Liquid metal thermal hydraulics R&D at European scale:achievements and prospects

A significant role for a future nuclear carbonfree energy production is attributed to fast reactors,mostly employing a liquid metal as a coolant.This paper summarizes the efforts that have been undertaken in collaborative projects sponsored by the European Commission in the past 20 years in the fields of liquid-metal heat transfer modeling,fuel assembly and core thermal hydraulics,pool and system thermal hydraulics,and establishment of best practice guidelines and verification,validation,and uncertainty quantification(UQ).The achievements in these fields will be presented along with the prospects on topics which will be studied collaboratively in Europe in the years to come.These prospects include further development of heat transfer models for applied computational fluid dynamics(CFD),further analysis of the consequences of fuel assembly blockages on coolant flow and temperature,analysis of the thermal hydraulic effects in deformed fuel assemblies,extended validation of three-dimensional pool thermal hydraulic CFD models,and further development and validation of multi-scale system thermal hydraulic methods.

An improved analysis method for assessing the nuclear-heating impact on the stability of toroidal field magnets in fusion reactors

The superconducting magnet system of a fusion reactor plays a vital role in plasma confinement,a process that can be dis-rupted by various operational factors.A critical parameter for evaluating the temperature margin of superconducting magnets during normal operation is the nuclear heating caused by D-T neutrons.This study investigates the impact of nuclear heat-ing on a superconducting magnet system by employing an improved analysis method that combines neutronics and thermal hydraulics.In the magnet system,toroidal field(TF)magnets are positioned closest to the plasma and bear the highest nuclear-heat load,making them prime candidates for evaluating the influence of nuclear heating on stability.To enhance the modeling accuracy and facilitate design modifications,a parametric TF model that incorporates heterogeneity is established to expedite the optimization design process and enhance the accuracy of the computations.A comparative analysis with a homogeneous TF model reveals that the heterogeneous model improves accuracy by over 12%.Considering factors such as heat load,magnetic-field strength,and cooling conditions,the cooling circuit facing the most severe conditions is selected to calculate the temperature of the superconductor.This selection streamlines the workload associated with thermal-hydraulic analysis.This approach enables a more efficient and precise evaluation of the temperature margin of TF magnets.Moreover,it offers insights that can guide the optimization of both the structure and cooling strategy of superconducting magnet systems.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

Core and blanket thermal-hydraulic analysis of a molten salt fast reactor based on coupling of OpenMC and OpenFOAM

In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released from the fuel salt and transferred to the second loop by fuel salt circulation.Therefore,the MSFR is characterized by strong interaction between the neutronics and the thermal hydraulics.Moreover,recirculation flow occurs,and nuclear heat is accumulated near the fertile blanket,which significantly affects both the flow and the temperature fields in the core.In this work,to further optimize the conceptual geometric design of the MSFR,three geometries of the core and fertile blanket are proposed,and the thermal-hydraulic characteristics,including the three-dimensional flow and temperature fields of the fuel and fertile salts,are simulated and analyzed using a coupling scheme between the open source codes OpenMC and OpenFOAM.The numerical results indicate that a flatter core temperature distribution can be obtained and the hot spot and flow stagnation zones that appear in the upper and lower parts of the core center near the reflector can be eliminated by curving both the top and bottom walls of the core.Moreover,eight cooling loops with a total flow rate of0.0555 m3 s-1 ensur an acceptable temperature distribusure an acceptable temperature distribution in the fertile blanket.

Development of CONTHAC-3D and hydrogen distribution analysis of HPR1000

An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the containment.CONTHAC-3D was developed using multiple models to simulate the features of the proprietary systems and equipment of HPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1%.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the accident.The amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,raising the installation height of these recombiners appropriately is recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.

Thermal–hydraulic analysis of space nuclear reactor TOPAZ-Ⅱ with modified RELAP5

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

Thermal hydraulic characteristics of helical coil once-through steam generator under ocean conditions

Owing to its advantages of high heat transfer efficiency and compactness, the helical coil once-through steam generator(HCOTSG) can be used in floating nuclear power plants and has been widely used in the design of small modular reactors. The helical tubular geometric structure of the HCOTSG allows heat transfer and local flow changes to occur under complex ocean conditions. In this study, theoretical models of ocean conditions are added to the RELAP5/MOD3.3 code and verified. Using the modified RELAP5 code, the thermal–hydraulic characteristics of the HCOTSG under ocean conditions are simulated. The results show that under rolling conditions, the flow oscillation amplitudes of the single liquid-phase, twophase flow, and single gas-phase regions are different. A circular change in the horizontal position of the helical tube causes the fluctuation of the parameters to change periodically. A phase difference of approximately 3.9 s at a flow rate of 23 kg/s is observed in the flow fluctuation along the axial direction. The driving force, period, and amplitude of rolling significantly affect the flow fluctuation in the HCOTSG. In natural circulation, the flow in the HCOTSG is complex, and the primary-side flow fluctuation can reduce the trough of the flow oscillation at the helical tube by approximately 24.3%.

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

The water-cooled ceramic breeder blanket（WCCB） is one of the blanket candidates for China fusion engineering test reactor（CFETR）.In order to improve power generation efficiency and tritium breeding ratio,WCCB with superheated steam is under development.The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions.In this paper,the coolant flow scheme was designed and one self-developed analytical program was developed,based on a theoretical heat transfer model and empirical correlations.Employing this program,the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions.The results indicated that the superheated steam water-cooled blanket is feasible.

Hydraulic and Thermal Calculation and Analysis of ITER Shield Block Module

ITER blanket design has progressed significantly since 2001, which resulted in a reduction in cost and an increase in performance with respect to FDR 2001. One of the most important improvements is the new coolant flow configuration in the shield block （ SB ） . In the current design TM, the cooling circuit in the SB is a matrix of radial holes which are arranged in eight poloidal rows. The rows are fed in parallel by front headers and back drilled collectors, and merge in four couples through the front header. These four couples of rows are linked in series by transverse holes. In the current design, a special shape of flow driver is mounted inside the radial hole, and coolant flows through clearance between the driver and drilled radial hole, which allows achieving a high coolant velocity,

Scaling Analysis of Thermal-Hydraulics for Steam Generator Passive Heat Removal System

Steam generator passive heat removal system(SG-PHRS) is used as a passively safe mode to provide decay heat removal in some advanced pressurized water reactors. Due to the structure characteristics of steam generator(SG), there are two natural circulation loops coupling in SG-PHRS in case of a safety-related event. The existing natural circulation scaling criteria could be used to simulate the natural circulation inside SG. Two-phase natural circulation loop is studied carefully, and the dominant effects of SG on behaviors of natural circulation in passive heat removal system are presented. Based on the understanding of SG-PHRS operation, system pressure transient scaling and two-phase natural circulation scaling are analyzed by establishing the relevant continuity,integral momentum and energy equations in one-dimensional area-averaged forms. With modified equations,similarity criteria for SG-PHRS are obtained for engineering application. In addition, equal height simulation and reduced height simulation are studied.

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module（TBM） program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R＆D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices（ITER relevant and DEMO）.The Indian Lead–Lithium Cooled Ceramic Breeder（LLCB） blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder（HCCB） is an alternative blanket concept that consists of lithium titanate（Li_2TiO_3） as ceramic breeder（CB） material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

Components Qualification for the Safe Operation of Nuclear Power Plants

AREVA operates a world-wide unique thermal hydraulic platform to ensure high safety standards in the nuclear industries. This platform is operated as an accredited test and inspection body according to ISO 17025 and 17020 to grant a high and independently confmned quality standard. The accreditation also ensures the independency of the organization and confidentiality to the individual stakeholders, for example research centers, utilities, components suppliers, engineering companies and vendors. Especially for nuclear power plants, it is very relevant to consider that reliability depends on the integrity of its components during its life time-from design through construction, operation and maintenance. For example, a typical NPP （nuclear power plant） has 1,000 to 2,000 large valves and 7,500 to 12,500 small valves, of which about 200 to 400 are designated Safety Class 1. The qualification of these Safety Class 1 components is relevant for reactor new builds but also for installed plants. This paper explains newly established qualification tasks, the corresponding testing infrastructure, and the state of the art of testing technology. By way of example, the paper describes the program and possible sequence of qualifying NPP safety-related components.

Geotechnical characterization of peat-based landfill cover materials

Natural methane （CH4） oxidation that is carried out through the use of landfill covers （biocovers） is a promising method for reducing CH4 emissions from landfills. Previous studies on peat-based landfill covers have mainly focused on their biochemical properties （e.g. CH4 oxidation capacity）. However, the utilization of peat as a cover material also requires a solid understanding of its geotechnical properties （thermal, hydraulic, and mechanical）, which are critical to the performance of any biocover. Therefore, the objective of this context is to investigate and assess the geotechnical properties of peat-based cover materials （peat, peat–sand mixture）, including compaction, consolidation, and hydraulic and thermal conductivities. The studied materials show high compressibility to the increase of vertical stress, with compression index （Cc） values ranging from 0.16 to 0.358. The compressibility is a function of sand content such that the peat–sand mixture （1：3） has the lowest Cc value. Both the thermal and hydraulic conductivities are functions of moisture content, dry density, and sand content. The hydraulic conductivity varies from 1.74 × 10^-9 m/s to 7.35 × 10^-9 m/s, and increases with the increase in sand content. The thermal conductivity of the studied samples varies between 0.54 W/（m K） and 1.41 W/（m K） and it increases with the increases in moisture and sand contents. Increases in sand content generally increase the mechanical behavior of peat-based covers; however, they also cause relatively high hydraulic and thermal conductivities which are not favored properties for biocovers.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

Ensuring the possibility of using thorium as a fuel in a pressurized water reactor(PWR)

The possibility of utilizing thorium as a fuel in a pressurized water reactor(PWR)has been proven from the neutronic perspective in our previously published work without assessing the thermal hydraulic(TH)and solid structure performances.Therefore,the TH and solid structure performances must be studied to confirm these results and ensure the possibility of using a thorium-based fuel as an excellent accident-tolerant fuel.The TH and solid structure performances of thorium-based fuels were investigated and compared with those of UO_(2).The radial and axial power peaking factors(PPFs)for UO_(2),(^(232)Th,^(235)U)O_(2),and(^(232)Th,^(233)U)O_(2)were examined with a PWR assembly to determine the total PPF of each one.Both Gd_(2)O_(3)and Er_(2)O_(3)were tested as burnable absorbers(BAs)to manage the excess reactivity at the beginning of the fuel cycle(BOC)and reduce the total PPF.Er_(2)O_(3)resulted in a more significant reduction to the total PPF and,therefore,a greater reduction to the temperature distribution compared to Gd_(2)O_(3).Given these results,we analyzed the effects of adding Er_(2)O_(3)to thorium-based fuels on their TH and solid structure performances.

Preliminary Design and Analysis of ITER In-Wall Shielding

ITER in-wall shielding （IIS） is situated between the doubled shells of the ITER Vacuum Vessel （IVV）. Its main functions are applied in shielding neutron, gamma-ray and toroidal field ripple reduction. The structure of IIS has been modelled according to the IVV design criteria which has been updated by the ITER team （IT）. Static analysis and thermal expansion analysis were performed for the structure. Thermal-hydraulic analysis verified the heat removal capability and resulting temperature, pressure, and velocity changes in the coolant flow. Consequently, our design work is possibly suitable as a reference for IT＇s updated or final design in its next step.

Energy modeling and optimization of building condenser water systems with all-variable speed pumps and tower fans:A case study

The emergence of building condenser water systems with all-variable speed pumps and tower fans allows for increased efficiency and flexibility of chiller plants in partial load operation but also increases the control complexity of condenser water systems.This study aims to develop an integrated modeling technique for evaluating and optimizing the energy performance of such a condenser water system.The proposed system model is based on the semi-physical semi-empirical chiller,pump,and cooling tower models,with capabilities of fully considering the hydraulic and thermal interactions in the condenser water loop,being solved analytically and much faster than iterative solvers and supporting the explicit optimization of the pump and tower fan frequency.A mathematical approach,based on the system model and constrained optimization technique,is subsequently established to evaluate the energy performance of a typical dual setpoint-based variable speed strategy and find its energy-saving potential and most efficient operation by jointly optimizing pumps and tower fans.An all-variable speed chiller plant from Wuhan,China,is used for a case study to validate the system model’s accuracy and explore its applicability.The results showed that the system model can accurately simulate the condenser water system’s performance under various operating conditions.By optimizing the frequencies of pumps and tower fans,the total system energy consumption can be reduced by 12%–13%compared to the fixed dual setpoint-based strategy with range and approach setpoints of 4℃and 2℃.In contrast,the energy-saving potential of optimizing the cooling tower sequencing is insignificant.A simple joint speed control method for optimizing the pumps and tower fans emerged,i.e.,the optimal pump and fan frequency are linearly correlated(if both are non-extremes)and depend on the chiller part load ratio only,irrespective of the ambient wet-bulb temperature and chilled water supply temperature.It was also found that the oversizing issue has further limited the energy-saving space of the studied system and results in the range and approach setpoints being inaccessible.The study’s findings can serve as references to the operation optimization of all-variable speed condenser water systems in the future.

Thermal Hydraulic Stability in a Coaxial Thermosyphon

The heat transfer and thermal hydraulic stability in a two-phase thermosyphon with coaxial riser and down-comer has been experimentally investigated and theoretically analyzed to facilitate its application in cold neutron source. The flow in a coaxial thermosyphon was studied experimentally for a variety of heat- ing rates, transfer tube lengths, charge capacities, and area ratios. A numerical analysis of the hydraulic balance between the driving pressure head and the resistance loss has also been performed. The results show that the presented coaxial thermosyphon has dynamic performance advantages relative to natural cir- culation in a boiling water reactor.

Thermal and hydraulic characteristics of a large-scaled parabolic trough solar field (PTSF) under cloud passages

To better understand the characteristics of a large-scaled parabolic trough solar field(PTSF)under cloud passages,a novel method which combines a closed-loop thermal hydraulic model(CLTHM)and cloud vector(CV)is developed.Besides,the CLTHM is established and validated based on a pilot plant.Moreover,some key parameters which are used to characterize a typical PTSF and CV are presented for further simulation.Furthermore,two sets of results simulated by the CLTHM are compared and discussed.One set deals with cloud passages by the CV,while the other by the traditionally distributed weather stations(DWSs).Because of considering the solar irradiance distribution in a more detailed and realistically way,compared with the distributed weather station(DWS)simulation,all essential parameters,such as the total flowrate,flow distribution,outlet temperature,thermal and exergetic efficiency,and exergetic destruction tend to be more precise and smoother in the CV simulation.For example,for the runner outlet temperature,which is the most crucial parameter for a running PTSF,the maximum relative error reaches−15%in the comparison.In addition,the mechanism of thermal and hydraulic unbalance caused by cloud passages are explained based on the simulation.

A Numerical Model Prediction for Boiling Multi Channel Flow Rate Distribution andApplication in 600MW Supercritical Variable－Pressure Once－Through Boiler with Vertical Tube Coils

Flow rate distribution is important in a multi channel system when the flow is heated non-uniformly.This paper describes a steady state approach for obtaining the flow distribution among various tubes of complex multi channel system. Based on the present approach,a program has been developed,which is directly applied in thermal hydraulic design and investigation of 600MW supercritical variable-pressure once through boiler.