Impact Of Nuclear Data Uncertainties Research Articles

Neutrons propagation in Lead: A feasibility study for experiments in the RSV TAPIRO fast research reactor

The increasing worldwide interest in Lead-cooled Fast Reactors (LFRs) substantiates the need to validate the analytical codes and methods used to support their design. For neutronic analyses, this is chiefly reflected in assessing the impact of nuclear data uncertainties on the integral and local parameters resulting from such analyses. The aim of refining nuclear data moves continuous efforts for more accurate measurements, be them differential or integral, for which adequate facilities are required. The availability at the ENEA's Casaccia research center of a fast source reactor – RSV TAPIRO – provides a unique opportunity to perform new integral experiments in support of fast reactors, including LFRs, owing to the well-characterized neutron spectrum of the thermal column. A series of experiments has been envisaged, dealing with the use of Lead in a reactor. The experiments concern the propagation of neutrons through blocks of materials representing relevant elements of a reactor core, and ranging from pure Lead to mixtures reproducing portions of the reflector and shield in LFRs. The paper is focused on the feasibility study of some of these experiments in which Lead and mixture blocks are inserted in the so-called thermal column of the RSV TAPIRO reactor and irradiated by the neutron flux emerging from the Copper reflector surrounding the reactor core.

Propagating neutronic uncertainties for FFTF LOFWOS Test #13

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.

Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors

In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse. The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite. The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients. These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA. An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures. It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in 235U fission, 235U ν, and 238U capture cross sections. The uncertainties of 239Pu and 238U capture cross sections become more significant as the fuel is irradiated. The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including 238U inelastic, 238U capture, and 239Pu capture cross sections. It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively. Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.

Nuclear data sensitivity and uncertainty analysis of critical VENUS-F cores with the Serpent Monte Carlo code

In the framework of the FP7 Euratom project FREYA, the Serpent Monte Code was used to characterize a number of critical VENUS-F core configurations. Several neutronic parameters calculated by Serpent were compared to the available experimental data and reported in a previous study. Although a generally reasonable agreement between the calculated and experimental values was obtained, the study also revealed some important issues related to the modeling results such as a systematic overestimation of reactivity and a systematic underestimation of U238 to U235 fission rate ratio (designated as F28/F25 spectral index).The objective of the current follow-up study is to assess the impact of nuclear data uncertainties on the Monte Carlo estimates of reactivity and spectral indices. The analysis is carried out for two critical VENUS-F cores using a recently upgraded version of Serpent which is capable to perform sensitivity analysis, based on the collision history method and to propagate and quantify nuclear data uncertainties using multi-group covariance libraries in an automated way. The criticality calculations were performed using the ENDV/B-VII.1 based cross sections while the uncertainties due to nuclear data were obtained using 56-group neutron cross section covariance library from the SCALE-6.2 package.The estimated uncertainties due to nuclear data are of the order of 2200 pcm and 8% on k-eff and F28/F25 spectral index, respectively. It was found that the major contributor to the k-eff uncertainty is the capture cross section of U235. In the F28/F25 case, the total uncertainty is dominated by inelastic scattering cross section of U238, fission spectrum of U235, and capture cross section of U235.

Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits

The impact of nuclear data uncertainties is studied for the reactor power and the reactivity during control rod withdrawal transients with reactivity insertions of 0.5$ and 0.97$, respectively, for a PWR mini-core model. Multi-group cross sections, the multiplicities of both prompt and delayed neutrons, and fission spectra are varied by the application of the random sampling-based method XSUSA with covariance data of SCALE 6.1 supplemented by JENDL-4.0. The varied multi-group data are used by TRITON/NEWT to generate varied 2-group cross sections, which are then applied in neutron-kinetic/thermal-hydraulic calculations with DYN3D-ATHLET. A significant impact on both the reactivity uncertainty and the power uncertainty is observed. Since the distributional properties of the output time series vary across the problem time, the distribution-free Wilks tolerance limit is applied as a robust uncertainty measure to complex time series patterns. The most contributing nuclide reactions to the power uncertainty are identified via sensitivity analysis.

Stress-testing the ALFRED design – Part I: Impact of nuclear data uncertainties on Design Extension Conditions transients

The advancement of the design of ALFRED beyond the conceptual phase, passes through the analysis of the impact of uncertainties, notably to what concerns safety-related conditions. Compliancy of plant safety to Design Extension Conditions is, according to IAEA and in line with the meaning itself of these beyond-design conditions, usually investigated by best estimates only. Due however to the demonstration nature of ALFRED, it was decided to assess the actual safety performances of this system even in ultimate conditions. To this regard, the emphasis was put on unprotected events like the UTOP (unprotected transient of over-power) and ULOOP (unprotected loss of offsite power, resulting from the combination of a loss of flow and loss of heat sink under unprotected conditions), pinpointed as the most challenging situations sought for the plant. The purpose of the present work, which has been divided in three parts, was then to assess the ultimate ALFRED safety margins against failure of the key core components and systems (Part III). To target this objective, the evaluation of uncertainties coming, on one hand, from nuclear data was performed at first, to retrieve their impact on the reactivity coefficients, thereby on the transient behavior driven by the latter (Part I); then, uncertainties from material properties, fabrication procedures, operation and computational tools were propagated to assess their influence on the thermal-hydraulics of the system (Part II). In this paper, presenting the first part of the work, the focus is on nuclear data. As such, a sensitivity/uncertainty analysis of the ALFRED core on key elementary reactivity effects, forming the basis for computing the feedback coefficients, has been performed. The sensitivity analysis allowed pointing out firstly the most relevant cross-sections for every response function. Uncertainty analysis allowed then establishing a possible range of confidence for the reactivity effects. The adjoint-based technique implemented in the TSUNAMI-3D module of the SCALE6 system was used. The confidence intervals identified for each reactivity effect have been combined then into confidence intervals for the feedback coefficients. Finally, the most conservative, yet physically sound (i.e., where correlations among coefficients, stemming from dependencies on common nuclear data, are taken into account), set of reactivity coefficients has been picked out from the confidence intervals, enabling transient calculations to propagate uncertainties into transient behavior. Using this off-nominal set, and comparing results with the reference ones, exploiting the system codes SIM-LFR and RELAP, a negligible effect of nuclear data uncertainties has been found for the ULOOP, while an increase of the maximum achieved power of around 6% has been computed for the UTOP. Overall, a modest contribution of nuclear data uncertainties for these transients has been found which, however, must be combined with the thermal-hydraulics one so to finally assess safety margins.

The path for innovative severe accident neutronics studies in ZPRs. Part II.2 - Interpretation of SNEAK-12B experiment for core disruption in LMFBRs impact of nuclear data uncertainties on reactivity coefficients

The present work details further information regarding a new benchmark to be introduced to the international community, for dealing with neutronic code validation in the frame of the analysis of severe accidents in fast reactors leading to core degradation and material relocation. This specific benchmark is based on further analysis of selected experiments performed at the Schnelle Null-Energie-Anordnung Karlsruhe (SNEAK). The SNEAK-12B core was loaded with plutonium fuel to better represent future fast systems and the experiments considered fuel relocation and redistribution of structural material. In this paper, the experimental results are analyzed by state of the art tools, Serpent-2 Monte Carlo and the ERANOS code for deterministic calculations. The paper presents a full sensitivity and uncertainty analysis based on the JEFF-3.1.1 and the associated covariance data COMAC-V01, which is performed in order to gain deeper insight into the governing phenomena related to geometrical changes of the core. A comparison of the propagated uncertainties between Serpent and ERANOS is made. The uncertainty propagation vary from code to code, and strongly disagree in most cases of axial fuel relocation. This is evident for small reactivity variation (< 1 cent), where the difference in the propagated uncertainties obtained from the two codes is vividly visible. The analysis provide valuable information on uncertainty propagation in a system where the overall material balance is not modified, and contributes to the design of future experiments. This work is done within the frame of new core design capacities and innovative experimental programs to be implemented in Zero Power Reactors, such as the ZEPHYR project led independently by CEA.

Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like sodium fast reactor

The EU 7th Framework ESNII+ project was launched in 2013 with the strategic orientation of preparing ESNII for Horizon 2020. ESNII stands for the European Industrial Initiative on Nuclear Energy, created by the European Commission in 2010 to promote the development of a new generation of nuclear systems in order to provide a sustainable solution to cope with Europe’s growing energy needs while meeting the greenhouse gas emissions reduction target. The designs selected by the ESNII+ project are technological demonstrators of Generation-IV systems. The prototype for the sodium cooled fast reactor technology is ASTRID (standing for Advanced Sodium Technological Reactor for Industrial Demonstration), which detailed design phase is foreseen to be initiated in 2019. The ASTRID core has a peculiar design which was created in order to tackle the main neutronic challenge of sodium cooled fast reactors: the inherent overall positive reactivity feedback in case of sodium voiding occurring in the core. Indeed, the core is claimed by its designers to have an overall negative reactivity feedback in this scenario. This feature was demonstrated for an ASTRID-like core within the ESNII+ framework studies performed by nine European institutions. In order to shift the paradigm towards best-estimate plus uncertainties, the nuclear data sensitivity analysis and uncertainty propagation on reactivity coefficients has to be carried out. The goal of this work is to assess the impact of nuclear data uncertainties on sodium voiding reactivity feedback coefficients in order to get a more complete picture of the actual safety margins of the ASTRID low void-core design. The nuclear data sensitivity analysis is performed in parallel using SCALE TSUNAMI-3D and the newly developed GPT SERPENT 2 module. A comparison is carried out between the two methodologies. Uncertainty on the sodium reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety margins conclusions are drawn.

The path for innovative severe accident neutronics studies in ZPRs. part I.2 - Interpretation of SNEAK-12A experiment for core disruption in LMFBRs impact of nuclear data uncertainties on reactivity coefficients

The present work details a further investigation of the SNEAK-12A experimental program, which aimed to study material relocation in Sodium Fast Reactors (SFRs) leading to core degradation. The further investigation include sensitivity and uncertainty propagation analysis. In this paper, a comparison is made using two codes, a Monte Carlo based code Serpent 2 and the deterministic system code ERANOS. A sensitivity analysis was made utilizing the two codes, with comparison of two nuclear data libraries (ENDF/B-VII.1 and JEFF-3.1.1). The code-to-code comparison resulted in a very good agreement, while the comparison of libraries showed large discrepancies, manly due to the differences in the sodium cross-section data. The sensitivity analysis, was followed by a complete propagated uncertainty analysis based on the covariance evaluated data available in the COMAC data evaluation. The results of the uncertainties show that there are still large discrepancies linked to the nuclear data. This work is done within the frame work of new core design capacities and new ways of conducting in Zero Power Reactors, such as the ZEPHYR project led independently by CEA.

PROPAGATION OF NUCLEAR DATA UNCERTAINTIES FOR PWR CORE ANALYSIS

An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for 235,238U, 239Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

Combining Total Monte Carlo and Benchmarks for Nuclear Data Uncertainty Propagation on a Lead Fast Reactor's Safety Parameters

Analyses are carried out to assess the impact of nuclear data uncertainties on keff for the European Lead Cooled Training Reactor (ELECTRA) using the Total Monte Carlo method. A large number of 239Pu random ENDF-formatted libraries generated using the TALYS based system were processed into ACE format with NJOY-99.336 code and used as input into the Serpent Monte Carlo neutron transport code to obtain distribution in keff. The mean of the keff distribution obtained was compared with the major nuclear data libraries, JEFF-3.1.1, ENDF/B-VII.1 and JENDL-4.0. A method is proposed for the selection of benchmarks for specific applications using the Total Monte Carlo approach. Finally, an accept/reject criterion was investigated based on χ2 values obtained using the 239Pu Jezebel criticality benchmark. It was observed that nuclear data uncertainties in keff were reduced considerably from 748 to 443 pcm by applying a more rigid acceptance criteria for accepting random files.

Propagation of Nuclear Data Uncertainties in Transmutation Calculations Using ACAB Code

Departamento de Ingeniera Energetica, Universidad Nacional de Educacion a Distancia (UNED), 28004 Madrid, Spain(Received 26 April 2010)The assessment of the accuracy of parameters related to the reactor core performance (e.g.,ke ) and fuel cycle (e.g., isotopic evolution/transmutation) due to the uncertainties in the ba-sic nuclear data (ND) is a critical issue. Di erent error propagation techniques (adjoint/forwardsensitivity analysis procedures and/or Monte Carlo technique) can be used to address by computa-tional simulation the systematic propagation of uncertainties on the nal parameters. To performthis uncertainty assessment, the ENDF covariance les (variance/correlation in energy and cross-reactions-isotopes correlations) are required.In this paper, we assess the impact of ND uncertainties on the isotopic prediction for a concep-tual design of a modular European Facility for Industrial Transmutation (EFIT) for a dischargeburnup of 150 GWd/tHM. The complete set of uncertainty data for cross sections (EAF2007/UN,SCALE6.0/COVA-44G), radioactive decay and ssion yield data (JEFF-3.1.1) are processed andused in ACAB code.

Impact of nuclear data uncertainties on the incineration of 237Np and 241Am targets

In this work, we have explored the impact of nuclear data uncertainties on the incineration of 237Np and 241Am targets and on the evolution of correlated safety quantities such as alpha and spontaneous fission activities and radiotoxicity. These studies were performed in three representative neutron flux spectra: EFR (fast reactor), GT-MHR (moderated reactor) and HTFR (thermal reactor). In order to get a common scale of comparison between these scenarios, the simulations were based on given incineration rate rather than irradiation time. Afterwards, the dominant contributors to alpha and spontaneous fission activities and radiotoxicity were identified in each case and the impact of nuclear data uncertainties on the formation of these isotopes was assessed. Finally, data which have the highest contributions were highlighted and their multi-group errors were detailed as a function of energy, to show on which energetic domains future measurements have to be focused on.

Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP–ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP–ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

Impact of Nuclear Data Uncertainties on Transmutation of Actinides in Accelerator-Driven Assemblies

The potential impact of nuclear data uncertainties on a large number of performance parameters of reactor cores dedicated to the transmutation of radioactive wastes is discussed. An uncertainty analysis has been performed based on sensitivity theory, which underlines the cross sections, the energy range, and the isotopes that are responsible for the most significant uncertainties.To provide guidelines on priorities for new evaluations or validation experiments, required accuracies on specific nuclear data have been derived, accounting for target accuracies on major design parameters. The required accuracies (mostly in the energy region below 20 MeV), in particular for minor actinide data, are of the same order of magnitude of the achieved accuracies on major actinides. Specific requirements also concern the improvement of minor actinide data related to decay heat and effective delayed-neutron fraction assessment.