[en] A Monte Carlo sampling code called SANDY for the nuclear data uncertainty propagation is described and tested against several reactor case studies. SANDY can efficiently read and process nuclear data files in the ENDF-6 format, commonly adopted for the storage and retrieval of evaluated nuclear data for applications of nuclear technology. In this work, SANDY is used to sample and propagate the covariance data available in ENDF-6 files. The uncertainties of multiple reactor responses are quantified, such as integrated cross sections, k_∞, k_eff, β_eff, reactivity worths and damage metrics. The reactor models taken into considerations were the VENUS-F fuel assembly with solid bismuth blocks and plates, a scale-down version of the ASTRID sodium fast reactor operated as a minor actinide burner and the MYRRHA multi-purpose facility in subcritical mode. Results show the flexibility of SANDY in propagating multiple uncertainties to different nuclear responses. Also, conclusions were drawn on the reliability of the current evaluated covariance data available. (authors)

[en] In this work, the objective is to perform an uncertainty analysis on a MYRRHA -Rev.1.6 irradiation cycle study, being applied to a depletion scenario of a single fresh fuel assembly while assuming reflective boundary conditions. Such analysis is statistically based on the application of Wilk’s method of building tolerance limits after 100 depletion calculations were performed with the SERPENT2 code. Due to the computational burden of such type of simulations, this propagation of nuclear data covariances study (allowed by the fast computational performance of SERPENT2) was done at constant power, constant flux and, in a final exercise, at constant power with the addition of fission yield uncertainties (all of these cases employed ENDF/B-VII.1 data). It was observed that while depleting at constant power, the statistical variation of key fission products such as ¹⁴⁸Nd is almost not present because of the normalization factor applied to the flux. In contrast, the irradiation at constant flux reveals dependence on burnup. Finally, the added fission yield uncertainties make clear the fact that they directly impact the degree of final uncertainty computed for fission products exemplified by ¹⁴⁸Nd and ¹³⁵Xe important for burnup estimation and reactor operation, respectively.

[en] In the framework of Phase I of the MYRRHA project implementation, the superconducting linear accelerator with proton beam parameters 100 MeV, 4 mA is going to be built. To stop the beam, a beam dump based on Al-6061 alloy is designed. The evaluation of radiological impact of an accidental radioactivity release requires the reliable estimates of primary radiation source terms with associated uncertainties. The article addresses the propagation of nuclear data uncertainties through beam dump activation calculations. The Total Monte Carlo approach was used to generate large number of random excitation functions for residual products of proton interactions with materials of Al-6061 alloy. The residual products do not impose any feedback on proton and neutron spectra in the beam dump, moreover the calculation of the production rates is sufficient to obtain uncertainties on final activities. This significantly accelerates the uncertainty quantification allowing to study the convergence of mean and higher moments (variance, variance of variance) for individual nuclides.

[en] MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multi-purpose research facility currently being developed at SCK·CEN. It will be able to work in both critical and subcritical modes and, cooled by lead-bismuth eutectic. It will play a key role in the development of the Pb-alloy technology needed for the lead fast reactor GEN IV concept. MYRRHA will demonstrate the Accelerator Driven System (ADS) full concept by coupling a proton accelerator, a spallation target and a sub-critical reactor at a reasonable power level to allow operation feedback. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Recently, a new core design with a longer active core has been proposed. This paper presents the neutronic analyses for this design improvement. The analyses have been done using the MCNP/X code and the in-house developed ALEPH2 depletion code. (author)

[en] A global sensitivity analysis of effective neutron multiplication factor k_eff to the change of nuclear data library revealed that JEFF-3.2T2 neutron-induced evaluated data library produces closer results to ENDF/B-VII.1 than does JEFF-3.1.2. The analysis of contributions of individual evaluations into k_eff sensitivity allowed establishing the priority list of nuclides for which uncertainties on nuclear data must be improved. Detailed sensitivity analysis has been performed for two nuclides from this list, ⁵⁶Fe and ²³⁸Pu. The analysis was based on a detailed survey of the evaluations and experimental data. To track the origin of the differences in the evaluations and their impact on k_eff, the reaction cross-sections and multiplicities in one evaluation have been substituted by the corresponding data from other evaluations. (authors)

[en] The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction β_eff is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of β_eff uncertainty. For validation of the calculated β_eff, a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified k-ratio method is superior to Bretscher's prompt k-ratio method in terms of reducing the statistical uncertainty in calculating not only β_eff but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified k-ratio method is the most practical for uncertainty quantification of β_eff. Finally, total β_eff uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of ²³⁵U. (author)

[en] Highlights: → The sensitivity and uncertainty analyses were performed to comprehend the reliability of the XT-ADS neutronic design. → The uncertainties deduced from the covariance data for the XT-ADS criticality were 0.94%, 1.9% and 1.1% by the SCALE 44-group, TENDL-2009 and JENDL-3.3 data, respectively. → When the target accuracy of 0.3%Δk for the criticality was considered, the uncertainties did not satisfy it. → To achieve this accuracy, the uncertainties should be improved by experiments under an adequate condition. - Abstract: The XT-ADS, an accelerator-driven system for an experimental demonstration, has been investigated in the framework of IP EUROTRANS FP6 project. In this study, the sensitivity and uncertainty analyses were performed to comprehend the reliability of the XT-ADS neutronic design. For the sensitivity analysis, it was found that the sensitivity coefficients were significantly different by changing the geometry models and calculation codes. For the uncertainty analysis, it was confirmed that the uncertainties deduced from the covariance data varied significantly by changing them. The uncertainties deduced from the covariance data for the XT-ADS criticality were 0.94%, 1.9% and 1.1% by the SCALE 44-group, TENDL-2009 and JENDL-3.3 data, respectively. When the target accuracy of 0.3%Δk for the criticality was considered, the uncertainties did not satisfy it. To achieve this accuracy, the uncertainties should be improved by experiments under an adequate condition.

[en] The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a k_eff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15% and 1%, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.

[en] Highlights: • High-energy nuclear data uncertainties have been propagated to the MYRRHA safety responses. • Monte Carlo sampling of random nuclear data files from covariance matrices generated online. • Proposed method allows to evaluate the convergence of uncertainties with the number of samples. • Total core power uncertainty is 14% and the nuclide concentration uncertainties range from 5 to 60%. - Abstract: Propagation of high-energy (above 20 MeV) nuclear data uncertainties on the safety related neutronic responses in accelerator driven systems has been assessed. The total core power and production of radionuclides contributing to radiation source terms were focused on. The article features a method based on the Monte Carlo sampling of random nuclear data files from the covariance matrices generated from the sets of reaction cross sections obtained with model calculations of high-energy particle interactions with matter or picked up from already existing nuclear data libraries. It has been demonstrated that nuclear data uncertainties do not need to be propagated through particle transport calculations to obtain uncertainties on the responses. This advantage allowed to investigate the convergence of the sample average to the best estimate. The number of random nuclear data file sets needed to obtain reliable uncertainty on the total core power is around 300 that results in the uncertainty of 14%. The uncertainties on the concentrations of nuclides most important for the safety assessment that are accumulated in lead–bismuth eutectic during irradiation, range from 5 to 60%. Concentrations of some nuclides exemplified by Tritium converge much slower than neutron multiplicities so that several thousands of samples are needed to ensure reliable uncertainty estimates.

[en] A new evaluated nuclear data library has been created. The library consists of two sub-libraries for neutron and proton incident particles. The neutron sub-library contains nuclear data for transport, heating and shielding applications for 242 nuclides with atomic numbers ranging from 8 to 82 in the energy region of primary neutrons from 10^-5 eV to 150 MeV. Data below 20 MeV are taken mainly from ENDF/B-VI (revision 8) and for some nuclides, from the JENDL-3.3 and JEFF-3.0 libraries. The proton sub-library should contain data for the same range of target nuclides and energies. Proton-induced evaluated cross-section files are available for 15 nuclides at the moment. The evaluation of emitted particle energy and angular distributions at energies above 20 MeV (for incident neutrons) and above the reaction threshold (for incident protons) was performed with the help of the ALICE/ASH code and the analysis of available experimental data. The total cross-sections, elastic cross-sections and elastic scattering angular distributions were calculated with the help of the coupled channel model. The results of the calculation were adjusted to the data from ENDF/B-VI, JENDL-3.3, or JEFF-3.0 at the neutron energy equal to 20 MeV. The library is written in ENDF-6 format using the MF=3/MT=5 and MF=6/MT=5 representations