[en] The conditions of the vessel ageing (damages through irradiation) that mostly determine the life time of a nuclear reactor depend on the dose received. For the determination of this dose we use two calculation methods: one exact method using the TRIPOLI code that solves the Boltzmann equation with the Monte-Carlo method and one simplified method based on the point-kernel method. The advantages of the second method which is fast (easy reproduction of the results) and deterministic (the effects of difference are possible) compared to the first one which is long and statistical make its development necessary. The qualifications of these methods are done by comparison with the experiment that we reach as following: for the first method, we check the programming of TRIPOLI code (representativity of the collision) and its alignment with SN codes, and we modify the second method in order to use variable linear attenuation coefficients inside each medium to represent better the effects of the spectrum and of the reflection. As part of the checking of the basic physical data and their mode of representation, we present a study of the influence of the energy group averaging of the cross sections and of the number of the groups, as well as study of the influence of the cross sections origin

[en] The dose rate measurements of the TN 12/2, TN 28 VT and FS 65 has been used to evaluate the calculational procedures of Transnuclaire (TN). The three-dimensional (3D) Monte Carlo code TRIPOLI-3.4 which is used to optimize the shielding of TN casks, is applied to the analysis of a series of benchmarks. In the same cases the one-dimensional (1D) Sn code SN1D and the point kernel code MERCURE-V (3D) used for the more simplified calculations, are checked by the comparison with the measurements. The multi-group approximation used by the above codes, in order to reduces nuclear data, introduces errors due to the neutron cross-sections resonance treatment and the repartition of the gamma-ray spectrum (discrete) into an energy group structure. For a cask consisting of an iron shell of 250 mm of thickness, neutron dose rates can been underestimated of 50% if the resonances of the iron cross sections for high energy (above 1 MeV) are not taken into account. Also, depending on the energy group structure, gamma-ray dose rates can be over-estimated or under-estimated by the repartition of the gamma rays. The comparisons between measured and calculated dose rates are closer than 20% for the Monte Carlo calculations, 50% for the Sn calculations (1D) and a factor of 2 for the point kernel calculations. (author)

[en] According to the latest IAEA recommendations (2012, SSR-6), shielding analyses under routine transport conditions have to be performed considering loading plans with the maximum allowed radioactive contents. In parallel, regulators of storage now request the definition of generic loading plans. Thus, cask designers are facing the challenge of implementing a method to define maximum generic loading plans and to avoid a performance decrease on new cask design or cask license renewal. (author)

[en] To provide a cask with the largest possible loading capacity of spent fuel assemblies with the largest practicable burnups and shortest cooling times within all safety requirements, AREVA TN has adapted its design process and developed a more elaborated shielding analysis method. Taking advantage of the potential heterogeneities between sources of fuel assemblies to be loaded, and the self-shielding of inner assemblies by the outer assemblies, the result of this method is expressed under the shape of a set of inequalities allowing to optimize the cask capacity and performances. (authors)

[en] To provide a cask with the largest possible loading capacity of spent fuel assemblies with the largest practicable burnup and shortest cooling time within all safety requirements, AREVA TN has adapted its design process and developed a more elaborated shielding analysis method. AREVA TN reconsiders the standard definition of the content in order to take advantage of potential heterogeneities between sources of the loaded fuel assemblies and of the self-shielding between the loaded fuel assemblies. The maximum authorized radioactive content is defined with only maximum neutron and gamma sources authorized in each basket lodgement. The result of this method is expressed under the shape of a linear inequalities system allowing to optimize the cask capacity and performance. The linear inequalities system assures that the radiation level limits are respected and presents a high implementation flexibility and allows us to take advantage of the explicit characteristics of the fuel assembly inventory. This method avoids the necessity of defining in the license a maximal burnup and a minimum cooling time, authorized in order to respect the radiation level limits and it does not require any dose rate calculation for each loading

[en] The authors used a new fast method to evaluate reaction rates in a pressurized water reactor. The new scheme is based on a line-of-sight point-kernel approximation, with a non-linear attenuation. The attenuation is a sum of a linear term and two logarithmic terms. The five parameters of the attenuation model are adjusted with two-dimensional flux calculations

[en] The author gives an overview of the various aspects, methods and trends related to the design, test and use of packaging used for the transportation of radioactive materials. He evokes the currently existing calculation methods and codes which are used to assess irradiation sources and to study the propagation of such irradiation through matter. He discusses the qualification process of a radioprotection calculation scheme and gives and comments comparison between computed and measured results of effective dose flow rate about packaging. He presents the 'Transportability' tool which uses a database of fuel assembly characteristics, a code to assess neutron and gamma sources, a packaging model used to compute the different contributions to effective dose flow rates, a validation stage, and a fixed margin

[en] Orano TN has implemented Burnup Credit (BUC) approaches for the demonstration of the sub-criticality for transport and dual purpose casks loaded with PWR uranium oxide (UO₂) used fuel assemblies. Usually, the BUC methodology uses Enriched Natural Uranium (ENU) to determine the isotopic composition of the fuel after irradiation for criticality calculations. Nevertheless, Enriched Reprocessed Uranium (ERU) may be used for the manufacturing of the PWR UO₂ fuel assemblies. As far as criticality is concerned, the main difference between ENU and ERU, for the same U²³⁵ content, is the presence of U²³², U²³⁴ and U²³⁶ in the ERU initial composition. This paper presents sensitivity calculations to assess the impact on transport cask reactivity when ERU initial composition is used for BUC applications. The results show that ENU considered as initial composition before irradiation is a bounding assumption for advanced BUC method compared to the use of ERU. Despite the important decrease of U²³⁴ during irradiation, the presence of U²³⁴ and U²³⁶ for ERU fuel is sufficient for balancing the production of fissile isotopes during irradiation which is more important for the Enriched Reprocess Uranium than for the Enriched Natural Uranium. Nevertheless, the important reactivity discrepancies between ENU and ERU fuels is due to the penalizing initial isotopic concentrations retained for ERU (U²³⁴, U²³⁶); for actual ERU fuel, the reactivity is obviously closer to the one of ENU fuel.

[en] Improvements of gamma-ray transport calculations in S_n codes aim at taking into account the bound-electron effect of Compton scattering (incoherent), coherent scattering (Rayleigh), and secondary sources of bremsstrahlung and fluorescence. A computation scheme was developed to take into account these phenomena by modifying the angular and energy transfer matrices, and no modification in the transport code has been made. The incoherent and coherent scatterings as well as the fluorescence sources can be strictly treated by the transfer matrix change. For bremsstrahlung sources, this is possible if one can neglect the charged particles path as they pass through the matter (electrons and positrons) and is applicable for the energy range of interest for us (below 10 MeV). These improvements have been reported on the kernel attenuation codes by the calculation of new buildup factors. The gamma-ray buildup factors have been carried out for 25 natural elements up to 30 mean free paths in the energy range between 15 keV and 10 MeV

[en] Exposure and energy absorption buildup factors for aluminum, iron, lead, and water are calculated by the SNID discrete ordinates code for an isotropic point source in a homogeneous medium. The calculation of the buildup factors takes into account the effects of both bound-electron Compton (incoherent) and coherent (Rayleigh) scattering. A comparison with buildup factors from the literature shows that these two effects greatly increase the buildup factors for energies below a few hundred kilo-electron-volts, and thus the new results are improved relative to the experiment. This greater accuracy is due to the increase in the linear attenuation coefficient, which leads to the calculation of the buildup factors for a mean free path with a smaller shield thickness. On the other hand, for the same shield thickness, exposure increases when only incoherent scattering is included and decreases when only coherent scattering is included, so that the exposure finally decreases when both effects are included. Great care must also be taken when checking the approximations for gamma-ray deep-penetration transport calculations, as well as for the cross-section treatment and origin