Conference Agenda

In this study, a mass and energy (M/E) release analysis was performed on the postulated main steam line break (MSLB) accidents in the Advanced Power Reactor 1000 (APR1000). The M/E release rate was calculated using the SPACE-ME 1.0 code, developed by KEPCO Engineering & Construction Company, Inc. (KEPCO E&C), for various break areas ranging from an area fraction (AF) of 0.1 to 1.0, where AF 1.0 corresponds to the maximum double-ended guillotine break area. The initial core power was evaluated at 102%, 75%, 50%, 20%, and 0% of full power (%FP). To ensure conservative results, the break flow phase separation model and wall heat transfer multiplier were adopted. A simplified conservative model for the passive auxiliary feedwater system was used. The containment pressure and temperature responses were analyzed using the CAP 3.1 code with the calculated M/E release rates. A single failure of containment spray system was assumed. The highest containment peak pressure and temperature were found to be 0.7925 and 0.9531, respectively, which are normalized values with respect to the design values. The design margins of 20.75% for pressure and 4.69% for temperature during the most limiting MSLB accident indicate that the APR1000 containment can maintain its integrity well during the MSLB accidents. In conclusion, the new M/E release analysis methodology using SPACE-ME code is expected to be highly applicable to analyzing the postulated MSLB accidents in the APR1000.

This study presents the experimental results of large-scale containment thermal-hydraulics phenomena driven by the combined effects of steam injection and spray activation in a postulated Design Basis Accident (DBA) scenario, specifically a main steam line break. This experimental campaign, named P1A1_5 and P1A1_6, is part of the OECD/NEA PANDA project series. These experimental data could contribute to the assessment and validation of advanced computational tools for containment analysis.

The experiments were conducted in Vessel 1 of PANDA, a cylindrical confinement with 8 m in height and 4 m in diameter. A compartment representing the steam generator tower model was inside the Vessel 1. Initial conditions involved pressurizing with air and steam at 2.5 bar. There were defined with two different steam superheating (P1A1_5, P1A1_6), and the spray was activated using a single nozzle. The phases of the experiment were as follows: steam injection (phase 1), combined steam and spray injection (phase 2), and spray-only injection (phase 3 for P1A1_6).

Results showed that during containment depressurization, steam remains superheated above the spray nozzle. In contrast, below the spray nozzle, the fluid and saturation temperatures are approximately same value. Thus, the primary effects of steam injection and spray activation are the depressurization of Vessel 1 and the cooling of fluid and gas temperatures. Upon spray activation, the pressure and temperature gradients, especially below the spray nozzle, decrease more sharply over time compared to the phase 2. This is due to the enhanced momentum mixing and droplet behavior, particularly below the spray nozzle.

The NEA Committee on the Safety of Nuclear Installations (CSNI) has long supported international collaborations to enhance confidence in nuclear safety codes and experimental validation. One such initiative is the International Standard Problem (ISP), which began in the early 1970s and continues today.

A new ISP-52 was proposed upon the recommendations of the WGAMA/WGFS report: “Analyses of Design Extension Condition without Significant Fuel Degradation (DEC-A) for Operating Nuclear Power Plants” which highlighted the need for computer code validations for DEC-A conditions. For this purpose, the ETHARINUS project provided experimental data related to the PKL III J5.1 Run1 and Run2 tests. The latter addressed DEC-A scenario of Multiple Steam Generator Tube Rupture (MSGTR), which may occur following a severe earthquake, with limited safety system availability. In Run1 two double-ended guillotine breaks were considered in three out of four steam generators (SGs), while in Run2 the scenario was extended to all four SGs. Both test results were made available to the ISP-52 participants, but only Run 2 was selected for the “blind” and “open” analytical exercises.

This paper presents the main steps that have been followed to carry out the PKL III J5.1 Run1 and Run2 experiments and provide a description of the main events that took place during the course of the transient as well as the effectiveness of the operator actions (primary bleed, and manual activation of the ECCS) and the available ECCS to bring the system to safe shutdown conditions.

The TRACE code is used at Gösgen NPP to perform scoping simulations of plant behavior in response to extreme, very unlikely, external events. This allows plant personnel to investigate how the plant changes and refurbishments under study increase the safety margins in the event of Design Extension Condition (DEC), such as a Station Black-Out (SBO) with failure of the on-site emergency power supply. The safety of the unit is ensured thanks to the special emergency safety systems, bunkered and thus SBO-proven.

This study analyzes the benefits from the refurbished SEFW (Special Emergency Feedwater) tanks. In addition, this study investigates the automatic partial cooldown via Atmospheric Relief Valves (ARVs) (plant change not yet realized). Three simulations are presented in the paper. The full autarky time of 10 hours without operator actions is considered. In the first two cases the unit is kept at hot-shutdown conditions (no cooldown) and the secondary pressure is limited, respectively, by the cyclic opening of the Safety Relief Valves (SRVs) and the ARVs with automatic partial cooldown. The third case implements, based on the second case with ARV partial cooling, the manual cooldown procedure (slow gradient, 10 K/h) to reach cold shutdown.

The results of the simulations show that the automatic partial cooldown reduces the amount of primary coolant released into the containment by opening of the pressurizer safety valves. The increased safety margins of the plant in case of SBO are determined in 53 h (hot-shutdown state) and 42 h (cold-shutdown state) without external water injection.

During a Loss Of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the core. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop, which was equipped successively with two types of 2 m² filters used in 900MWe NPP’s. The debris carried to the filter generate at 80°C (with chemistry) a very quick increase of the pressure drop across the filter (≈ 1 to 7 kPa according to the debris source term) that could be due to rapid chemical effects further to fibers corrosion. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior (debris transferred to the core). The increase of the duration of the corrosion of zinc in acidic conditions (as a sensitivity study) lead to increase head losses by a factor 4 the which indicates the formation of chemical precipitates.

In the first decades of the construction of the nuclear reactors, the attention in the field of nuclear safety analyses was focused on major accidents starting from full power (with maximum initial energy in the system and decay heat). Later on, however, accidents occurring during shutdown were found to be as important as those occurring at full power. Abnormal operational events, postulated accidents and design extension conditions occurring during shutdown operational modes represent a significant contribution to the NPP risk due to the fact, that both preventive and mitigatory capabilities of the plant are partially or fully unavailable. Deactivation of safety features, equipment under maintenance or repair, reduced amount of coolant in some regimes, some instrumentation and measurements switched off or non-functionable; open primary circuit (loss of one barrier); and open containment (loss of another barrier) are the causes of the specific risk of accidents in the shutdown mode.

The core of the paper concentrates on the deterministic thermal-hydraulic (TH) safety analyses of the events starting from the shutdown operating modes of VVER-1000. Number of the performed analyses are long-term analyses specifying time windows for the operator (in situation with reduced availability of safety systems and their automatic actuation). Specification of VVER-1000 shutdown modes accompanied, availability of safety systems, methodology basis for the safety analyses, acceptance criteria, computer codes and their validation, list of scenarios analyzed for the VVER-1000, examples of analyses results, and incorporation of new analyses into Safety Analysis Report (SAR) will be described step by step.

System-scale TH (STH) analysis codes have extensively been used in WCNR safety evaluation along with quantifying the prediction uncertainties in close conjunction with adopting the best estimate (BE) safety analysis. There still exist some deficiencies in the BE safety evaluation, however, originating mainly from our limited knowledge or poor understanding of underlying fundamental physics on key TH phenomena associated with two-phase flow hydrodynamics and heat transfer, which are broadly relevant to WCNR safety concerns.

TH experiments and analyses for WCNR performance analysis and safety evaluation, in general, need to be carefully checked in terms of their scalability to assure whether they are realistically representative of prototypic situations. The basic concern of ‘scalability’ originates from the differences or gap existing between the prototypic and down-scaled systems due to their idealization and/or simplification.

The scalability of experimental data used for validating TH analysis codes will be discussed, focusing on STH codes with their application to WCNR safety evaluation accompanied by the uncertainty quantification.

Discussion will be focused mainly on our unsatisfactory understanding of fundamental physics associated with the constitutive relations adopted in STH codes, many of which were developed based on unrealistic observation under non-prototypic geometric and TH conditions, and partly on the limited numerical capabilities of STH codes in describing multi-dimensional features of dominant phenomena. Then the perspectives of advanced TH safety evaluation are introduced aiming at improving the modelling and simulation (M&S) capabilities.

Results of a Best-Estimate Plus Uncertainty analysis of the LOFT L2-5 experiment performed with millions of cases is presented. The results are used to examine how the techniques traditionally used in analyses are equipped to handle and address likelihoods significantly less probable than at the 95%/95% level. The paper describes the required changes to the underlying probability distribution functions that were required to ensure physical results. The paper presents changes to the model required to achieve sufficient robustness for the process and changes to the typically used uncertainty distributions.
The analyses show that the simulation results do not form a single population of data. The input distributions that lead to the various clouds of data are presented. Physical explanations are provided that link the specific changes in uncertainty parameters to the various outcomes.

Best Estimate Plus Uncertainties (BEPU) approaches are often perceived as complex in licensing processes by licensees. The complexity of the BEPU approach is generally considered justified, as it a priori offers the potential for a more accurate estimation of safety margins. Since it tends to be less conservative than deterministic methods, it raises legitimate questions about its maturity from a regulatory standpoint, particularly given the challenges of nuclear safety assessments.

A relevant example is the case of Loss of Coolant Accidents (LOCA), where BEPU methodology can play a crucial role in assessing fuel behavior (e.g., peak cladding temperature, rupture…). Analyzing these transients involves multiscale (from sub-channel to reactor level) and multiphysics phenomena (multiphase thermohydraulics, fuel and cladding thermomechanics, neutronics, etc.).

Recently, BEPU methodologies have been proposed by licensees in France for the safety demonstration of operating and newly designed reactors. IRSN’s analyses have led to two key questions:

• Can numerical modeling be considered "best estimate" given the diversity of phenomena involved?
• Can we fully rely on the estimation of uncertainties?

This paper highlights some of IRSN’s concerns regarding these aspects, drawing on expert judgments or explicit CATHARE modeling. IRSN believes that BEPU methodologies could play a role in safety demonstration due to their ability to naturally incorporate different combinations of multiscale and multiphysics phenomena. Nevertheless, using BEPU does not exclude some penalties to be required for covering certain limitations.

Since the 1980s, the Best Estimate Plus Uncertainty (BEPU) methodology has been a cornerstone for deterministic safety analysis of design basis accidents in nuclear power plants. Despite endorsements from the International Atomic Energy Agency (IAEA) and various national regulatory bodies, its industrial application remains limited. At the BEPU 2024 conference in Lucca, Italy, a panel of global experts convened to discuss strategies for expanding BEPU’s industrial use. The panel session focused on:

• Enhancing understanding of uncertainty analysis and Input Uncertainty Quantification (IUQ).
• Addressing challenges and exploring future advancements in BEPU methodology.

• Sharing insights and experiences in BEPU application.

• Fostering dialogue on extending BEPU to broader industrial applications.

Key outcomes of the discussions included:

• Advancements in UQ/IUQ methods by NEA, highlighting the need for international cooperation.
• The role of IUQ in boosting confidence in BEPU.
• Evolution of BEPU methodologies for LBLOCA, addressing high burnup fuels and ECCS rule revisions.
• Expansion of BEPU applications from LOCA safety analyses to non-LOCA scenarios, severe accidents, containment integrity, and dose consequences.
• The necessity of a graded approach for broader BEPU application.
• Limited BEPU use in some countries due to regulatory conservatism and high development costs.
• Potential benefits for non-LWR applications, necessitating new experimental data.

This paper summarizes the main contents and outcomes of the panel discussions.

In order to solve the problem that the planned life extension of Qinshan No.3 Nuclear Power Co., Ltd. in China (hereinafter referred to as Qinshan No.3 Nuclear Power Plant) leads to the increase of spent fuel, and the capacity of existing spent fuel dry storage modules is insufficient, based on the original 1~6 (QM-400) spent fuel storage modules, the intensive spent fuel dry storage facilities (M1 and M2 spent fuel storage modules) have been developed. Compared with QM-400 spent fuel storage module, M1 and M2 modules have larger storage capacity and higher energy density. In order to demonstrate the thermal safety of M1 and M2 modules, a thermalhydraulic program is used to establish the thermal analysis model of M1 and M2 modules based on conservative initial assumptions, and calculate the temperature of each region under normal operation and accident analysis of the module under extreme weather conditions. At the same time, the three-dimensional fluid CFD program is used to verify the calculation results of the thermalhydraulic program, and the calculation results of thermalhydraulic program and CFD program are integrated, The thermal safety of M1 and M2 modules is demonstrated. Finally, from the perspective of engineering feasibility and the condition of nuclear power plant site, M1 module is adopted as the implementation plan of intensive spent fuel dry storage facility.

There is currently a growing interest in analysing the behaviour of Advanced Technology Fuels (ATF) under development. Among the new evolutionary ATF designs, the FeCrAl and Cr coated claddings are the most promising. On the other hand, the LBLOCA sequences in Pressurized Water Reactors (PWR) are among the most demanding for safety systems and have a small safety margin. To perform this analysis, NFQ and UPM developed an in-house version of the TRACE5P6 system code for FeCrAl cladding, as TRACE5P6 is not designed to simulate this cladding material. Then, a BEPU analysis of LBLOCA sequences in a PWR was performed for Zry, FeCrAl and Cr-coated Zry cladding and the safety margins were obtained for each case. The results show that the safety margins for ATF materials are greater than those for the Zry case.

The steam system downstream of the main steam isolation valves (MSIV) is normally not structurally verified for the hydraulic loads that can occur following a steam line break (SLB). Also, the turbine trip is normally not classified according to nuclear safety grade standards. Mechanical failure of the steam system, or a failure of the turbine trip system, can therefore not be excluded following a SLB. This could lead to additional steam outflow in addition to the break flow. As a consequence, blowdown of two steam generators could occur if a single failure is assumed on one MSIV. Also, considering extreme external events such as an earthquake or antagonistic actions, the integrity of the turbine building itself, along with the whole steam system outside containment, could be questioned, potentially leading to blowdown of all three steam generators.

In NUREG-0138 it is expected that the assumption of a stuck control rod would compensate for any penalties associated with the blowdown of two steam generators. In the present study, this statement has been investigated using the system thermal hydraulics code TRACE with its built in neutronic point kinetics model. SLBs with blowdown of one, two and three steam generators are analyzed. The effect of the stuck control rod assumption is also studied. An increase in maximum power is observed when blowdown of several steam generators is assumed. However, as expected in NUREG-0138, a large decrease in maximum power is seen if no stuck rod is postulated. Finally, the impact on DNBR is discussed.

Swiss NPPs undergo a Periodic Safety Review (PSR) every ten years. In this framework, the deterministic safety analyses must be updated following the requirements of the regulatory body ENSI. In September 2018 ENSI put a revision of the guideline for technical safety analyses (ENSI-A01) into effect. As a result, new events must be analyzed at DBA level (safety level 3) and BDBA level (safety level 4a). Besides, the fulfilment of the safety goals (acceptance criteria) must be proved also for operating conditions other than full power (such as zero-power or start and shutting-down conditions).

In preparation of the next PSR, the Gösgen NPP, a 3-Loop PWR; is working in tight collaboration with its vendor Framatome GmbH to evaluate and update the existing accident analyses. The present paper reports on the main findings from the new safety analyses, which are being carried out with the proprietary system code S-RELAP5. Attention is given on the update of the plant operational procedures, which are optimized for increasing safety margins and reducing the radiological impact of the accident.

A SBLOCA (break of a measuring line connected to main coolant line or pressurizer) is analyzed implementing a new fast secondary-side cooldown, which allows the stable shutdown without core uncovering (DBA). Operational procedures have been optimized in case of SGTR, preventing the interruption of natural circulation in the affected loop (DBA). New analyses have been performed for ATWS sequences (BDBA). The importance of operator measures is highlighted in the accident mitigation to reach the cold-shutdown state.

Helical coil steam generators play a critical role in the operation of small modular reactors, primarily due to their superior heat transfer capabilities. However, the intricate design of these systems makes it challenging to fully understand the boiling phenomena occurring within the helical coil tube, which is crucial for reactor safety. During boiling, it is common to observe oscillatory flows, which can have a substantial impact on the operating conditions of the reactor and introduce potential safety risks. While thermal-hydraulic system codes have been widely used over the past decades, they often fall short of accurately capturing reverse flows, staggered grid issues, and simplistic spatial discretization. These limitations might result in discrepancies between experimental data and code predictions.

In response to these challenges, a new system code (in-house code for helical coil steam generator) is currently under development, designed to bridge the gap between theory and experimental observation. The goal of this enhanced approach is to provide a more accurate representation of the oscillatory movements induced by two-phase flow within the helical coil tube. By improving the depiction of complex bubble dynamics, this new system code aims to advance the understanding of these oscillations, ultimately contributing to more effective reactor safety analysis and providing a solid foundation for the validation of reactor designs.

The NOUS is a Python™ based software which is capable of setting the initial plant data for the VVER-1200 type reactor for coupled thermo-hydraulic and neutron kinetic (TH-NK) calculation. The program can define the thermo-hydraulic (TH) parameters both for the primary and secondary circuit and can vary the spatial discretization.

The neutron kinetic coupling can be selected between point kinetic, 3D Fuel Assembly (FA) or even pin-wise resolution.

A further option is that the user can choose the availability of the safety and non-safety system trains during an initiating event and the corresponding safety functions.

The software can be used for testing of maneuvering modes, for deterministic safety analysis and for functionality analysis, as well.

Airborne radioactive materials are inevitably released through stacks during nuclear power plant operation. Accurate monitoring is crucial to assess environmental impact and ensure regulatory compliance, as it depends on samples that accurately represent stack radioactivity.

According to the ISO 2889-2023 standard, the study simulated the flow field of a 5:1 scaled model of a stack. The diameter and the height of the scaled model are of 0.6 meters and 12 meters. The gas carrying aerosols in the scaled stack model enters through a horizontal mixing pipeline, with the Reynolds number ranges from 400,000 to 700,000 in the vertical main stack. The Standard k-epsilon Model was employed to calculate the flow field, while the Species Transport Model was used to simulate the mixing processes of tracer gas and air. After achieving convergence, the Discrete Phase Model was employed to compute the trajectories of aerosol particles, thereby obtaining the characteristics of the motion and distribution.

By analyzing the airflow streamlines, aerosol particle trajectories and contour plots at different sampling sections, the study reveals the flow characteristics, the concentration distribution patterns of tracer gases and aerosol particles. The results were further processed to evaluate whether the sampling sections at different elevations meet the well-mixed criteria based on five parameters: average resultant angle, velocity variation coefficient, maximum tracer gas concentration deviation, tracer gas concentration variation coefficient and aerosol particle concentration variation coefficient. The findings of this study serve as a reference for selecting sampling sections and evaluate the mixing performance of the model.

Sodium-cooled fast-neutron reactors are currently considered to be the most mature type of reactor able to closing the fuel cycle. In France and throughout the world, pool-type reactors are selected to build generation IV power plants. Complex liquid sodium flows are known to occur in reactors in several conditions. In order to predict these phenomena, CEA developed the MATHYS code (Multi-scale Advanced Thermal-HYdraulics Simulation). This tools enables the coupling of the system thermal hydraulics code CATHARE, the sub-channel code TrioMC and of the 3D thermal-hydraulics code TrioCFD. Thanks to this coupling approach, the entire primary side of a reactor can be modelled, accounting for the feedbacks for the different scales (core, inter-wrapper flow, pools). In the late 1980s, the CLAUDINA test facility was operated at the CEA Cadarache research centre. The experimental campaigns aimed at the characterisation of the behaviour of an intermediate heat exchanger (IHX) under a variety of operation conditions. The CLAUDINA facility is a mock-up of a sodium –sodium IHX. Tests at different flow conditions were performed. The experimental data from these tests are very valuable for the validation of the intermediate heat exchanger modelling in MATHYS. In this paper, the CLAUDINA facility is first introduced. The CATHARE, TrioCFD and Neptune_CFD codes are then presented, and the models of the CLAUDINA facility are described. The results of these different modelling approaches for several tests are presented and discussed. Conclusions and recommendations are proposed.

The inherent safety features make lead-cooled fast reactors (LFRs) an attractive solution for the increased energy demand and the development of advanced nuclear power plants. Due to the limited operational experience with these reactors, simulation and analysis with system thermal-hydraulic (STH) codes become crucial to study the plant behaviour under safety-relevant conditions and to support the reactor design. In this paper, the LFR modelling has been carried out with different STH codes, such as RELAP5 Mod 3.3 version Beta, modified by University of Pisa to account for lead as working fluid, ASYST-LM and ATHLET codes.

The simulation activity aimed at assessing the code capabilities to reproduce selected phenomena occurring in LFRs under normal and accidental conditions, derived from some of the most representative transient scenarios, such as unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient of overpower (UTOP) and unprotected loss of offsite power (ULOHS+ULOF). Particular attention has been paid to the establishment of natural circulation following the loss of primary pumps, which affects the sizing of the safety features derived from such operating conditions.

The obtained results will support the verification and validation efforts of the STH codes applied to LFRs. The investigated codes show a good agreement and the comparison proposes some open perspectives and future improvements.

Passive safety systems employing physical processes and phenomena are increasingly applied to contemporary nuclear reactor design. Assessment of the performance of these systems under various scenarios relies heavily on numerical analysis using codes from 1D to 3D depending on different levels of the design and safety demonstration purposes. Thermal-hydraulic (TH) phenomena in pool-type Lead-cooled Fast Reactors (LFRs) often exhibit multi-dimensional characteristics such as the development of thermal stratification and mixing during natural circulation. Accurate prediction of mutual interaction between these phenomena in the pool and its effects on loop dynamics requires 3D analysis.

Computational Fluid Dynamics (CFD) provides high-fidelity 3D TH analysis but is computationally expensive for analysis of prototypical conditions. System Thermal-Hydraulic (STH) codes (e.g., RELAP5) offer efficient calculation but are inadequate to resolve 3D phenomena. A compromised solution is to use system-level TH codes with 3D features, e.g., (GOTHIC, CATHARE). The recent development of GOTHIC enables the modeling of Lead-Bismuth Eutectic (LBE) flow while its suitability and validity for safety analysis need to be confirmed.

Therefore, this work aims to assess GOTHIC predictive capabilities for LBE 3D phenomena through code-to-code and code-to-experiment comparisons. Validation data is obtained from a forced to natural circulation transient produced in TALL-3D facility which is a 7m LBE loop featuring a 3D pool-type test section. Simulations are performed using CFD code ANSYS Fluent and system TH code GOTHIC. The focus will be pool thermal stratification and mixing in the 3D test section.

Current major international nuclear reactor system analysis codes predominantly utilize the two-fluid six-equation model to study the behavior of nuclear power plants under accident conditions, which presents considerable limitations. Most studies of the two-component two-fluid model have focused on water-steam systems, while liquid metal-gas systems at high temperatures have received relatively less attention. This paper studies the two-component two-fluid model and its rapid solution method to address the demands of full-scale simulations for both existing and conceptual nuclear reactor systems. The conservation equations of the two-component two-fluid model are discretized using a first-order upwind semi-implicit method, based on the staggered grid and finite volume difference. A system of linear equations is derived by substituting the equation of state and solved using the NRLU method. The mathematical suitability of the model is enhanced by introducing a virtual mass force. When one phase of the two-component two-fluid model is absent, the fraction of the virtual phase is assigned a small value to prevent the coefficient matrix from becoming singular. By simulating the natural circulation and gas injection-enhanced circulation conditions at varying power levels on the lead-bismuth loop test bench NACIE, the numerical accuracy and computational stability of the semi-implicit numerical algorithm for lead-bismuth-gas two-component two-fluid model are successfully demonstrated. It lays the foundation for further research on the two-component two-fluid model and the development of related code.

Lead-cooled Reactors (LFRs) are considered a promising concept in the framework of designing new Generation-IV reactors. The analysis of the thermal-hydraulics phenomena can be performed by means of numerical RANS simulations. This paper aims to evaluate the best practices for the mesh generation of a grid spaced fuel-bundle assembly. The results are compared to the experimental results provided in the LFR-T/H benchmark promoted by OECD NEA. The work focuses on different strategies to generate the background mesh and alternative modelling tools (e.g baffles) for capturing detailed geometry and dealing with the contact points in OpenFOAM. Initially, the bundle geometry without the grid is simulated under isothermal conditions and the results in terms of pressure drop are compared with existing correlations. In the second part, a single grid is included in the numerical domain, and it is characterized in terms of pressure drop as function of the flow velocity.

Liquid metal cooled fast reactors is one of the most promising fourth-generation nuclear systems. Y-junctions are commonly adpot in piping system. Non-Isothermal fluids frequently mixed in these components, lead to thermal pulsation on the solid-wall, and may induce thermal fatigue to piping system.To understand the mechanism of thermal pulsation and thermal fatigure, we independently set up a non-isothermal mixing test platform of the working fluid, and obtained the temperature distribution of the working fluid during the non-isothermal mixing process in the 90° Y-shaped component. This verifies the correctness of the Dynamic Smagorinsky Sublattice model in the non-isothermal mixing simulation, so as to correctly simulate the flow and heat transfer of liquid metal. On this basis, large Eddy Simulation (LES) approach for liquid metals in Y-junctions was applied. Angle and velocity pulsation behavior caused by the mixing of hot and cold fluids in Y-junctions under different incident angles (θ=30-90°), and momentum ratios (M_R=0.25–4.51) were discussed. The results show that the momentum ratio and the angle significantly influence the mixing characteristics of hot and cold fluids. At a 90-branch angle, fluid mixing is uniform, the thermal pulsation peak is larger. As the momentum ratio increases, the peak temperature pulsation gradually decreases. The findings offer valuable insights for the thermal-hydraulic design of future liquid metal cooled fast reactors.

Natrium® reactor vessel design utilizes the multi-dimensional computational fluid dynamics (CFD) to investigate the various flow phenomena and heat transfer mechanisms to predict the temperature distribution and flow velocity of the sodium. CFD is one of the many tools utilized during the design phase to inform various engineering teams including but not limited to transient and safety analysis, structural design and analysis etc. As part of the validation and investigation of the prediction capability of the CFD, multiple legacy data is being investigated. MONJU reactor trip benchmark by IAEA-CRP is one of them as it is investigated within the present paper. It investigates specifically thermal stratification behavior in the upper plenum of sodium cooled reactor. Previous studies investigated uncertainties on the flow hole geometry and turbulence modeling. Present paper investigates a more recent second-generation URANS closure (STRUCT−ε) model. The approach aims at advancing the robustness of hybrid turbulence models by relying on the efficiency of an extensively validated anisotropic k−ε method, while locally including the optimum resolution of complex unsteady flow structures.

In Sodium-cooled Fast Reactors (SFRs), the sensitivity of neutron cross-sections is essential for understanding the complex relationship between core neutronics, thermal hydraulics, and reactor safety. SFRs utilize fast neutrons and liquid sodium as a coolant, which introduces specific challenges in heat transfer and neutron interaction. Sensitivity analysis of neutron cross-sections in these systems quantifies the effects of uncertainties in nuclear data on parameters like reactivity, neutron flux distribution, and power peaking factors. These parameters significantly impact the core’s thermal hydraulic behavior.

The interplay between neutronics and thermal hydraulics in SFRs is crucial due to the fast neutron spectrum and sodium’s high thermal conductivity. Variations in neutron flux and cross-section values influence localized heat generation, while changes in coolant temperature and flow affect cross-sections through feedback mechanisms. Proper modeling of these interactions ensures effective heat removal from the core, preventing excessive fuel temperatures and avoiding material degradation or fuel failure, especially during transients and accident scenarios.

In safety analysis, sensitivity calculations are vital for predicting the reactor’s behavior under normal and off-normal conditions, including critical events like Loss of Flow (LOF) or sodium boiling accidents. These analyses assess how cross-section uncertainties affect thermal hydraulic margins, guiding the development of design strategies to ensure safe reactor shutdown and decay heat removal. Sensitivity analysis thus plays a key role in optimizing SFR performance and safety by offering insights into how nuclear data uncertainties impact overall system behavior, leading to more robust safety measures.

Liquid Metal Fast Reactors (LMFR) are a promising technology for expanding nuclear energy to reduce carbon emissions from the energy sector. The ultimate goal of the project is to provide validation data for the Clad Damage Propagation (CDAP) module of SAS4S/SASSYS-1, a reactor safety code system. The Experiment on Pin Failure for LMFRs (ExPL) project at Oregon State University (OSU) aims to provide data on the heat transfer impingement due to fission gas ejection from an initial cladding failure event. The experiments in liquid sodium will consist of a 19-pin assembly with electrically heated surrogate pins in a liquid sodium flow loop. The test section will be instrumented with High-Definition Fiber Optic Temperature Sensors, placed inside capillaries in place of a solid wire wrap consistent with the simulated fuel assembly geometry. The temperature within the capillaries, determined by the heat transfer through the capillary, will necessarily be different from the heater rod surface. This paper details experiments in water and air that investigate the presence and magnitude of both spatial and temporal discrepancies and potential modes for mitigating observed error. These alternative, low-risk fluids, were utilized to establish a baseline understanding of this application of fiber optic sensors and to inform future experiments in liquid sodium.

The primary cooling loop in sodium-cooled fast nuclear reactors is achieved using a centrifugal pump immersed in liquid sodium. Under certain conditions, fluid vortices can be generated and develop into bubbles of the cover gas present on the sodium coolant free surface. This phenomenon, known as Gas Entrainment (GE), may have an impact on the reactor vessel design and on the core reactivity. The GE is difficult to predict and parameters influencing its occurrence are still poorly known. Simulations using Computational Fluid Dynamics (CFD) could help to better understand such phenomenon and identify the parameters that govern its occurrence. In this work, free surface flow simulations based on the Large-Eddy Simulations (LES) for flow hydrodynamics prediction combined with the Front-tracking method for interface modeling, were performed. It figured out that predictions of the interface dynamics is greatly influenced by the the element mesh size, on which depends the accuracy of the flow hydrodynamics prediction. Finer meshes allowed to better capture the instantaneous small eddies and local velocities, which enhanced the generation of the vortices. The coarse simulation predicted less intense pressure variations near the vortices, and smoother pressure distribution throughout the domain. On the contrary, the fine simulation exhibited more distinct clusters of high positive and negative vorticity, associated with more distinct low-pressure cores. This enhanced the development of large vortices moving along the free surface.

The Liquid Metal Cooled Fast Reactors (LMFRs) are one of the technologies being considered by the Generation IV International Forum (GIF). The unique geometric design of the rod bundle channels causes an uneven circumferential temperature distribution on fuel rod surfaces, further intensified by liquid metal's high thermal conductivity. This temperature variation may induce thermal fatigue damage to the cladding, threatening reactor safety. It is identified that the subchannel heat transfer characteristics in liquid metal reactors are predominantly influenced by the Peclet number (Pe) and the pitch-to-diameter ratio (P/D). Usually, Reynolds-Averaged Navier-Stokes (RANS) computational fluid dynamics (CFD) with turbulent models are employed to study heat transfer in liquid metal rod bundles, which fail to capture the anisotropic thermal transfer in liquid metal rod bundles, ignoring circumferential temperature differences. Conversely, Large Eddy Simulation (LES) offers detailed insights into flow and heat transfer phenomena. Accordingly, this study conducts a numerical investigation on hexagonally arranged fuel bundles using LES to explore the circumferential temperature distribution under varying Pe and P/D conditions. The LES results show that Pe and P/D can affect circumferential temperature non-uniformity. Moreover, considering the experimental costs, smaller hexagonally arranged fuel bundles with a non-prototypical cold wall are selected. Due to the cold wall effect, the central fuel rods exhibit smaller circumferential temperature differences compared to the outer rods. These findings highlight the critical impact of Pe and P/D on the circumferential temperature distribution of fuel rods, providing valuable theoretical guidance for the optimized design of liquid metal-cooled fast reactors.

Direct Numerical Simulation (DNS) is regarded as an accurate and reliable approach to generate high-resolution data. However, due to the high computational costs involved, DNS investigations are typically limited to simple geometries and scaled-down conditions. In addition to providing an insight into the flow physics, fully-resolved DNS data is often used as a valuable reference for development, validation and improvement of Computational Fluid Dynamics (CFD) models. It is shown in this study, Under-resolved DNS (UDNS) based on spectral element methods (SEM) offers a more cost-effective alternative for generating high-quality reference data, while still providing sufficiently accurate flow statistics with significantly fewer degrees-of-freedom. Moreover, calculating the statistical quantities of the flow using UDNS does not require the additional effort involved in including sub-grid modeling contributions.

The resolution and accuracy of DNS and UDNS simulations are evaluated for an internally-heated natural convection flow at Rayleigh number of 10¹¹. Three UDNS simulations are performed at progressively coarser grid resolutions, thereby reducing the computational costs. It is shown that high-accuracy DNS can only be achieved at a grid resolution criteria of Δ/η_k ~ π. The UDNS approach is shown to achieve sufficiently accurate results for the mean and RMS velocity and temperature fields, as well as the turbulent kinetic energy budgets. Such UDNS statistical data can serve as a cost-effective alternative for reference for the validation of engineering CFD models. Furthermore, high-quality reference data can be obtained at more realistic flow conditions at higher Reynolds or Rayleigh numbers, where fully-resolved DNS may be infeasible.

Direct Numerical Simulations (DNS) are considered an accurate and reliable approach to generate high-resolution data. However, due to higher computational costs involved, DNS is generally performed in scaled and/or simpler geometry, which is representative of the real industrial-scale scenario. In order to represent the flow and heat transport between an array of fuel rods in a reactor core, simulations are conventionally performed in a smaller computational domain with periodic boundaries – often limited to a single interstitial subchannel space or domain around a single pin.

The present study examines DNS results for several different configurations – a periodic domain around a single pin, a single subchannel space, and arrays of 2×1 and 2×2 subchannels. It is shown quantitatively that the proximity of the periodic boundaries in the smaller domains significantly alters the flow physics. The periodic boundaries of the smaller domains, being highly-correlated, cannot faithfully represent the large-scale flow interaction across subchannels. Higher-order flow statistics, which may be used for development and validation of lower order models, are shown to be significantly affected by the periodic boundaries of the smaller domain. The distribution of wall shear stress and Nusselt number around the pin are also seen to be affected. A comparison of secondary flows in the different geometry representations is also presented. Flow pulsations in the narrow gap, typically associated with low P/D ratios, are also observed for the present P/D ratio of 1.326. The frequency of these pulsations is also shown to be affected by the size of the domain.

The supercritical water-cooled reactor is one of the proposed designs under further
study and development by the Generation IV International Forum (GIF). Like any
emerging technology, it presents technical challenges that must be resolved and better
understood. This paper addresses one of these challenges, from a thermal hydraulics
perspective, by focusing on the effect of idealised surface roughness on flow and heat
transfer in mixed convection of supercritical water.
To investigate this phenomenon, we perform six DNS simulations for fully developed,
asymmetrically heated channel flows. Three cases feature channel walls (top and
bottom) tiled with pyramids to represent an idealized surface roughness of Rt (peak-to-
valley height) 0.05δ (where δ is the half-channel height), while the remaining three have
smooth walls for comparison. We examine three flow conditions: (i) vertical flow with
mixed convection, (ii) stably-stratified mixed convection horizontal flow, and (iii) forced
convection. Constant temperature boundary conditions are applied at the channel walls,
with the hot side having a temperature greater than the pseudocritical temperature and
the cold wall temperature smaller than it. The temperature difference between the walls
is kept at 15 K. The aim of this work is to study the impact of roughness on heat transfer
deterioration and enhancement. Comparisons are made between mixed and forced
convection flows in both vertical and horizontal orientations, and the DNS data provided
will be useful for benchmarking.
The simulations employ the open-source DNS solver CHAPSim 2.0, with the prescribed
surface roughness modelled using an Immersed Boundary Method (IBM). This method
allows for the modelling of complex geometries by introducing a body force to the solid
nodes.

CFD codes used in the nuclear industry require an extensive validation, over a large range of thermal hydraulic conditions. In the context of PWR scenarios with shutdown of the primary pumps, buoyancy-affected flows in the core rod bundle can be encountered at low flow rate. However, scarce experimental data are available for this type of flow.

To address this issue, a Wall-Resolved LES (quasi-DNS using high-order schemes) of an upward flow within a heated rod bundle subjected to lateral power skews has been performed by Vicente Cruz et al.[1]. This numerical experiment of a complex turbulent mixed convection flow using a Boussinesq approximation, exhibiting cross flows, a mixing layer and a local re-laminarization, provides a highly accurate database for code validation.

In the present work, following the recommendations by Benhamadouche[2], the same configuration is investigated using code_saturne[3]. The predictions show good agreement with the reference database which demonstrates the capability of this in-house industrial code to carry out high-fidelity LES computations for this type of configuration.

[1] R. Vicente Cruz et al., “Numerical investigation of the mainly axial flow in mixed convection regime within tube bundles”, Proceedings of the 18th UK Heat Transfer Conference (2024)

[2] S. Benhamadouche, “On the use of (U)RANS and LES approaches for turbulent incompressible single phase flows in nuclear engineering applications”, Nuclear Engineering and Design, 312, pp. 2–11 (2017)

[3] code_saturne is an open-source incompressible CFD code developed by EDF (https://www.code-saturne.org/cms/web/).

Twisted elliptical tubes are a proposed heat transfer enhancement (HTE) for use in Molten Salt Reactors (MSRs) due to their enhanced thermal performance compared to plain tubes. The twisting ellipsoid geometry causes the fluid to swirl as it passes over the tubes, enhancing mixing and inducing turbulence at lower Reynolds numbers. This effect increases the overall convective heat transfer; however, this is at the cost of increased frictional pressure losses. In this study, the Nek5000 computational fluid dynamics (CFD) code is used to perform Large Eddy Simulations (LES) of external flow around twisted elliptical tube bundles. Simulations were performed for three different tube cross-sectional aspect ratios (maximum versus minimum diameter), Reynolds numbers of 1,000 and 7,000, and Prandtl numbers ranging from 1-25. The aim of this study is to further characterize the heat transfer phenomena seen around twisted elliptical tubes when varying the tube aspect ratio, which is a dependency that has not been historically accounted for. Previous work by the authors has investigated the dependency of cross-sectional aspect ratio (AR=1.1 – 2.1) and below unity Prandtl numbers (Pr=0.001 – 1) for this geometry, and this study aims to extend this work by investigating above unity Prandtl numbers and varying Reynolds numbers.

This study presents a computational fluid dynamics (CFD) methodology for simulating boiling at bubble scale without prescribed flowrate in the context of the quenching fabrication process. The approach is based on the Volume Of Fluid (VOF) method and the incorporation of phases sources allowing the water/steam phase change. This method is based on the use of an adaptive mesh allowing a very high refinement along the water/steam interface during the development of the bubble. The primary objective was to validate this methodology by comparing CFD results with experimental data in pool boiling cases. The quantity of interest is the wall temperature at different heat fluxes and for different surface topologies (roughness is modeled explicitly). Analysis shows promising agreement between CFD results and measurements regarding the wall temperature for low heat fluxes. This work demonstrates the model’s ability to create and develop multiple interacting bubbles as well as predict relevant wall temperatures for low heat fluxes.

This work represents a significant advancement towards developing a methodology to numerically assess the heat removal capability of a given surface as a function of its topology.

The strengthening of pool boiling heat transfer capacity is of great significance. As an active enhancement technology, ultrasound can effectively enhance the boiling process by influencing the growth and detachment of bubbles. With low cost and simple operation, it has a broad application potential while the mechanism still needs to be studied.

This study presents a multi-physical model which considers acoustics and fluid dynamics based on Multiphysics software. The volume force term is added to describe the nonlinear effects caused by the sound field; and the level-set method is used to track the phase interface that brings the saturated boiling bubble behavior under the influence of ultrasound. As a commonly used numerical value in the industry, the ultrasonic frequencies are set to 20,28 and 40 kHz. Numerical simulation has found that the acoustic streaming caused by ultrasound can cause the enhancement of fluid flow which generate shear forces on the bubbles. The acoustic streaming also make perturbations on the surface, which can accelerate bubble detachment and further enhancing the surface heat transfer capacity. As a result, the heat transfer efficiency has a considerable increase. Increased frequency and ultrasonic power can effectively enhance the acoustic streaming then act on the heat transfer process. Meanwhile, experimental research has been carried out to verify the results of numerical simulations. The derived conclusions could be useful for the application of ultrasonic treatment on boiling heat transfer.

Nucleate pool boiling is a useful heat transfer technique present in many engineering applications such as space and aircraft industries, thermal design of electronic components, refrigerants, and nuclear reactors. Despite its wide use, the physical mechanisms linking heat transfer to bubble dynamics and turbulence remain largely unexplored. Depending on the subcooling temperature, which is the difference between the wall temperature and the liquid saturation temperature, bubbles may either depart, merge rapidly, coalesce slowly or shrink. These complex dynamics affect heat transfer and lead to intricate vortex patterns in the flow, influencing velocity and temperature statistics.

The present work uses Direct Numerical Simulations with an in-house solver for incompressible multiphase flow to study the effects of subcooling on boiling behavior. The level set method captures the interface between liquid and vapor phases, while a computer vision algorithm was developed to track bubble properties. The findings show that subcooling temperature significantly impacts heat flux and bubble behavior, which in turn alters the flow's vorticity structures, producing ring vortices or irregular patterns. Statistical analysis is provided to better understand these complex interactions, shedding light on the relationship between bubble dynamics and heat transfer in boiling flows.

The authors are grateful for the financial support by the National Aeronautics and Space Administration (NASA) Grant number: 80NSSC21K0470 monitored by Dr. David F. Chao. We thank NASA Ames Research Center and NASA Advanced Supercomputing Division (NAS) for their generous allocation on Pleiades to perform three dimensional CFD simulations.

Predicting maximum sheath temperature accurately is important from the safety perspective (to demonstrate fuel and fuel channel integrity). The geometry and dryout mechanisms for CANDU fuel channels differ significantly from the light water fuel bundle assemblies. A new Eulerian multiphase computational fluid dynamics wall boiling model is being developed to model liquid film thickness-induced dryout in CANDU fuel channels. The operating conditions in CANDU fuel bundles are especially challenging because they span over a range of two-phase flow regimes. The proposed liquid film thickness-induced dryout model, which leverages advanced boiling closures for water at high pressures, was assessed in an earlier study using single-element CHF tests performed at Stern Laboratories, and encouraging results were obtained. This paper documents the model development for CANDU fuel bundles, its implementation in the STAR‑CCM+ software, and a qualitative assessment of the predicted dryout power using data from heated tests on the modified 37-element CANDU fuel bundle configuration. The approach adopted in the analyses is anticipated to yield advanced predictive capabilities that can be leveraged to improve traditional reactor safety analyses.

High resolution computational fluid dynamics (CFD) studies of turbulent flow boiling are challenging due to the simultaneous requirements that turbulence and bubbles are resolved adequately. These interface-capturing simulations can support the understanding of the mesoscale mechanisms of the heat removal process. Detailed analysis of such simulations has been demonstrated to uncover important physics of the heat transfer process and support macroscopic heat transfer model development. One area in which CFD can support the understanding of complex fluid mechanics is high heat flux boiling conditions. In pursuit of this goal, a benchmark problem has been developed to mimic the conditions of a previously completed high resolution experiment. The topic covered in this paper is a scoping study designed to assess the performance of the selected approach applied to this problem. The domain in this case is accurate to the experiment, with realistic Reynolds number, inlet subcooling, and heat flux. However, for simplicity and computational cost concerns, only a limited number of nucleation sites are considered. The mesh design is covered in detail to illustrate the consideration of both liquid turbulence and nucleate boiling. The bubble dynamics and associated wall temperature are compared against experimental values to assess the suitability of the approach for future multiple nucleation site studies and to which surface conditions the approach is generalizable. Based on the observed results, CHT & microlayer evaporation models are expected to introduce additional physics to improve predictions. Therefore implementation of these models is evaluated for future work.

This research aims at simulating and contributing to understand the generic physics which could occur in nuclear spent fuel pools during loss of cooling accidents. Based on the limitations inherent to the Direct Numerical Simulation approach in terms of Rayleigh number and geometry, we also intend to provide relevant reference results for RANS simulations.
This paper presents in a stepwise layout the different necessary components to simulate the natural convection of water in a cuboid domain combining vapor bubbles dynamic and evaporation through the free surface resulting into a descent of the water level.

Heat transfer due to evaporation is accounted for using the model presented by W. H. Hay et al. (2021), while the related mass transfer relies on a new remeshing procedure which attributes the descent proportionally to all cells by re-meshing the grid at each time-step. This allows to avoid any field changes at the boundaries whilst distributing the error along the height. This remeshing procedure, although apparently simple, involves a change in the temporal discretization of the governing equations. On the other hand, an Eulerian-Lagrangian approach is implemented and allows to compute the motion and growth/shrinkage of vapor bubbles while the effect of the bubbles on the fluid is accounted via momentum and energy exchanges between the two phases in a two-way coupling.

First, we detail the different models implemented. We then present a validation and verification procedure against analytical, experimental and numerical results. Finally, we present and discuss results of both models separately and combined.

In the case of a LOCA in the PWR, during the long-term cooling stage, boric acid in the coolant may be carried with steam or entrained droplets due to steam discharge, which may affect the reactivity of the core. There are a number of research that make it possible to determine the value of droplet entrainment leaving the reactor. However, the loss of boric acid with steam is related to the solubility at steam. It is necessary to study dissolution process of boric acid in steam under different parameters.

This paper conducts an experimental study of the distribution coefficient of boric acid between the steam and liquid phases of the solvent. The test device of the solubility of boric acid with steam has been established, including a heated test section, steam separator, and steam condenser. The experiments are conducted under a certain pressure of 0.1-0.4MPa, with an initial boric acid solution concentration range of 1000-10000 ppm. The concentration of boric acid in the discharged steam and the remained solution are obtained. The effects of pressure, temperature and initial concentration on the solubility of boric acid in steam are summarized. The empirical relationship of distribution coefficient of boric acid in the steam phase and liquid phase are obtained by fitting the experimental data.

This study investigates the flow characteristics in a 3x3 Helical Cruciform Fuel (HCF) bundle using Magnetic Resonance Velocimetry (MRV) experiments and Computational Fluid Dynamics (CFD) analysis. The HCF bundle, designed to enhance thermal-hydraulic performance in nuclear reactors, features a unique geometric configuration with helically twisted cruciform-shaped fuel elements. To validate the numerical predictions, experimental measurements were conducted using MRV technology, which provides three-dimensional velocity field data without intrusive flow disturbance.

The experimental facility consisted of a full-scale 3x3 HCF bundle model operating at high Reynolds numbers exceeding 10,000. MRV measurements captured the complex flow structures, including secondary flows and vortex formations in the sub-channels. The CFD analysis employed the Reynolds stress model with a refined mesh containing approximately 2.5 million elements, validated through rigorous mesh sensitivity studies.

The study revealed distinct flow patterns characterized by enhanced mixing due to the helical geometry. Secondary flows were particularly pronounced in corner sub-channels, exhibiting higher tangential velocities compared to interior sub-channels. These findings provide crucial validation data for CFD methodologies in nuclear fuel bundle analysis and contribute to understanding the thermal-hydraulic behavior of advanced fuel designs.

The validated CFD model can serve as a reliable tool for future HCF bundle optimization studies and thermal-hydraulic characteristic analyses, potentially leading to improved nuclear reactor fuel efficiency and performance.

Thermal fatigue cracking may initiate at a T-junction where high and low temperature fluids flow in. In this study, the flow structure and fluid temperature fluctuations in a branch pipe of a T-junction were investigated under flow patterns where the main pipe flow penetrates into the branch pipe. The test section consists of a horizontal main pipe with an inner diameter of 150 mm and a vertical branch pipe with an inner diameter of 50 mm. A 45º elbow was installed upstream on the branch pipe side in order to study its effect on the penetration flow pattern and temperature fluctuations. To simulate penetration flow, the experiment was conducted under conditions where the inlet flow velocity on the branch pipe side was much smaller than the inlet flow velocity on the main pipe side. The flow pattern was visualized using the tracer method. The flow in the branch pipe was classified into three flow patterns: no penetration; entrained penetration; and impinged penetration. These patterns depended on the momentum ratio of the main and branch pipes, regardless of the presence of the elbow. The maximum penetration depth into the branch pipe increased when the upstream elbow was installed. Fluid temperature distribution along the branch pipe was measured with eight sheathed thermocouples. The fluid temperature fluctuations also increased, especially in the range of relatively small momentum ratios where the hot mainstream intermittently penetrated into the branch pipe.

Fatigue and defect tolerance assessments of high integrity PWR pressure boundary components require transient temperature fields to be defined to produce thermal stress predictions. These temperatures are often produced using a heat transfer coefficient and bulk temperature boundary condition approach. This can be imprecise and inaccurate for components with complex flows and geometries. An alternative approach is to directly predict the temperatures using conjugate heat transfer CFD, where the solid temperature field is predicted directly and simultaneously with the adjacent flow. This approach removes the uncertainties of using an intermediate model to transfer the information, but since it requires flow predictions at all times, the computational cost is impractical for the large number of multi-hour plant transients that need to be considered.

The cost of the CFD solution can be reduced by using an 'infrequent updates' approach, where the flow-field is considered to change slowly and is 'frozen' for intervals where only the thermal fields are solved. This is cheaper to calculate and can use longer time steps. The flow is solved in brief update intervals throughout. This method has been applied to the assessment of a large number of transients for the feedwater nozzles for the steam generators at the Sizewell B PWR in the UK.

The setup considerations and accuracy of the infrequent updates approach are discussed, as well as the effects of finite domain size and buoyancy driven flow instabilities.

Hydroaccumulators are designed to provide fast water injection into the vessel to ensure proper core cooling. One of its main function is that during LOCA cases, when no HPSI pumps are available, the hydroaccumulator supplies cooling water until LPSI pumps can operate and maintain the long term cooling. In VVER-440 reactors the hydroaccumulator empties at higher pressure than the LPSI can operate. The core should survive that injection hiatus. In that sense the primary pressure when hydroaccumulator injection terminates is very important

In PMK-2 facility several experiments were conducted utilizing various ECCS configurations and these tests were later calculated using RELAP5 thermal-hydraulic system code. During these post-test calculations a difference between the measured and calculated hydroaccumulator injection characteristic and terminating pressure value was noticed.

The hydroaccumulator is pressurized using nitrogen gas. Under pressure, some of this gas gets dissolved into the coolant in the hydroaccumulator water. During injection, the dissolved gas is reintroduced into the gas dome increasing its pressure. The RELAP5 system code does not consider this effect, leading to the observed differences. This is an unconservative deviation since the code predicts lower pressures at HA emptying thus shorter injection hiatus.

To address the phenomena experiments were performed. Using the PMK-2 facility more than 70 separate effect tests were conducted using one of the hydroaccumulator vessels. The tests were done at several different pressure levels and coolant temperatures. RELAP5 post-test calculations were carried out for each test, and the effect of introducing additional gas into the vessel was studied.

A significant percentage of reactors in operation, under construction or recently commissioned are VVER reactors. In parallel, there is a growing interest in analyzing the behavior of the Advanced Technology Fuels (ATF) under development in this type of nuclear reactor. Among the new ATF designs, the FeCrAl and Cr-coated claddings are the most promising evolutionary options. In addition to a relatively high level of technology readiness, these evolutionary cladding materials offer improved oxidation and hydride resistance, as well as improved mechanical strength. All these properties are essential in accident sequences where high core temperatures are reached. In the present study, core damage sequences have been analyzed with a model of a VVER-1000 reactor for the thermal-hydraulic code TRACE. In addition, an in-house version of the TRACE5P6 system code for FeCrAl cladding has been developed by NFQ and UPM.The selected sequences are SBO and LOCA sequences. The results show that the core damage temperature for the Zircaloy cladding cases is reached well before that for the FeCrAl and Cr coating cladding cases. The performance of these new cladding materials provides additional time for recovery of the safety systems responsible for core cooling and replenishment of the reactor coolant system inventory.

In molten-salt-fuelled reactor systems, the fluid may experience substantial volumetric heat generation in addition to heat removal from surrounding structures. To quantify these effects, we investigate developed channel flow with internal heating using a systematic multi-scale approach comprising Direct Numerical Simulation (DNS), Large Eddy Simulation (LES), and a semi-analytical solver (SAS). First, DNS and LES are compared in a turbulent parallel-plate configuration at different Prandtl and Reynolds numbers, demonstrating excellent agreement in flow and thermal fields, with the SAS method showing acceptable accuracy. Building on this benchmarking, the SAS tool is then employed to explore a broad parameter space, offering insights into how internal heat deposition modifies the temperature distribution across Reynolds and Prandtl numbers. Comparisons are also drawn against the canonical wall-heating scenario, revealing that volumetric heating often remains a secondary effect in turbulent regimes but can become more pronounced at lower Reynolds numbers, higher Prandtl numbers, or when nearly all heat is deposited in the fluid. These findings establish guidelines for reduced-order modeling in liquid-fuel reactor analyses and highlight conditions under which internal heating warrants particular attention. The paper concludes by outlining ongoing and future research directions, including refinements for variable fluid properties and complex geometry extensions.

Moltex Energy is developing the technology and design for a stable salt reactor-waste burner (SSR-W). It is a static fueled chloride molten salt reactor (MSR) with a fast neutron spectrum and is designed to be fueled with transuranic elements recovered from spent fuel of CANDU and light-water reactors. Argonne, University of Michigan, and Moltex are developing three multi-physics plant digital twins (DT) which leverage advanced computational methods to achieve reactor performance optimization and enable predictive maintenance to reduce the plant operation and maintenance (O&M) costs. DT-1 provides methods for long-term fuel cycle modeling and optimization and provides the operator with an indication of refueling time and position. DT-2 comprises of an integrated system plant model that can be utilized in simulating normal operation as well as in assessing the safety performance of SSR-W during postulated and bounding accident conditions. DT-3 develops a conceptual structural health monitoring strategy for an innovative matrix heat exchanger, which involves machine learning-based classification of distributed temperature sensing for detection and localization of faults. Additionally, a separate effects loop (SEL) is being constructed at Argonne to support the SSR-W in three technical areas: (1) thermal-hydraulic heat transfer in MSR, (2) natural circulation phenomenon pertinent to passive decay heat removal, and (3) redox control in salt to minimize corrosion of structural materials. This paper describes the methodologies implemented in the three DTs and shows some illustrative results and outputs. The paper also describes the SEL, the test campaign and presents some preliminary thermal hydraulic and heat transfer data.

Aqueous homogeneous reactor is a new type of reactor that dissolves nuclear fuel in a liquid. The aqueous homogeneous reactor have significant advantages in extracting medical isotopes such as I-125 and Sr-89 due to their liquid fuel characteristics. However, during the operation of the aqueous homogeneous reactor, water molecules in the fuel aqueous homogeneous will collide with fission fragments, decompose to produce hydrogen and oxygen. Under accident conditions, the rapidly increasing nuclear power of the reactor will exacerbate this phenomenon. The generated hydrogen will accumulate in the gas space of reactor, posing a threat to the structural integrity of the reactor. Therefore, the system design of an aqueous homogenous reactor needs to take the hydrogen elimination into consideration.This article considers a gas recombination system(GRS) , a nitrogen blow system(NBS) and constructs a 200kW aqueous homogeneous reactor model in RELAP5.The impact of the gas Recombination system on the volume fraction of hydrogen produced by the aqueous homogeneous reactor in a typical reactivity introduction accident was analyzed under different working conditions. The results indicate that the presence of a gas recombination system can significantly reduce the volume fraction of hydrogen during a reactive accident in a aqueous homogeneous reactor. The hydrogen volume fraction can be guaranteed to be less than 4% during the accident , which make sure meets the requirement of hydrogen risk guideline. This study contributes to the design and construction of medical aqueous homogeneous reactor.

Due to the high-temperature operation expected from Molten Salt Reactors (MSRs), thermal radiation can significantly influence the thermal-hydraulic evaluation of these systems, and the intricate coupling of multiple physical phenomena in this kind of reactor necessitates the development of high-fidelity simulation codes. This work presents a new library, developed in OpenFOAM using C++ object-oriented programming, to model thermal radiation coupled with fluid mechanics and other physical phenomena. This library employs the SP₃ method to solve the Radiative Transfer Equation (RTE) and is compatible with all OpenFOAM sub-models for absorption, emission, and scattering, thanks to its inheritance from the default radiationModel library. Additionally, an innovative boundary condition is implemented to model metals such as Stainless Steel and Hastelloy, which are commonly used in the reflector of MSRs. This code can accurately calculate the power density distribution within the system through the strong coupling between neutronics and thermal-hydraulics calculations. This work considers two case studies: the 2D axisymmetric EVOL geometry and the 3D one-sixteenth MSFR geometry. In both cases, the effects of the thermal radiation on the temperature fields are evaluated. Additionally, the effects of temperature, fluid flow type (laminar or transient), and radiation properties (emission and absorption coefficients) on incident radiation are analyzed.

Kairos power is developing a pebble fuel based molten salt reactor. As a part of development, pressure drop through the pebble bed is tested, and a converging and diverging conical part of the core is experimentally studied in current scope. Test facility is scaled down with pebble’s Reynolds number. Test results are compared to KTA correlation, which is widely used correlation in cylindrical pebble bed pressure drop. Since KTA correlation is based on cylindrical geometry, parameters are taken at inlet or outlet of each truncated conical interval. For converging cone, there are four measurement intervals along flow direction. Most of the test results show good accordance with KTA correlation based on parameters from outlet of the truncated cone. However, the last interval, which has the smallest cross section, KTA underestimates the pressure drop. The difference between KTA and test at the last interval increases with Rep. For diverging cone, there are three measurement intervals along the cone. For the middle interval, KTA and test result match well. However, for the other two intervals, KTA shows significant overestimation. Considering manifold like geometry of the two cones and interval-sensitive predictability of KTA, a new correlation is needed.

Subchannel (SC) analysis has historically supported core design numerical simulations for a wide variety of concepts encompassing thermal and fast reactors. The suitability of SC codes to heavy liquid metal coolants and extreme operating conditions have been the object of a work package in the framework of the IAEA Coordinated Research Project ‘Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)-CRP-I31038’, based on the experimental data provided by the NACIE-UP loop located at the ENEA.

The facility features 19, electrically heated, Fuel Pins Simulator (FPS) arranged with a triangular pitch and spaced by a wire wrap, instrumented with 67 thermocouples. The thermal hydraulic problem in the FPS assembly has been simulated by eleven institutions with nine different SC codes. The focus of the simulations is on the steady states of both forced and natural circulation conditions, as well as the in-between transition. In this paper two extreme cases ahave been analysed, one with all pins heated (ADP10) and one with only the seven central pins active (ADP06). ADP10 test is more representative of a condition which could be found in power reactors, while ADP06 test is a challenging power-profile condition, which allows for unprecedented physics insights, both cases can be used for validation of the SC codes .

The comparison demonstrates that SC codes can reliably capture the temperature profile within a wire-wrapped pin assembly, though it also highlights a need for modelling improvements of the wire effect with extreme intra-bundle temperature gradients.

TerraPower is developing a subchannel analysis code, Mongoose++, to support the design of the Natrium^® demonstration reactor. Development of this code was initially motivated by the need to predict core-wide duct temperature distributions to calculate reactivity feedback from radial expansion and assembly bowing. Such calculations require resolution of local coolant flow and temperature distributions within an assembly as well as global inter-assembly heat transfer effects. While Computational Fluid Dynamics (CFD) remains computationally prohibitive for routine design and analysis calculations at the core-wide scale, intermediate fidelity subchannel methods are well-suited for these tasks and have a proven record in the industry for licensing calculations.

Mongoose++ is written in C++ and traces its lineage to the legacy COBRA series of subchannel codes. Mongoose++ utilizes a similar formulation of the subchannel conservation equations but with several advancements to support the specific needs of the Natrium^® project. The subchannel equations are discretized using the finite volume method on a staggered mesh and solved iteratively using a variant of the SIMPLE algorithm. Each iteration of the SIMPLE algorithm is parallelized across assemblies. While more costly than traditional axial-marching schemes, the SIMPLE algorithm is more robust when modeling assemblies with significant buoyancy-induced flow redistribution and localized flow recirculation; such conditions may arise in SFR non-fuel assemblies during off-normal operating conditions at reduced flow.  

This paper provides a detailed overview of the design and implementation of Mongoose++, discussing current capabilities and planned developments. Benchmark comparisons against legacy experimental data and recent CFD calculations are presented.

Recent efforts have focused on at establishing high-fidelity, multi-physics safety analysis methodologies to assess realistic safety margins. However, these approaches still exhibit conservatism by employing conservative initial conditions, such as assuming a hot rod with maximum power for the whole reactor core. This study presents the development of a coupled code for pin-wise reactor core analysis, coupling the subchannel analysis code CUPID, neutron kinetics code MASTER, and fuel performance code GIFT. The objective is to generate accurate pin-wise fuel rod conditions during normal operation, providing more realistic initial conditions for safety analysis. The coupled code performs thermal-hydraulic analysis of the reactor core, accounting for fuel deformation, and simulates realistic power distributions with reactivity feedback from coolant and fuel temperatures. Coupling between the codes was achieved using socket communication and dynamic link library (DLL). A practical simulation was performed on an OPR1000 reactor core during the first cycle, and key results from the steady-state simulation were evaluated. The impact of the fuel performance code was also examined by comparing the results of the coupled CUPID/MASTER and CUPID/MASTER/GIFT codes. Finally, the effect of pin-wise initial conditions on safety analysis was investigated for a reactivity-initiated accident (RIA) scenario with results compared between conservative initial conditions and pin-wise initial conditions.

Hitachi has been developing subchannel analysis codes to predict thermal-hydraulic characteristics of newly designed BWR (Boiling Water Reactor) fuel bundles. The steady-state subchannel analysis code SILFEED (Simulation of Liquid Film Evaporation, Entrainment, and Deposition) with an updated film flow model has been mainly utilized for mechanistic film dryout predictions of various fuel bundle designs. Next-generation fuel bundles, such as Hitachi's RBWR (Resource-Renewable BWR) and GNF3, feature tight lattice configurations, axially varying water rod, and partial-length rods. For these designs with complex geometry, accurate evaluation of thermal hydraulic behavior for each subchannel is required.

To address these demands, Hitachi is developing a new subchannel analysis code based on a transient two-fluid three-field model enabling more advanced evaluations of void fraction and pressure drop by directly solving the void fraction. While its current capabilities are limited to steady-state and transient void fraction and pressure drop, future developments aim to extend its capability to fuel temperature and critical power predictions.

The first step of the code validation, we performed calculations of void fraction and pressure drop in NUPEC 8×8 bundle and φ5.2 to 10.1 mm tube, and compared with experimental data. The results demonstrate that the code achieves prediction accuracy of void fraction within ±15% and pressure drop also within ±15% under BWR operating conditions, which are comparable to those of other subchannel analysis codes.

Accurate modelling of the annular film flow regime in BWR fuel assemblies is of paramount importance for the prediction of the minimum film thickness, dryout location and duration, and crud deposition. Recent decades have seen the introduction of complex assembly structures such as part-length rods and spacer grids, whose complex effects on the flow must also be modelled appropriately for accurate estimation of safety parameters. This paper presents the initial steps towards an improved modelling package for churn-turbulent and annular flow at a subchannel scale. The subchannel code CTF, which uses a two-phase three-field approach, is modified to implement updated models for the droplet entrainment and deposition rates and a new solver (SCARF) has been developed, verified, and applied to partition the subchannel flow rate of the liquid field from CTF during annular flow into distinct films on adjacent rods. This method enables the assessment of film depletion on the sides of each rod and preserves the relative isolation of the films. The results of the two codes are then compared and show that the SCARF produces similar estimates of global flow parameters significant differences at the subchannel scale are observed. The effect is exacerbated for non-uniform radial power profiles. Future research will expand on this by incorporating new models for turbulence and void drift. Additionally, these models will consider how the distribution of film flow within the assembly affects inter-channel and inter-field transfer rates.

Plate-fuel reactors are one of the most common types of research reactors. The thin fuel plates have characteristics such as enhanced thermal conductivity and heat transfer properties, optimized neutron flux, and fuel performance, which makes them an attractive alternative. CTF is a state-of-the-art sub-channel code used for reactor thermal-hydraulics, initially developed for rod bundles and core analysis. In the current version of CTF, the heat generative solid structures are limited to cylindrical shaped rods or tubes. This work aims to develop a new heat structure model in CTF for the plate-type fuel geometry expanding the current capabilities of the code. This new feature extends the applicability of CTF to research reactors, mini- and micro reactors utilizing this fuel design. This development supports the growing demand for accurate thermal-hydraulic modeling and simulation of diverse fuel types and configurations to support the feasibility and safety analysis of small modular reactors. The proposed paper details the implementation process of the plate-fuel heat structure in CTF together with a detailed verification and validation process based on the analytical solution and the experimental available data, respectively. The verification process includes heat conduction within the plates and the convective heat transfer between the plate and the coolant. Thermal expansion and fuel performance models have been analyzed and they are currently suggested as a future improvement of the code.

The high-temperature gas-cooled reactor (HTGR), combined with a helium turbine cycle, represents a cutting-edge application of advanced nuclear energy technology. A critical component in this system is the printed-circuit heat exchanger (PCHE), which features microchannels (1–2 mm wide) for efficient heat transfer. However, these systems face challenges, particularly from dust particles generated within the reactor. These particles tend to deposit on PCHE surfaces due to the microchannels' narrow dimensions and frequent turns, potentially degrading heat transfer performance and blocking the channels. Understanding the long-term effects of particle deposition on PCHE performance is essential for the sustainable operation of HTGR systems. This study investigates this issue using a near-wall drag model to incorporate shear flow effects and an EA rebound model to simulate particle deposition behavior. A dynamic mesh method was employed to track the evolution of deposition morphology over time. Key findings reveal a non-linear relationship between particle size and deposition fraction, with deposition increasing and then decreasing as particle size grows. Smaller particles deposit primarily at upstream bends, where flow dynamics promote adherence, while larger particles settle downstream after sufficient energy dissipation. Additionally, upstream deposition significantly reduces downstream particle accumulation, influencing deposition distribution patterns.

The shell-and-tube heat exchanger (STHE) is crucial for industrial safety and efficiency. Refined simulations of its operational characteristics often use three-dimensional (3D) and one-dimensional (1D) models to calculate outlet temperatures.3D models are complex and time-consuming meanwhile 1D models are simple but less accurate, and data-driven models lack interpretability and have limited application. To overcome these limitations, this paper proposes a hybrid "mechanism-data" driven model for STHE.Firstly, based on heat transfer and fluid mechanics, a 1D simulation model is built using MWORKS/Modelica, analyzing outlet temperature and pressure changes under steady, transient, and fault conditions. Secondly, operational data from STHE test benches, including flow rates, temperatures, pressures, and outlet temperature simulations from mechanism-driven models, are collected to build a data-driven model using deep learning algorithms. This model captures nonlinear relationships and dynamic characteristics, addressing the mechanism model's inability to observe and describe factors like corrosion and baffles.Combining both models, a hybrid "mechanism-data" driven model is established, offering physical interpretability and high accuracy. By simulating test bench operations in real-time, it detects potential faults, identifying their types and severity, supporting maintenance and management.Experimental validation shows the hybrid model outperforms single models in simulation accuracy and fault diagnosis, accurately reflecting STHE operational status and fault characteristics. Future work will optimize and enhance this model for broader applicability and accuracy across different conditions and STHE types.

The supercritical carbon dioxide (sCO₂) Brayton cycle, as a core component in the design of next-generation nuclear energy systems, emphasizes safety, operational efficiency, and non-proliferation characteristics. Within this framework, the Printed Circuit Heat Exchanger (PCHE) plays a key role in optimizing the heat transfer process under high temperature and high pressure conditions. This paper proposes a numerical design framework for the PCHE, focusing on the reduction of flow non-uniformity through a combination of secondary heads and porous baffles. Computational Fluid Dynamics (CFD) methods are employed to simulate the thermal-hydraulic performance of the heat exchanger, assessing the impact of different geometric parameters on flow distribution and heat transfer efficiency. The results demonstrate that the combination of secondary heads and optimized porous baffles significantly improves flow uniformity, thereby enhancing heat transfer efficiency and reducing pressure drop. This study provides valuable insights for optimizing the thermal-hydraulic performance of heat exchangers in supercritical CO₂ Brayton cycles.

Surrogate-based optimization (SBO) is a powerful approach for the design of the airfoil fin printed circuit heat exchanger (PCHE), which maximizes heat transfer and simultaneously minimizes pressure loss. Existing optimization studies of the airfoil fin PCHE commonly focus on the channel configuration and the fin arrangement. However, a meticulous optimization of the airfoil PCHE may consider the parameterization of the shape of the airfoil fin, while existing optimization methods are insufficient in such cases. To address this issue, the free form deformation (FFD) method is applied to parameterize the airfoil fin shapes and design variables are extracted to control the deformation of the shapes. The airfoil fins are divided into two groups according to the upstream and downstream of the channel. Fins within the same group deform synchronously. The heat transfer rate and pressure drop are employed as the objective functions of the optimization. To improve the optimization efficiency, the Kriging surrogate model is adopted to approximate the relations between the design variables and objective functions. Then, a multi-objective optimization using Non-dominated Sorting Genetic Algorithm-II (NSGA-II) is conducted and the Pareto solutions are obtained. Comprehensive optimal designs are selected on the Pareto front, and the thermal and hydraulic characteristics of the optimized designs have the advantage over those of the PCHE with original airfoil fin shape.

The water-cooled SMR is generally designed for flexible operation. In the performance analysis of nuclear power plants, system codes are used to evaluate plant performance under various Performance-related Design Basis Event (PRDBE) conditions. The load-following operation is one of the major PRDBE conditions. Many water-cooled SMRs employ a helical type once-through steam generator (SG), which produces superheated steam on the secondary side. Unlike conventional steam generators, the helical SG has no concept of water level, making superheated steam pressure a key control parameter in balancing plant pressure during load-following operation.

The helical SG comprises thousands of helical coils surrounding the riser, each with a different helical geometry. These geometric differences affect the heat transfer characteristics, potentially altering the outlet conditions. However, modeling every tube’s unique geometry would be inefficient. Instead, the SG is typically divided into multiple units, and system code calculations are performed on this segmented model.

This study explores how the fineness of the nodalization, or the number of divisions of the SG, affects performance analysis results. Using MARS-KS code, the reference helical steam generator is divided into 5, 10, 15, and 20 units, and steady-state calculations are performed for each case. The focus is on comparing the steam condition at the secondary outlet. It is expected that the outlet conditions of superheated steam are similar across different nodalizations, suggesting that coarse nodalization does not significantly impact analysis results. This is expected to allow for more efficient calculations in large-scale performance scenarios.

The IAEA has been a key supporter of the development of advanced power-reactor technologies and nuclear fuel technologies for many decades. Its efforts include providing platforms for information exchange, organizing meetings, issuing publications, coordinating research activities and maintaining databases (for advanced reactor designs, fuels, fuel cycle and post irradiation examination facilities).

This presentation will highlight the IAEA’s on-going programmes and near-term plans to support the development of new reactor technologies and advanced fuels for both operating and innovative power reactors. This includes IAEA’s efforts in developing accident tolerant and advanced technology fuels (ATF), fuels for recycling/multi-recycling of nuclear materials, and advanced fuels for GEN-IV and small modular reactors, as well as advanced reactor designs.

Special emphasis will be paid to key Coordinated Research Projects (CRPs) including “Testing and Simulation of Advanced Technology and Accident Tolerant Fuels (ATF-TS)”, “Fuel Materials for Fast Reactors”, “Standardization of Subsized Specimens for PIE and Advanced Characterization for SMR and Advanced Applications”, “Fuel Modelling Exercises for Coated Particle Fuel for Advanced Reactors Including Small Modular Reactors”, “Developing a Phenomena Identification and Ranking Table and a Validation Matrix, and Performing a Benchmark for In-Vessel Melt Retention”, and “Advancing Thermal-Hydraulic Models and Predictive Tools for Design of SCWR Prototypes”.

IAEA Member States are strongly encouraged to participate in the IAEA’s topical meetings and new CRPs, which provide unique opportunities to engage with cutting-edge reactor and fuel technologies critical to the future of nuclear energy.

CRUD is one of the major considerations from the perspective of operation and safety, especially in PWR nuclear power plants. CRUD, which consists of corrosion products in the reactor coolant system, is known to induce thermal resistance, distortion of power distribution, local corrosion, boron hide-out, etc. Some of these adverse effects imply that the CRUD can affect the plant economics and core integrity hindering the thermal limit of nuclear power plants. These kinds of challenges significantly highlight that the CRUD effect should be investigated to calculate accurate operational margin of nuclear power plants in operation as well as in development. Various academic efforts have been made to figure out the mechanical and chemistry characteristics related to the deposition mechanisms and implications, trying to reflect them on the nuclear power plants. A reliable database of CRUD is necessary for solutions since deposit experiments are too hard to generate quantitative results under high-pressure/high-temperature PWR operational conditions. Hence, this paper investigates the CRUD effects utilizing the thermal properties obtained from experiments under actual PWR conditions. In this method, the CRUD layer is arranged on the surface of the fuel clad composing an active core in a simulated PWR plant model. Based on the results, the guideline can be made to calculate the local heat flux on the nuclear fuel considering the high burn-up rate of the reactor core. Further research will be conducted for the better quality of the database, expanding the test conditions and results.

Passive safety systems are widely used in the advance nuclear reactors. As an important component in the passive safety system, the core makeup tank (CMT) plays a key role in the safety injection and the core decay heat removal during the transient process of small-break loss of coolant accident (SBLOCA) . In this study, the thermal hydraulic behaviors of the CMT injection process were investigated by simulating different accident scenarios of the ACME test facility with different break sizes which includes the 1inch，2inch，4inch and 8inch breaks. Through the comparative analysis of the transient simulation under different break conditions along with visualization results, the switching process between the two working modes of CMT and its interaction with other safety injection component (such as ACC) were studied. Moreover, the mechanism of thermal stratification in CMT and the related influencing factors of CMT safety injection were analyzed. This work can provide guidance for the safety design and performance qualification of advanced passive nuclear reactors.

Abstract – In the framework of the OECD/NEA experimental projects, like ATLAS-3, a set of DEC-A experiments were carried out aiming at assessing the nuclear power plants capabilities to deal with complex accidental scenarios and evaluating the design provisions of the safety systems and the adequacy of the accident management measures. DEC-A scenarios are generally based on events and combinations of events which may lead to severe fuel damage in the core. The safety assessments are normally carried out using best estimate tools with the objective to demonstrate the fulfilment of the safety criteria and the design robustness. In this paper, a DEC-A experimental scenario carried out in the ATLAS test facility is considered. The ATLAS C3.2 test concerns a steam line break scenario relying on the passive auxiliary feedwater and operator action to cooldown the primary system. The transient involves complex interacting natural circulation phenomena including natural circulation flow interruption, steam condensation and heat exchange in large pool. In this framework the CATHARE code is used to simulate the course of the transient and the related natural circulation phenomena. It is shown, on the one hand, that the safety features of the design together with the operator actions are capable to bring the primary system to a safe end state and on the other hand, the CATHARE code prediction capabilities, for such complex scenario, are generally good. Nevertheless, additional efforts should be carried out to enhance the simulation under passive natural circulation cooling conditions.

During a Loss of Coolant Accident (LOCA) with failure of Emergency Coolant Injection (ECI) in a Pressurized Heavy Water Reactor (PHWR), the convective cooling is compromised, leading to an increase in the fuel channel temperature. Initially, the fuel temperature rises due to the decay heat and the energy stored in the fuel. The chemical reaction between the cladding and steam further escalates the temperature, causing the cladding to embrittle from oxygen diffusion. This can result in cladding rupture and the release of fissile materials. Additionally, this reaction produces hydrogen gas, which threatens the structural integrity of the containment. The heat from the fuel bundle is transferred to the Pressure Tube (PT) and the rising temperature of the PT leads to deformation, such as ballooning, sagging, or both, due to the rapid degradation of its thermo-mechanical properties, influenced by internal pressure. Given the significant risks associated with such accidents, it is crucial to study the behavior of fuel channels under LOCA conditions. This paper investigates the thermal performance of Indian PHWR under an oxidizing environment that simulates a late-phase accident scenario. The temperature profiles of the fuel element simulators, PT, and CT under steady-state conditions are obtained. A 37-element fuel bundle simulator is used, with the PT mounted eccentrically inside the CT, maintaining a 4 mm eccentricity.