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The joints connecting the ITER magnet busbars and coils utilize the twin-box “shaking hands” concept: inside a helium leak tight box, a bare cable is pressed in an indium-tinned copper base. To form the joint, two boxes are tightly compressed against each other on the copper side with a layer of indium in between. This concept was chosen to address the different requirements of the joints: to provide low electrical resistance without degradation and need of maintenance during ITER lifecycle, to sustain cyclic electromagnetic and pressure loads, to be easily dismountable in case of failure, to be tolerant to manufacturing and assembly misalignments, to provide low coolant flow impedance, and to limit ac losses and at the same time to facilitate current redistribution in the busbars. This paper presents the design; the key results of the electromagnetic, thermal, and stress analysis; and the major manufacturing and qualification steps. The latter includes cryogenic fatigue test of the joint welds and joint resistance measurements.

The results of experimental investigations of local hydrodynamics of a coolant flow in fuel rod assembly (FA) of KLT-40C reactor behind a plate spacer grid have been presented. The investigations were carried out on an aerodynamic rig using the gas-phase diffusive tracer test. An analysis of spatial distribution of absolute flow velocity projections and distribution of tracer concentration allowed specifying a coolant flow pattern behind the plate spacer grid of the FA. On the basis of obtained experimental data the recommendations were provided to specify procedures for determining the coolant flow rates for the programs of cell-wise calculation of a core zone of KLT-40C reactor. Investigation results were accepted for the practical use in JSC “OKBM Afrikantov” to assess heat engineering reliability of cores of KLT-40C reactor and were included in a database for verification of CFD programs (CFD-codes).

Test results for models of particle coagulation and settling on a surface in an aerosol module in analytical tests and experiments are presented. A module is being developed for modeling the dynamics of polydisperse multicomponent aerosols in the first loop of VVER during a serious accident with emission of fission products from fuel elements into the coolant flow.

The cooling and solidification of a continuous-cast ingot in transient casting conditions are studied by simulation. The influence of the method of controlling the secondary cooling on the thermal state of the ingot is shown. Different approaches to the control of the coolant flow rates in various operations are compared.

The paper presents results of experimental research, the purpose of which was to simulate the phenomenon of mixing of loop coolant flows inside the model of the pressure chamber of the water-cooled reactor. The study was conducted at the high-pressure aerodynamic test bench of NNSTU n.a. R.E. Alekseev. The scale model consisted of elements characteristic of water-cooled nuclear reactors (lowering annular channel, lower pressure chamber). Experimental studies were carried out in the range of Reynolds numbers from 20,000 to 50,000. The axial velocity field was studied at the entrance to the reactor core simulator, using a pneumometric research probe. The temperature field was simulated using the impurity diffusion method by introducing a contrast tracer into one of the loops of the model. Propane was used as a passive tracer. Tracer concentration at the inlet core simulator allowed to evaluate the degree of flow mixing and thereby studied the direction of coolant flow. As a result of the studies, the spatial distribution of the tracer concentration in the coolant flow in the lower annular channel, as well as in the lower pressure chamber, was determined. It was shown that in the lower pressure chamber, an axial central vortex influences the mixing intensity. The estimation of the parameters of mixing of the passive tracer in the model of reactor pressure was given.

A modernized experimental setup diagram is presented for studying the cooling of a high-temperature surface with a dispersed coolant flow. The results of an experimental study of unsteady heat removal from a target with a temperature exceeding the Leidenfrost temperature are presented. The cooling curves are obtained, the regions of effective heat transfer are determined. The primary experimental data were obtained during stationary heat transfer in the following range of operating parameters of the coolant: mass water flow rate 0.0038 ÷ 0.0145 kg/s; water pressure 0.4 ÷ 0.2 MPa, induction heating power 7 ÷ 20 kW.

The main directions of pneumodynamic system usage for storage of bulk agricultural products are bubbling and pneumatic mixing; loading and unloading of technological masses. During pneumodynamic impulse wave formation with alternation of supersonic and subsonic zones, which are limited by shock waves, sound force increasing on the technological mediumabout 150…170 dB, pressure increasing on the product mass due to compression-stretching at the pulse limits to amplitudes of 6…15 barare caused, that promotes implementation of loosening processes during the grain material storage. The developed scheme of the studied process involves the opposite arrangement of pneumatic impulse generators, which allows the formation of a stationary wave, that spreads the pulse energy in both longitudinal and transverse directions, covering the maximum space of the free flowing technological medium. Using the method of D’Alembert for solving the equations, methods of mathematical analysis and their processing in the MathCAD mathematical environment graphical and analytical dependences of force parameters of oscillatory system have been obtained. Choosing an effective mode of pneumodynamic loosening, the force pulse magnitude and the regularities of its change under conditions of the highest mechanical resistance of the free-flow mass have been used as the evaluation criteria the standing wave energy.

An analysis of statistical data of diagnostic measurements of two parameters determining the performance of the RBMK-1000 SHADR-8A flowmeters – the minimum value of the negative amplitude half-wave at the transistor flow measuring unit (TIBR) input and the mean-square deviation over the flowmeter ball rotation period – made it possible to develop a mathematical model of the flowmeter parametric reliability. This mathematical model is a random process, which is a superposition of two delayed renewal processes. Studying the flowmeter operational reliability model provides an exponential estimate of the probability that the parameters determining the flowmeter performance will not exceed the specified levels. Using the Bernoulli scheme and the probability-estimating relationship for the flowmeter performance parameters, it is possible to calculate the probability of failure-free operation of both a single reactor quadrant and the coolant flow measurement system. In addition, it becomes possible to estimate the quadrant failure rate. Important for practice is the possibility of predicting the number of failed flowmeters depending on the system operation time. An indicator of the system reliability can be the average number of failed flowmeters, the relation for which is given in the paper. All the research results were obtained without any additional assumptions about the random values distribution laws. The obtained results can be easily generalized for the cases when the vector dimension of the determining parameters is greater than two. The use of the results of this study is illustrated by calculated quantitative values of the flowmeter parametric reliability indicators and the coolant flow measurement system.

The choice of the shape and number of cooling channels is critical when seeking for the best design of a liquid rocket engine thrust chamber. Trade-off has to be considered among the different constraints that must be satisfied in terms of thrust chamber mass and resistance to mechanical and thermal loads, while keeping acceptable the power spent to make the coolant flow in the channels. In this study attention is focused on the phenomenon of thermal stratification that occurs in liquid rocket engine cooling channels due to their asymmetric heating. Its role is studied with a suitable engineering approach that allows to emphasize when rectangular cross section channels become inefficient due to stratification. A parametric analysis is carried out on a reference configuration showing the margins required to satisfy specific wall temperature constraints.

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.

Capillarytubes are widely used as coolant flow controllers in small cooling systems. Inthis study, the effect of capillary tube sizes and refrigerant mass charge onthe room temperature was investigated by using different refrigerants.Experiments were performed by passing R410 and R32 refrigerants from differentcapillary tubes in different mass charge. Experiments were carried out in atest setup which can work with R410 and R32 refrigerants. As a result of theexperiments, room temperature values were presented depending on the diameterof the capillary tube and the refrigerant mass charge distribution in asteadily cycling domestic refrigerator., Kılcal borular küçük soğutma sistemlerinde soğutucu akışkontrol cihazı olarak yaygın olarak kullanılmaktadır. Bu çalışmada, farklısoğutucu akışkanları kullanılarak soğutucu akışkan şarj miktarının ve kılcalboru ebatlarının oda sıcaklığına etkisi incelenmiştir. Farklı çaplı kılcalborulardan, farklı kütlelerde R410 ve R32 soğutucu akışkanları geçirilerekdeneyler yapılmıştır. Deneyler, R410 ve R32 soğutucu akışkanlar ile çalışabilenbir deney düzeneğinde yapılmıştır. Deneyler sonucunda, sürekli döngülü bir evbuzdolabında kılcal boru çapına ve soğutucu kütle şarjına bağlı olarak odasıcaklığı değerleri sunulmuştur.

The theoretical results of mathematical modelling of the operation of a porous compact heat exchanger with the results of experimental studies are checked. Methods of comparing the results are given. Satisfactory errors in the thermohydraulic characteristics were obtained.

In this study we used Nek5000, an open-source, high-order spectral element CFD code developed at Argonne National Laboratory (ANL), to model the coolant flow in spacer grids. Two fuel assembly configurations were studied: 2 × 2 and 5 × 5 fuel rod arrangements. The simulations for the 2 × 2 case were based on previous studies, simulating one span of the 2 × 2 fuel rod configuration including a surrogate spacer grid and mixing vane design with typical features of spacers for energy production. Dual periodic boundary conditions were applied in the spanwise direction to take the crossflow into consideration. The study of the 5 × 5 fuel assembly was performed as part of the ANL–Framatome collaboration for advancing computational fluid dynamics (CFD) tools. An advanced numerical model was developed to simulate the experimental setup provided by Framatome. For the 5 × 5 fuel assembly study, two cases of flow geometry were simulated with Nek5000: balanced and unbalanced configurations. In the balanced flow the coolant was entering the fuel rod assembly through 121 uniformly spaced inlet holes arranged in an 11 × 11 matrix. The unbalanced case, on the other hand, featured 14 larger holes placed on only one side of the horizontal plane. Nek5000 accepts only hexahedral meshes, which bring a great challenge to the meshing process for a spacer grid fuel assembly. A tet-to-hex meshing strategy was applied to handle the complex geometric features. A tetrahedral mesh was created first, and then each tetrahedral element was converted into four hexahedral elements. Boundary layers were extruded to fit to the exact geometry. In order to account for transient flow characteristics, the large eddy simulation approach was applied in this study. The employed subgrid-scale model relies on explicit filtering, which has been proven valid for many engineering-scale simulations. We present here the simulation results obtained for both the 2 × 2 and 5 × 5 fuel assemblies.

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17 × 17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12 × 12 and 13 × 13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13 × 13 and 12 × 12 arrays) are several times larger than that of the referenced solid fuel (17 × 17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.

Following an accidental event in a nuclear pressurized water reactor, involving the loss of primary-side forced coolant flow, the core decay heat is generally removed through a natural circulation ...

The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in fuel-rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.

To develop tritium breeding technology for a Fusion Reactor, Korea has participated in the Test Blanket Module (TBM) program in ITER. The He Cooled Ceramic Reflector (HCCR) TBM consists of functional components such as First Wall (FW), Breeding Zone (BZ), Side Wall (SW), and Back Manifold (BM) and it was designed based on the separate analyses for each component in 2012. Based on the each component analysis model, the integrated model is prepared and thermal-hydraulic analysis for the HCCR TBM is performed in the present study. The coolant flow distribution from BM and SW to FW and BZ, and resulted structure temperatures are obtained with the integrated model. It is found that the non-uniform flow rate occurs at FW and BZ and it causes excess of the design limit (550 °C) at some region. Based on this integrated model, we will perform the design optimization for obtaining uniform flow distribution for satisfying the design requirements.

Results of computer modeling of coolant flow in the fuel assembly of the reactor of a floating nuclear power plant using the LOGOS CFD programs have been given. The possibility of using the obtained results to improve models built into the engineering programs of thermohydraulic calculation of nuclear-reactor cores has been considered.

CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the result of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analyses, it was found that the MCST does not exceed 1260 °C, meets the design safety requirements. The 2nd MCST are higher than 1st MCST. The isolation of main steam line has the least safety margin.

Designing cooling channels for the thermoplastic injection process is a very important step in mold design. A conformal cooling channel can significantly improve the efficiency and the quality of production in plastic injection molding. This paper introduces an approach to generate spiral channels for conformal cooling. The cooling channels designed by our algorithms has very simple connectivity and can achieve effective conformal cooling for the models with complex shapes. The axial curves of cooling channels are constructed on a free-form surface conformal to the mold surface. With the help of boundary-distance maps, algorithms are investigated to generate evenly distributed spiral curves on the surface. The cooling channels derived from these spiral curves are conformal to the plastic part and introduce nearly no reduction at the rate of coolant flow. Therefore, the channels are able to achieve uniform mold cooling. Moreover, by having simple connectivity, these spiral channels can be fabricated by copper duct bending instead of expensive selective laser sintering.

Understanding the effects of spacer grids on the coolant flow through a nuclear reactor core is required for best-estimate design and analysis of the plant. The impact of a spacer grid on two-phase...

It is important to consider powertrain cooling performance, when engine is applied to new vehicle. If theperformance is poor, engine will be damaged by overheating. But, the development timing of engine is faster thantiming of vehicle, it is difficult to test the cooling performance of new engine and vehicle. In this study the powertaincooling performance was estimated with some test and calculation data. First, the heat rejection test was conducted. From this test, the heat rejection data at required rpm and load was acquired. Second, coolant flow test was conducted.From this test coolant flow rate to radiator was measured. Then engine torque and rpm was calculated from vehicle load and speed. Vehicle load and speed was calculated from test mode. Then by comparing these data, the powertraincooling performance was estimated. Key Words : Coolant Flow Data, Heat Rejection Rate, PTC (Powertrain Cooling Performance) * Corresponding Author : Sunki Min(Ajou Motor College)Tel: +82-10-9072-2658 email: skmin@motor.ac.krReceived January 29, 2015 Accepted May 7, 2015Revised (1st March 20, 2015, 2nd May 6, 2015)Published May 31, 2015

The present work has been carried out in the framework of the European FP7 project SEARCH, in which the MYRRHA demonstrator reactor is designed to be able to operate both in ADS mode and in critical mode using lead–bismuth eutectic (LBE) as primary coolant. According to the project task definition, the pin failure and fuel dispersion scenarios in severe accidents had to be extensively studied for reactor safety analysis. In this paper, the unprotected severe transients analyses for the MYRRHA–FASTEF critical core were performed using the SIMMER-III code. The aim of the current work is to obtain a deeper understanding of core material redistribution processes after pin damage. Since the fuel has almost the same density as the coolant, its pellets, chunks and particles will essentially be carried by the coolant flow, thus moving upwards out of the core and redistributing into the upper pool region and peripheral structures. Starting the simulations from the steady state configuration, relevant parameters reflect good agreement with the design operational conditions. For the transients, the most severe accident scenario proposed, that may possibly lead to pin failure and furthermore core damage, is the unprotected blockage accident (UBA). The calculation results show that after pin failure, the mobile fuel starts to re-distribute. In the meantime, the reactor stabilizes to shut-down status because of the fuel loss. Our results show that the blockage propagation is impossible thanks to the gap between fuel subassemblies.

핵연료를 이용한 원자력 발전에서 가장 중요한 것 중 하나는 핵분열중 발생하는 열을 효과적으로 제거하는 일이다. 핵연료의 열이 냉각수를 통해 효과적으로 제거되지 못할 경우 과도한 열에의해 핵연료봉 피복관이 녹아 폭발로 이어질 위험이 있다. 연소중인 가압 경수로 핵연료의 중심온도는 약1400℃까지 상승하며, 냉각수는 300℃, 150기압의 고온 고압 상태에서 1.6 kg/s 의 유속으로 순환된다. 새로운 핵연료를 개발하여 성능을 검증하기 위해서는 조사리그를 제작하여 연구로의 테스트 루프에서 연소시험을 수행해야 한다. 이때 핵연료 연소시험 중 발생하는 열을 제어하기 위하여 일정한 압력, 온도를 갖는 냉각수가 조사리그를 통과하여 루프내에서 순환하도록 설계되어 있으며, 특정 중성자속에서 조사 시험중인 핵연료의 발열량 측정은 순환중인 냉각수의 핵연료봉 주변 통과 전후 온도변화와 냉각수의 유속을 측정함으로써 간Key Words: Nuclear Fuel Irradiation Test(핵연료 조사시험), Coolant Flow Simulation System(냉각수 유동 모의장치), Nuclear Fuel Test Rig(핵연료조사리그) 초록: 핵연료 연소시험 도중 핵연료봉에서 발생하는 열을 효괻적으로 제거하기 위해서는 핵연료의 발열량을 정확하게 괃산하괟 충분한 유속을 갖는 냉각수를 순환시켜야 한다. 하나로는 개방형 수조 형태로서 핵연료 연소시험을 위한 별도의 냉각수 순환 루프를 갖추괟 있는데, 여기에 핵연료 조사리그를 장착하괟 냉각수를 순환시킴으로써 조사중인 핵연료봉의 온도를 일정온도 이하로 유지시킨다. 특히 순환되는 냉각수의 유속이 매우 높은 상태에서 조사리그 내에 부착된 부품이나 센서들이 유체유발 진동에 의해 파손되거나 기능을 상실하는 과우 매우 큰 기회비용을 야기한다. 본 연궫에서는 조사리그 부품의 건전성 사전 검토 및 괟속 유동에서의 센서 동작 특성에 대한 사전검토를 위해 냉각수 모의 순환장치를 개발하였다. Abstract: To remove heat generated during a burn-up test of nuclear fuels, the heat generation rate of nuclear fuels should be calculated accurately, and a coolant should be circulated in the test loop at an adequate flow rate. HANARO is an open pool–type reactor with an independent test loop for the burn-up test of nuclear fuels. A test rig is installed in the test loop, and a coolant is circulated through the test loop to maintain the temperature of the nuclear fuel rods within a desired temperature during an irradiation test. The components and sensors in the test rig can be broken or malfunction owing to the flow-induced vibration. In this study, a coolant flow simulation system was developed to verify and confirm the soundness of components and sensors assembled in the test rig with a high flow rate of the coolant.

In this paper, a new-type coolant chamber with higher film cooling effectiveness is proposed based on the vortex reconstruction. Three different kinds of coolant chamber configuration based on the cylindrical hole are selected to develop the swirling flow structure of coolant, and the comparative investigations have been carried out to study the effect of different coolant chambers at blowing ratios ranging from 0.5 to 2.0. The results show that the coolant jet momentum is small at low blowing ratio, and the difference of the film cooling effectiveness for three kinds of coolant chamber configuration is little, but the advantage of swirling inflow coolant film cooling becomes obviously with the increase of blowing ratio. When the blowing ratio is 2.0, the jet momentum with original coolant chamber configuration is large and uniform, which leads to the lowest cooling effectiveness due to the formation of a strong kidney vortex. The first coolant chamber configuration has a low jet momentum region at upstream of the film hole, the coolant in this region interacts with high temperature mainstream and bypasses the large jet momentum coolant to attach cooling surface at downstream, the cooling effect is obvious at downstream. The second coolant chamber configuration is sprayed with the structure of unidirectional vortex, which forms a vortex pressing on other vortex, making the coolant in pressed vortex attach surface better. The coolant laterally velocity is large, producing the best coverage and the higher film cooling effectiveness. The average film cooling effectiveness of the first and second coolant chamber configuration are larger than original by about 10% and 25%, respectively (M = 1.0), or 50% and 550% (M = 1.5). From the distribution of average film cooling effectiveness of different blowing ratios, it can be concluded that the optimal blowing ratio of swirling coolant flow film cooling is in the range of 1.8 to 2.1.

The authors examine dynamic and metrological characteristics of a vortex-acoustic flowmeter, a classification, a brief description, and the operation principle of widely used flowmeters. They present the results of research and exploitation of vortex-acoustic and electromagnetic flowmeters within the calibration interval (CI), obtained on an automated laboratory research test bench. The authors give some practical advice on choice of devices for flow metering based on the procedure of metrological reliability.

To safely and efficiently load the fuel assemblies of the VVER 440 / V 213 nuclear reactor, the relation between the temperature of the coolant at the outlet of the fuel assembly, measured by a thermocouple in the assembly’s axis, and the mean coolant temperature, present in the plane of the thermocouple, must be analysed. Based on the analysis of the coolant flow at the output of the physical model of the fuel assembly I. [1] and published CFD simulations [2,3,4] it was shown, that a special attention has to be paid to the influence of the water flow in the central tube on the temperature and velocity profile of the coolant at the thermocouple’s plane in the fuel assembly. For this reason, an experimental device with a physical model of the fuel assembly II. of the nuclear reactor VVER 440 / V 213 was designed, manufactured, and operated at the Faculty of Mechanical Engineering STU in Bratislava.

A conversion of the TRIGA reactor Bandung has been planned by replacing fuel from rod type to plate type. The fuel replacement causes the fuel arrangement or reactor core configuration should be changed as well as the cooling system. Currently, the reactor cooling system takes place by natural convection with the flow direction from the bottom to the top of the core. The replacement of plate-type fuel with a small distance between plates causes the core cooling process to decrease, therefore to make more effective the forced convection cooling system with the direction of upward flow is required. It is expected that the cooling process will increase. The ability of the cooling system depends on flow distribution in the core at the time of reactor in operating, therefore it is necessary to analyze the distribution of coolant flow and the velocity distribution of the cooling system from the top reactor pool to the reactor core to determine that the coolant flow from the inlet pipe mostly goes to the core. The purpose of this research is to analyze the distribution of coolant flow and flow velocity to the reactor core. The analysis was carried out in a simulation using computational fluid dynamics FLUENT software by creating reactor core modeling. The result shows that the cooling water flows from above along the reactor pool, then the stored energy is reduced so that the flowing slowdown. In this case, resulting in evenly flow distribution to the reactor core.

EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the experimental results provided by IAEA (International Atomic Energy Agency) obtained for a series of transient tests initiated by step insertion of different magnitudes of positive reactivity with varying degrees of different controlled parameters such as reactor initial power, coolant temperature and coolant flow condition. 20 out of 39 tests that fall under forced convection mode have been considered for the present simulation provided the reactor scram system is disabled. Peak power and peak clad temperature due to transient have been calculated and it was found that although peak clad temperature values agreed, the peak power values seem to underestimate the experimental values. Further study appears to be needed to identify the limitations in modeling or examining the effect of input parameters during modeling to obtain the better simulation results.

Limited results of studies on coolant channel blockage accidents in a prismatic-gas cooled reactor core are available, although coolant channel blockage of the fuel block in a prismatic gas-cooled reactor has been regarded as a highly important phenomenon. In this study, therefore, intensive and extensive thermal analyses were carried out to enhance the understanding and knowledge about the thermal performance of multi-hole type prismatic fuel blocks under coolant channel blockage accidents. A series of thermo-fluid calculations were performed using the CORONA code. In contrast to existing works, a full core model was used to consider an accurate coolant flow redistribution after the blockage in this work. The results of the CORONA calculations show that the single channel blockage accidents of PMR200 are not a safety concern regardless of the location. However, the maximum fuel temperature exceeds the safety limit in the case of more than ∼10 channel blockage at the same region of the fuel column. It was also found that the selection of the computational domain is crucial. A full core analysis is required for accurate results in the case of severe blockage accidents.

Analysis has shown that the current research on the technology of heavy liquid-metal coolants includes scientific-research and experimental development work on the validation, design and development of methods and means as well as the direct preparation of the coolant, the startup of circulation loops with heavy liquid-metal coolants and routine operation and decommissioning. The current methods and means include hydrogen purification of the coolant and loop by removal of slag-forming impurities, regulation of dissolved oxygen in the coolant for anticorrosion protection of steel, filtration of the coolant and protective gas and monitoring of the coolant under reactor and non-reactor conditions. The current processes presuppose careful preparation of heavy liquid-metal coolants before and during loading of the circulation loop, a complex of measures in startup regimes and minimal worker diversion for maintenance during routine operation of the facility.

Results of research works on studying local hydrodynamics and mass transfer for coolant flow in the characteristic zones of PWR reactor fuel assemblies in case of using belts of mixing spacer grids are presented. The investigations were carried out on an aerodynamic rig using the admixture diffusion method (the tracer-gas method). Certain specific features pertinent to coolant flow in the fuel rod bundles of Kvadrat-type fuel assemblies were revealed during the experiments. The obtained study results were included in the database for verifying computation fluid dynamics computer codes and detailed cell-wise calculations of reactor cores with Kvadrat-type fuel assemblies. The obtained results can also be used for more exact determination of local coolant flow hydrodynamic and mass transfer characteristics in assessing thermal reliability of PWR reactor cores.