High Performance Light Water Reactor Research Articles

Experimental and numerical study on the oxidation and melting behaviors of annular fuel rod under high temperature steam condition

In the safety analysis of nuclear reactor severe accident, the oxidation and melting behaviors of fuel rods under high temperature steam condition is a crucial issue that need to be concerned. As a new type of high-performance light water reactor nuclear fuel, the annular fuel is composed of outer cladding, annular fuel pellet, inner cladding and two coolant channels. At present, the oxidation and melting behaviors research of annular fuel under high temperature steam condition is insufficient. Thus, this paper builds a visual experimental platform for out-of-pile analysis, and carries out the high temperature oxidation experimental research of annular fuel rod based on real cladding material-Zircaloy. The transient temperature characteristic curve and visualization process of annular fuel rod under the conditions of high temperature melting, steam oxidation and re-flooding are obtained. In addition, this paper establishes a seven steps numerical analysis method for oxidation and melting behaviors analysis of annular fuel rod. Based on the three tests results, the accuracy of the model is verified, and the temperature prediction error is within 10%, which lays a foundation for the further calculation of the whole reactor core.

Concept of a Small Modular SCWR With Horizontal Fuel Assemblies

Abstract Based on the concept of the high performance light water reactor (HPLWR), an innovative concept of a small modular reactor is presented, which shall have smaller peak cladding temperatures than the HPLWR due to better coolant mixing, and which shall enable a passive residual heat removal by natural convection. The core is designed with horizontal fuel assemblies including an internal water channel to improve moderation. Supercritical water is heated up in seven steps, when running through these assemblies, and the outer reflector around the core is used as mixing channels. More design optimization and analyses will be needed, however, to assess the feasibility of such a concept.

Heat transfer analysis in sub-channels of rod bundle geometry with supercritical water

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 °C. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner sub-channel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.

Coupled Analysis of Thorium-based Fuels in the High-Performance Light Water Reactor Fuel Assembly

One of the six selected concepts to be part of Generation IV nuclear reactors is the Supercritical Light Water Cooled Reactor. The High-Performance Light Water Reactor (HPLWR) is the European version and it is a very promising design. In recent years, interest in the study of thorium-based fuel cycles has been renewed and its possibilities for current LWRs have been evaluated. The use of thorium-based fuels will be fundamental in the future sustainability of nuclear energy, since in addition to its abundance in nature, thorium has an important group of advantages. In this paper, performance of thorium-based fuels in the typical fuel assembly of the HPLWR reactor is evaluated, using a computational model based on CFD and Monte Carlo codes for the neutronic/thermal-hydraulic coupled analysis. The volumetric power density profiles, coolant temperature profiles, fuel temperature profiles and others are compared with those obtained for standard UO 2 fuel. When the thorium-based fuels are used, the obtained infinite multiplication coefficients are smaller than the value obtained when UO 2 is used, since the 232 Th isotope has a lower contribution to the multiplicative properties of the medium than 238 U. As a result, a difference of approximately 12 000 pcm was observed. The results verified that the HPLWR is a thermal reactor with a hard spectrum. There are no notable changes in the neutron spectrum if the mass fraction of thorium is slightly varied. With coupled analysis, the potential benefits of the utilization of thorium-based fuels were verified. Moreover, a significant temperature decrease by 136 K on the center line of the fuel elements was observed. When the mass fraction of thorium increases in the oxides mixture, the weighted average temperature on the fuel elements decreases .

Further Details of a Numerical Analysis on the Thermal Hydraulic Effect of Wrapped Wire Spacers in Fuel Bundle

Abstract The application of relatively simple and cheap wrapped wire spacer in the European supercritical water-cooled reactor (SCWR) (high-performance light water reactor (HPLWR)) has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss. The wires cause twisting flow in the fuel assembly, which means the coolant not only flows straight in the axial direction but also has a significant transverse velocity component, and strong mixing between neighboring subchannels occurs. The aim of this ongoing research is to numerically investigate the effect of wrapped wire spacers on thermal hydraulics of the turbulent coolant flow and its heat transfer in a small bundle of four fuel rods. One bare and six-wired geometries with varying wire pitches (1–6 turn(s) of wires) have been studied. It was found that the wires generate significant amount of transverse velocity, decrease the wall temperature, and increase the heat transfer coefficient mostly in corner subchannels which were the hottest in bare geometry. Thus, the presence of wires enhances heat transfer where it is most needed. Temperature hot spots with moderate values have been identified on the cladding wall of fuel rods. Based on the results, a technically optimal choice of number of wire turns from thermal hydraulic sense has been proposed.

Coupled Monte Carlo-burnup and CFD analysis of coated UN and UC fuels in an HPLWR fuel assembly

Ceramic nuclear fuels may have potential benefits to be considered as nuclear fuels in future Supercritical Water Reactors (SCWR), they offer the advantage of high thermal conductivity and high-density values compared to uranium dioxide. In recent years, uranium nitride (UN) and uranium carbide (UC) are being considered as viable options for alternative nuclear fuels in SCWR, but an important issue is related to their chemical reactivity with water and nickel, which forces the use of coatings for fuel pellets, reducing heavy metal volume and reducing their thermal and neutronic performance. The main purpose of the study is the analysis of the burnup dependent performance of coated ceramic fuels in the SCWR compared with conventional UO2 fuel. Particularly, it focuses on the study of the fuel centerline temperature and the cladding surface temperature. Furthermore, the evolution of the neutron multiplication factor, the fissile inventory ratios and the main heavy isotopes composition along burnup are analyzed and compared to asses proliferation resistance issues. The use of UN fuel coated with Zirconium Carbide (ZrC) and UC fuel coated with Titanium Nitride (TiN) in the fuel assembly of the High-Performance Light Water Reactor is analyzed, and the results are compared with those of the conventional UO2 fuel. The comparative study was conducted for a hypothetic operation time of 540 days. Due to the large coolant density variation along with the active height of the fuel assembly, coupled calculations are performed using the Monte Carlo MCNP6 code and the CFD code ANSYS-CFX 19, taking into account the capabilities of CFD codes to describe the heat transfer mechanisms at supercritical conditions. The main results show significantly lower values of the centerline temperature distribution in UN_ZrC and UC_TiN fuels compared to UO2 fuel. The fuel burnup values obtained at the end of the 540 days are practically the same, a slightly higher burnup is obtained for UC_TiN fuel because of its lower amount of heavy metals in the initial composition. The evolution of the fissile inventory is practically the same for all the analyzed fuel options, around 65% of fissile inventory is still present in the core at EoC. Total plutonium production shows a decrease of 1.85% (UN–ZrC) and 4.53% (UC–TiN) with respect to UO2 fuel. Coated ceramic fuels show a slightly improved proliferation resistance compared with UO2 fuel, since a slightly smaller fraction of 239Pu is obtained in their compositions.

Study on thermal hydraulic characteristics under startup of SCWR

To investigate the startup characteristics of the supercritical pressure water-cooled reactor (SCWR) system, a complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the high-performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with a circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, turbine, reheaters, steam extraction and heaters at each stage were obtained. The calculation results showed that the startup sequence and startup thermal parameters could fully meet the expectations: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays superheated; the core inlet temperature can reach 280 °C after the adoption of the high-pressure and low-pressure heaters; and the inlet pressure of the high-pressure turbine can be kept constant. During the startup process, the maximum cladding surface temperature (MCST) remains below the limit of 650 °C, suggesting that the entire startup procedure is safe and reliable.

Accident analysis of supercritical water reactors during startup

To analyze the typical characteristics of accidents occurring during the startup of a supercritical water reactor (SCWR), a comprehensive SCWR system model was established with the SCTRAN analysis code, based on models of the high-performance light-water reactor (HPLWR) steam cycle, SCWR circulation startup loop, passive safety systems, and CSR1000 core parameters. According to the sequence of the loss of coolant flow accident (LOFA), without a safety system and trigger signal of the safety system under the rated condition, a new trigger signal during the startup was designed and then used to assess “LOFA,” “uncontrolled CR withdrawal accident,” and “loss of coolant accident (LOCA)” during the startup process. The results show that the new trigger signal can ensure the effective and timely response of the safety system, as well as reactor safety during the startup process. Moreover, the maximum cladding surface temperature (MCST) of the reactor peaks (850 °C, well below the safety criterion of 1260 °C) at the end of the fourth stage of the startup process in the case of LOCA.

Coupled Monte Carlo-CFD analysis of heat transfer phenomena in a supercritical water reactor fuel assembly

In this paper coupled calculations with the CFD code ANSYS-CFX-19.0 and the Monte Carlo neutronics code MCNP6 were performed to analyze the heat transfer in supercritical water flowing through the typical fuel assembly of the high-performance light water reactor (HPLWR), in order to improve the characterization of the heat transfer phenomena in supercritical water under non-uniform axial heat flux distributions that is characteristic of this type of reactor. To check the capability of the CFX model to predict the thermal-hydraulic behavior of supercritical water, the computational results were compared with two experimental data. The Shitsmańs experiment in the presence of heat transfer deterioration (HTD) using four low-Re turbulence models (SST, k-ω, BSL-k-ω, and ω-Reynolds Stress) and the Wanǵs experiment in absence of HTD, using the low-Re-SST and the scalable-wall-function-SSG turbulence models. In the presence of the HTD phenomenon, results showed the high dependency of the wall temperature with the turbulence model and the turbulent Prandtl number selected. In the absence of HTD, both turbulence models studied adequately predicted the behavior of the wall temperature distribution.For the coupled neutronic/thermal-hydraulic analysis of the typical HPLWR fuel assembly, the low-Re SST turbulence model and the Prt = 1.5 were used. Different axial profiles of heat flux generated in the fuel rods were obtained for the different power values studied. For the analyzed conditions, the presence of HTD in the lower zone of the fuel assembly was observed. In addition, the results showed a strong non-uniformity of the circumferential surface cladding temperature distribution in the sub-channel located at the corner of the fuel assembly; a new curvature radius of the assembly box corner was proposed to obtain a well homogenized circumferential wall temperature distribution.

Design and test of an update of TRAC-BF1 for the modelling of supercritical water reactors

Some designs of nuclear power plants are bringing the working fluid to supercritical conditions (SCW) to improve their efficiency. But the evaluation of nuclear plants' response to transient events or postulated accidents is complex. In the academic environment, analysis is even more complex since the codes are generally restricted access. The possibility of improving routines and interfaces with the TRAC-BF1 code to allow the modelling of complete plants that use SCW during operational events or postulated accidents, was established as the main objective of this investigation. The new tools allow the analysis of models with steam cycles including supercritical regions such as the High-Performance Light Water Reactor, in which core nominal conditions were modelled for verification purposes. Also, a benchmark with experimental data was performed as part of the tests for new models. Results showed that our code modification is capable of modelling through the critical point without defining a pseudo-two-phase region.

Thermal analysis of ceramic nuclear fuels for the HPLWR

The High Performance Light Water Reactor (HPLWR) is one of the most promising designs of the Super-Critical Water Reactor (SCWR). The use of uranium nitride (UN) and uranium carbide (UC), as alternative nuclear fuels for the SCWR, offer the advantage of high thermal conductivity compared to uranium dioxide (UO2). For the analysis of these alternative nuclear fuels in SCWRs, some important design features must be considered. One of them is the porosity; the nuclear fuel is manufactured with different porosity values, as high as 20%, to reduce its hardness and swelling, influencing its thermal conductivity. Another issue is related to the chemical reactivity of UN and UC with water and nickel, which forces the use of coatings for fuel pellets. In this paper, a preliminary thermal-hydraulic analysis, using CFD codes, of the HPLWR fuel assembly, focused on the thermal behavior of ceramic fuel, was carried out. The use of UN coated with zirconium carbide layers and UC coated with titanium nitride layers were analyzed. The impact of porosity induced changes in fuel temperature profiles was analyzed. The radial and axial fuel temperature distributions were obtained for all cases. It was found that the maximum temperature values obtained using UN and UC, both coated and uncoated, were lower than those obtained with UO2. It was found that the porosity changes on the proposed fuels have a small influence on the maximum fuel temperature.

Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Since 2002 a generation IV nuclear energy system development project has been initiated to enhance the future role of nuclear energy systems. Challenging technology goals for generation IV nuclear reactor systems were defined in 4 areas: sustainability, economics, safety and reliability and proliferation resistance. Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The high-performance light water reactor (HPLWR) is the European version of the supercritical-pressure water cooled reactor (SCWR). The aim of this study is to reliability analysis of heavy water emergency core cooling system (ECCS) (instead of light water ECCS) in HPLWRs during LOCA (loss of coolant accident) through neutronics calculations and applied feedback.. Reliability analysis (neutronic calculations) are performed using WIMS and CITATION codes. Changes in neutronic parameters, such as neutron cross-sections, effective multiplication factor and power peaking factor during LOCA, are discussed by injecting different concentrations of the heavy water coolant by ECCS.

New high-performance light water reactor core concept with mixed cycle length operation

New high-performance light water reactor core concept with mixed cycle length operation

CFD analysis of thermal-hydraulic behaviour of the high performance light water reactor fuel assembly

The High-Performance Light Water Reactor (HPLWR) is the European Supercritical Water-cooled Reactor design. In this paper, a thermal-hydraulic study of the HPLWR fuel assembly using CFD codes was carried out. The capability of the Reynolds Stress model of Speziale (SSG) and the k-ω Shear Stress Transport model (SST) for predicting the supercritical water heat transfer was evaluated. The axial temperature distributions of the fuel, cladding, coolant and moderator in the fuel assembly were obtained. Numerical results of the fuel temperature profiles were compared with that obtained by Waata (2006) and a good agreement was achieved. The cladding surface temperature profiles calculated with SSG and SST turbulence models are below the prescribed limits; however, hot spots in one sub-channel were found. The difference in the average thermal-hydraulic properties of the supercritical water calculated with SSG and SST was negligible. The fuel and cladding surface temperatures are higher when using the SST model.

CFD analysis of thermal-hydraulic behaviour of the high performance light water reactor fuel assembly

The High-Performance Light Water Reactor (HPLWR) is the European Supercritical Water-cooled Reactor design. In this paper, a thermal-hydraulic study of the HPLWR fuel assembly using CFD codes was carried out. The capability of the Reynolds Stress model of Speziale (SSG) and the k-ω Shear Stress Transport model (SST) for predicting the supercritical water heat transfer was evaluated. The axial temperature distributions of the fuel, cladding, coolant and moderator in the fuel assembly were obtained. Numerical results of the fuel temperature profiles were compared with that obtained by Waata (2006) and a good agreement was achieved. The cladding surface temperature profiles calculated with SSG and SST turbulence models are below the prescribed limits; however, hot spots in one sub-channel were found. The difference in the average thermal-hydraulic properties of the supercritical water calculated with SSG and SST was negligible. The fuel and cladding surface temperatures are higher when using the SST model.

Development of coupled neutronics/thermal–hydraulics test case for HPLWR

The High-Performance Light Water Reactor (HPLWR) is the European concept of a supercritical water reactor (SCWR) which is one of the most promising and innovative designs of the Generation IV nuclear reactor concepts. The thermal–hydraulics behavior of supercritical water is significantly different from water at sub-critical pressure because of the difference in the specific heat value. Coupled analysis of HPLWR assembly neutronics and thermal–hydraulics has become important because of the strong influence of the water density on the neutron spectrum and power distribution. Programs MCU (Monte-Carlo Universal) and ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) were used for better estimation of power and temperature distribution in HPLWR assembly.

NEUTRONIC INVESTIGATION OF SEMI-HEAVY WATER APPLICATION IN HPLWR NEW FLOW PATTERN

The high-performance light-water reactor (HPLWR) is the European version of the supercritical-pressure water-cooled reactor. Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The variable moderation high performance light-water reactor (VMHWR) is the new version of HPLWR conceptual design. In this design, the fluid passing through the moderator channels and the assembly gaps (known as a moderator) flows in channels detached from coolants. This paper presents a neutronic analysis of semi-heavy water in VMHWR. Neutronic calculations were made through WIMSD5 and CITATION codes. Neutronic parameters such as effective multiplication factor, power peaking factor, and the effect of semi-heavy water on these parameters were investigated.

Axial enrichment profile in advance nuclear energy power plant at supercritical-pressures

Abstract The High-Performance Light Water Reactor (HPLWR) is the European version of the advance nuclear energy power plant at Supercritical-pressure. A light water reactor at supercritical pressure, being currently under design, is the new generation of nuclear reactors. The aim of this study is to predict the HPLWR neutronic behavior of the axial enrichment profile with an average enrichment of 5 w/o U-235. Neutronic calculations are performed using WIMS and CITATION codes. Changes in neutronic parameter, such as Power Peaking Factor (PPF) are discussed in this paper.

Design of an In-Pile SCWR Fuel Qualification Test Loop

In the development of the supercritical water-cooled reactor (SCWR), an in-pile fuel assembly test loop has been designed within the framework of the joint Chinese–European project, called SCWR-FQT (Fuel Qualification Test). This paper presents the basic design of the loop with its auxiliary and safety systems, which has been examined in detail by thermal-hydraulic analyses in order to achieve operation of the loop above the thermodynamic critical point of water (374°C, 22.1 MPa) and checked by stress analyses to assure safe operation. The designed experimental loop for fuel qualification in supercritical water consists of a closed pressurized water circuit with forced circulation of the coolant through the test section—the active channel which is intended to be installed into the existing research pool-type reactor LVR-15. The active channel will be operated at temperatures and pressures which are typical for the high-performance light water reactor (HPLWR). A thick-walled pressure tube made from austenitic stainless steel, which is able to withstand the high system pressure, encloses the active channel. It contains four fuel rods with UO2 (enrichment of 19.7% U235) with a total heating power of ∼64 kW and a recuperator in order to achieve hot channel conditions as they are expected to occur in the evaporator of the HPLWR. The internal flow is realized so as to prevent the creep condition of the pressure tube. An internal U-tube cooler serves as heat sink and is connected to the secondary circuit. The entire active channel is isolated from water of the reactor pool by an air gap between the pressure tube and an aluminum displacer. The test section with fuel is connected to a 300°C closed loop and to a primary pump located outside the reactor building as well as safety systems and auxiliary systems, such as purification and measurement circuits, which are all connected with the primary circuit.

European Project “Supercritical Water Reactor–Fuel Qualification Test”: Overview, Results, Lessons Learned, and Future Outlook

The supercritical water reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). One of the major challenges in the development of a SCWR is to develop materials for the fuel and core structures that will be sufficiently corrosion resistant to withstand supercritical water conditions and to gain thermal-hydraulic experimental data that could be used for further improvement of heat transfer predictions in the supercritical region by numerical codes. Previously, core, reactor, and plant design concepts of the European high-performance light water reactor (HPLWR) have been worked out in great detail. As the next step, it has been proposed to carry out a fuel qualification test (FQT) of a small-scale fuel assembly in a research reactor under typical prototype conditions. Design and licensing of an experimental facility for the FQT, including the small-scale fuel assembly, the required coolant loop with supercritical water, and safety and auxiliary systems, was the scope of the recently concluded project “Supercritical Water Reactor–Fuel Qualification Test” (SCWR-FQT) described here. This project was a collaborative project cofunded by the European Commission, which took advantage of a Chinese–European collaboration, in which China offered an electrically heated out-of-pile loop for testing of fuel bundles. The design of the facility, especially of the test section with the fuel assembly, and the most important results of steady-state and safety analyses are presented. Material test results of the stainless steels considered for the fuel cladding are briefly summarized. Finally, important outcomes and lessons learned in the “Education and Training” and “Management” work packages are presented.