over-conservatism of the “fresh-fuel” approach. The presented research is focused on creating an innovative, best estimate approach of the burnup-credit method for boiling water reactor (BWR) spent fuel based on the results of neutronic/thermal-hydraulic coupled full core simulations. The analysis is performed using a Polaris/DYN3D sequence. Four different shuffling procedures were used to estimate the possible range of the BWR fuel discharged burnup variation. The results showed a strong influence of the shuffling procedure on the final burnup distribution. Moreover, a comparison of the 2D lattice and 3D coupled nodal approaches was conducted for the criticality estimation of single fuel assemblies. The analysis revealed substantial improvement in criticality curves obtained with the full-core model. Finally, it was shown that the benefit from the burnup-credit method is larger in the case of more optimal fuel utilisation-based shuffling procedures. The new approach developed here delivers a promising basis for future industrial optimisation procedures and thus cost optimisation. Full article

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 k_inf 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. Full article

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO₂ spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, $k_{eff}$ , of the canister would comply with the imposed criticality safety criterion. Full article