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AbstractUncertainty in the figure of merit (FOM) parameters is a central feature of the best-estimate plus uncertainty (BEPU) method, which provides insight into the analysis margins not available from other analysis methods. The FOM uncertainty distributions are formed from propagation of the variations and uncertainty distributions in the operational and modeling parameters used in simulations of a design-basis accident (DBA) scenario for a nuclear power plant.To compute an accurate FOM uncertainty distribution, it is critical to accurately quantify and account for the input parameter prediction uncertainties. The coolant flow rate through fuel channels, or more precisely, the hydraulic resistance, including the impact of two-phase flow and its distribution in the primary heat transport system and the critical heat flux (CHF) of the fuel, are two key parameters for the limiting postulated accident scenarios in a CANDU reactor for various DBAs.Prediction uncertainty distributions for these parameters can be derived by directly validating code predictions against in-reactor measurements of flow rate and experimental measurements of CHF, respectively. Such code validation circumvents the convoluted and complex approach of decomposing computer models of physical phenomena into microscopic parameters, such as interfacial mass, momentum, and heat transfer correlations, and propagation of their uncertainty distributions to obtain an overall parameter uncertainty distribution of interest. Uncertainties associated with predictions of the coolant flow rate and CHF arise due to temporal and spatial variations and uncertainties in reactor conditions, limitations of physical models and their implementation in the codes, and in the case of CHF, measurement uncertainties associated with full-scale experiments.Careful assessment of key uncertainties, specifically their magnitudes, is important for ensuring uncertainty magnitudes are not unnecessarily over- or underestimated. These uncertainties also need to be characterized properly, e.g., whether uncertainties are common to a group of reactor fuel channels or vary independently for each fuel channel. Inadequate identification or incorrect classification or characterization of uncertainties would result in an inaccurate FOM uncertainty distribution.One important focus area for this study is the distinction between apparent prediction uncertainty (the difference between code prediction and measurement) and actual prediction uncertainty (the difference between code prediction and the true value). The actual code prediction uncertainty can be calculated from the apparent code uncertainty, provided there is adequate knowledge about the measurement uncertainty. The uncertainty models developed using this approach will be used as part of the BEPU analysis for slow loss-of–reactor power regulation accidents. [ABSTRACT FROM AUTHOR]

Reactor thermal-hydraulic codes can be categorized into system codes, sub-channel codes and CFD (computational fluid dynamics) codes according to the spatial discretization and simplification methods. As a traditional reactor system safety analysis code, system code is able to analyze the overall performance of the reactor quickly, while it cannot capture the mixing effects between coolant channels in the core accurately. The sub-channel code is currently the most suitable code for core analysis, with higher fidelity than system code and less computation resources than CFD code. To perform analysis of coupling effects between thermal-hydraulics characteristics of the reactor system and those of core, the in-house system code ASRAC and the in-house sub-channel code CORTH are coupled based on a generic coupling architecture. This generic coupling architecture comprises the generic coupling interface concept ICoCo (Interface for Code Coupling) and the generic data exchange model MED (Model for Exchanging field Data). In order to evaluate the accuracy and capability of the coupling code, the LOFT experiment case is chosen and analyzed. According to the validation results, compared with ASRAC code standalone, the ARSAC-CORTH coupling code is able to better analyze the coupling effects of loop system and core, meanwhile capturing the coolant mixing between coolant channels. [ABSTRACT FROM AUTHOR]

The High Temperature Test Unit (HTTU) was an experimental set-up to conduct separate and integral effects tests of the Pebble Bed Modular Reactor (PBMR) core. The annular core consisted of a randomly packed bed of uniform spheres. Natural convection tests using both nitrogen and helium, and forced convection tests using nitrogen, were conducted. The maximum material temperature achieved during forced convection testing was 1200 °C. This paper presents the numerical analysis of the flow and temperature distribution for a forced convection test using 3D CFD as well as a 1D systems-CFD computer code. Several modelling approaches are possible, ranging from a fully explicit to a semi-implicit method that relies on correlations of their associated phenomena. For the comparison between codes, the analysis was performed using a porous media approach, where the conduction and radiative heat transfer were lumped together as an effective thermal conductivity and the convective heat transfer was correlated between the solid and gas phases. The results from both codes were validated against the experimental measurements. Favourable results were obtained, in particular by the systems-CFD code with minimal computational and time requirements.

AbstractMost of the neutronic core calculation schemes used by industrialists for nuclear reactor studies are based on a two-step deterministic scheme: A two-dimensional transport calculation at the assembly level produces homogenized and condensed nuclear data used by a full three-dimensional core solver under the diffusion approximation. The validity domain of such schemes is driven by the hypotheses of the diffusion approximation, implying wide meshes (10 to 20 cm), and is hence limited to studies that do not require a thinner description of neutron behavior. Consequently, local phenomena and heterogeneous interfaces are not yet fully validated with industrial two-level calculation schemes. For instance, the axial core-reflector interface, which is characterized by extra thermalization of neutrons leading to a local (≈2-cm height) increase of the neutron flux at the axial edges of fuel pins, is specifically challenging for deterministic methods. To cope with this issue, specific safety studies are performed with reference Monte Carlo simulations. This paper shows that enhancing the equivalence method enables flux discrepancies to be reduced from 12% to 6% for mixed oxide fuels and from 9.3% to <1% for uranium oxide fuel (impacting the power discrepancy from 5% to 3% and from 0.3% to almost 0.0%, respectively) between Monte Carlo and deterministic simulations (SCIENCE V2). The improved equivalence method uses dedicated discontinuity factors and constants, according to an optimized mesh composed by a mesh for each medium and a refined mesh in the fuel region. [ABSTRACT FROM AUTHOR]

In order to support the learning of novice students in Java programming, the web-based Java Programming Learning Assistant System (JPLAS) has been developed. JPLAS offers several types of exercise problems to foster code reading and writing skills at different levels. In JPLAS, the code writing problem (CWP) asks a student to write a source code that will pass the test code given in the assignment where the correctness is verified by running them on JUnit. In this paper, to reduce the teacher's workload during the marking process, we present the answer code validation program that verifies all the source codes from a large number of students for each assignment and reports the number of passing tests for each source code in the CSV file. Besides, to test a source code with various input data, we implement the test data generation algorithm that identifies the data type, generates new data, and replaces it for each test data in the test code. Furthermore, to verify the correctness of the implemented procedures in the source code, we introduce the intermediate state testing in the test code. For evaluations, we applied the proposal to source codes and test codes in a Java programming course in Okayama university, Japan, and confirmed the validity and effectiveness. [ABSTRACT FROM AUTHOR]

Background: The positive predictive value (PPV) of the International Classification of Diseases, Ninth Revision—Clinical Modification (ICD-9-CM) code for "essential and other specified forms of tremor" in identifying essential tremor (ET) cases was found to be less than 50%. The ability of the ICD-10-CM G25.0 code for "essential tremor" to identify ET has not been determined. The study objective was to determine the PPV of the G25.0 code. Methods: Patients in a tertiary health system with a primary care encounter associated with ICD-10-CM code G25.0 in 2022 underwent medical record review to determine if the consensus criteria from the International Parkinson and Movement Disorder Society for an ET diagnosis were met. Results: 442 patients were included. The PPV of G25.0 in identifying probable ET cases was 74.7% (95% confidence interval (CI) 70.4–78.5%). Among patients prescribed propranolol, the PPV improved to 87.8% (95% CI 78.0–93.6%). Discussion: Compared to the ICD-9-CM code 333.1, G25.0 is superior for identifying ET cases. A potential limitation of this study is that the consensus criteria applied relies on nonspecific physical exam findings which may lead to an overestimation of the PPV of G25.0. Highlights: The ICD-10-CM diagnosis code for essential tremor has not been previously validated. The objective of this study was to determine the PPV of the G25.0 code. The PPV in identifying essential tremor cases was 74.7%. The PPV improved among patients prescribed propranolol. [ABSTRACT FROM AUTHOR]

The detailed behavior of neutrons in a rapidly changing time-dependent physical system is a challenging computational physics problem, particularly when using Monte Carlo methods on heterogeneous high-performance computing architectures. A small number of algorithms and code implementations have been shown to be performant for time-independent (fixed source and k-eigenvalue) Monte Carlo, and there are existing simulation tools that successfully solve the time-dependent Monte Carlo problem on smaller computing platforms. To bridge this gap, a time-dependent version of ORNL's Shift code has been recently developed. Shift's history-based algorithm on CPUs, and its event-based algorithm on GPUs, have both been observed to scale well to very large numbers of processors, which motivated the extension of this code to solve time-dependent problems. The validation of this new capability requires a comparison with time-dependent neutron experiments. Lawrence Livermore National Laboratory's (LLNL) pulsed sphere benchmark experiments were simulated in Shift to validate both the time-independent as well as new timedependent features recently incorporated into Shift. A suite of pulsed-sphere models was simulated using Shift and compared to the available experimental data and simulations with MCNP. Overall results indicate that Shift accurately simulates the pulsed sphere benchmarks, and that the new time-dependent modifications of Shift are working as intended. Validated exascale neutron transport codes are essential for a wide variety of future multiphysics applications. [ABSTRACT FROM AUTHOR]

The detailed behavior of neutrons in a rapidly changing time-dependent physical system is a challenging computational physics problem, particularly when using Monte Carlo methods on heterogeneous high-performance computing architectures. A small number of algorithms and code implementations have been shown to be performant for time-independent (fixed source and k-eigenvalue) Monte Carlo, and there are existing simulation tools that successfully solve the time-dependent Monte Carlo problem on smaller computing platforms. To bridge this gap, a time-dependent version of ORNL’s Shift code has been recently developed. Shift’s history-based algorithm on CPUs, and its event-based algorithm on GPUs, have both been observed to scale well to very large numbers of processors, which motivated the extension of this code to solve time-dependent problems. The validation of this new capability requires a comparison with time-dependent neutron experiments. Lawrence Livermore National Laboratory’s (LLNL) pulsed sphere benchmark experiments were simulated in Shift to validate both the time-independent as well as new time-dependent features recently incorporated into Shift. A suite of pulsed-sphere models was simulated using Shift and compared to the available experimental data and simulations with MCNP. Overall results indicate that Shift accurately simulates the pulsed sphere benchmarks, and that the new time-dependent modifications of Shift are working as intended. Validated exascale neutron transport codes are essential for a wide variety of future multiphysics applications.

The uncertainty quantification is an indispensable part for the validation of the nuclear safety best-estimate codes. However, the uncertainty quantification usually requires the combination of statistical analysis software and nuclear reactor professional codes, and it consumes huge computing resources. In this paper, a design method of coupling interface between DAKOTA Version 6.16 statistical software and nuclear reactor professional simulation codes is proposed, and the integrated computing workflow including interface pre-processing, code batching operations, and interface post-processing can be realized. On this basis, an integrated framework of uncertainty quantification is developed, which is characterized by visualization, convenience, and efficient computing. Meanwhile, a typical example of small-break LOCA analysis of the LOBI test facility was used to validate the reliability of the developed integrated framework of uncertainty quantification. This research work can provide valuable guidance for developing an autonomous uncertainty analysis platform in China. [ABSTRACT FROM AUTHOR]

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

The burn-up code MOTIVE is a 3D code for fuel assembly inventory determination developed at GRS in recent years. It modularly couples an external Monte Carlo neutron transport code to the in-house inventory code VENTINA. In the present publication, we report on the validation of MOTIVE with respect to full-assembly decay heat measurements of light water reactor fuel. For this purpose, measurements on pressurized water reactor and boiling water reactor fuel assemblies from different facilities have been analyzed with MOTIVE. The calculated decay heat values are compared to the measured data in terms of absolute and relative deviations. These results are discussed and compared to other published validation analyses. Moreover, the observed deviations between measurements and calculations are analyzed further by taking into account the results of the validation of nuclide inventory determination with MOTIVE. The influence of possible biases of calculated nuclide densities important to decay heat at the given decay times are investigated and discussed.

This paper presents the results of a calculation code approach providing a solution to the point kinetics problem for the IVG.1M research reactor of the National Nuclear Center of the Republic of Kazakhstan and allowing the simulation of dynamic processes going on during reactor start-ups, including changes in the thermal state of all its elements, reactor regulator displacement, accumulation of absorbers in the fuel, and the beryllium reflector. A mathematical description of the IVG.1M point kinetics model is presented, which provides a calculation of the reactor neutron parameters, taking into account the dependence of reactivity effects on the temperature, changes in the isotopic composition of materials, and thermal expansion of core structural elements. An array of data values was formed of reactivity added by separate elements of the core when changing their thermal state and other reactor parameters, as well as an array of data with the parameters of heat exchange of coolant-based reactor structural elements. These are used in the process of solving the point kinetics problem to directly replace formal parameters, eliminating the need to calculate the values of these parameters at each calculation step. Preliminary calculations to form an array of values of reactivity effects was applied to the reactor by separate structural elements when their temperature changes were performed using the IVG.1M precision reactor calculation model. The model was validated by the reactor parameters in the critical state. Preliminary calculations to form an array of data with the parameters of heat exchange of coolant-based reactor structural elements were performed in ANSYS Fluent software using the calculation model that describes the IVG.1M reactor fuel element in detail. Validation of the developed calculation code based on the results of two start-ups of the IVG.1M reactor was performed and its applicability for the analysis of transient and emergency modes of reactor operation and evaluation of its safe operation limits was confirmed. [ABSTRACT FROM AUTHOR]

BackgroundThe kilowatt reactor using stirling technology (KRUSTY) is a heat-pipe-cooled reactor experimental system that uses a Stirling engine to convert thermal energy to electricity, it is the only one published experimental data for heat-pipe-cooled reactor systems. The KRUSTY experimental data under different working scenarios include the cold startup and load change processes, heat pipe failure, reactivity insertion, and heat sink loss.PurposeThis study aims to validate the self-developed system transient analysis code named TAPIRS-D for the heat-pipe-cooled reactor concept using KRUSTY experimental data.MethodsFirstly, an in-house system code for a heat-pipe-cooled reactor named TAPIRS-D was introduced, with the main theoretical module briefly explained, including the reactor power calculation module, heat transfer module for fuel assembly, and heat pipes. Then, the TAPIRS-D was applied for the first time to the simulation of the key processes of the KRUSTY prototypic reactor test under normal operation and accident conditions. Finally, comparison between the simulation data and experimental data was conducted for the validation of this analysis code.ResultsComparison results demonstrate that the maximum relative prediction error for the fuel temperature is less than 2%, and the reactor power average prediction error is less than 10%.ConclusionsThe prediction trend of the numerical simulation by TAPIRS-D fits well with the experimental data on key parameters such as core power and the temperature of fuel and heat pipes, which indicates that TAPIRS-D is well developed and is capable of conducting safety analysis for heat pipe cooled reactor concepts. The validation of this system analysis code provides a good reference for other newly developed system codes for heat pipe reactors.

For JET L-mode plasmas in low-recycling conditions (electron temperature at the outer strike point, Te,ot≳30eV), SOLPS-ITER simulations agree within the error bars for the experimental profiles at the low-field side (LFS) divertor target. The peak Balmer-α (Dα) emission in the LFS divertor agrees within the error bars of the KS3 filterscope diagnostic, but is approximately 30% lower than the peak value of the KT1 spectrometer. Simulations have been performed with fluid, kinetic, and hybrid models for the neutrals. The large fluid-kinetic discrepancies of more than a factor 2 are successfully corrected by using a hybrid fluid-kinetic approach, for which kinetic atoms are transferred to the fluid population when the local Knudsen number of the atom becomes smaller than a user-defined transition Knudsen number Knt. The hybrid-kinetic discrepancies are limited to a few % for Knt≤100. When increasing the upstream density to high-recycling conditions, at the onset of detachment (Te,ot≈5eV), the simulations predict more than a factor 2 lower peak ion saturation current to the LFS divertor than the experiments. Also the Dα emission is underpredicted with approximately a factor 2. For these high-recycling conditions, the fluid-kinetic discrepancies are limited to maximum 50%, which are again corrected by using the hybrid approach.

• A part of the flow coming from the fuel assemblies cross the Upper Core Structure (UCS). • Pressure loss versus flow rate in the guide tubes was determined in a mimic UCS. • The flow rate is calculated in the actual USC by measuring the height of water in the guide tubes. • The results are in the same order as those determined using the velocity integration method. • The errors due to the sensors are high but may be improved by a better calibration. In the purpose of the design of SFR reactors, the CEA is developing codes, which must be validated from experimental data. Since experiments with sodium are complicated, a part of the studies is performed on small scale mock-ups using water thanks to the dimensional analysis. The mock-up MICAS, representative of the ASTRID upper plenum at a scale 1/6th, allows studies of the thermal hydraulics behavior in the vessel for the code validation. Numerical codes usually model the Upper Core Structure (UCS) as porous media because of its complexity. This methodology requires data about the pressure loss coefficients of the different components of the UCS. They are evaluated using correlations from the literature, but this method leads to uncertainties owing the complex geometry of the UCS. One aim of the work carried out on the MICAS mock-up was to obtain data about the flow rate crossing the UCS to compare them with the numerical results. In former studies, it has been calculated by measuring and integrating the velocity around the core and the UCS. Another technic, based on pressure losses measurements, was performed in the aim of comparison. The pressure drop coefficients were calculated in a dedicated mock-up replicating the UCS geometry. These results and the measurements of the pressure losses in the MICAS UCS allowed determining the flow rate crossing this component. Comparisons with numerical and former method results show some discrepancies. [ABSTRACT FROM AUTHOR]

Nuclear power plants play a critical role in providing clean and sustainable energy. Ensuring the safety of these plants is of utmost importance to the nuclear industry. In this regards, thermal-hydraulic computer codes are essential for the simulation and understanding of the behavior during both normal and accidental conditions. In this context, LOCUST 1.2 is a newly developed thermal-hydraulic code by China General Nuclear, which aims at simulating the steady-state and accidental behavior of HPR-1000, a Pressurized Water Reactor design. This paper presents the validation of LOCUST 1.2 using the Advanced Nuclear Technologies assessment methodology from the Universitat Politècnica de Catalunya. The validation focuses on a nodalization created for the Large Scale Test Facility from the Japan Atomic Emergency Agency. In particular, four tests from the OECD/NEA ROSA 1 and 2 projects were selected. The selected tests encompass scenarios such as Anticipated Transient Without Scram, Intermediate Break Loss-Of-Coolant Accident, Steam Generator Tube Rupture, and Main Steam Line Breaks. The ANT-UPC methodology provides a comprehensive phenomenological assessment combining qualitative and quantitative analyses with the help of Best Estimate Plus Uncertainty calculations. All four tests are divided in phenomena through a Phenomena Identification and Ranking Tables (PIRT) to then perform the assessment of each phenomena individually, finally yielding an assessment matrix. Due to length constraints, this article only presents the detailed description of four key phenomena. However, the full PIRT table together with the assessment for each phenomenon is provided. The overall findings indicate that LOCUST 1.2 demonstrates a good capability to accurately reproduce most of the phenomenology observed in all four tests. • Validation of LOCUST 1.2 thermal-hydraulic code. • Use of the ANT Assessment Methodology for phenomenological assessment of LSTF tests. • Use of BEPU to determine the uncertainty of LOCUST 1.2 thermal-hydraulic code. [ABSTRACT FROM AUTHOR]

Present study provides a simple analytical formula, the “Klingel-like formula” or “Pascal's Formula” that can be used as a reference to test some results of existing railway codes and specifically those using rigid contact. It develops properly the 3D Newton-Euler equations governing the 6 degrees of freedom (DoF) of unsuspended loaded wheelsets in case of zero wheel-rail friction and constant conicity. Thus, by solving numerically these equations, we got pendulum like harmonic oscillations of which the calculated angular frequency is used for assessing the accuracy of the proposed formula so that it can in turn be used as a fast practical target for testing multi-body system (MBS) railway codes. Due to the harmonic property of these pendulum-like oscillations, the square ω2 of their angular frequency can be made in the form of a ratio K/M where K depends on the wheelset geometry and load and M on its inertia. Information on K and M are useful to understand wheelsets behavior. The analytical formula is derived from the first order writing of full trigonometric Newton-Euler equations by setting zero elastic wheel-rail penetration and by assuming small displacements. Full trigonometric equations are numerically solved to assess that the formula provides ω2 inside a 1% accuracy for usual wheelsets dimensions. By decreasing the conicity down to 1 × 10−4 rad, the relative formula accuracy is under 3 × 10−5. In order to test the formula reliability for rigid contact formulations, the stiffness of elastic contacts can be increased up to practical rigidity (Hertz stiffness × 1000).

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user's assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt. [ABSTRACT FROM AUTHOR]

The experimental investigation of a prototypical set-up simulating a loss of flow accident in a helium-cooled breeding blanket first wall mock-up under typical heat load conditions is presented. The experimental campaign reproduces the expected DEMO thermal-hydraulics conditions during normal and off-normal situations and aims at providing some insight into the fast transients associated with the loss of flow in the blanket first wall. The experimental set-up and the definition of the experimental matrix are discussed, including pre-test analysis performed in support of these activities. The major experimental results are discussed, and a procedure of using the acquired data for validating and calibrating the RELAP-3D model of the mock-up is introduced. All these activities contributed to the creation of a relevant theoretical and practical experience that can be used in further studies concerning incidental transients in real-plant scenarios in the framework of DEMO plant fusion safety activities. [ABSTRACT FROM AUTHOR]

Based on the requirement of Hexagonal Casing Type Fuel Reactor (HCTFR) nuclear design and the critical physical experiment design method introduced by a single factor, 11 core critical physical experiments are proposed to validate the calculation accuracy and reliability of the nuclear design code CPLEV2. The experiment loading scheme fully takes into account the various components and more than one irradiate hole in the HCTFR core, which is used as critical physical experiment schemes successfully. According to the critical physical experiment data, the reactivity calculation deviations of all critical physical experiments are within ±1.0%. The validation results show that the nuclear design code CPLEV2 has high calculation accuracy and reliability for the core of hexagonal casing type fuel, and it can be used for HCTFR nuclear design.

Gamma spectroscopy measurements can be simulated using a pulse-height tally functionality of Monte Carlo particle transport codes. Such a functionality must account for the complete simulation history of a particle's energy deposited in a particular volume. OpenMC, an open-source application for neutron and photon particle transport which is widely used in the nuclear engineering community, previously lacked adequate simulation capabilities for gamma spectroscopy. In this paper, we outline the implementation of the photon pulse-height tally for OpenMC. Additionally, we validate the new function by comparing results to analytical calculations, other software, and experimental data. [ABSTRACT FROM AUTHOR]

Passive safety systems mostly working on the principles of natural circulation are given much priority in increasing the reliability of reactors’ inherent safety features. In view of this, many researchers carry out numerical and experimental studies in single-loop natural circulation model to study the natural circulation behaviors. In the present study, the experimental loop consists of three loops; two of which operate on natural circulation while the third on forced circulation. The experimental loop has successfully established natural circulation under different pressure conditions and the experimental results obtained are compared with those of the RELAP5 code simulation for validating the multi-loop natural circulation RELAP5 model. Single-phase natural circulation flow rates and temperatures at various locations give reasonable agreements between the code and the experiment. The results show that height difference between the heat source and the heat sink is more influential than system pressure in increasing natural circulation flow rate. The system pressure maintains the stability of the system, thus keeping it within single-phase region with increase in heat. In addition, temperature difference in the first loop being higher than that of the second loop at all pressures, did not influence the first loop flow rate to be higher due to the advantage of height difference in the second loop. However, extending the conditions beyond those of the experiment, that is higher power and system pressure, the first loop flow rate increased more than that of the second loop as predicted by the RELAP5 code.

Tungsten sputtering rates and density profiles predicted using the edge plasma codes EDGE2D-EIRENE and DIVIMP are found to agree within a factor of 4 with measurements of neutral and singly-ionized W spectral line emission in the JET low-field side (LFS) divertor, and within a factor of 2 with SXR, VUV, and bolometric calculations of the W density in the main plasma. The edge plasma W predictions are extended to the core plasma using JINTRAC integrated core-edge modelling. Prompt redeposition of W is identified as the primary reason for the discrepancy between predicted and measured W emission in the divertor. The studied plasmas include attached divertor conditions in L-mode and type-I ELMy H-mode plasmas typical for JET.To more accurately reproduce the spectroscopically inferred W sputtering rates in EDGE2D-EIRENE, imposing the experimentally observed Be concentration of order 0.5% in the divertor is necessary. However, the W density in the main plasma is predicted to be insensitive to whether or not W is sputtered by Be at the divertor targets. Instead, the majority of the predicted core W originated in L-mode from sputtering due to fast D charge-exchange atoms at the W-coated tiles above the LFS divertor, and in H-mode due to D and W ions at the targets during ELMs.

The average tungsten concentrations in the pedestal region (cW) predicted by the Monte Carlo code DIVIMP and the coupled multi-fluid plasma/kinetic neutral code EDGE2D-EIRENE are found to agree within a factor of 2 for a range of JET-ILW L-mode and H-mode plasma conditions. Under attached divertor conditions with cW exceeding 10−6, the cW predicted by DIVIMP is consistently ~50% higher than by EDGE2D-EIRENE. In colder plasma scenarios with cW

Abstract Asteroids and comets have the potential to impact Earth and cause damage at the local to global scale. Deflection or disruption of a potentially hazardous object could prevent future Earth impacts, but due to our limited ability to perform experiments directly on asteroids, our understanding of the process relies upon large‐scale hydrodynamic simulations. Related simulations must be vetted through code validation by benchmarking against relevant laboratory‐scale, hypervelocity‐impact experiments. To this end, we compare simulation results from Spheral, an adaptive smoothed particle hydrodynamics code, to the fragment‐mass and velocity data from the 1991 two‐stage light gas‐gun impact experiment on a basalt sphere target, conducted at Kyoto University by Nakamura and Fujiwara. We find that the simulations are sensitive to the selected strain models, strength models, and material parameters. We find that, by using appropriate choices for these models in conjunction with well‐constrained material parameters, the simulations closely resemble with the experimental results. Numerical codes implementing these model and parameter selections may provide new insight into the formation of asteroid families (Michel et al., 2015, https://doi.org/10.2458/azu_uapress_9780816532131‐ch018) and predictions of deflection for the Double Asteroid Redirection mission (Stickle et al., 2017, https://doi.org/10.1016/j.proeng.2017.09.763).

Description of computation model for Kyoto University Critical Assembly (KUCA) developed with the help of Monte-Carlo Serpent code was presented in this paper. The simulation of criticality and subcriticality condi-tions for KUCA was carried out. The effective multiplication factors for different critical experiments were calcu-lated for KUCA. The presented obtained results were considered and compared with the experimental data and with computation results from other Monte Carlo codes.

Asteroids and comets have the potential to impact Earth and cause damage at the local to global scale. Deflection or disruption of a potentially hazardous object could prevent future Earth impacts, but due to our limited ability to perform experiments directly on asteroids, our understanding of the process relies upon large‐scale hydrodynamic simulations. Related simulations must be vetted through code validation by benchmarking against relevant laboratory‐scale, hypervelocity‐impact experiments. To this end, we compare simulation results from Spheral, an adaptive smoothed particle hydrodynamics code, to the fragment‐mass and velocity data from the 1991 two‐stage light gas‐gun impact experiment on a basalt sphere target, conducted at Kyoto University by Nakamura and Fujiwara. We find that the simulations are sensitive to the selected strain models, strength models, and material parameters. We find that, by using appropriate choices for these models in conjunction with well‐constrained material parameters, the simulations closely resemble with the experimental results. Numerical codes implementing these model and parameter selections may provide new insight into the formation of asteroid families (Michel et al., 2015, https://doi.org/10.2458/azu%5fuapress%5f9780816532131‐ch018) and predictions of deflection for the Double Asteroid Redirection mission (Stickle et al., 2017, https://doi.org/10.1016/j.proeng.2017.09.763). Plain Language Summary: Asteroid and comet impacts into Earth are a low‐probability but high‐consequence risk. Given that the risk exists, we prepare ahead of time by researching ways to stop a potentially hazardous object from hitting our planet. Conducting experiments in space on actual asteroids or comets to practice mitigation tactics is possible but limited. In the meantime, the planetary defense community uses codes to simulate different ways of stopping these potentially dangerous objects. But this begs the question, how do we know our codes are correct? In an effort to gain confidence in our codes, this work compares our simulation results to data from a well‐known laboratory‐scale experiment to assess the accuracy of our models. We find that our code can produce results that closely resemble the experimental findings, giving assurance to the planetary defense community that our code can correctly simulate asteroid or comet mitigation. Key Points: We investigated the accuracy of our code by comparing our simulation results to data from a 1991 hypervelocity experimentThe simulation results indicate that our code can produce results that closely resemble the experimental findingsThis work provides insight into model and material parameter selections in our code, potentially applicable in other numerical models [ABSTRACT FROM AUTHOR]

The experimental investigation of a prototypical set-up simulating a loss of flow accident in a helium-cooled breeding blanket first wall mock-up under typical heat load conditions is presented. The experimental campaign reproduces the expected DEMO thermal-hydraulics conditions during normal and off-normal situations and aims at providing some insight into the fast transients associated with the loss of flow in the blanket first wall. The experimental set-up and the definition of the experimental matrix are discussed, including pre-test analysis performed in support of these activities. The major experimental results are discussed, and a procedure of using the acquired data for validating and calibrating the RELAP-3D model of the mock-up is introduced. All these activities contributed to the creation of a relevant theoretical and practical experience that can be used in further studies concerning incidental transients in real-plant scenarios in the framework of DEMO plant fusion safety activities.

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt.

The input uncertainties propagation methods are the most frequently applied statistical methods in uncertainty analyses. Among them, particularly popular are the methods based on Wilks' formula. Numerous studies on uncertainty analyses show that the identification and quantification of input uncertainties is a major problem with uncertainty analyses. Among input uncertainties evaluation, the identification and quantification of physical model uncertainties in thermal-hydraulic codes appear to be particularly difficult. This paper deals with this problem by proposing inherent model uncertainties quantification by code developers in the frame of code development and validation. The introduction of the extended code validation would not only contribute to potential uncertainty analyses, solving to a large degree the problem of model uncertainties quantification, but also contribute to code validation, and as a consequence, improve the safety issues. A not-negligible factor is also better management of the resources. Instead of uncertainty quantification repeatedly performed by each user, the quantification could be performed once and, in addition, by experts having the required know-how. Introducing this new standard in code validation would require additional effort from the code developers but integral quantification of the model uncertainties would be profitable also for code development. In fact, by code development, in particular if the model is own development of the team, such an accuracy (or uncertainty) evaluation is usually performed. The additional effort, in this case, would be to describe the present information in the form of probability distribution functions or at least in the form of ranges. [ABSTRACT FROM AUTHOR]

• Three-dimensional approach for simulating BWR core melt progression. • Validation with CORA-BWR experiment. • Development of a detailed code for BWR core degradation. A three-dimensional simulation code has been developed for a Boiling Water Reactor (BWR) to perform two missions: realistic simulation of the phenomena induced in in-vessel core melt progression and evaluation of the composition and condition of molten materials/debris. The current stage of the code development focuses on simulation of the core melt progression from increase in cladding temperature to molten materials relocation, in order to decrease the computational uncertainties. To simulate the core melt progression more realistically, the code has several features: (i) a multi-phase, multi-component and multi-velocity-field, (ii) a three-dimensional geometrical configuration of complicated internal components in the core and lower plenum zones in a reactor pressure vessel of BWR, and detailed modelling of (iii) multiple core materials melt and chemical interactions, (iv) a candling phenomenon and (v) melt blockage in narrow zones. Subsequently to the previous qualitative validation of these physical models, more detailed validation was conducted in the present study for the CORA-BWR experimental series. In the validation, the code has semi-quantitatively simulated the temperature escalation and maximum temperature, the melt progression, the post-test configuration of relocated materials and the hydrogen generation of CORA-16 as the reference case first. Furthermore, the CORA-18, CORA-28 and CORA-33 have been evaluated in terms of the effects of scale (18 versus 48 fuel-rod mockup), pre-oxidation and wet/dry core condition. Additionally, through the CORA-16 and -18 validation, the code showed the capability for predicting the post-test configuration of relocated materials visually by comparing the photos. [ABSTRACT FROM AUTHOR]

The aim of this study is to validate the reactor physics lattice transport code DRAGON5 and the Monte Carlo code OpenMC by neutronic analysis of critical experiments of light water cores. In this work the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-1, BAPL-2, and BAPL-3 is achieved based on evaluated nuclear data library ENDF/B-VII.1. BAPL and TRX experiments provide experimental buckling and are suitable benchmark lattices for validating the deterministic reactor physics lattice transport code DRAGON5 and the stochastic OpenMC code as well as evaluating nuclear data library. The integral parameters of the abovementioned critical experiments are calculated using DRAGON5 and OpenMC codes. To assess our calculation scheme the calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0. Our calculations led to results in good agreement with the experiment and the earlier MCNP calculation results. Therefore, this study reveals the potential validation of the reactor physics lattice transport code DRAGON5 and the Monte Carlo code OpenMC using ENDF/B-VII.1 nuclear data library. [ABSTRACT FROM AUTHOR]

• Eleven major integral thermal-hydraulic test facilities are summarized in detail. • Different scaling methodologies are compared and evaluated. • The scaling distortion and evaluation applicability of integral test facilities are analyzed. The integral thermal-hydraulic test facility is an important part of the design certification and safety review of nuclear power plants. During the past few decades, a considerable number of resources and efforts have been devoted to establishing integral test facilities and carrying out experimental programs throughout the world. These test facilities provide a large database for computer code assessments and a better understanding of thermal-hydraulic phenomena for postulated accident transients. In this paper, some main integral thermal-hydraulic test facilities are summarized from the perspective of scaling design and applicability evaluation. Scaling analysis methods are presented for the rational design of scaled-down test facilities. The main characteristics of these thermal-hydraulic test facilities are illustrated, including their construction history, scaling design, test matrix, verified codes and related research works. The scaling distortion and experimental applicability of different integral test facilities are compared and evaluated. This review is expected to provide the reference guidance for the rational design and application evaluation of integral thermal-hydraulic test facilities. [ABSTRACT FROM AUTHOR]

• Experimental study of lithium-lead/water interaction in WCLL BB "in-box LOCA". • Set-up and description of LIFUS5/Mod3 experimental facility. • Identification of experimental test matrix for SIMMER-III code validation. The new experimental facility LIFUS5/Mod3 has been designed, manufactured and installed to investigate the phenomena connected with the thermodynamic and chemical interaction between lithium-lead and water in case of in-box LOCA (Loss of Coolant Accident) of the WCLL breeding blanket concept and to validate the chemical model implemented in SIMMER code for fusion application. In order to fulfill these objectives, the necessary step is to obtain data, suitable to code validation, by means of an experimental campaign in LIFUS5/Mod3 facility, executed with controlled initial and boundary conditions. Thus, specific instrumentation and dedicated data acquisition system are installed on the facility to provide meaningful and reliable data. The final aim of the LIFUS5/Mod3 campaign is the SIMMER code validation, applying the standard methodology to post-test analyses. Besides, the expected outcomes of the tests are the improvement of the knowledge of physical behavior and of understanding of the phenomena, the investigation of the dynamic effects of energy release towards the structures and of the chemical reaction with the consequent hydrogen production, and the enlargement of the database for code validation. [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 thermal hydraulic transient analysis code for OFNPs is developed. • Ocean condition models are established by considering the effects of ship motions. • The developed code is verified by experimental data under rolling motion. • Effects of the coupled ship motion on natural circulation system are studied. Abstract Offshore floating nuclear power plants (OFNPs) can effectively solve the offshore energy supply problem in marine resource development and island construction. Affected by ocean waves and other ocean conditions, the OFNPs can generate different kinds of ship motions, which can oscillate the thermal hydraulic parameters and threaten the reactor safety. In the present study, ocean condition theoretical models are established by considering the effects of three basic movement forms (static inclining, linear motion and rotation motion) as well as the coupled ship motions. A thermal hydraulic transient analysis code for OFNPs is developed by adding ocean condition theoretical models into the RELAP5/SCDAPSIM/MOD3.4 code. The experimental data obtained by zero power loading experiment and single-phase natural circulation experiment under rolling motion are used to verify the ocean condition theoretical models as well as the code modification strategy. Results show that the flow fluctuation behaviors caused by rolling motion can be well simulated by the developed code. The calculation capability of modified RELAP5 code under static inclining and heaving motion is also verified by comparing with RETRAN-02/GRAV code. Besides, the effects of the coupled ship motions on natural circulation system are studied by the modified RELAP5 code. Compared with the basic movement forms, the coupled ship motions can cause greater flow fluctuation and obviously reduce the core flow rate, which means the influence of the coupled ship motions is necessary to be considered in the safety analysis of OFNPs. [ABSTRACT FROM AUTHOR]

In this paper neutron transport simulation results of experimental configurations used during the 2019 IPNDV measurement campaign at the Belgian Nuclear Research Centre SCK CEN in Mol are presented. As these are a good basis for benchmarking simulations, results of the Monte-Carlo simulation codes Geant4, SCALE/KENO-VI, MCNP and openMC are compared with the aim of validating Geant4. With fissile material present as plutonium–uranium mixed oxide, (α ,n) emissions are included in the simulations. A Geant4 extension to calculate neutron multiplication factors k eff was developed and applied to the IPNDV configurations and a set of OECD/NEA benchmark experiments. Neutron fluxes of the IPNDV configurations through a reference volume are simulated. The Geant4 (α ,n) simulation toolkit SaG4n is compared to the SCALE code ORIGEN and small deviations are observed. Geant4 calculations of k eff for criticality benchmark experiments deviate in the mean by ＋0.4%, for MCNP and openMC deviations are ≤ 6%, for KENO-VI ≤ 8%. Geant4's total neutron fluxes through a reference volume next to the IPNDV configurations agree within a margin of ± 5% with MCNP and openMC and −14 % with KENO-VI. Considerable differences between all codes were observed for thermal neutron scattering but Geant4 results are still within the other codes' variability. Differences to Geant4 for epithermal energies can be addressed to a not yet included "Doppler Broadening Rejection Correction". By comparing Geant4 simulations of a complex set-up with various reference codes and demonstrating its good performance as well as some discrepancies, this study contributes to the validation of Geant4 neutron physics in fissile material systems and for nuclear disarmament verification simulations. • Validation of Geant4 with three other Monte Carlo codes: MCNP, KENO-VI, openMC. • Development of Geant4 extension to calculate neutron multiplication factor. • Validation of benchmark experiments and nuclear disarmament verification set-up. • Capability demonstration of Geant4 simulating neutronics of plutonium configurations. [ABSTRACT FROM AUTHOR]