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• A novel multi-scale domain overlapping method is applied to the transient analysis of Xi'an pulsed reactor. • When the maximum time step of involved codes RELAP5 and CTF is 0.1 s, the stability and efficiency are the best. • The results of coupling calculation are closer to the experimental data than the standalone system code. • The trend of global parameters and detailed distribution of local are obtained by coupling calculation. Multi-scale thermal–hydraulic coupled calculation is widely applied in the design and safety analysis of reactor. It can provide the thermal–hydraulic results with high-fidelity and high-resolution. According to natural circulation flow requirement of Xi'an pulsed reactor (XAPR), which is a pool-type reactor with natural circulation, a novel multi-scale coupling method of domain overlapping for XAPR is proposed. The system code RELAP5 and subchannel code CTF are coupled. The explicit coupling strategy is applied to the pulsed operation transient condition of XAPR. The results show that the transient coupling strategy is effective, and the results of coupled calculation are closer to the experimental data than those of standalone system code. Meanwhile, the coupled calculation has advantages in obtaining local detailed thermal–hydraulic characteristics. Furthermore, the method is applied to the research of station blackout (SBO) accident of XAPR, and sensitivity analysis of the coupling time-step is carried out. The results indicate that the stability and efficiency are best when the maximum time-step is set as 0.1 s. Meanwhile, in the transient analysis of accident, the multi-scale coupled calculation could provide high-fidelity, high-precision and high-resolution thermal–hydraulic results, which proves the validity and advantage of coupled calculation. [ABSTRACT FROM AUTHOR]

• The transient response of FuSTAR in five design basis accidents are analyzed with the transient simulation code FUSY. • The sensitivity analysis is performed for some key parameters that may affect the transient behavior. • FuSTAR has a great ability to survive in URIA and ULOF. • The maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. FuSTAR is a small, modular fluoride-salt-cooled high-temperature advanced reactor, proposed by Xi'an Jiaotong University (XJTU), aiming at fulfilling the requirements of Gen IV nuclear system and meanwhile ensuring the inherent safety and economics. In order to evaluate the safety performance of FuSTAR, five typical design basis accidents are selected for safety analysis with different initial events, which are unprotected reactivity insertion accident (URIA), unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected overcooling (UOC), and the accident combing ULOF with ULOHS. The sensitivity analysis is also performed. FuSTAR is modeled by modified system analysis code and the simulations of the steady-state are verified by the results with CFD method. The results of transient accidents indicate that safety of FuSTAR has been guaranteed in URIA and ULOF. Meanwhile, the maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. In the ULOF and ULOHS, the peak of fuel temperature, cladding surface temperature and core outlet temperature of coolant are below their safety limits. Therefore, FuSTAR exhibits acceptable safety under design basis accidents. This study will provide supports in verifying the rationality of the design and the subsequent optimization of FuSTAR. [ABSTRACT FROM AUTHOR]

Severe nuclear accidents which result in substantial damage to the reactor core have the potential to release radiation into the environment if the integrity of the reactor vessel is not maintained. In the event of a full loss of core cooling, the expected failure mechanism for Canada Deuterium Uranium (CANDU) calandria vessels is thermal stresses, which are influenced by the heat transfer within the corium. The current study uses computational fluid dynamics (CFD) to simulate the fluid flow, heat transfer, and crust formation in the corium using of full-scale CANDU calandria vessel. Uncertainties in the boundary conditions are considered with a sensitivity analysis. Results indicate that about 90% of the corium volume is in the mushy and solid form. The heat flux over the majority of the vessel wall is below the critical heat flux values and reaches the maximum value at the edge of the step region. [ABSTRACT FROM AUTHOR]

In nuclear reactor safety assessment studies, void formation in the coolant, both in nominal operation and in accidental scenarios, must be considered since bubble nucleation can lead to fast reactivity changes, thus putting a strain on the reactor control. The evaluation of such a phenomenon is not trivial due to the non-linearities related to the competing phenomena (e.g. neutron absorption and scattering) and the spatial effects. This work aims to analyse the impact of void formation on the multiplication factor in the Pavia (Italy) TRIGA Mark II reactor, focusing on spatial effects. A model of the TRIGA has been developed using Serpent Monte Carlo code and validated against experimental measurements conducted at the Pavia reactor. Two approaches have been adopted. The first consists of directly evaluating the multiplication factor of several configurations, each featured by a different water density to mimic a homogeneous void formation. The second approach consists of a perturbative procedure, i.e., a first-order sensitivity analysis, which allows the gathering of more information on the aforementioned competing phenomena. Both cases subdivide the core into several radial and axial regions to recover the spatial effect. The two approaches appear to be complementary in the information they provide. The results show that the void coefficient is negative in the core and strongly dependent on both position and void fraction. Moreover, the results show the strong influence of the fuel elements type (101 and 103), their location, and the experimental methods adopted on the void coefficient. • Monte Carlo analysis of void feedback coefficient is performed. • The sensitivity tool has been exploited. • Experimental data have been used to validate the numerical model. • Experimental procedures for assessing void coefficient have been analysed. [ABSTRACT FROM AUTHOR]

• Activation analysis was performed according to impurities in the reactor structure. • OPR-1000 was modeled with MCNP, and activation analysis through ORIGEN-S was performed using the calculated neutron flux. • It was calculated that Co increases the total specific activity the most in metal, and Li increases the most in concrete. Various kinds of radioactive wastes are expected to be generated resulting from the decommissioning of nuclear power plants. For the safe management of decommissioning wastes, the preliminary classification of radioactive wastes such as very low level, low level, intermediate level, and regulatory clearance has to be performed. In this regard, we intend to conduct an activation analysis and radioactive waste level estimation using a computational simulation code for the structure materials of an OPR-1000, a standard nuclear power plant in South Korea. In addition, the structure material that constitutes the reactor vessel, reactor vessel internals, and bio-shield concrete of the nuclear power plant contains small amounts of unintended impurities. The results of the activation analysis may vary depending on the presence of those impurities. In this study, a detail modeling of nuclear reactor and the activation analysis are carried out combined the MCNP code with the ORIGEN-S code. Based on this methodology, the sensitivity test is done for impurities of structure materials. From the result of the specific activity of activated structure material, the cobalt impurity is found to be the most important contributor in the level determination of radioactive wastes. However, the bio-shield concrete has the highest specific activity when Li impurity is contained. From this study, some impurities affect radioactive waste level, thus it is recommended to consider structure material impurities for decommissioning the nuclear power plants. [ABSTRACT FROM AUTHOR]

Resolved resonance region evaluations of the major isotopes of natural lead, 206Pb, 207Pb, 208Pb, have been performed to support the development of Generation IV reactors. Validation of nuclear data for lead fast reactors was performed with simulations of shielding benchmarks, integral critical benchmarks, and quasi-differential scattering measurements. Sensitivity analyses of these systems showed that elastic scattering reactions above 100 keV were the dominant reactions driving system performance. The resolved resonance regions (RRRs) of the lead isotopes extend past 100 keV, making the RRR an ideal starting point to evaluate lead cross sections. Since the R-matrix requires knowledge of bound, distant, and observed resonances, it was necessary to evaluate from 1 0 − 5 eV up to the respective limit of the RRR. The 208Pb RRR evaluation was extended to 1.5 MeV in order to obtain resonance parameters used to calculate new elastic scattering angular distributions up to 1.5 MeV. Resonance parameter uncertainties and covariance were generated using the R-matrix code SAMMY. The new RRR parameters show a direct improvement to the scattering kernel below 1.5 MeV which in turn greatly improves fast critical experiments over ENDF/B-VIII.0. [ABSTRACT FROM AUTHOR]

• The eigenvalue sensitivity to P 1 -moment is largest for 238 U , 56 Fe , and. 16 O. • The void reactivity sensitivity to P 1 -moment is largest for 56 Fe and. 23 Na. • A clear lack of data is shown excepting JENDL-5. • 16 O of ENDF/B-VIII.0 and 23 Na of TENDL-2021 have issues. • The eigenvalue P 1 -moment uncertainty for the top sources does not exceed 0.05%. Nuclear data uncertainties are considered a main limiting factor in proving safety of advanced reactors. A reasonable amount of available information should be applied to sensitivity and uncertainty analyses for this, and this work considers the influence of usually omitted angular distribution uncertainties. A sensitivity and uncertainty analysis of MOX3600 is conducted using Serpent. The angular distribution influence is assessed by applying covariances of the state-of-the-art libraries. The sensitivity analysis shows that quantifying the influence of higher-order Legendre moment uncertainties on the eigenvalue and void reactivity is of minor importance for MOX3600. The uncertainty analysis shows the P 1 -coefficients for 238 U and 16 O introduce uncertainties to the eigenvalue similar to 56 Fe and should be accounted for. A library comparison reveals TENDL-2021 and ENDF/B-VIII.0 introduce uncertainties of a few hundred pcm to the eigenvalue, and the data should be reassessed, while the other libraries give an uncertainty of several tens of pcm. [ABSTRACT FROM AUTHOR]

In nuclear reactor physics, the integral quantities (reaction rates, k e f f , etc.) are estimated by calculation codes (Monte Carlo and deterministic codes). The determination of the bias and uncertainty that may exist between the calculated results and the reality is necessary. These uncertainties are mainly due to the approximations used during modeling, the approximation associated with the calculation method, as well as the uncertainties of the nuclear data. It is necessary to perform the Sensitivity/Uncertainty Analysis using a set of benchmarks in order to provide an overview of nuclear data uncertainty propagation and to identify isotopes and reactions which can contribute to the uncertainty of integral parameters. In this study, a set of benchmarks contained 233 U were selected in order to estimate the impact of the uncertainties related to the reaction of this isotope on k e f f and to validate the ENDF/B-VIII.0 library in comparison to ENDF/B-VII.1 nuclear data. • Study of eigenvalue sensitivity due to the different reactions of 233U isotope using MCNP code. • Analysis of the K eff uncertainty caused by several nuclear data. • Nuclear data uncertainties comparison between ENDF/B-VIII.0 and ENDF/B-VII.1 libraries are performed. • 233U uncertainties of ENDF/B-VIII.0 is significantly reduced compared with that of ENDF/B-VII.1. [ABSTRACT FROM AUTHOR]

• MUSA project aimed at quantifying uncertainties and sensitivities in SA analyses. • WP4 was a joint attempt to apply BEPU to simplified but representative SA scenario. • WP4 provided key insights into BEPU methodology when applied to SA scenarios. The Management and Uncertainties of Severe Accidents (MUSA) project was funded in HORIZON 2020 and is coordinated by CIEMAT (Spain). The project aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) analysis, focusing on source term figures of merit. The Application of Uncertainty Quantification (UQ) Methods against Integral Experiments (AUQMIE – Work Package 4 (WP4)), led by ENEA (Italy), was devoted to apply and test UQ methodologies adopting the internationally recognized PHEBUS FPT1 test. FPT1 was chosen to test UQ methodologies because, even though it is a simplified SA scenario, it was representative of the in-vessel phase of a severe accident initiated by a break in the cold leg of a PWR primary circuit. WP4 served as a platform to identify and discuss the issues encountered in the application of UQ methodologies to SA analyses (e.g. discuss the UQ methodology, perform the coupling between the SA codes and the UQ tools, define the results post-processing methods, etc.). The purpose of this paper is to describe the MUSA PHEBUS FPT1 uncertainty application exercise with the related specifications and the methodologies used by the partners to perform the UQ exercise. The main outcomes and lessons learned of the analysis are: scripting was in general needed for the SA code and uncertainty tool coupling and to have more flexibility; particular attention should be devoted to the proper choice of the input uncertain parameters; outlier values of figures of merit should be carefully analyzed; the computational time is a key element to perform UQ in SA; the large number of uncertain input parameters may complicate the interpretation of correlation or sensitivity analysis; there is the need for a statistically solid handling of failed calculations. [ABSTRACT FROM AUTHOR]

The accurate calculation and uncertainty quantification of the characteristics of spent nuclear fuel (SNF) play a crucial role in ensuring the safety, efficiency, and sustainability of nuclear energy production, waste management, and nuclear safeguards. State of the art physics-based models, while reliable, are computationally intensive and time-consuming. This paper presents a surrogate modeling approach using neural networks (NN) to predict a number of SNF characteristics with reduced computational costs compared to physics-based models. An NN is trained using data generated from CASMO5 lattice calculations. The trained NN accurately predicts decay heat and nuclide concentrations of SNF, as a function of key input parameters, such as enrichment, burnup, cooling time between cycles, mean boron concentration and fuel temperature. The model is validated against physics-based decay heat simulations and measurements of different uranium oxide fuel assemblies from two different pressurized water reactors. In addition, the NN is used to perform sensitivity analysis and uncertainty quantification. The results are in very good alignment to CASMO5, while the computational costs (taking into account the costs of generating training samples) are reduced by a factor of 10 or more. Our findings demonstrate the feasibility of using NNs as surrogate models for fast characterization of SNF, providing a promising avenue for improving computational efficiency in assessing nuclear fuel behavior and associated risks. • A neural network can replace CASMO5 decay heat calculations of spent nuclear fuel. • A simulation using the neural network can be up to 106 times faster than CASMO5. • Network predictions are within 3% of measured decay heat, comparable to CASMO5. • Uncertainty quantification (hundreds of simulations) with network in under 1 s. [ABSTRACT FROM AUTHOR]

• CALMET-LAPMOD can be used for radionuclide atmospheric dispersion simulation. • Summarizing the unique advantages of source terms based on recognized events. • Verifying the releases caused by certain recognized events. • Kernel function concentration calculation may not require massive released particles. • Wind direction substantially influences the temporal tendency of simulated results. This paper investigates a local-scale radionuclide dispersion modeling, focusing on the 20 km range of the Fukushima accident. The California Meteorological Model (CALMET) was used to calculate the diagnostic wind fields for simulation using on-site meteorological observations, and the Lagrangian Particle Model (LAPMOD) was used to simulate the dispersion of the 137Cs. The simulation was validated against the measurement from two 137Cs airborne concentration stations (Futaba and Naraha). A comprehensive sensitivity analysis was implemented to the source term, the number of particles per second, and the disturbance in wind speed and direction. The results show that the simulation agrees well with the measurement at Futaba (Geometric Mean Bias: 0.79, Fractional Mean Bias: 0.58) than that at Naraha (Geometric Mean Bias: 1.01, Fractional Mean Bias: 0.52), especially for the peak values. The sensitivity analysis indicates that the simulation is highly affected by the source term and the wind direction. [ABSTRACT FROM AUTHOR]

• An integrated methodology for ranking reactor physics uncertainties is presented. • An eigenvalue decomposition of the nuclear data covariance matrix is used. • Perturbed data is propagated through lattice and core calculations. • Comparison to a direct Sobol sample gives consistent results for cases examined. Nuclear data uncertainties lead to substantial uncertainties in the predictions of LWR core calculations. Such problems are typically solved using lattice calculations and nodal codes, such as the WIMS-PANTHER system. It is desirable to derive uncertainties of core simulation output parameters (e.g. enthalpy deposition in an REA) to nuclear data uncertainties for specific nuclide-reaction pairs. This allows the key uncertainty contributions to be identified and ranked and facilitates use of data assimilation techniques. Key challenges in ranking the contributions to uncertainty are (1) the large number of input parameters (2) burnup-dependence of nuclide populations and microscopic cross sections, which challenges the use of adjoint methods (3) propagation of uncertainties through a multi-step calculation sequence. In this paper, an integrated methodology for deriving such uncertainties is presented. This comprises the following steps: (1) sensitivity analysis to identify the key reactions in the problem and exclude insignificant reactions (2) orthogonal decomposition of the nuclear data covariance matrix to calculate and rank eigenvectors (3) direct and consistent perturbation of the microscopic cross sections for the eigenvectors with significant eigenvalues (4) separate lattice (WIMS) and core (PANTHER) calculations with nuclear data perturbed for each eigenvector. This method (which assumes linearity of perturbations) is compared to a direct Sobol sample of the nuclear data and for the cases examined gives consistent results for the importance ranking of reactions, while greatly reducing the number of calculations employed. Hence, it can complement sampling-based approaches in developing an overall understanding of the magnitude, distribution, and contributors to uncertainty in core calculations. Both the linear method and the Sobol sampling method are integrated within the Visual Workshop graphical user environment to automate the process with the WIMS and PANTHER codes. [ABSTRACT FROM AUTHOR]

• A detailed analysis of the steam explosion characteristics is performed with the MC3D software. • A comprehensive review of the up-to-date knowledge and modeling issues is given. • 1-D steam explosion calculations are performed with sensitivity of melt and void fractions. • The impacts of major modeling uncertainties are discussed. Steam Explosion is a destructive event that may arise from the mixing of the molten fuel (corium) with the liquid coolant (water, sodium) during the course of a severe accident. Due to its high complexity and very specific related conditions, the phenomenon is still subject to numerous uncertainties. MC3D is a CFD software for the simulations of 3D multiphase flows and is used to evaluate steam explosion loads through a specific application called EXPLO. The present paper provides a comprehensive analysis of the characteristics of steam explosion, as computed with MC3D-EXPLO, using a simple one-dimensional test case. The ultimate goal is to better understand the behavior of the code itself in order to give orientations for improvements. An in-depth analysis of the current knowledge and main points of modeling is at first provided, highlighting the major uncertainties. The analysis of pressure shock propagation in two-phase liquid/gas flows is then performed. It allows to propose an explanation of the difficulties encountered to obtain strong explosions with UO 2 /ZrO 2 melts from a distant trigger in one-dimensional geometries such as in the KROTOS installation. The general cases of 3-phase flows are then examined with details, as a function of the melt and vapor volume fractions. A comparison of the behavior with the classical analytical detonation model is given and the competition between the fragmentation and heat transfer in terms of time scales is outlined. Finally, the impact of major modeling issues are discussed and conclusions are provided for recommendations of improvements. [ABSTRACT FROM AUTHOR]

Reactor design and uncertainties propagation request repeated execution of neutron diffusion calculations. To address this, one can rely on the Generalized Perturbation Theory which is restricted to single integral values such as reaction rate ratios. To overpass such limitations, the perturbed flux could be calculated with the modal expansion method. The difficulty to calculate more than one eigenvector has limited its use. Hence, modal expansion lacks a clear explanation of its strengths and weaknesses. Using recent advances in eigenvalues solvers, we have successfully applied modal expansion to the 2D-BIBLIS and IAEA-3D benchmarks. Our results are promising for fission and capture cross sections perturbations. Still, modal expansion is faulty for local perturbations or when it concerns the reflector. In terms of computational cost, modal expansion is more efficient than solving the direct problem with a calculation time divided by 8 for a basis of 20 eigenvectors. [Display omitted] • Calculation time is a critical issue in reactor physics computations. • The modal expansion method can drastically reduce the calculation time. • The modal expansion method gives promising results. • Modal expansion method is faulty when perturbations are found to be local. [ABSTRACT FROM AUTHOR]

• Two series HEU-THERM benchmarks are interpreted using OpenMC for nuclear data analysis; • Analysis of ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, and JEFF-4 T1 is performed; • 9Be evaluation (especially for ENDF/B-VIII.0) should be reviewed; • Thermal Scattering Law (TSL) of BeO has a great influence on HMT-027 benchmarks (up to 2500 pcm); • TSL of metal Be is negligible for the HMT-026 experimental benchmarks. This study aims to provide additional validation of nuclear data performance through a comparative analysis of the evaluated nuclear libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, and JEFF-4 T1. The study utilizes the OpenMC Monte Carlo neutron transport code and integral benchmarks provided by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) which have not been interpreted since their first modeling. The criticality benchmarks assemblies used in this study cover 2 series which are fueled with Highly Enriched Uranium (HEU) in Metal form (MET) and have Thermal neutron spectra (THERM), referred to as HMT-026 and HMT-027, respectively. The results show that for the HMT-026 benchmarks, ENDF/B-VII.1 and ENDF/B-VIII.0 have results close to experiments. The best global agreement with the HMT-027 experiments was found to be ENDF/B-VII.1. Through relevant critical and nuclear data sensitivity analysis, it is found that Be and BeO are respectively of crucial importance for HMT-026 and HMT-027 and the nuclear data of 9Be (especially for ENDF/B-VIII.0) should be reviewed. Additionally, it is worth noting that the Thermal Scattering Law (TSL) of BeO has a great influence on the k eff calculation for the HMT-027 benchmarks, while the influence of TSL of metal Be is negligible for the HMT-026 experimental benchmarks. [ABSTRACT FROM AUTHOR]

• Conducting a benchmark case study on FFTF LOFWOS Test#13 by utilizing the modern system code SAM. • Performing comprehensive uncertainty quantification using the sampling method in conjunction with DAKOTA. • Implementing statistical and machine learning methods to conduct global sensitivity analysis using Sobol indices. The development and deployment of advanced reactors, such as the sodium-cooled fast reactor (SFR), relies on sophisticated modeling tools to ensure the safety of the design under various transients. The predictive capability of these advanced modeling tools requires validation to garner trust in supporting the licensing of the advanced reactors. For this reason, 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. In this study, we present and discuss the benchmarking efforts of the modern system code SAM on the FFTF LOFWOS Test #13. The SAM baseline model was developed according to the benchmark specification, which included a detailed core model with reactivity feedback. Generally, good agreement was observed between the baseline results and benchmark measurements; however, discrepancies persisted, particularly in predicted fuel assembly coolant outlet temperatures. Utilizing the baseline model, uncertainty quantification (UQ) and sensitivity analysis (SA) were conducted with the assistance of various statistical learning and machine learning methods, including kernel density estimation, Gaussian processes, and Sobol indices. Following the baseline model prediction and UQ and SA results, we discuss the reasons for the simulation discrepancies and propose further improvements to the model. This benchmarking effort adheres to the best-estimate plus uncertainty approach and can serve as a valuable example for supporting risk-informed licensing of advanced reactors. [ABSTRACT FROM AUTHOR]

• Mesh sensitivity analysis and validation study had been conducted. • Three different types of bowing fuel assembly are simulated. • The results of cross-sectional average velocity and temperature, as well as local velocity, pressure, and temperature are obtained. To investigate the effect of fuel assembly bowing on coolant flow and heat transfer, three-dimensional thermal–hydraulic characteristics of 5×5 bowing fuel assembly are investigated using CFD simulation. The effects of different bowing sizes (1 mm and 2 mm) on the coolant thermal–hydraulic characteristics under three patterns of bowing shapes (C-shape, S-shape, and W-shape) of fuel assembly are revealed. The results show that in case 6 bowing of the fuel assembly reduces the average coolant outlet temperature by 0.37 k. Compared with normal conditions, the pressure, velocity, and temperature of the corner channel and the side channel are more affected by bowing than the middle channel. The research has reference value for fuel assembly design. [ABSTRACT FROM AUTHOR]

• Karlsruhe Tool for Uncertainty and Sensitivity Analysis (KATUSA) tool (as part derived from the Fast Source Term Calculation (FSTC) tool) is described; • Medium Break Loss-of-Coolant Accident (MBLOCA) scenario simulated with the ASTEC code up to the basemat rupture; • Consistency of the simulations results is proven – using two sets of simulations: up to the basemat rupture and up to the lower head vessel failure + 6000 s; • Results for the release of the different fission products (FPs) into the environment and containment are presented; • Spearman correlation values for different "FPs release – input parameter" pair are presented; • Role of the input parameters governing aerosols behavior shown for the later stages of the severe accident (SA); • Influence from the pre-correlation set for the input parameters is discussed. This paper presents the results of uncertainty and sensitivity (U&S) analysis of Medium Break Loss-of-Coolant (MBLOCA) severe accident (SA) scenario simulations performed with the ASTEC code with a generic input deck prepared for a KONVOI-1300 nuclear power plant (NPP). The analysis was done with the in-house Fast Source Term Calculation Tool (FSTC) (currently: KArlsruhe Tool for Uncertainty and Sensitivity Analysis (KATUSA)). Part of the presented here results – analysis of MBLOCA scenario simulations up to the basemat rupture - were already shown at the ERMSAR conference in 2022. Here an extension related to the analysis of shorter MBLOCA simulations (up to 6000 s after lower head vessel failure (LHVF)) and investigation of the influence from adding correlations between uncertain input parameters is given. The analysis allows to identify which uncertain input parameters influence the release of fission products (FPs) in the containment and environment at the different stages of SA progression. [ABSTRACT FROM AUTHOR]

This work studies the reproducibility of passive gamma spectroscopy measurements for spent nuclear fuels (SNFs). The fifty assemblies used for this study span over a variety of initial enrichments, burnups, and cooling times. These SNFs have been measured in two different gamma axial measurement campaigns. The net peak counts are determined for Cs-137, Eu-154 and Cs-134. Furthermore, a sensitivity analysis of the relative position of the SNF and the detector is performed. Most importantly, this work describes a methodology using an intrinsic self-calibration procedure that can be used to compare the relative activities of the radionuclides without the need for detailed knowledge about the measurement set-up and its properties. The reproducibility of the Cs-137 net peak count rate ranges between 2% and 4%. Systematic reproducibility of the ratio of Eu-154 and Cs-134 to Cs-137 is between 0,4% - 5 % using the intrinsic self-calibration method. • Reproducibility study of gamma measurements for entire spent nuclear fuel assemblies. • Development of intrinsic self-calibration curves. • Impact of spent nuclear fuel assemblies' position during the gamma measurements. • New experimental gamma measurements data included for spent nuclear fuel assemblies. [ABSTRACT FROM AUTHOR]

• Using sensitivity analysis leads to improve cross section evaluation. • High resolution data are needed to derive resonance parameters. • Neutron capture data are of paramount importance in the 233 U evaluation. • Issues on the k eff bias are addressed using IRSN computer tools. The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is the French public service expert in nuclear and radiation risks with activities that cover all the related scientific and technical issues. IRSN provides expertise in support of radiation protection and nuclear safety risks. A great deal of research work carried out at IRSN concerns the improvement of computational tools and data used on a daily basis by nuclear safety assessment experts. Particularly in this paper the approach used to improve data quality and its uncertainties is presented. Computation tools and nuclear data evaluation methodology are discussed. [ABSTRACT FROM AUTHOR]

• The mathematical model for reactor core coolant fill rate is built. • The sensitivity analysis model adopts the GUM method. • Sensitivity factors of cross validation of fill rate are determined. • The optimal window and system condition setting for the redundancy cross validation is proposed. The coolant fill rate setting of CPR1000 unit reactor core has train A/B of redundancy measuring channels, whose function is to monitor the water inventory of reactor core under the working conditions of water loss and other accidents to ensure the safety of reactor core. In order to confirm the reliability of train A/B of redundancy measuring channels, a technical solution for the cross comparison of fill rate during the reactor refueling cycle is proposed to validate the consistency of train A/B of redundancy measuring and computing channels. Through the cross validation model of fill rate of reactor core coolant monitoring system, the sensitivity analysis on such input variables as relative pressure of primary coolant circuit, temperature of reactor core, differential pressure of reactor core, absolute pressure of containment, etc. shall be conducted. The sensitivity analysis model adopts the GUM method for the law of propagation of uncertainty, through which the proportion of uncertainty components of various input variables can be obtained. Moreover, the aided analysis based on the engineering statistic data from the nuclear power units of various bases shall be conducted to determine the sensitivity variables of cross validation of fill rate. Based on the combination of variable sensitivity analysis and process variables influence analysis with the switching process of the unit uplink status and the objective law of system change, the optimal window and system condition setting for the redundancy cross validation is proposed, including such factors and statuses as system pressure, temperature, running condition of primary pump, etc. Such solution has been applied to engineering sites, and has achieved good effects which can prove its feasibility and effectiveness. [ABSTRACT FROM AUTHOR]

• The gamma-eigenvalue is the spatial decay constant in an exponential experiment. • A new Monte Carlo method for sensitivity coefficient (SC) of gamma is proposed. • The method is applicable to SC of alpha-eigenvalue (time decay constant) as well. • The differential operator sampling method is used for calculating SC. • The SC is obtained by the ratio of two SCs using the k-alpha or k-gamma algorithm. A new Monte Carlo method is developed to calculate the sensitivity coefficients of the α -eigenvalue (the time decay constant) and the γ -eigenvalue (the spatial decay constant in an exponential experiment) with respect to nuclear data. A method that was previously developed for the sensitivity analyses of the α -eigenvalue, which is not based on the normal k - α algorithm, is not applicable to the γ -eigenvalue due to its inability to obtain a converged source distribution. Then, a two-step method in which two sensitivity coefficients are separately calculated using the k - α or k - γ algorithm and the differential operator sampling method is newly developed. The sensitivity coefficient of the α - or γ -eigenvalue is represented by the ratio of the two sensitivity coefficients. Some numerical tests for three-energy group problems are performed using the new method. The sensitivity coefficients that are obtained by the new method are verified by comparing them to the solutions of deterministic transport calculations or to the approximate results that are obtained from the direct perturbations of the cross-sections. [ABSTRACT FROM AUTHOR]

• A fine Control Volume division was applied to the core region of Fukushima Unit-3. • Sensitivity analysis of SRV operations during core slumping has been performed. • Sensitivity analysis of the alternative water injection has been performed. The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions resulting from loss of offsite power and AC power. It is believed that the Station Blackout (SBO) and loss of heat sink led to core meltdown in all Units 1–3. Despite past research efforts on the severe accident progression in Fukushima NPP Units 1–3, there are still knowledge gaps and uncertainties existing in understanding of the severe accident scenarios and consequences. Hence, this study aims at identifying the modeling uncertainties and addressing the sensitivity parameters in Fukushima NPP Unit 3. A more detailed Control Volume (CV) division model of the reactor core region has been developed to better simulate the thermal-hydraulic behavior of liquid water and steam, which is considered to be crucial in simulating the core uncovery and degradation process. The boundary conditions such as the water injection rates by the Reactor Core Isolation Cooling (RCIC) system, the High Pressure Core Injection (HPCI) system and Alternative Water Injection (AWI) to the reactor core were determined based on the available reactor water level and pressure measurement data. The current study suggested that the local vapor heatup behavior could influence the core melting and relocation behavior, which can lead to different core degradation scenarios. With the current modeling assumptions in MELCOR, the best estimate conditions for RPV pressure history of Unit 3 suggested that 6 SRVs could have remained open when the major core slumping took place at ca. 45:20 h (ca. 12:00, March 13) with 50 to 80 tons of water inventory in the lower plenum. The current analysis also suggested that the efficiency of the AWI to the reactor core could have been only 15% as of reported by TEPCO with the current modeling conditions if debris dryout was assumed to have occurred at around ca. 54.0 h (20:40 h, March 13th). As for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-Miller creep rupture model in the current MELCOR modeling. [ABSTRACT FROM AUTHOR]

• Nuclear technology needs better evaluation of gadolinium cross sections for safety and economics. • Experiments at n_TOF (CERN) facility obtained measures of gadolinium thermal capture cross sections. • Sensitivity analysis to gadolinium capture cross-sections for ZED-2 facility with MCNP6.1 code. • New gadolinium capture cross-sections perform better in criticality experimental benchmarks. This paper presents the results of a sensitivity and uncertainty (S&U) analysis on k eff performed on a BWR fuel assembly (FA) poisoned with gadolinia, as well as of a criticality and sensitivity study of the ZED-2 research reactor, in whose moderator has been dissolved. The evidence collected from the ZED-2 analysis, the BWR FA S&U analysis, and from other criticality benchmarks indicates that improvements should be introduced in the 157Gd (n, γ) cross section evaluation, especially in the thermal and near-thermal ranges, and in its uncertainty. In 2016, to cope with this need, the n_TOF Collaboration conducted an experimental campaign to obtain new 157Gd and 155Gd cross section measurements. The preliminary information available from the analysis of the experimental data suggests that these new data have the potential to correct some of the discrepancies found between some experimental benchmarks and their simulation. [ABSTRACT FROM AUTHOR]

• Inverse uncertainty quantification is implemented based MLE and formulation. • Mesh refinement effect on statistical parameters of physical models is investigated. • Methodology implemented for RSTART code based on the BFBT benchmark. • Noticeable effect of mesh refinement on the estimates of the mean was observed. • Estimations based on most refined mesh led to most satisfactory results. Proper quantification of the uncertainties in the input parameters of a thermal-hydraulics code (e.g. physical models) is an essential part of the forward Uncertainty Quantification (UQ) problem. Such quantification can be systematically achieved by solving the Inverse Uncertainty Quantification (IUQ) problem using data from code predictions and experimental measurements. The IUQ problem is highly dependent on several factors such as geometry and discretization. In this paper, we study the effect of mesh refinement on the statistical parameters of three uncertain physical models (interfacial friction coefficient, wall to liquid friction coefficient and critical heat flux). A mathematical framework based on MLE and MAP algorithms is implemented to perform IUQ for the physical models of the thermal-hydraulics code RSTART based on experimental data from the OECD/NEA BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark. Sensitivity analysis required for the IUQ problem was achieved by implementing discrete adjoint methods in the thermal-hydraulics code. The statistical parameters of the three physical models were obtained by implementing the MLE and MAP methods using code predictions that correspond to three different mesh sizes. Results from MLE and MAP showed a noticeable effect of mesh refinement on the estimates of the mean and a negligible effect on the estimates of the variance of the physical models. The compensation of error due to mesh refinement was studied by running the code with perturbation values calculated using the three different mesh sizes. It was shown that the results based on the most refined mesh led to best agreement between experimental data and code prediction. [ABSTRACT FROM AUTHOR]

• A structural method for uncertainty evaluation of constitutive models is proposed. • Several tools for model uncertainty quantification are investigated and discussed. • Uncertainties of important models during the LBLOCA are quantified. • Best estimate plus uncertainty analysis is applied to the LP-02-6 experiment. Best estimate (BE) codes are developed to carry out realistic safety analysis of the nuclear reactor, and generally hundreds of constitutive models are comprised in a BE code. Nevertheless, uncertainties of these constitutive models are often not properly handled in the best estimate plus uncertainty (BEPU) analysis. It is not sufficient to fully evaluate the uncertainties of different sorts of models with only one method. Thereby a structural method for uncertainty quantification (UQ) of constitutive models is proposed and relevant description is presented. Based on the method, constitutive models will be classified into two categories according to the characteristics, namely the independent model and the dependent model, and different methods will be adopted for different sorts of model. Several statistical methods for UQ of constitutive models such as the non-parametric curve estimation method, the Bayesian calibration method and the coverage calibration method are evaluated and the characteristics of these methods are discussed. In addition, several methods for the construction of surrogate model are utilized to reduce the computational cost, and a model selection technique is adopted to opt the optimal model among all alternative models. The large break loss of coolant accident (LBLOCA) experiment LOFT LP-02-6 is utilized to verify the proposed structural method, and the BEPU analysis of the LP-02-6 experiment is carried out. The results show that uncertainty intervals of the identified models obtained through the structural method are reasonable, uncertainties of the peak cladding temperature (PCT) as well as the accumulator injection time (AIT) are quantified, and the sensitivity analysis is carried out to evaluate the influence of different input parameters on the 1st PCT, 2nd PCT and the AIT. [ABSTRACT FROM AUTHOR]

• The PSA methodology is very effective in improving reactor safety in a conceptual design phase. • Risk-informed design (RID) is highly suitable way to find the deficiencies of a reactor under design and to improve the reactor safety by solving them. This paper describes the effort to improve the design and procedure of a research rector with regard to safety, using a probabilistic safety assessment (PSA) at the conceptual design stage. A PSA was undertaken to assess the level of safety for the design of a research reactor and evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA reported here is a Level 1 PSA, which addresses the risks associated with core damage. The technical objectives of this study were to identify accident sequences leading to core damage and to derive design and procedure improvements from the dominant accident sequences through sensitivity analysis. The preliminary result indicates a point estimate of 6.79E-05/yr for the overall core damage frequency (CDF), attributable to internal initiating events for the research reactor under design. Based on the dominant accident sequences from the PSA, six kinds of sensitivity analysis were performed, and some design and procedure improvements were derived. When four of these six ways to improve safety were applied to the reactor design and procedure, risk was reduced to approximately 1.21E-06/yr. The safety of the research reactor was significantly improved, and the risk was considerably more reduced than that observed before the design and procedure improvements gained from the sensitivity analysis were adapted. The present study indicated that the research reactor has well-balanced safety with regard to each initiating event contributing to the CDF. [ABSTRACT FROM AUTHOR]

• A new PIRT of ATWS-LONF sequences is proposed. • A review of ATWS DSA/PSA, experiments, human actions and sensitivities is made. • 20 phenomena are assessed in two different time phases. Anticipated Transient Without Scram (ATWS) sequences belong to the beyond design basis events. Within ATWS sequences, the most limiting for PWRs is the Loss Of Normal Feedwater (LONF). This sequence produces a pressure peak in the reactor cooling system that can compromise its integrity. To cope with this sequence, systems like the AMSAC were installed in PWR Westinghouse reactors, helping to manage the transient. However, the consequences of this sequence are still strongly dependent on several variables such as the systems availability, reactor coolant system characteristics and the human actions performance. In order to study this sequence, a sensitivity analysis focused on the most influential parameters is performed. The parameter selection for these sensitivities is based on the literature found, and the analysis is made using a 3-Loop PWR-W TRACE model. Moreover, a proposal of a Phenomena Identification and Ranking Table (PIRT) for the ATWS-LONF sequence is developed. The sequence is analyzed for two different time phases (before and after the pressure peak), focusing on 6 different systems and assessing 20 phenomena. In order to rank the importance of the different phenomena, a review of the experiments, proceduralized human actions, PSA and DSA is performed, with the additional information obtained in the sensitivity analysis. The results show that the important phenomena are those related to reactivity feedback, the energy stored in the core, the AMSAC actuation, and the mass and energy release through pressurizer safety and relief valves. [ABSTRACT FROM AUTHOR]

Nuclear containment structure is the most important shielding structure in nuclear power plant, the main function of the containment structure is to prevent the leakage of radioactive materials into environment during the accident conditions, it is of great significance to gain insight into the performance of the containment under severe accidents. This paper presents fragility analysis of prestressed concrete containment vessels under severe accidents. Firstly, detailed three-dimensional finite element model of the containment is established, material nonlinearity and non-uniform effective prestressing distribution along the prestressed tendons is also considered. Secondly, Latin hypercube sampling technique combined with MATLAB program and python script is developed to update the material parameters in ABAQUS input files to generate 30 random samples. Fragility curves of five critical locations with consideration of aleatory uncertainty and epistemic uncertainty and fragility curves with different confidence levels had been thoroughly investigated. Finally, sensitivity analysis for input random variables had been conducted to provide better understanding of the significance of the various random parameters. Results indicate that functional failure of containment is dominated by the location of equipment hatch. The difference between the HCLPF pressure capacity obtained from the composite fragility curve and the lower bound pressure capacity of the containment obtained from fragility curve with 95% confidence is small. The lower bound pressure capacity and upper bound pressure capacity is 0.837 MPa and 1.259 MPa, respectively. Sensitivity analysis of the input random variables reveale that modulus of elasticity of prestressed tendon make most contribution to the pressure capacity of the containment. [ABSTRACT FROM AUTHOR]

• The physics adjoint and the mathematical adjoint are compared based on the 2-D/1-D solver. • Different adjoint flux functions are used to perform S&U analysis for comparison. • The SF96 problem and the PB-2 assembly problem are selected for verification. Sensitivity and Uncertainty analysis (S&U) based on the Classical Perturbation Theory (CPT) requires the solution of the adjoint flux. The adjoint flux is used as a weight function for sensitivity calculation. In the previous study, the MOC-based adjoint flux in the 2-D/1-D transport solver was given and applied to S&U analysis in the UAM benchmarks. The eigenvalue k obtained by the forward calculation is different from that obtained by the adjoint calculation. In order to illustrate the phenomenon, three different adjoint flux functions, the MOC physics adjoint, the CMFD physics adjoint, and the CMFD mathematical adjoint are solved based on the 2-D/1-D solver. Next, three different adjoint flux functions are used to perform S&U analysis. The SF96 problem and the PB-2 assembly problem are selected for verification. Recommendation is made for the proper application of the CMFD mathematical adjoint flux to sensitivity calculation due to less computation time. [ABSTRACT FROM AUTHOR]

• Sensitivity analyses were performed to determine an optimized source intensity for a numerical reactor model. • Sensitivity analyses were performed to determine an optimized time bin size for a numerical reactor model. • When the source intensity is greater than 105 s−1, the Rossi-alpha method has errors of 200 pcm or less. • When the source intensity is smaller than 105 s−1, the change in accuracy was shown tobe greater. • When the time bins were smaller than 10−5 s, the Rossi-alpha method showed errors than 200 pcm. • When the time bins were greater than 10−5 s, the change in accuracy was shown to be greater. • In this study, setting the neutron source intensity to 105 s−1 or higher and the time bin unit to 10−5 s or less provides converged accuracy and standard deviation. Until now, various studies focusing on subcriticality measurements have been conducted using the Rossi-alpha method. However, no guidelines have been provided for the source intensity and time bin size of the Rossi-alpha analysis. In this study, sensitivity analyses were performed to determine an optimized source intensity and time bin size for a numerical reactor model. Using the MCNP6 Monte Carlo code, fission count signals in the fuel region were tallied 50 times with different random seeds in real-time mode. These signals were applied to the Rossi-alpha method as detector signals. The estimated k eff values were compared to the reference k eff values generated by the MCNP6 criticality calculation. As a result of the sensitivity test for the neutron source intensity, the Rossi-alpha method has errors of 200 pcm or less when the source intensity is greater than 105 s−1. It is shown that the standard deviations of 50 analyses are converged when the source intensity is greater than 105 s−1. At the same time, as a result of the sensitivity test for the time bin size, the Rossi-alpha method shows errors less than 200 pcm when the time bin size is smaller than 10−5 s. It is shown that the standard deviations of 50 analyses are converged when the time bins are smaller than 10−5 s. For the numerical reactor model in this study, the threshold time bin size was set to 10−5 s. However, it generally depends on the neutron generation time of the problem. Thus, it is recommended to set the neutron source intensity to 105 s−1 or higher and the time bins to the time of the neutron generation or less when the Rossi-alpha method is applied to subcriticality measurements in nuclear reactor cores. [ABSTRACT FROM AUTHOR]

Highlights • Analyzing uncertainty on keff due to cross section uncertainties in KRITZ-2:19. • Sensitivity vectors are calculated using the Monte Carlo code MCNP6.1 - Ksen card. • Cross sections were processed by the recent nuclear data processing system NJOY2016. Abstract In criticality and stability studies of the nuclear reactor, it is important to evaluate the impact of the uncertainties of the basic nuclear data (cross sections) on the different neutron parameters. In this work sensitivity and uncertainly analysis of KRITZ-2:19 experiment at 21.1 °C was carried out with respect to cross-section uncertainties of 1H, 10B, 16O, 235U, 238U, 239Pu, 240Pu, 241Pu and 241Am isotopes on the effective multiplication factor " keff " for JENDL-4.0 and ENDF/B-VII.1 evaluated libraries. Sensitivity vectors calculated by card KSEN/MCNP6.1 have been combined with the covariance matrices (from ENDF/B-VII.1 , JENDL-4.0 , and/or ENDL-3.2) generated by the ERRORJ module of NJOY2016 to produce nuclear data uncertainties. Comparison between JENDL-4.0 and ENDF/B-VII.1 evaluated libraries consistency among the calculated-to-experiment values of keff and the overall computational uncertainties is discussed. The keff was found to have the largest sensitivity to fission and capture cross-sections of 239Pu for the two libraries JENDL-4.0 and ENDF/B-VII.1. Neutron fission reaction of 239Pu for the two evaluated libraries was found to contribute to the major part of the keff uncertainty due to the cross-sections. The discrepancy of the calculated and measured keff values was strongly correlated with the uncertainty in the calculated keff due to the 239Pu neutron fission cross-section uncertainty. [ABSTRACT FROM AUTHOR]

Highlights • Application of Game theory principles in variance-based sensitivity analysis. • Shapley effect is compared to Sobol indices under correlated parameters. • Application is demonstrated using a lattice depletion problem. • Analysis of operating conditions effects on variance contribution. Abstract Analyzing the variance of complex computer models is an essential practice to assess and improve that model by identifying the influential parameters that cause the output variance. Variance-based sensitivity analysis is the process of decomposing the output variance into components associated with each input parameter. In this study, we applied a new concept of variance-based sensitivity analysis inspired by the game theory proposed by Shapley. The technique is called the Shapley effect, and it investigates the contribution of each input parameter as well as its interactions with every other parameter in the system by exploring all possible permutations between them. The Shapley effect is compared to the common Sobol indices technique (first order and total effects) to investigate their performance under correlated and uncorrelated parameters. The Shapley effect demonstrated superior performance when compared to the Sobol indices for correlated input parameters. Shapley effect captured the correlation between the input parameters, expressing the variance contribution in a single index instead of two indices, and normalization of the fractional indices is preserved without over or underestimation. On the other hand, the two algorithms we selected to calculate Sobol indices under correlated inputs experienced different issues including: over/underestimating the output variance, first order effect could be larger than the total effect, possibility of negative indices, unnormalized fractional indices, and difficulty to interpret the results. However, Sobol showed satisfactory performance when the inputs are uncorrelated as the numerical values and input ranking were in good agreement with Shapley effect. The main disadvantage of Shapley effect is its large computational cost especially for high dimensional problems where the number of possible input subsets becomes very large. The results of our tests showed that the thermal fission cross-section carried most of the uncertainty at BOL, and its contribution declines after fuel burnup, which is replaced by the uncertainty contribution of the fast cross-section parameters. [ABSTRACT FROM AUTHOR]

• A sensitivity and uncertainty analysis of two SFRs is conducted; • The main breeding ratio uncertainty drivers are 239 Pu n , γ and 238 U n , γ ; • The 238 U n , n ' reaction is a significant uncertainty source in the breeding ratio; • 16O introduces the uncertainty in the breeding ratio chiefly indirectly; • A softer fast spectrum is able to increase the total breeding ratio uncertainty. Nuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239 Pu n , γ , 238 U n , γ , and 238 U n , n ' . The uncertainty sources between the models are compared, and 16 O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum. [ABSTRACT FROM AUTHOR]

• The knowledge and technical understandings on the corium cooling mechanisms especially including the corium behaviors in pre-flooded cavity have been much improved. • The sensitivity analysis was performed for the corium coolability, and it was confirmed that the identification of the required amount of water for flooding can provide higher confidence for BMT-prevention strategy. • Estimation methodology to enhanced cavity flooding strategy was proposed in this paper. To mitigate the concrete ablation by the relocated molten corium in cavity, the decay heat from molten corium should be removed sufficiently to prevent BMT by flooding the cavity. In practical aspect, the negative effects of this cavity flooding strategy, such as threats of the containment over-pressurization and the steam explosion, may compete with the effectiveness of the strategy itself at the real accident situation. According to recent researches, the understanding the cavity corium behaviors has been advanced and it can be utilized in improving the cavity-flooding strategy, and based on this consideration, the evaluation methodology to enhance cavity flooding strategy was proposed in this study. The main idea is how to materialize the technical aspects of the enhanced ex-vessel corium coolability into the existing cavity flooding strategy together with minimizing the negative effects. Application of this evaluation methodology can provide an optimal way to prevent BMT more effectively. [ABSTRACT FROM AUTHOR]

Highlights • A more realistic simulation of the ocean conditions was provided. • Uncertain samples related to ocean motions were generated. • An ocean-based thermal hydraulics analysis code has been developed based on RELAP5. • Uncertainty analysis of natural circulation under ocean conditions was carried out. Abstract The stability of natural circulation is more susceptible to the uncertainties involved in the thermal-hydraulic process because the driving force is relatively small. Especially in the marine nuclear power plants, the complex ocean motions will introduce more uncertainties in the natural circulation operation. In this work, the joint probability distribution function of wave parameters was investigated, which provided a more realistic simulation of the ocean conditions. On that basis, it was applied to generate ocean motion parameters as uncertain input via the Metropolis-Hastings algorithm. A thermal-hydraulic code that is capable of simulating system behaviors under ocean conditions was developed to propagate the uncertainties. Additionally, the sensitivity analysis was carried out to identify the parameters that have a significant impact on natural circulation characteristics. Results show that the intense ocean motions will introduce large uncertainties in the natural circulation operation, especially at the startup stage. The rolling amplitude contributes the most on the natural circulation flow rate and the coolant temperature, while the heaving motions have less influence. Uncertainties introduced by the ocean motions will damage the natural circulation ability and stability, which need further considerations during the design optimization. [ABSTRACT FROM AUTHOR]

Highlights • Pu and TRU composition when recycled in FRs was obtained by 3D burnup simulating. • Sensitivity of fuel out-of-pile time and actinides recovery ratio were studied. • The installed capacity and HLW generation were sensitive to the key parameters. • The recommended target for the key parameters in future R&D was proposed. Abstract The sustainable development of nuclear energy depends on maximization of uranium utilization and minimization of waste produced. Sensitivity analysis was made on the influence of out-of-pile residence time and the actinides recovery ratio on nuclear energy system performance; a series of recommended values were given; the effect of closed fuel cycle on the reduction of high-level waste(HLW) was analyzed. The following conclusions were drawn: 1) when Pu or transuranics(TRU) multi-cycled in fast reactors(FRs), the composition would reach equilibrium; for the referenced CFR1000, the fissile Pu took an equilibrium fraction of 70%, and the minor actinides(MA) took an equilibrium fraction of 3%(in TRU recycle); 2) the installed FRs capacity was sensitive to out-of-pile residence time and recovery ratio; the HLW generation was only sensitive to recovery ratio; the recommended value was, out-of-pile time be no more than 5a and actinides recovery ratio be no less than 99.9%; 3) the synergistic development of pressurized water reactors(PWRs) and FRs could, not only improve uranium utilization, but also effectively reduce HLW generation; compared with PWRs with once-through cycle, FR with closed-fuel-cycle could reduce the long-term radioactive toxicity of HLW to 1/5–1/6 for Pu recycle and 1/7–1/8 for TRU recycle (both with 99.9% recovery). [ABSTRACT FROM AUTHOR]

Highlights • A Kriging model was used as a surrogate for complex T-H code. • Reliability assessment of PRHRs in IPWR was carried out. • Contributions of different failure modes were analyzed. • Global sensitivity analysis was performed to identify key parameters. Abstract Passive safety systems have been incorporated into new design nuclear power plants to enhance the power plants' inherent safety. However, due to the lack of the experimental data and operational experience, evaluating the reliability of passive systems is a challenging task. Many uncertainties are involved in the thermal hydraulic process during startup and operation. This may make the system unable to accomplish its expected function even though the hardware is available, known as functional failure. The evaluation of system functional failure probabilities based on the direct Monte Carlo method may be computationally impractical. In order to reduce the computational load, the Kriging regression model was constructed to avoid a large number of thermal hydraulics simulations. Compared with other metamodels, the Kriging regression model obtained better accuracy. The proposed method was applied to evaluate the reliability of a passive residual heat removal system of IPWR200 during a station blackout accident. The Kriging regression model not only ensures the accuracy of the reliability assessment, but also greatly reduces the runs of T-H codes. Furthermore, global sensitivity analysis aimed at determining the contributions of input parameters was carried out. Results indicate that the probability of passive residual heat removal system failing functionally is estimated to be 1.94e-4 and that insufficient decay heat removal is the most likely failure mode. Sensitivity analysis results identified five key uncertain parameters, which is significant for system performance. The sensitivity information can provide guidance for enhancing passive safety systems design and operation, which is of great importance to the safety and reliability of passive safety systems. [ABSTRACT FROM AUTHOR]

Highlights • A hypothetical multi-failure MSLB accident was simulated by RELAP5 code. • Through uncertainty analysis, the 95/95 uncertainty bands of some key outputs were obtained. • The simulation results were compared with experiment data to assess the code's capacity. Abstract The RELAP5/MOD3.3 code was used to simulate a Main Steam Line Break (MSLB) Loss of Coolant Accident (LOCA) on a Generation III type Reactor with passive safety features. The MSLB is a hypothetical multi-failure large break LOCA accident scenario. The results of the base case calculation have been compared with the available integral testing data to assess the code's capability to predict the large break LOCA phenomena in this generic reactor, particularly with passive safety systems. Various flows such as break flow and gravity injection flow were simulated with a high level of accuracy. In addition, both sensitivity cases and uncertainty analysis were carried out to build the best estimate plus uncertainty practice. The sensitivity study was performed with different initial conditions and adjusted containment cooling capacities. The results confirmed that the code model held its stability and repeatability against a few variations of initial and boundary conditions. In the uncertainty quantification process, the 95/95 uncertainty bands of key output parameters were obtained using the Wilks statistic methods. Ten input parameters were selected with imposed uncertainty distribution to perform uncertainty propagation analysis. The contribution weight of input parameters for the uncertainty of minimum core level were identified with the Spearman rank correlation coefficient in sensitivity analysis. [ABSTRACT FROM AUTHOR]

Highlights • A cheaper way to validate the reactor-physics code SARAX for application in CFR is introduced. • The nuclear-data adjustment to extrapolate measurements to CFR is calculated and analyzed. • Different correlation coefficients between measurements are compared. Abstract The validation works have been implemented to a newly-developed code SARAX for China Fast Reactor (CFR) in this paper. Different with the conventional way to validate the code using the dedicated experimental data, a theoretical approach was proposed and implemented by using the existing similar experimental data. This theoretical approach is based on the technology of sensitivity/uncertainty analysis, similarity analysis and nuclear-data adjustment. In our previous works, detailed introduction towards sensitivity/uncertainty analysis and similarity analysis have been implemented to distinguish the existing experiments which are 'similar' to the CFR's criticality characteristics. This paper focus on the method to extrapolate the 'similar' experimental data to predict the best-estimate measurements of CFR with application of nuclear-data adjustment, providing reference values to validate the SARAX code. From the numerical results, it can be observed that through nuclear-data adjustment, the simulation results of the existing experiments can agree well with corresponding measurements, with the bias reduced notably to be within 25 pcm. Moreover, the nuclear-data adjustment can also improve the nuclear-data uncertainties notably, with the relative uncertainties of the reactor k eff due to nuclear data having been reduced from the value exceeding 1.0% to the values about 0.15%. [ABSTRACT FROM AUTHOR]

Highlights • A Monte Carlo method for sensitivity coefficient of alpha-eigenvalue is proposed. • The differential operator method is used for calculating sensitivity coefficient. • "Time source method" is used instead of the power iteration of "fission source". • A newly proposed method for void region can overcome the difficulty in "time source". • The superhistory method is introduced to reduce the computer memory. Abstract A method for Monte Carlo sensitivity analyses of α -eigenvalue (prompt neutron time decay constant) in a subcritical system is developed using the first-order differential operator sampling (DOS) method. The first-order derivative of α -eigenvalue with respect to nuclear data is calculated using the DOS method that includes the capability of calculating perturbed source effect. This paper is an extension of the author's previous work for development of the sensitivity analysis method for k eff -eigenvalue. Unlike the conventional Monte Carlo method for α -eigenvalue calculation that uses the power iteration of fission sources, this paper introduces a recently developed "time source method". The "time source method" has a weakness for a void-containing subcritical system, which is overcome by assigning a virtual total cross section in the void region. The perturbed source effect, which is caused by the change of nuclear data in a subcritical system, can be calculated by two methods, the source perturbation iteration method and the superhistory method. The source perturbation iteration method is superior in terms of computation efficiency, but a huge computer memory is required. The superhistory method dramatically reduces the memory requirement, although it worsens the variance of the sensitivity coefficients. The method developed in this paper is applied to some numerical tests that use multi-group constants, and it is verified by comparing to the results obtained by a deterministic perturbation theory. [ABSTRACT FROM AUTHOR]

Highlights • Second-order sensitivities for uncollided particles are obtained by ray-tracing. • Responses are uncollided angular flux integrals on the system boundary. • Inputs are densities, cross sections, and source emission rates. • In a homogenous spherical test problem, results agree with analytic results. • In a two-region spherical test problem, results agree with those from PARTISN. Abstract The Second-Level Adjoint Sensitivity System (2nd-LASS) that yields the second-order sensitivities of a response of uncollided particles with respect to isotope densities, cross sections, and source emission rates has previously been derived and solved using the off-the-shelf discrete ordinates code PARTISN. In this paper, the 2nd-LASS equations are solved using ray-tracing for the case when the response is evaluated at a point on the boundary (the partial current density or scalar flux). The resulting second derivatives, including all mixed partial derivatives as well as derivatives involving material mass densities, are compared with analytic and PARTISN results for a homogeneous sphere and PARTISN results for a two-region sphere. Second derivatives computed using ray-tracing are within 0.003% of those computed using PARTISN and are over 100 times faster to compute for the spherical test problems. [ABSTRACT FROM AUTHOR]

Highlights • The prompt method and the 1/v insertion method have been implemented in SUMMON. • Methods to derive the sensitivity to nuclear data of β eff and Λ eff are presented. • The sensitivity methods have been verified against well-validated codes. • Good agreement between SUMMON and SUSD3D and SERPENT has been obtained. Abstract Accurate and reliable tools for sensitivity analysis and uncertainty quantification are needed to estimate the uncertainties in reactor key parameters (k eff , β eff , Λ eff , safety coefficients, ...) and to identify possible nuclear data weaknesses. A Sensitivity and Uncertainty Methodology for MONte carlo codes (SUMMON) has been developed to calculate the sensitivities of the criticality safety parameters as well as their uncertainties due to uncertainties in nuclear data, addressing limitations of existing tools. The methodology is currently based on the use of the KSEN card of MCNP6 code to perform the eigenvalue sensitivity calculations, although any Monte Carlo code that can provide sensitivity coefficients can be used. The sensitivity coefficients of a reactivity response are calculated using the eigenvalue definition of reactivity, which is equivalent to applying the Equivalent Generalized Perturbation Theory. Moreover, the effective delayed neutron fraction sensitivity coefficients are derived from Bretscher's approximation, whereas the sensitivity coefficients of the effective neutron generation time are obtained using the 1/v insertion method and the Equivalent Generalized Perturbation Theory. Uncertainties are propagated using the "Sandwich Rule" of the "Propagation of Moments" method employing state-of-the-art covariance libraries. The implementation in SUMMON of Bretscher's approximation and the 1/v insertion method has been validated using integral experiments from ICSBEP and IRPhE databases, while the methods for calculating the sensitivity coefficients have been verified against consolidated codes, such as SCALE, SUSD3D, SERPENT and XSUSA and good agreement has been obtained. [ABSTRACT FROM AUTHOR]

Highlights • Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM) Roussopoulos-functional. • First-order sensitivities of the Roussopoulos-functional to system parameters. • Second-order sensitivities of the Roussopoulos-functional to model parameters. Abstract The Roussopoulos-functional has long been used for selecting trial functions for the forward and adjoint fluxes when computing reaction rates in nuclear systems and/or particle detector responses, because this functional was considered in previous works and textbooks to be accurate to second-order in variations of the model's forward and adjoint fluxes. However, by applying the Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM), this work shows that the first-order sensitivities of the Roussopoulos-functional to the system's parameters (including isotopic number densities, microscopic cross sections, fission spectrum, forward and adjoint sources) are not identically zero. This means that the Roussopoulos-functional is not accurate to second-order in the flux variations if the system parameters are imprecisely (rather than perfectly well) known, as is invariably the case in practice. This work also illustrates the application of the 2nd-ASAM for computing efficiently the correct expressions of the second-order sensitivities of the Roussopoulos-functional to model parameters. The results presented in this work indicate the correct path for future possible uses of the Roussopoulos-functional for performing sensitivity and uncertainty analyses of both forward and inverse problems in nuclear systems. [ABSTRACT FROM AUTHOR]

Highlights • The sensitivity of k eff of parts of existing evaluated experiments is analyzed. • Different methods to distinguish the similar experiments are compared. • Similarity characteristics of sodium-cooled fast reactor are discussed. Abstract We conducted a validation of the newly developed SARAX code for the China Fast Reactor (CFR). Unlike the usual way to validate code using dedicated experimental data, we proposed and implemented a theoretical way using the existing experimental data. This paper introduced a method to select the suitable existing experiments, which are "similar" to the CFR's criticality characteristics. We based our method on the sensitivity and uncertainty (S&U) analysis and the similarity analysis. We analyzed different experiments and discussed different ways to distinguish these experiments. Our results showed that a large-size mixed-oxide-fueled experimental core was necessary to represent the critical characteristics of CFR for the code validation. The method using both sensitivity and covariance was better able to distinguish the difference between the objective core and the other cores. [ABSTRACT FROM AUTHOR]

Highlights • Simulations of resolved bubbles rising in a highly viscous liquid. • Zinc bromide – corn syrup mixture used as a simulant for molten borosilicate glass. • Projected bubble area at surface validated against experimental observations. • Uncertainty of 15% in fluid viscosity and 20% in surface tension. • Uncertainty in viscosity has a larger effect on bubble size than surface tension. Abstract Computational fluid dynamic and heat transfer models are being developed for the waste glass melters that will be used to vitrify legacy tank waste at the Hanford site. The objective of this study is to validate the ability of these models to predict the forced convection bubbling in the molten glass at prototypic conditions and scale. Bubbling is used to homogenize the melt pool and increase the melt rate of the batch. Simulations were compared with experimental data obtained with a surrogate liquid under isothermal conditions in an acrylic tank. The simulation results quantify the projected bubble area at the free surface as a function of flow rate and nozzle depth and show agreement with experimental data. Additionally, the local sensitivities of the surface tension and viscosity are evaluated for their effects, since in the melter this parameter directly affects the heat transfer from the molten glass to the reactive batch (or cold cap) layer. From the sensitivity study performed, the viscosity exhibits twice as much influence on bubble projected surface area as the surface tension. In comparison to the grid resolution used in this study, these glass properties produce higher levels of uncertainty. [ABSTRACT FROM AUTHOR]

Highlights • The impact of the SFP fuel assembly arrangement is more limited than expected. • The offload power does not present a crucial role in fuel damage progression. • User effects on the building modelling greatly impacts the SFP response. Abstract This work performs an analysis of the inherent nuclear spent fuel pool response to a loss of pool cooling accident. The influence of fuel offload times, fuel assembly arrangement, fuel assembly binning for modelling, and building compartmentalization are evaluated. These impacts are assessed through sensitivity calculations performed with the MAAP code. Results are compared in terms of selected significant safety quantities such as fuel uncovery time, hydrogen generation, building temperature, relocated mass to the floor, and integrated total radiological dose at 1 km distance from the spent fuel pool. According to the results, the impact of the fuel assembly arrangement on the output safety quantities is more limited than initially expected. Also, the offload power (as a function of elapsed time between shutdown and removal from the reactor pressure vessel to the spent fuel pool), relevant in the pre-uncovery phase, does not present a crucial role in fuel damage progression. Somewhat unexpectedly, user effects on the building modelling hosting the pool significantly affect its inherent response. This is due to the large influence that natural circulation currents of gas have on mitigating fuel damage as such currents significantly rely on the modelling of the building in terms of nodalization, thermodynamic imposed conditions and heat sinks configuration. Follow-on work is identified to address uncertainties in the calculation of natural convective flows and heat transfer. [ABSTRACT FROM AUTHOR]

Highlights • The numerical modeling for a repository was validated using an analytical model. • Thermo-hydraulic behavior of a bentonite buffer in a repository was investigated. • The effect of design factors for a buffer was evaluated through sensitivity analysis. • A case study of sensitivity analysis was conducted according to a buffer property. Abstract A buffer in a geological repository minimizes groundwater inflow from the surrounding rock and protects the disposed high-level waste against any mechanical impact. As decay heat of spent fuel causes temperature variation in the buffer, which affects the mechanical performance of the system, an accurate estimation of the temperature variation is substantial. The peak temperature of the buffer, which is a critical design criterion of the repository, is affected by not only the thermal properties but also the hydraulic properties of the buffer, and these properties also affect each other. Thus, the effect of these factors should be properly evaluated during the system design. Hence, in this study, a sensitivity analysis of seven input factors was conducted using a numerical model based on a finite element method, and the model was verified using an analytical solution. In addition, a case study of the sensitivity analysis was conducted on two different property relationships of the buffer. The sensitivity indices of thermal conductivity, intrinsic permeability, and viscosity, which were determined as major factors, were 0.387, 0.215, and 0.163 in Case 1, and 0.510, 0.145, and 0.110 in Case 2, respectively, which means that these properties should be carefully considered during the design process of repositories. Furthermore, the case study shows that the relationship of the properties of a bentonite buffer has a significant effect on the design properties influencing the temperature variation. [ABSTRACT FROM AUTHOR]

Highlights • Sensitivity analysis based on CPT is applied to the 2-D/1-D whole-core transport code KYADJ. • 2-D MOC and 1-D Sn coupled forward-adjoint neutron transport equations are solved by an advanced iteration approach. • A new covariance data file is processed from ENDF-B/VII.1 for uncertainty quantification. Abstract An evaluated nuclear data file is one of the important footstones in reactor physics design. Nuclear data can be viewed as arguments of output parameters. Therefore, the uncertainty of nuclear data can directly influence on the uncertainty of output parameters. A process of Sensitivity and Uncertainty (S&U) analysis is used to quantify the influence. In this study, we choose 2-D/1-D transport code to develop S&U analysis in order to reduce the impact of numerical model uncertainty. We can solve 2-D/1-D adjoint transport equation and apply the classical perturbation theory to sensitivity analysis of eigenvalue with respect to nuclear data. In addition, we obtain the uncertainty of eigenvalue using the "Sandwich" rule. The Peach Bottom-2 (PB-2) BWR cell benchmark and the Three Mile Island-1 (TMI-1) PWR cell benchmark are applied to verification. The Reactor Monte Carlo code RMC is used as a reference code. The results show that the code has ability of eigenvalue sensitivity and uncertainty analysis with high accuracy. [ABSTRACT FROM AUTHOR]

Estimating the changes in a nuclear system due to perturbations in the input nuclear data by separate Monte Carlo calculations might be extremely cumbersome for reactor applications. The Iterated Fission Probability (IFP) method has recently paved the way for the application of first-order standard perturbation theory in continuous-energy Monte Carlo codes. In this work, we detail the reactivity perturbation and k -eigenvalue sensitivity analysis capabilities of the Monte Carlo code T ripoli-4 ®. Simulation results obtained by using the newly implemented IFP algorithm of T ripoli-4 ® are compared to findings coming from other Monte Carlo methods (such as the differential operator and the correlated sampling) and codes (such as MCNP6 and KENO ). For this purpose, we select some benchmark configurations (Godiva, Stacy, Jezebel, Flattop and a fuel lattice) and we test some of the most common perturbation and sensitivity methods currently available in production codes. Their respective advantages and drawbacks are analyzed, and possible future improvements are suggested. Our main finding is that T ripoli-4 ® produces very similar results to MCNP6 when the same techniques are used. Uncertainty propagation based on the obtained sensitivity profiles and on the COMAC nuclear data covariance matrices is finally discussed. [ABSTRACT FROM AUTHOR]