Coolant Loop Research Articles

Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig

Fuel rods in a pressurized water nuclear reactor (PWR) experience coolant flow-induced relative motion against the grid support features, a phenomenon called grid-to-rod fretting (GTRF), which may cause progressive fretting wear damage. Wear-through of the cladding on fuel rods will leak the radioactive fuel into the coolant loop to significantly increase the radiation level in PWR primary loop and often trigger expensive Post Irradiation Examination. In this study, a unique autoclave fretting test rig was designed and fabricated to allow studying GTRF using actual cladding and grid materials. Water temperature was up to 220 °C, which was designed to mimic the environment in an industry full-assembly reactor core simulator. The contact load (0.1–1 N), oscillation frequency (20–30 Hz) and stroke (50–150 µm), and work rate (1–3 mW) were defined based on simulations of GTRF in an actual reactor. Using this autoclave GTRF rig, tests were conducted to learn the effect of water temperature and to investigate the wear behavior of different cladding-grid material combinations currently used in actual PWRs, including Zr alloy claddings without and with pre-oxidation against Zr alloy and Inconel grids. Higher water temperature evidently increased the cladding wear, and pre-oxidation of the Zr alloy and/or using the Inconel grid effectively reduced the wear rate. Results were correlated with the data of a full-size industry reactor core simulator in terms of both wear rate and wear scar morphology.

Modeling and Simulation for Coolant Activity and Dose Rate Estimation in a Pool-Type Nuclear Research Reactor

Coolant activity in the primary coolant system and reactor pool in an open pool-type research reactor is very important in view of operational and radiological safety considerations. As pool water acts as the shielding medium to minimize the radiation dose at the top of the reactor pool, an estimation of the activities of radionuclides in pool water is essential to establish the safety of the operating personnel and researchers working at the reactor pool top. A system is provided to create a hot water layer (HWL) at the top of the pool by supplying water at a temperature more than that of the pool water so that the dose rate at the reactor pool top can be minimized. This HWL system helps in breaking the natural convection current of reactor pool water by maintaining a higher temperature at this layer so that high-density pool water below this layer cannot replace this low-density HWL. Therefore, pool water that is comparatively more radioactive will not be able to cross this HWL by convection. Hence, diffusion will be the only mechanism by which radioactivity can reach the pool top. So eliminating the convection current keeps the activity at this topmost layer of the pool at a minimum value. The estimation of the activity of the radioactive nuclides is required to assess the radiation field at different locations in the primary coolant loop for designing proper shielding requirements of the system. The radionuclides of interest are the activation products of aluminum (24Na, 27Mg, 28Al) and 41Ar. In this technical note, a transient code is presented for estimating the activity of radioactive nuclides in the coolant loop and reactor pool of a nuclear research reactor. The reduced activity level at the pool top is estimated considering the presence of the HWL at the top of the pool. It is observed that purification flow plays a major role on the activity level of radioactive nuclides in reactor pool water. The variation of the activity dose rate at the reactor pool top with purification flow is also discussed.

Design and Analysis of CFETR CSMC Cooling Loop

The central solenoid model coil (CSMC), a crucial component of China Fusion Energy Test Reactor, is currently in the design and manufacturing process. CSMC assembly consists of a number of steps such as winding pack, an outer NbTi coil, a middle Nb3Sn coil, an inner Nb3Sn coil, and a preload structure. The highest field of the model coil is 12 T with the maximum change rate of a magnetic field of the conductor is 1.5 T/s. In this paper, length of the cooling channels together with thermal–hydraulic parameters such as inlet pressure, temperature, and mass flow rate are optimized. In addition, thermal–hydraulic analysis for the cooling loop located in the worst condition of the model coil was conducted to recognize the temperature and mass flow rate change over time. The hydraulic model, the material properties, and the heat loads associated with analysis are given, and the results of the analysis are presented.

Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method

In detailed previous work by the authors, an innovative decay heat removal (DHR) system has been proposed and designed for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The current paper studies the inadvertent actuation of one DHR system train during I2S-LWR normal operation due to a false signal or operator action. The RELAP5 code is used to perform a one-dimensional study, and important thermal-hydraulic characteristics, including primary loop coolant flow rate, pressure, temperature, DHR primary-side flow rate, and coolant temperature, are achieved during this transient. Then, a detailed computational fluid dynamics simulation utilizing STARCCM+ is carried out to investigate the coolant mixing characteristics in the downcomer and lower plenum and obtain the local thermal-hydraulic conditions at the reactor core inlet. It is found that as a consequence of inadvertent DHR actuation, the maximum overcooling at the reactor core inlet is about 3 K, which would not result in significant reactivity insertion. Furthermore, a more severe transient of inadvertent DHR operation with intermediate loop break is studied, and the results show that this would not lead to more significant overcooling to the I2S-LWR core compared with inadvertent DHR operation without intermediate loop break. This work is an indispensable supplement for DHR system comprehensive assessment in the I2S-LWR project.

Thermal Characteristics of High Temperature Naturally Circulating Helium Cooling Loop

Abstract The paper deals with the process properties in terms of the heat transfer, i.e. the thermal performance of the thermal-process units within a helium loop intended for the testing of the decay heat removal (DHR) from the model of the gas-cooled fast reactor (GFR). The system is characterised by a natural circulation of helium, as a coolant, and assume the steady operating conditions of the circulation. The helium loop consists of four main components: the model of the gas-cooled fast reactor, the model of the heat exchanger for the decay heat removal, hot piping branch and cold piping branch. Using the thermal calculations, the thermal performance of the heat exchanger model and the thermal performance of the gas-cooled fast reactor model are determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Abstract A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the Primarkreislauf-Versuchsanlage (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

Study of tritium transport characteristics in a transportable fluoride-salt-cooled high-temperature reactor

Tritium management is one of the most critical issues that limit the development of fluoride-salt-cooled high-temperature reactor (FHR); therefore, it is important to figure out the tritium transport characteristics in FHRs. In this paper, 3 works concerning about tritium in FHR are conducted: first, the tritium transport characteristics in the primary loop of FHRs are introduced, including tritium production and speciation, the absorption and desorption by graphite, dissolution and diffusion in molten salt, and permeation through structural materials. Second, the physical and mathematical models are established for tritium transport characteristic analysis in a transportable FHR (TFHR). The tritium transport characteristic analysis code (TAPAS) for TFHR is developed and benchmarked. The results prove the fidelity and accuracy of TAPAS. Finally, the tritium transport characteristics in the TFHR are analyzed systematically by TAPAS. Three conclusions are obtained: (1) tritium in the primary coolant loop is mainly in the form of T2; (2) when TFHR operates at steady state, the permeation rate of T2 can be regarded as a constant (9.03 × 109 Bq ⋅ EFPD−1); and (3) 7Li enrichment and redox potential of molten salt have great influence on the tritium distribution. This work might provide contribution to the tritium control in FHRs.

Analysis of Radiation Transport due to Activated Coolant in the ITER Neutral Beam Injection Cell

Detailed spatial distributions of the biological dose rate due to a variety of sources are required for the design of the ITER tokamak facility to ensure that all radiological zoning limits are met. During operation, water in the Integrated loop of Blanket, Edge-localized mode and vertical stabilization coils, and Divertor (IBED) cooling system will be activated by plasma neutrons and will flow out of the bioshield through a complex system of pipes and heat exchangers. This paper discusses the methods used to characterize the biological dose rate outside the tokamak complex due to 16N gamma radiation emitted by the activated coolant in the Neutral Beam Injection (NBI) cell of the tokamak building.Activated coolant will enter the NBI cell through the IBED Primary Heat Transfer System (PHTS), and the NBI PHTS will also become activated due to radiation streaming through the NBI system. To properly characterize these gamma sources, the production of 16N, the decay of 16N, and the flow of activated water through the coolant loops were modeled. The impact of conservative approximations on the solution was also examined. Once the source due to activated coolant was calculated, the resulting biological dose rate outside the north wall of the NBI cell was determined through the use of sophisticated variance reduction techniques. The AutomateD VAriaNce reducTion Generator (ADVANTG) software implements methods developed specifically to provide highly effective variance reduction for complex radiation transport simulations such as those encountered with ITER. Using ADVANTG with the Monte Carlo N-particle (MCNP) radiation transport code, radiation responses were calculated on a fine spatial mesh with a high degree of statistical accuracy. Advanced visualization tools were also developed and used to determine pipe cell connectivity, to facilitate model checking, and to post-process the transport simulation results.

The Plasma Parameters and Geometry of Cool and Warm Active Region Loops

Abstract How the solar corona is heated to high temperatures remains an unsolved mystery in solar physics. In the present study we analyze observations of 50 whole active region loops taken with the Extreme-ultraviolet Imaging Spectrometer on board the Hinode satellite. Eleven loops were classified as cool loops (<1 MK) and 39 as warm loops (1–2 MK). We study their plasma parameters, such as densities, temperatures, filling factors, nonthermal velocities, and Doppler velocities. We combine spectroscopic analysis with linear force-free magnetic field extrapolation to derive the 3D structure and positioning of the loops, their lengths and heights, and the magnetic field strength along the loops. We use density-sensitive line pairs from Fe xii, Fe xiii, Si x, and Mg vii ions to obtain electron densities by taking special care of intensity background subtraction. The emission measure loci method is used to obtain the loop temperatures. We find that the loops are nearly isothermal along the line of sight. Their filling factors are between 8% and 89%. We also compare the observed parameters with the theoretical Rosner–Tucker–Vaiana (RTV) scaling law. We find that most of the loops are in an overpressure state relative to the RTV predictions. In a follow-up study, we will report a heating model of a parallel-cascade-based mechanism and will compare the model parameters with the loop plasma and structural parameters derived here.

Tritium aspects of the fusion nuclear science facility

The Fusion Nuclear Science Facility (FNSF) is intended to bridge the technical gaps from ITER to DEMO, and provide a fully nuclear environment for the advancement of materials and blanket science, among other missions. Though smaller (∼500MW fusion power) than the envisioned DEMO, our FNSF design still seeks to test DEMO-relevant components and technologies, and to that end one of its goals is to achieve a tritium breeding ratio near one using a dual-cooled lead lithium (DCLL) blanket. We explore here the migration of bred tritium through in-vessel components and coolant loops, which is exacerbated by the relatively low solubility of tritium in PbLi and the high operating temperatures envisioned in later phases of operation. Some features inherent to the design of the DCLL help limit tritium losses, namely the relatively high PbLi flow rates; coupled with efficient tritium extraction, the DCLL potentially achieves low tritium permeation losses without the use of permeation barriers. We outline here the design of such an extraction system, and some of the materials challenges that must be met in order to realize it. The efficacy of these and other design features, such as coaxial piping, is demonstrated via a series of parameter studies. We also investigate the effect of parameter uncertainties, and find that the imprecise knowledge of the tritium solubility in PbLi in particular leads to very disparate estimates of tritium losses. Depending on the experimental value used, losses range from ∼0.05 to over 6g of tritium per year. The former is acceptable based on estimated environmental releases and resultant dose to members of the public; the latter exceeds regulatory limits

The coolant purification system in DEMO: Interfaces and requirements

The blanket concepts investigated under the EUROfusion program rely on water or helium as primary or secondary coolant medium; main duty of the coolant is to recover the thermal power from the first wall and the blanket units and drive it into the Primary Heat Transfer System (PHTS). The coolant path goes through three different systems: the breeder, the tritium plant and the PHTS. In the breeder region, part of the produced tritium can permeate into the coolant loop, from the coolant into the steam generator and thus reach the environment. The Coolant Purification System (CPS), located inside the tritium plant, is responsible to recover tritium from the coolant and to control the coolant chemistry.The paper provides the status of the CPS design: the technology review was completed last year, while now the CPS interfaces and requirements are defined. Main issues arise from the definition of the amount of coolant to be routed inside the CPS due to the large uncertainties related to the amount of tritium permeation rate from blanket into coolant and to the presence and effectiveness of anti-permeation barriers.

Improving Battery Safety in Electric Vehicles By Monitoring Cooling Loop Temperature

As automotive industry continues its embrace of high capacity Li-ion batteries, increasing safety monitoring measures are needed. The use of Li-ion battery in electric vehicles presents itself far more strenuous conditions than other applications such as cell phones, cameras, computers and other stationary applications. Not only the power requirements for electric vehicles are high but also the operating environment is much more uncertain due to temperature variability, vibration, high rates of charge and discharge. To dissipate heat away from batteries, cooling loop of water is used. To protect a battery bank from failure, effective thermal management is absolutely essential. To this end, we have developed a temperature monitoring method which supplement existing function of cooling with that of temperature monitoring. We have achieved this by introducing aqueous based redox couple potassium ferri/ferro cyanide in water and using graphite electrodes in the path of flowing electrolyte. The ions reduce/oxidize based on temperature and produce voltage proportional to temperature difference. As a result, we are able to obtain temperature profile of the entire cooling loop and provide additional information to battery management system for reliable operation.

Thermal characterisation of compact heat exchangers for air heating and cooling in electric vehicles

The use of air conditioning in all-electric cars reduces their driving range by 33% in average. With the purpose of reducing the energy consumption of the vehicle and optimising the performance of the batteries, the mobile air-conditioning can be integrated with the temperature control system of the powertrain by means of a coolant loop. In such layouts, the air-to-coolant heat exchangers must operate efficiently in both air heating and cooling modes. Dynamic simulation tools comprising the entire thermal system are essential to assess its performance. In this context, fast but accurate models of the system components are required.This paper presents the thermal characterisation of a commercial compact louvered-fin flat-tube heat exchanger (heater core) for this novel application, based on an experimental campaign comprising 279 working points that reflect real air-conditioning (heating and cooling) working conditions. A general methodology to fit a single correlation of the global heat transfer coefficient for both dry and wet working conditions is explained. The semiempirical correlation developed is employed in a single-node model of the heat exchanger that requires minimal computation time. The present model predicts the heat transfer rate with an average deviation of 3.5% in the cases with dehumidification and 1.9% in the cases when the heat exchanger remains dry.

VECTOR MAGNETIC FIELD MEASUREMENTS ALONG A COOLED STEREO-IMAGED CORONAL LOOP

ABSTRACT The variation of the vector magnetic field along structures in the solar corona remains unmeasured. Using a unique combination of spectropolarimetry and stereoscopy, we infer and compare the vector magnetic field structure and three-dimensional morphology of an individuated coronal loop structure undergoing a thermal instability. We analyze spectropolarimetric data of the He i λ10830 triplet ( ) obtained at the Dunn Solar Telescope with the Facility Infrared Spectropolarimeter on 2011 September 19. Cool coronal loops are identified by their prominent drainage signatures in the He i data (redshifts up to 185 km s−1). Extinction of EUV background radiation along these loops is observed by both the Atmospheric Imaging Assembly on board the Solar Dynamics Observatory and the Extreme Ultraviolet Imager on board spacecraft A of the Solar Terrestrial Relations Observatory, and is used to stereoscopically triangulate the loop geometry up to heights of 70 Mm (0.1R Sun) above the solar surface. The He i polarized spectra along this loop exhibit signatures indicative of atomic-level polarization, as well as magnetic signatures through the Hanle and Zeeman effects. Spectropolarimetric inversions indicate that the magnetic field is generally oriented along the coronal loop axis, and provide the height dependence of the magnetic field intensity. The technique we demonstrate is a powerful one that may help better understand the thermodynamics of coronal fine-structure magnetism.

Direct Liquid Cooling in Low-Power Electrical Machines: Proof-of-Concept

This paper evaluates the feasibility of a direct liquid cooling approach in the thermal management of an axial flux permanent-magnet machine. To demonstrate the cooling method, a test motor was fitted with helical tooth coil windings formed from a hybrid conductor comprising a stainless steel coolant conduit tightly wrapped with a stranded Litz wire. The motor is a 100-kW permanent-magnet, axial-flux, double-stator, single-rotor machine. The proof of concept integrated the motor with a closed liquid coolant loop, appropriate instrumentation, and a data acquisition system. The general performance of the motor was examined at various power levels using polyalphaolefin oil as the cooling fluid. The results show the proposed cooling method to be feasible, and furthermore, to provide significant improvements to the machine thermal management.

Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments

Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corroded region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA − which suffered the least corrosion −retained some Al in the corroded region. The results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.

Design Summary of the Mark-I Pebble-Bed, Fluoride Salt–Cooled, High-Temperature Reactor Commercial Power Plant

The University of California, Berkeley (UCB), has developed a preconceptual design for a commercial pebble-bed (PB), fluoride salt–cooled, high-temperature reactor (FHR) (PB-FHR). The baseline design for this Mark-I PB-FHR (Mk1) plant is a 236-MW(thermal) reactor. The Mk1 uses a fluoride salt coolant with solid, coated-particle pebble fuel. The Mk1 design differs from earlier FHR designs because it uses a nuclear air-Brayton combined cycle designed to produce 100 MW(electric) of base-load electricity using a modified General Electric 7FB gas turbine. For peak electricity generation, the Mk1 has the ability to boost power output up to 242 MW(electric) using natural gas co-firing. The Mk1 uses direct heating of the power conversion fluid (air) with the primary coolant salt rather than using an intermediate coolant loop. By combining results from computational neutronics, thermal hydraulics, and pebble dynamics, UCB has developed a detailed design of the annular core and other key functional features. Both an active normal shutdown cooling system and a passive, natural-circulation-driven emergency decay heat removal system are included. Computational models of the FHR—validated using experimental data from the literature and from scaled thermal-hydraulic facilities—have led to a set of design criteria and system requirements for the Mk1 to operate safely and reliably. Three-dimensional, computer-aided-design models derived from the Mk1 design criteria are presented.

Thermal-hydraulic analysis of the improved TOPAZ-II power system using a heat pipe radiator

A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process.

A Numerical Model of a Reciprocating-Mechanism Driven Heat Loop for Two-Phase High Heat Flux Cooling

An effective design option for a cooling system is to use a two-phase pumped cooling loop to simultaneously satisfy the temperature uniformity and high heat flux requirements. A reciprocating-mechanism driven heat loop (RMDHL) is a novel heat transfer device that could attain a high heat transfer rate through a reciprocating flow of the two-phase working fluid inside the heat transfer device. Although the device has been tested and validated experimentally, analytical or numerical study has not been undertaken to understand its working mechanism and provide guidance for the device design. The objective of this paper is to develop a numerical model for the RMDHL to predict its operational performance under different working conditions. The developed numerical model has been successfully validated by the existing experimental data and will provide a powerful tool for the design and performance optimization of future RMDHLs. The study also reveals that the maximum velocity in the flow occurs near the wall rather than at the center of the pipe, as in the case of unidirectional steady flow. This higher velocity near the wall may help to explain the enhanced heat transfer of an RMDHL.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300–500 μm thick cladding.