Thermal-hydraulic Field Research Articles

Heat transfer enhancement of PCM in the triple-pipe helical-coiled latent heat thermal energy storage unit and complete melting time correlation

Secondary flows induced in helical-coiled pipes significantly enhance the thermal storage performance of latent heat thermal energy storage units. This paper presents a three-dimensional model of a triple-pipe helical-coiled system to investigate the melting characteristics of phase change material within it. A correlation for the complete melting time was developed based on the simulation results. The findings indicate that the thermal-hydraulic fields of the heat transfer fluids and phase change material deflect from the vertical central axis at cross-sections. The secondary flows within the inner pipe are more effective in enhancing melting than those in the outer pipe. Melting performance improves with increasing inner pipe diameter, coil diameter (from 100 mm to 160 mm), inclination angle, or heat transfer fluid temperature, or decreasing helical pitch. The inner pipe diameter, helical pitch and heat transfer fluid temperature exhibit more pronounced influences. An inner pipe diameter of 30 mm, a coil diameter of 160 mm, a helical pitch of 80 mm, an inclination angle of 90°, a heat transfer fluid temperature of 360 K, and a heat transfer fluid Reynolds number of 2000, respectively produce a faster melting process. The developed correlation for the triple-pipe helical-coiled pipe thermal storage unit accurately predicts 96 % of the 45 data within ±11 %.

Scaling in nuclear thermal hydraulics: Advances, applications and future perspectives in Spain

The demonstration of the safety, high performance and reliability of any man-made device relies heavily in experimental programs. For nuclear reactor safety demonstration, several experimental facilities are used for both testing the design and the validation of computational codes. In the field of thermal hydraulics, most experimental facilities are built at a smaller scale, thus the scaling analysis is a requirement in the licensing process of the nuclear power plant. This includes the analysis of the scaling capabilities of the computer codes, the scale and design differences with the licensed reactor and the methodologies applied for licensing.In collaboration with the Spanish regulatory Body, several Spanish university groups have contributed considerably to the development and application of scaling methodologies. This article provides an overview of the most relevant contributions presented in the last decades. The activities deal mainly with the use of system codes in scaling analysis. Specifically the works can be classified in the use of Kv scaling analyses, analysis of counterpart experiments, plant application and the scaling of uncertainties. Finally, the future challenges that will be addressed in relation to scaling are highlighted with emphasis to the contribution to the safe operation of current and future nuclear power plants.

On the Use of the Jacobian-Free Newton Krylov Method to Generate One-Group Discontinuity and Super Homogenization Factors for Full-Core Neutron Diffusion Simulations

Coupled Monte Carlo (MC) and thermal-hydraulic analysis is valuable as a design or reference tool but can be slow, especially when implemented in a Picard iteration. Previous work has developed a novel prediction block to achieve convergence with fewer MC simulations. The prediction works in two stages: (1) a surrogate-like model predicts macroscopic cross sections on the fly and (2) a reduced-order neutronic model predicts the flux response to the updated cross sections. The main challenge with the prediction block is that the reduced-order neutronic model cannot reproduce the spatial flux distribution with high fidelity. This paper investigates the well-established Jacobian-free Newton Krylov (JFNK) method to preserve equivalency between a homogeneous (nodal diffusion) solution and a high-fidelity transport (MC) solution. Instead of performing multiple computationally consuming MC simulations, the nonlinear iterative approach iterates on correction parameters, e.g., assembly discontinuity factors (ADFs) or super homogenization (SPH) factors, using unexpensive nodal solutions. The JFNK approach does not require additional overhead from the MC solver to generate flux tallies. Further, the approach iterates on diffusion solutions produced directly from a desired code, thus ensuring that the parameters are compatible with that code. The approach is applied to correct a nodal diffusion solution for a realistic three-dimensional pressurized water reactor core. The results obtained in the paper show that the method is very successful in reproducing the heterogeneous solution (up to 2.5% difference in assembly flux for SPH and 0.3% for ADFs) without needing to modify the source code of the nodal diffusion solver. In addition, the results show that ADFs yield the best agreement and are also stable (i.e., weakly varying) when thermal-hydraulic fields are perturbed.

Full scope scaling analysis and recommendations for PWR SGTR scenarios

Scaling studies were conducted in the field of nuclear thermal-hydraulics to analyze the scaling distortions in a Steam Generator Tube Rupture (SGTR) scenario. Test 4 of the NEA/OECD/ROSA-2 Large-Scale Test Facility (LSTF) was chosen as the basis for the analysis of scaling distortions. Firstly, a post test calculation of the experiment was performed with the RELAP5 system code along with a phenomenological assessment. This model was then scaled over a wide range of sizes to investigate the possible scaling distortions by means of the Power-to-Volume Scaling Tool (PVST) which is an in-house tool developed and validated by the Universitat Politècnica de Catalunya (UPC) for scaling nodalizations. The paper demonstrates that the main scaling distortion observed in Test 4 is attributed to the lack of preservation of the L/D ratio in the SGTR assembly nozzle. This rationale is quite important in the design of break units in experimental facilities to preserve the friction and the mass flows of the reference (prototype) plant and scenario. Subsequently, the paper presents a series of guidelines for minimizing scaling distortions and reports the results obtained by applying them, the proportional scaling distortion was reduced by more than tenfold. Additionally, the paper analyzes the RELAP5 models involved in the conservation of momentum equation for the SGTR assembly nozzle, and their potential scaling distortions.

Steady-state and transient simulations of the MTR research reactor using the high-fidelity N-TH coupling method

The tremendous advancement in computer technology makes it feasible to perform high-fidelity and multi-physics numerical simulations of reactors, allowing for a more accurate and comprehensive description of the phenomena that occur within nuclear reactors. The application of the high-fidelity coupled neutronic and thermal-hydraulic codes to execute the safety analysis and assessment of the fresh highly enriched uranium core of the IAEA-10 MW MTR research reactor is presented in this research. A three-dimensional coupled neutronic and thermal-hydraulic code is realized using the open-source C++ library OpenFOAM. The coupled code used in this study utilizes an internal coupling approach, where the identical program and grid system are utilized to solve both the neutron physical and thermal-hydraulic fields. The neutron physical field in the reactor is solved by the three-dimensional multi-group time-dependent neutron diffusion equations, and the CFD methodology is applied in solving the thermal-hydraulic governing equations. The global and local steady-state calculation results of MTR are obtained, which are in accordance with the results issued from the IAEA TECDOC and other previous relevant studies. The unprotected loss of flow accident transient condition is considered to conduct the simulation and analysis of the MTR research reactor. The evolutions of temporal and spatial distributions of important parameters of the MTR research reactor are acquired under the ULOFA transient condition, revealing the responses of the reactor core during the transient accident without SCRAM. The results obtained show that the high-fidelity N-TH coupling code can be utilized to analyze the safety of the MTR research reactor.

Fast reactor thermal hydraulics in the Dutch PIONEER program

The multi-year research program carried out by NRG and funded by the Dutch ministry of economic affairs and climate is called ‘Program for Innovation and cOmpetence development for NuclEar infrastructurE and Research’ (PIONEER). The program comprises seven themes, i.e. long term operation, nuclear modelling and simulation, nuclear safety and compliance, fuels & materials, radioactive waste management, radiation protection, and innovative nuclear systems. One of the pillars in the theme of innovative nuclear systems is fast reactor research, particularly in the field of thermal hydraulics. This paper provides an overview of all fast reactor thermal hydraulics activities in the program, covering development and validation of System Thermal Hydraulics (STH) and 3D (engineering as well as high-fidelity) Computational Fluid Dynamics (CFD) codes and simulation approaches. Applications range from fundamental turbulent heat transport to core, pool and system thermal hydraulics. With the recent improvements in computational infrastructure and power, also further developments of multi-scale and multi-physics computational approaches are being integrated in the PIONEER program. A generic coupling tool ‘myMuscle’ is under development which is introduced in this paper. Recent results and current developments are presented together with an outlook for the results to be expected at the end of the current multi-year program and beyond.

Recently revisited TH issues and associated R&D A French perspective

This paper presents an overview of some recent evolutions in the field of thermal-hydraulics (TH) within nuclear safety studies in France, therefore restricted to Pressurized Water Reactors (PWR) design, from the perspective of the French Technical Support Organization IRSN (Institut de Radioprotection et de Sûreté Nucléaire). These evolutions can concern either the hypothetical accident scenario (recently raised questions, modified rulemaking, new methodology) or the development of recent sources of knowledge (experimental programs, advanced numerical simulation tools).

The development of multi-physics approach with Monte Carlo and computational fluid dynamics coupling for reactor cores

Highly accurate solutions for neutronic and thermal hydraulic phenomena in reactor cores certainly improve the safety conclusions about the reactors. The coupling methodology of governing physics of these two fields by using Monte Carlo and computational fluid dynamics (CFD) methods was described in this study with its application to Istanbul Technical University (ITU) TRIGA Mark II Research Reactor. As a different approach, instead of not modelling top and bottom grid plates of the reactor that is the common practice for geometrical simplification, the effect of grid plates on flow hydrodynamics was included into the conjugate heat transfer calculations as boundary conditions. The investigation on the effect of core element temperature, cross sections, and S(α,β) thermal scattering data showed the need of proper coupling of neutronic and thermal hydraulic fields since 30–45% relative error on excess reactivity value of the ITU research reactor was observed. With the developed loose coupling model, 483 individual cross section and 483 different S(α,β) thermal scattering data were generated by using NJOY nuclear data generation code and MAKSXF and OTF utilization tools of MCNP6.2 code. The reactivity loss against power increase in research reactor value of this study showed 6% relative error with loose coupling comparing to maximum 50% without coupling.

RELAP5 validation for core makeup tank analysis based on separate effect tests

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during accident and has been widely used in the field of nuclear thermal-hydraulics and safety analysis. The core make-up tank (CMT) is a crucial device of the passive core cooling system in the AP series pressurized water reactor nuclear power plants. Since the CMT not only provides full-pressure injection but also triggers auto depressurization system by its level signal during small-break loss of coolant accident. Therefore, it is necessary to carry out CMT separate effect tests to verify the feasibility and accuracy of RELAP5/MOD 3.4 to simulate CMT. The previous studies found that the lower version of the RELAP5 program has problems in simulating CMT, and the applicability of the latest RELAP5/MOD3.4 has not been evaluated. In this paper, a CMT separate effect test facility were presented and then RELAP5/MOD 3.4 was used to establish a CMT test bench model to evaluate the applicability, and obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. It’s found that RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time and the flowrate of delayed drainage stage is not affected by the heat transfer coefficient of steam direct contact condensation. The simulation of delay time depends on experimental data and the CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model.

Thermal hydraulic modeling of solid-fueled nuclear thermal propulsion reactors part II: Full-core coupled neutronic and thermal hydraulic analysis

This paper focuses on the extension and application of the previously validated thermal–hydraulic engineering code ntpThermo to perform multi-channel analysis for full-core coupled neutronic and thermal–hydraulic simulations. This paper also introduces the Basilisk code, which enables coupled full-core neutronic and thermal–hydraulic simulations by integrating Serpent, the continuous energy Monte-Carlo neutron transport code, and ntpThermo within a flexible object-oriented framework. Today, limited data exists pertaining to coupled multiphysics analysis of low-enriched uranium nuclear thermal propulsion reactors. More specifically, the impact of thermal–hydraulic feedback on the neutronic solution for full core applications is not well quantified. In this paper full-core coupled neutronic and thermal–hydraulic simulations are performed to ascertain the importance of thermal–hydraulic feedback on the predicted reactivity and local power distributions. The reactor design analyzed adheres to the current industry ground rules in an effort to provide useful insights for the current industry design effort. The results in this paper demonstrate that thermal–hydraulic feedback must be considered to accurately predict reactivity; otherwise, errors on the order of hundreds to thousands of pcms can be observed. Additionally, neglecting thermal–hydraulic feedback can trigger inconsistencies between the radial power distribution and the fuel orificing pattern. A design approach that doesn’t account for thermal–hydraulic feedback can lead to a final design that may experience thermal failure via excessive fuel temperatures, and would require a reduction of reactor power. The latter will cause a subsequent decrease in exit propellant temperatures and thus specific impulse of the rocket engine. Sensitivity studies were conducted to determine the importance of specific thermal–hydraulic fields and revealed that the local moderator temperature has a notable impact on the local power distributions.

In Memoriam: Professor Pavel Leonidovich Kirillov (Aug. 20, 1927–Oct. 10, 2021)

Abstract Professor Pavel L. Kirillov has died on October 8th, 2021, on his 95th year after a life as a husband, father, and an internationally renowned scientist, researcher, and educator in the field of nuclear engineering, thermalhydraulics, heat transfer, and two-phase flow. He was a passionate and dedicated in everything, what he has done and leaves an incredible legacy to the profession. He was born on August 20th, 1927 in Russia, and received his M.A.Sc. degree in thermal physics in 1950 (Moscow Power-Engineering Institute (MPEI) (МосковскийЭнергетическйИнститут (МЭИ)), Faculty of Physics and Power Engineering (Физико-ЭнергетическийФакультет), Ph.D. and Doctor of Technical Sciences degrees in 1959 and 1969, respectively. Professor Pavel Leonidovich Kirillov was a respected technical leader, mentor, and friend to innumerable students, researchers, scientists, and engineers, and he will be sadly missed by all, who had the privilege to know him. He was an outstanding contributor in every aspect of his prolific work and career in the true traditions of technical excellence and critical thinking, and his irreplaceable loss is deeply felt worldwide.

A Freezing-Thawing Damage Characterization Method for Highway Subgrade in Seasonally Frozen Regions Based on Thermal-Hydraulic-Mechanical Coupling Model

Seasonally frozen soil where uneven freeze–thaw damage is a major cause of highway deterioration has attracted increased attention in China with the rapid development of infrastructure projects. Based on Darcy’s law of unsaturated soil seepage and heat conduction, the thermal–hydraulic–mechanical (THM) coupling model is established considering a variety of effects (i.e., ice–water phase transition, convective heat transfer, and ice blocking effect), and then the numerical solution of thermal–hydraulic fields of subgrade can be obtained. Then, a new concept, namely degree of freeze–thaw damage, is proposed by using the standard deviation of the ice content of subgrade during the annual freeze–thaw cycle. To analyze the freeze–thaw characteristics of highway subgrade, the model is applied in the monitored section of the Golmud to Nagqu portion of China National Highway G109. The results show that: (1) The hydrothermal field of subgrade has an obvious sunny–shady slopes effect, and its transverse distribution is not symmetrical; (2) the freeze–thaw damage area of subgrade obviously decreased under the insulation board measure; (3) under the combined anti-frost measures, the maximum frost heave amount of subgrade is significantly reduced. This study will provide references for the design of highway subgrades in seasonally frozen soil areas.

On the validation of BEPU methodologies through the simulation of integral experiments: Application to the PKL test facility

Validation is a key step in the field of thermal hydraulics and has played a crucial role in the implementation of system codes. These codes are then included in licensing methodologies being those either best estimate plus uncertainties (BEPU) or conservative approaches. However, only a small amount of efforts has been done on the validation of the methodologies as a whole. Such validation would imply the comparison of the bounding limits and the experimental data. Due to the lack of actual data of accidental situations, such validation should be carried out with data from integral test facility (ITF) experiments.The OECD/NEA PKL-4 test programme is investigating safety issues relevant for current pressurized water reactor (PWR) plants as well as for new PWR design concepts focusing on complex heat transfer mechanisms under two-phase flow. Within the project, a blind benchmark activity has been launched where an IBLOCA experiment has been simulated by 12 different organizations. The Universitat Politècnica de Catalunya (UPC) participated in the activity with a blind BEPU approach with the main objective of assessing the validity of the methodology as a whole.At first, the scalars that characterize the phenomenology of the particular scenario have been considered. Secondly, in order to broaden the extent of the validation, the results of the different participants have been interpreted as if they were blind experiments as well. Finally, different output parameters have been considered providing a matrix of results that is used to evaluate the validity of the methodology.

Theoretical Study on the Characteristics of Critical Heat Flux in Rectangular Channel of Natural Circulation under Motion Conditions

With the wide application of sea-based reactors, the impact of ocean conditions on the safety performance of reactors has gradually attracted attention. In this paper, by establishing the thermal hydraulic transient analysis model and the critical heat flux (CHF) model of natural circulation system, the CHF characteristics in the rectangular channel of natural self-feedback conditions under ocean conditions are studied. The results show that the additional acceleration field generated by ocean conditions will affect the thermal hydraulic parameters of the natural circulation system, that is, the external macroscopic thermal hydraulic field. On the other hand, the boiling crisis mechanism will be affected, that is, the force on the bubble and the thickness of the liquid film. Within the parameters of the study, ocean conditions have a great impact on CHF of natural circulation, and the maximum degradation of CHF is about 45%. The obtained analysis results are significant to the improvement of design and safety operation of the reactor system.

Numerical study of external inertia and compressible volume on nuclear coupled density wave oscillations in a supercritical pressure channel

In this study, the effect of external inertia and compressible volume on density wave oscillations (DWOs) is analyzed, for a virtual supercritical water reactor (SCWR), including the effect of neutron coupling on the thermal–hydraulic field variables. An existing mathematical model is extended for the intended purpose. The model developed is validated against the available experimental and numerical results. Next, the model is applied for the analysis of a pressure vessel SCWR which is similar to the design and operating conditions of the U.S. SCWR. Extensive numerical simulations are carried out to obtain marginal stability boundaries (MSBs) at different operating conditions. The simulation results indicate that the stability of the system decreases if any one of the external outlet inertia or the inlet compressible volume increases; whereas the increase in any one of the external inlet inertia or the outlet compressible volume, increases the stability of the system. Finally, a case study is undertaken to demonstrate how to stabilize a SCWR system with a large upstream compressible volume by increasing the inlet external inertia which, otherwise, would have been unstable.

Numerical investigation of heat transfer from a plane surface due to turbulent annular swirling jet impingement

The heat transfer characteristic of an annular turbulent swirling jet impingement on a heated plane surface is investigated numerically. The annular jet configuration causes instabilities and fluctuations in the flow. Depending on the combinations of different parameters, the annular jet impingement may have positive or negative effects on the heat transfer from the impingement surface. Swirl, on the other hand, introduces vorticity and fluid mixing in an impinging jet, which is desirable in some applications. The axisymmetric two-dimensional flow domain on a radial-axial plane is considered in the numerical model. Both non-swirling and swirling jet impingement is studied. The numerical computations are performed using the ANSYS Fluent CFD code. The Realizable k-ε turbulence model with enhanced wall treatment is used in the computation. The computational process is validated against other published data on similar flow configuration for non-swirling annular impinging jets. The flow and geometric parameters are the jet exit Reynolds number, Re (5000 to 25,000), the swirl strength Sw (0–0.77), the jet exit to the impingement surface distance, H (0.5–4.0), and the moderate blockage ratio of the annular jet, BR (0.4–0.6). The thermal-hydraulic field in the domain is computed for various combinations of these parameters. The effects of these parameters on the Nusselt number distribution on the impingement surface are analyzed. For short separation distance (H = 0.5), the swirling motion positively affects the overall heat transfer, and the average Nu is increased as high as 8% for certain combinations of Re, Sw, and BR; compared to the non-swirling annular jet impingement. For higher separation distances, the average and peak Nusselt number is initially reduced and then increased with increasing swirl strength.

Low-frequency modes in the fluid-structure interaction of a U-tube model for the steam generator in a PWR

Abstract In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

Velocimetry during depressurized conduction cooldown events in the HTTF

Abstract The High Temperature Test Facility (HTTF) is a quarter-scale integral-effect facility designed to study pressurized and depressurized conduction cooldowns (PCC and DCC, respectively) in high temperature, gas-cooled reactors (HTGR). This study focuses on DCC and aims to characterize the gas exchange between the reactor core and the reactor cavity to assess the risk for air ingress and subsequent potential for natural circulation core cooldowns. High-resolution velocity profile measurements are performed at the coolant pipe break location over a long time-scale using molecular tagging velocimetry. This non-intrusive, laser-based technique is applied for the first time in the field of nuclear thermal hydraulics, providing new insights for understanding the underlying phenomena at play during a DCC event. Such data are also valuable for validation of numerical simulations such as Reactor Excursion and Leak Analysis Program (RELAP). Results are reported here for various gas mixtures of helium (the coolant) and nitrogen (surrogate for air) to study the effect of density ratio. The short transient (30 s) lock-exchange phase is well captured with helium flow velocity between 0.8 and 2 m/s. The flow is shown to persist over much longer time-scales (hours) at a velocity of about 0.2 m/s, a likely consequence of gas mixing in the hot leg and in the lower plenum of the reactor. Flow visualization at the exit of the hot leg shows shear instabilities, which supports the hypothesis of significant mixing.

Analysis of the Thermophysical Processes and Operating Modes of Electrothermic Reactor Using a Computer Model

Based on the proposed algorithm for the computer simulation of thermophysical processes in an electrothermal ore-reduction reactor, the finite element method was used to analyze the regime parameters of the working volume. In this case, the computer calculation of the interconnected thermal-hydraulic fields has been carried out for the entire volume of the reactor. A comparison of the results of the computational experiment with the experimental data on the reactor survey and physical modeling data has shown quite satisfactory agreement. Based on an analysis of the reactor operation modes, regime maps have been presented that provide an idea of the change in technical and economic indicators depending on the control parameters. This information allows operators to know the trends in the behavior of the reactor when the regime changes.

Efficient simultaneous solution of multi-physics multi-scale nonlinear coupled system in HTR reactor based on nonlinear elimination method

Pebble bed HTR reactor core is a globally coupled system involving neutronics and thermal-hydraulic field. At the same time, the locally coupled relations exist in fuel sphere temperature, local neutron flux and pebble bed temperature. What’s more, several types of fuel spheres with different burnup and fission power must be treated separately. Due to the essential difference between local and global coupling properties, the method, named nonlinear elimination (NlEm), is developed in this paper to solve the multi-physics/multi-scale system of HTR in a tightly coupled, two-level nonlinear form. In the NlEm framework, the local coupled variables and the global coupled variables are treated separately at two different levels to enhance the computational efficiency. For comparison, the Jacobian-free Newton Krylov (JFNK) method and the Picard method are utilized to solve the multi-physics/multi-scale system respectively. The numerical results reveal that the computational efficiency of NlEm is 6.7 times higher than that of Picard method and 3.5 times higher than that of JFNK method for five transient cases of a two-dimension simplified reactor model. And with the increase of the computational scale and time step size, the speedup ratio of NlEm is further raised.