[en] Highlights: • A benchmark dataset is collected for steady-state single-phase natural circulation. • Experimental conditions are simulated with SAS4A/SASSYS-1. • Solution verification, uncertainty quantification and sensitivity study are covered. • Validation of SAS4A/SASSYS-1 proves its satisfactory prediction capability. The validation of system analysis codes for nuclear reactor systems is required for the development and application of these computational tools. Designed as a comprehensive system analysis code for advanced nuclear reactors, SAS4A/SASSYS-1 requires validation of its physics model for capturing single-phase natural circulation behavior. To support the validation of SAS4A/SASSYS-1, high-precision experiments are performed capturing steady-state single-phase natural circulation on a scaled facility with comprehensive instrumentation. Dedicated tests are performed quantifying the critical modeling parameters, and a single-phase natural circulation benchmark dataset is obtained with well-documented uncertainty and comprehensive facility description. The validation is then performed against the dataset examining the capability of SAS4A/SASSYS-1 in simulating steady-state single-phase natural circulation. The experimental facility is modeled in the candidate code. Solution verification is performed using Richardson-extrapolation-based estimators which quantify and restrict numerical errors from discretization. Input uncertainty provided by the benchmark dataset is forward propagated through the candidate code, quantifying the output uncertainty in a Monte Carlo approach. The composition of the output uncertainty is also quantified through a variance-based sensitivity analysis. With the uncertainty quantified for each individual condition, a detailed comparison between the simulation results and experimental data is performed covering the whole dataset. The results show consistent agreement for all primary parameters. The current validation activity provides a valuable benchmark dataset for the validation of system analysis codes in capturing single-phase natural circulation and demonstrates satisfactory prediction capability of SAS4A/SASSYS-1 for steady-state single-phase natural circulation.

[en] Highlights: • A 1/3 concrete containment building mock-up was built by EDF in France. • An international simulation benchmark has been organised. • The mock-up behaviour (temperature, humidity, strain and leak-tightness evolution) has been predicted by participants. Electricité de France (EDF) operates a large fleet of nuclear reactors and is responsible for demonstrating the safety of facilities, including concrete containment buildings (CCB), which are non-replaceable components. The leak-tightness of CCBs is assessed every 10 years during integrated leak-rate tests (IRLT). For double-wall containments, which have no metallic liners, the leak-tightness is strongly influenced by the degree of cracking of concrete and opening of the cracks, which mostly depends on (a) the prestress decrease due to the delayed strains of concrete and to a lesser extent due to relaxation of tendons steel, and (b) the saturation degree of the Powered by Editorial Manager® and ProduXion Manager® from Aries Systems Corporation concrete wall. Therefore, to optimize the maintenance programs on CCBs, it is important to predict the evolution of drying, creep and shrinkage strains of concrete to be able to correctly assess the pre-stress losses, and finally the air leak-tightness at a structural level during pressure tests or under accidental loadings. To improve our understanding and identify the best modelling practices on this issue, a large experimental program called VERCORS was launched in 2014. VERCORS is a 1/3 mock-up of a 1300 MWe nuclear reactor CCB. It has been widely instrumented, and its concrete thoroughly characterized. A specific attention has been paid to ensure it is consistent with real CBBs features in EDF's nuclear fleet. To complement its internal R&D efforts, EDF decided to associate external partners to this program. One of the means for this is the organization of benchmarks, where all teams are given data and information about the mock-up and are asked to quantitatively predict its behaviour. The present paper reports the organization and findings of the 2nd benchmark which was organized in 2018 and gathered several international teams around the same objective: improve the confidence in the modelling of structural behaviour as well as the leak-tightness of concrete in containment walls under pressure test loading. The benchmark has shown once again that predicting the mechanical and leakage behaviour of containment buildings is a difficult task. The benchmark also yielded interesting information about the possibility to use spatially reduced models to predict the mechanical behaviour and leakage and underlined the fact that more research must be done to better predict the localization of cracks and leakage. Some lessons have been learnt for the next benchmark: EDF will ask to clarify further the calibration methods, will give more data (including drying, creep and shrinkage at different temperatures and moisture measurements in the mock-up), and will help the participants using local leakage data by projecting the raw measurements on a regular grid, so that the local leakage models can be improved.

[en] Highlights: • Materials, irradiation parameters and research methods. • Fuel pins with nitride fuel irradiated in the BN-600 reactor. • Fuel pins with nitride fuel irradiated in BOR-60. Today the investigations have been completed on helium-bonded fuel pins with mixed uranium-plutonium nitride of BN-600/BN-800, BN-1200 and BREST reactors types after irradiation as a part of ten EFAs of BN-600 reactor up to maximum burn-up of 7.5; 6.0 and 4.5at%, respectively. Also the PIE of mixed nitride pins of BREST type with helium and lead sub-layers after intermediate tests to maximum burn-up of 4,8 and 3,9at% as part of dismountable EFAs of BOR-60 reactor are over. As a result of postirradiation examinations, the main intra-fuel pin processes that affect on materials properties change and fuel pins state, their relationship with the initial nitride state were identified. The regularities and quantitative characteristics of nitride swelling, the behavior of fission products, cladding corrosion and mechanical properties change are revealed. The average for the irradiation time rate of fuel swelling in cross sections near the core midplane in gas-bonded pins of different EFAs irradiated in BN-600 reactor is equal to (1.6 - 2.0) %/at%. According to the results of irradiation in BOR-60 reactor, the fuel swelling rate in lead-bonded pins is lower than in helium-bonded pins: 1.4 ± 0.2 and 1.7 ± 0.2 %/at%, respectively. The presence of carbon and oxygen impurities in the fuel can lead to local areas of claddings carburization and oxidation, randomly distributed along the height and perimeter of its inner surface. The key characteristics of the fuel that determine the cladding carburization and oxidation, and ways to prevent them, are determined. Cladding corrosion in lead-bonded pin is caused by selective dissolution of cladding steel components. The maximum depth of the corrosion zone is located in the upper part of the fuel column and is equal to 40 microns. The results of cladding mechanical tests obtained using longitudinal segment samples and tubular samples loaded by internal gas pressure showed the significant margin of strength and ductility of cladding materials along the entire height of fuel pins both at nominal operating temperatures and at room temperature. High values of strength characteristics indicate a weak influence of corrosion on the strength of studied claddings materials. The mechanical cladding properties of a lead – bonded pins have the same pattern of change along the pin height depending on irradiation and testing temperatures, as of fuel pins with helium sublayer, keeping enough ductility in the area of low-temperature irradiation embrittlement. So, for example, at irradiation and test temperatures of 350-380 °C, the values of the strength limit from 1070 to 1200 MPa were obtained with the values of the total elongation of not less than 2 %. Thus, the results of post-irradiation examinations confirmed the performance of mixed nitride pins at the achieved test parameters and the possibility of their irradiation prolongation to higher fuel burn-up.

[en] Highlights: • A numerical simulation method for liquid metal HCOTSG is proposed. • The wall heat flux and comprehensive performance are compared. • The radial pitch has a significant effect on the wall heat flux. • The HCOTSG has the best performance when P_r /D is 0.167 and P_a /D is 0.137. Lead-bismuth Fast Reactor (LFR) has attracted attentions due to its intrinsic characteristics, such as sustainability, safety, and economics. Helical coil once-through tube steam generator (HCOTSG) is a proposed form of steam generator. Due to its unique advantages, HCOTSG is widely used in various reactor power systems, including LMFR. In this paper, FLUENT CFD commercial code is adopted to simulate heat transfer from the primary side (liquid metal) and secondary side (water/steam). The robustness of the method is validated by comparing it with relevant experimental research, and the maximum error is less than 25%. Based on this method, 9 models with different geometrical parameters were established to analyze the effects of geometrical parameters on the performance of liquid metal HCOTSG. The wall heat flux and comprehensive performance factor (En) of different models is compared. The optimal geometric arrangement scheme for the working conditions is recommended. The geometrical parameters of the tube side and shell side are analyzed as sensitivity parameters. Among the established geometric models, the HCOTSG has the best comprehensive performance when P_r /D is 0.167 and P_a /D is 0.137 under the working conditions in this paper. This study provides a numerical simulation method on structural design optimization of liquid metal HCOTSG.

[en] Highlights: • A framework is proposed to integrate inverse uncertainty quantification and quantitative validation. • The framework can improve model predictions while accounting for all sources of uncertainties. • Bayesian hypothesis testing is used to calculate a quantitative validation metric called the Bayes factor. • This framework is an initial step towards addressing the ANS Nuclear Grand Challenge on “Simulation/Experimentation”. The Best Estimate plus Uncertainty (BEPU) approach for nuclear systems modeling and simulation requires that the prediction uncertainty must be quantified in order to prove that the investigated design stays within acceptance criteria. A rigorous Uncertainty Quantification (UQ) process should simultaneously consider multiple sources of quantifiable uncertainties: (1) parameter uncertainty due to randomness or lack of knowledge; (2) experimental uncertainty due to measurement noise; (3) model uncertainty caused by missing/incomplete physics and numerical approximation errors, and (4) code uncertainty when surrogate models are used. In this paper, we propose a comprehensive framework to integrate results from inverse UQ and quantitative validation to provide robust predictions so that all these sources of uncertainties can be taken into consideration. Inverse UQ quantifies the parameter uncertainties based on experimental data while taking into account uncertainties from model, code and measurement. In the validation step, we use a quantitative validation metric based on Bayesian hypothesis testing. The resulting metric, called the Bayes factor, is then used to form weighting factors to combine the prior and posterior knowledge of the parameter uncertainties in a Bayesian model averaging process. In this way, model predictions will be able to integrate the results from inverse UQ and validation to account for all available sources of uncertainties. This framework is a step towards addressing the ANS Nuclear Grand Challenge on “Simulation/Experimentation” by bridging the gap between models and data.

[en] Highlights: • Experiments on coolant mixing during boron dilution scenarios at the ROCOM test facility. • Boundary conditions from the integral test facility PKL. • Validation of ANSYS against boron dilution scenarios with the presence of density differences. This paper compares CFD simulation results of a boron dilution scenario with experimental results. The simulation was carried out with ANSYS CFX using the SST turbulence scheme, experimental data were produced in the ROCOM test facility, experiment E2.3. The main features of the scenario are asymmetric, transient mass flow conditions in the affected loops 1 and 2 of a KONVOI-type reactor vessel and reduced density of the underborated coolant slugs fed into the reactor trough the cold legs of both loops. The CFD simulation was able to capture the density stratification in the loops affected by underboration and the mixing in the downcomer and in the lower plenum. Good agreement with the experiment was obtained for the temporal evolution of average boron concentration in several measuring sections of the reactor vessel and of the boron distribution in the core inlet. In general, minimum values of boron concentration were found to be lower in the simulation than in the experiment.

[en] Highlights: • Proposed a TRU started-ZrH-MSR core design for TRU burning. • Investigated the core critical and burnup performances with consideration of TRU solubility limit for different fuel salts. • Optimized the TRU burning capacity of the ZrH-MSR under different neutron spectrums. • Evaluated the temperature feedback coefficient in the optimized ZrH-MSR during operation. Inventory reduction of transuranium (TRU) nuclides can effectively reduce the long-lived and high-level radioactive hazards in the spent nuclear fuel (SNF) from the current light water reactors (LWRs), which is of importance to the long-term development and deployment of nuclear energy. Considering its outstanding features such as excellent neutron spectrum, high fuel utilization and compact core structure, a molten salt reactor moderated by zirconium hydride rods (ZrH-MSR) is considered as one of the potential reactor types for TRU burning. In this work, several fuel salts and molten salt volume fractions (SVFs) are analyzed to assess the TRU burning capability in a ZrH-MSR. The fuel salt of 77.5 mol% LiF-22.5 mol% (ThF₄ + TRUF₃) is selected since it can meet the requirement of TRU solubility limit during 50 years operation and can achieve a high initial TRU loaded mass in the core to improve TRU burning capacity. The optimized SVF is 0.5, and the TRU burning rate and the Support factor can achieve about 28.76 kg/TW_thh and 2.9, respectively. The TRU burning mass, TRU fuel utilization ratio and ²³³U production after 50 years operation are about 12.6 tons, 70.4 %, and 0.7 tons, respectively. The total TRU radiotoxicity after discharge is about 63.9 % smaller than that without burning. This optimized core is beneficial to get a high TRU burning rate to reduce the storage of TRU and to minimize the long-term radiotoxicity risk in a relatively short reactor life. The total temperature feedback coefficient can remain negative during the whole operation time, which ensures the safety of the ZrH-MSR.

[en] This paper introduces a modified internal energy equation derived for multiphase flow in various flow conditions for a staggered mesh system. The pressure drop and heat dissipation terms in the internal energy conservation equation currently being developed based on the assumption that the scalar and momentum variables, which originate from the total energy and mechanical energy equations, respectively, are all located in the cell center of the control volume were redefined such that two different pressures and velocities stored in both cell centers and faces were imposed for the internal energy conservation. To achieve this, first, a modified internal energy conservation equation for a staggered mesh was derived by subtracting the mechanical energy equation from the total energy equation. The equation was then discretized classifying each term by its origin; variables that originated from the total energy equation were defined in the cell center, whereas terms that came from the mechanical energy equation were identified at faces. Since the discretized form of the proposed equation contained face velocity and cell pressure for the heat dissipation and pressure drop terms, respectively, these two terms were calculated implicitly leading to enhanced numerical stability. The accuracy of the modified internal energy equation in predicting the system pressure, fluid temperature, and heat balance for various flow channels was assessed. The verification of the proposed equation was completed through simulations of multiple theoretical problems including saturated liquid depressurization, adiabatic expansion of hydrogen, heat transfer in a helical steam generator, and energy transfer in a converging pipe. An improved result was obtained with the modified equation as the numerical calculation results agreed well with the analytic solutions with relative deviations less than 0.1% for most cases, while the solution obtained by the conventional internal energy equation showed a significant amount of deviation from the analytic solution.

[en] Highlights: • Quantitative thermal imaging for Intermediate Level Waste container “fingerprinting”. • Determination of Intermediate Level Waste container internal temperatures. • Reliable determination of Special Nuclear Material containers surface temperature. • Non-radiance surface thermometry for the long term monitoring of SNM containers. • Remote identification of anomalous temperature 3 m³ ILW packages in stores. Many established nuclear power producing countries are currently decommissioning first and increasingly second-generation power producing plants and fuel processing facilities. This has led to a growing inventory of different containers and packages containing radioactive waste and other nuclear materials, as well as storage of spent fuel. Here we describe research to establish in-situ yet remote health monitoring techniques based on novel temperature measurement methods for different containers and racks used to hold different nuclear waste forms, special nuclear materials and spent fuel.

[en] Highlights: • Limited experimental evidence affects prediction accuracy of severe accident progression. • Modeling consistency and proper understanding of uncertainties are critical issues. • Main paper objectives include the following issues: • Importance of identifying dominant core meltdown phenomena, • Consistency between models and inherent randomness of underlying physics/chemistry. The complexity of phenomena occurring during severe accidents in nuclear reactors, combined with a limited amount of experimental evidence, do not allow for formulating detailed reliable models or making highly accurate predictions of accident progression. Thus, the modeling consistency and a proper understanding of the uncertainties associated with the results of any computer simulations, including those caused by the imperfection and inherent limitations of the available experimental data used in model validation, are critical for improving accident mitigation capabilities and for enhancing the safety of current and future generations of nuclear reactors. The objective of this paper is to give an overview of selected issues illustrating the importance of: (a) identifying the dominant phenomena governing the progression of core meltdown accidents, and (b) formulating models which are consistent with our understanding of the underlying physics and chemistry and of the increasing level of randomness as the accident progresses. The results used as examples, in particular those pertaining to the research performed by the authors and their collaborators, have been obtained over past several years and documented in a several reports (in particular, in the USNRC NUREG series), but have never been included in copy-righted publications. The sources of any other experimental data used in the discussion of specific coupled experimental/modeling issues (typically, also reports of various agencies or labs) are clearly identified in the text. Since some of the examples include recent unpublished results of computer simulations performed using updated versions of the models which have already been published in other journals, only brief information about such models is presently shown, assuming that the reader can find details in the corresponding references.