• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.22 no.3_1spc
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• pp.573-579
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• 2013

Once-through helically coiled steam generator tubes subjected to external pressure are of interest because of their application to advanced small- and medium-sized integral reactors, in which a primary coolant with a relatively higher pressure flows outside the tubes, while secondary water with a relatively lower pressure flows inside the tubes. Another notable point is that the values of the mean radius to thickness ratio of these steam generator tubes are very small, which means that a thick-walled cylinder is employed for these steam generator tubes. In the present paper, the maximum allowable pressure of helically coiled and thick-walled steam generator tubes with through-wall cracks under external pressure is investigated based on a detailed nonlinear three-dimensional finite element analysis. In terms of the crack orientation, either circumferential or axial through-wall cracks are considered. In particular, in order to quantify the effect of the crack location on the maximum external pressure, these cracks are assumed to be located in the intrados, extrados, and flank of helically coiled cylinders. Moreover, an evaluation is also made of how the maximum external pressure is affected by the ovality, which might be inherently induced during the tube coiling process used to fabricate the helically coiled steam generator tubes.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.197-204
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• 1994

In a PWR rod cluster control assembly(RCCA) for shutdown is released upon action of control rod drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as fluid resistance caused by the RCCA movement, buoyance and mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly(KOFA). The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreement with the test results, the computer program developed in this paper can be employed to modify the exiting design features of the RCCA and guide thimble and to develope their new design features for advanced nuclear reactors.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.223-228
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• 2018

AREVA NP (Courbevoie, Paris, France) is actively developing several enhanced accident-tolerant fuels cladding concepts ranging from near-term evolutionary (Cr-coated zirconium alloy cladding) to long-term revolutionary (SiC/SiC composite cladding) solutions, relying on its worldwide teams and partnerships, with programs and irradiations planned both in Europe and the United States. The most advanced and mature solution is a dense, adherent chromium coating on zirconium alloy cladding, which was initially developed along with the CEA and EDF in the French joint nuclear R&D program. The evaluation of the out-of-pile behavior of the Cr-coated cladding showed excellent results, suggesting enhanced reliability, enhanced operational flexibility, and improved economics in normal operating conditions. For example, because chromium is harder than zirconium, the Cr coating provides the cladding with a significantly improved wear resistance. Furthermore, Cr-coated samples exhibit extremely low corrosion kinetics in autoclave and prevents accelerated corrosion in harsh environments such as in water with 70 ppm Li leading to improved operational flexibility. Finally, AREVA NP has fabricated a physical vapor deposition prototype machine to coat full-length cladding tubes. This machine will be used for the manufacturing of full-length lead test rods in commercial reactors by 2019.

• Advances in environmental research
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• v.9 no.4
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• pp.233-249
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• 2020

To date, there is no central wastewater (WW) treatment plant in Erbil city, Kurdistan region, Iraq. Therefore, raw WW disposes to the environment and sometimes it used directly for irrigation in some areas of Erbil city. Disposal of the untreated WW to the natural environment and using for irrigation it causes problems for the people and the environment. The aims of the current work were to study the characteristics, design of primary and different secondary treatment units and reusing of produced WW. Raw WW samples from Ashty city-Erbil city were collected and analyzed for twenty three quality parameters such as Total Suspended Solids (TSS), total dissolved solids, total volatile and non-volatile solids, total acidity, total alkalinity, total hardness, five-day Biochemical Oxygen Demand (BOD5), Chemical Oxygen Demand (COD), biodegradability ratio (BOD5/COD), turbidity, etc. Results revealed that some parameters such as BOD5 and TSS were exceeded the standards for disposal of WW. Design and calculations for primary and secondary treatment (biological treatment) processes were presented. Primary treatment units such as screening, grit chamber, and flow equalization tank were designed and detailed calculation were illustrated. While, Conventional Activated Sludge (CAS), Sequencing Batch Reactor (SBR) and Moving Bed Biofilm Reactors (MBBR) were applied for the biological treatment of WW. Results revealed that MBBR was the best and economic technique for the biological treatment of WW. Treated WW is suitable for reusing and there is no restriction on use for irrigation of green areas inside Ashty city campus.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.516-525
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• 1994

Since the TMI and Chernobyl accidents, many passive safety features are suggested in advanced reactors in order to enhance the safety in future nuclear power plants. In order to verify the effectiveness and provide the data for detailed design of passive cooling system, in the present work, the effects of air inlet position and external condition on the natural circulated air flow rate and the natural and forced convective heat transfer coefficient have been investigated for the one-side heated closed path such as the passive containment cooling system of the Westinghouse's AP-600. A series of experiments have been peformed with the 1/26th scaled segment type test facility of the AP-600 passive containment. Under natural and forced convection, the air velocities and temperatures are measured at several points of the air flow path. The experimental result are compared with a simple one-dimensional model and it shows a good agreement.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1563-1570
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• 2023

In this study, a DeCART/MIG uncertainty quantification (UQ) analysis code system with a multicorrelated cross section stochastic sampling (S.S.) module was established and verified through the UAM (Uncertainty Analysis in Modeling) and the BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) benchmark calculations. For the S.S. calculations, a sample of 500 DeCART multigroup cross section sets for two major actinides, i.e., ²³⁵U and ²³⁸U, were generated by the MIG code and covariance data from the ENDF/B-VII.1 evaluated nuclear data library. In the three pin problems (i.e. TMI-1, PB2, and Koz-6) from the UAM benchmark, the uncertainties in k_inf by the DeCART/MIG S.S. calculations agreed very well with the sensitivity and uncertainty (S/U) perturbation results by DeCART/MUSAD and the S/U direct subtraction (S/U-DS) results by the DeCART/MIG. From these results, it was concluded that the multi-group cross section sampling module of the MIG code works correctly and accurately. In the BEAVRS whole benchmark problems, the uncertainties in the control rod bank worth, isothermal temperature coefficient, power distribution, and critical boron concentration due to cross section uncertainties were calculated by the DeCART/MIG code system. Overall, the uncertainties in these design parameters were less than the general design review criteria of a typical pressurized water reactor start-up case. This newly-developed DeCART/MIG UQ analysis code system by the S.S. method can be widely utilized as uncertainty analysis and margin estimation tools for developing and designing new advanced nuclear reactors.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1140-1151
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• 2023

Fully Ceramic Microencapsulated (FCM) fuel is emerging advanced fuel material for the future nuclear reactors. The fuel pellet in the FCM fuel is composed of matrix and a large number of TRistructural-ISOtopic (TRISO) fuel particles which are randomly dispersed in the SiC matrix. The minimum layer thickness in a TRISO fuel particle is on the order of 10^-5 m, and the length of the FCM pellet is on the order of 10^-2 m. Hence, the heat transfer in the FCM pellet is a multi-scale phenomenon. In this study, three multi-scale heat conduction models including the Multi-region Layered (ML) model, Multi-region Non-layered (MN) model and Homogeneous model for FCM pellet were constructed. In the ML model, the random distributed TRISO fuel particles and coating layers are completely built. While the TRISO fuel particles with coating layers are homogenized in the MN model and the whole fuel pellet is taken as the homogenous material in the Homogeneous model. Taking the results by the ML model as the benchmark, the abilities of the MN model and Homogenous model to predict the maximum and average temperature were discussed. It was found that the MN model and the Homogenous model greatly underestimate the temperature of TRISO fuel particles. The reason is mainly that the conventional equivalent thermal conductivity (ETC) models do not take the internal heat source into account and are not suitable for the TRISO fuel particle. Then the improved ETCs considering internal heat source were derived. With the improved ETCs, the MN model is able to capture the peak temperature as well as the average temperature at a wide range of the linear powers (165 W/cm~ 415 W/cm) and the packing fractions (20%-50%). With the improved ETCs, the Homogenous model is better to predict the average temperature at different linear powers and packing fractions, and able to predict the peak temperature at high packing fractions (45%-50%).