Evaluation of Feasibility of Gamma-Spectrometer NaI PAK-01 and SAS Na M3 Software Application for Regular Radionuclide Analysis of NPP Permissible Wastes

The feasibility of correct NaI gamma-spectrometry activity measurement for each nuclide in 131I, 132I, 133I, 134I, 135I, 133Xe, 135mXe, 135Xe and 222Rn composition is presented. To get this result the special matrix method M3 and SAS Na M3 software were used for spectra processing. SAS Na M3 software was developed for complex NaI gamma-spectra processing. Special algorithms and auxiliary software are used to overcome the problems of the classic spectra processing matrix method. Being used for spectrum processing SAS Na M3 software determines the nuclide composition of the sample, activity of nuclides identified and activities uncertainties. The activity values estimation is made for nuclides not identified in the sample measured but included in SAS Na M3 software nuclides library. The values of minimal detectable activities for NaI ∅3''× 3'' gamma-spectrometer and 1 hour measuring time are ~ 0.6 Bq for 131I, 132I, 133I, 134I and ~ 2 Bq for 135I.

Neutron resonance transmission analysis of cylindrical samples used for reactivity worth measurements

A characterisation of cylindrical samples by Neutron Resonance Transmission Analysis (NRTA) at the GELINA facility of JRC Geel (Belgium) is presented. The samples were designed and produced for reactivity worth measurements in the MINERVE reactor of CEA Cadarache (France). NRTA was applied to determine the nuclide composition of UO2, Al2O3 and liquid samples that were doped with silver. The volume number densities of 238U, 107Ag and 109Ag obtained by NRTA are within 2 % fully consistent with the values that are quoted by the manufacturer. In addition, the NRTA data reveal a tungsten contamination which is not reported by the provider. It is shown that such a contamination contributes by up to 5.7 % to the reactivity worth.

ROCOCO system of combined neutron constants – current status and results of testing using geometrical module of the MMK code

Results of studies aimed at the further refinement of the ROCOCO system (routine for calculation and organization of combined constants including cross-sections in group and subgroup representation with detailed description of energy dependence of neutron cross-sections) (Zherdev et al. 2018, Kislitsina et al. 2016) are presented in the paper. Inclusion of this system as a physical module into a set of Monte Carlo calculation codes with OOBG geometric module from the MMK code (Zherdev et al. 2003) is discussed. OOBG module is designed for calculation of neutron multiplication systems with heterogenous cores arranged as hexagonal grids with different degrees of complexity. The name ROCOCO-MMK was assigned to the complex. Results of testing the complex in the calculations of multi-zone neutron multiplication systems (including those with zones containing neutron moderator, zones with close composition but with different temperature, etc.) are described. Accounting for the dependence of constants for one and the same nuclide in the zones with different compositions and temperatures required substantial modernization of routines for preparation of constants for calculation described in (Zherdev et al. 2018). Algorithm for preparation of subgroup constants was modified, methodology for taking into account resonance self-screening of cross-sections within the range of unresolved resonances was improved, and other changes were introduced in the process of this modernization. Results of calculations are compared with data obtained using the MCNP-5 precision program (MCNP 1987), which is linked to the same library of evaluated neutron data ROSFOND as that used in ROCOCO. The ROCOCO-MMK includes procedures for registering different neutron flux functionals (also based on ROCOCO data), which allowed including it in the SCALA computation complex (Zherdev et al. 2003, Zherdev 2005), and performing step-by-step calculation of evolution of fuel nuclide composition during the fuel residence campaign. Directions for further development of the system are outlined in conclusion and, in particular, some possibilities of using the created software for further improvement of methods for preparation of few-group constants for calculations in diffusion approximation are examined.

Features of Burnup Calculations in the SERPENT 2 Software Package Using the Example of a BWR Fuel Assembly

In this paper, we used the data from “OECD/NEA Burnup Credit Criticality Benchmark Phase IIIB: Nuclide Composition and Neutron Multiplication Factor of BWR Spent Fuel Assembly” ([1]) for the verification of the SERPENT 2 code. The results obtained which were compared with the results of other authors, which were also given in “OECD/NEA Burnup Credit Criticality Benchmark Phase IIIB: Burnup Calculations of BWR Fuel Assemblies for Storage and Transport” ([2]). Investigations of the influence of the detailed model of pins and pins with gadolinium, as well as various methods of burn-up calculations were also carried out.

generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data

Depletion of RBMK-1500 spent nuclear fuel (SNF) with and without an erbium burnable absorber was modelled, and one-group burn-up dependent cross-section libraries for Origen-ARP were created. Depletion calculations for the generation of cross-section libraries were performed using the SCALE 6.1 code package with the TRITON control module, which employs the NEWT deterministic 2D transport code with the 238-group energy library based on the ENDF-B VII library and the ORIGEN-S nuclide composition calculation code. Concentrations of the most important actinides for criticality safety were calculated using the created libraries and were compared with the available experimental data and the newest modelling results. Available experimental data of fission products (Nd and Cs isotopes) were also compared to the modelling results. Composition differences were evaluated for several fuel enrichments and water densities. The comparison shows an acceptable agreement between the values obtained using new one-group cross-section libraries and experimental data except for238Pu and241Am, as well as the causes of discrepancy are discussed. It has been found that the enrichment and presence of the burnable absorber play an important role in the SNF composition. At the highest evaluated burn-up (29 GWd/tU) isotopic composition differences between 2% enrichment fuel and 2.8% burn-up for actinides important to burn-up credit (BUC) applications varied from 11 to 52%.