• Journal of The Korean Society of Grassland and Forage Science
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• v.16 no.1
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• pp.27-38
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• 1996

This study was conducted to evaluate the potential of Festulolium braunii(Festuca pratensis Huds. $\times$ Loliurn mulrijZorum Lam.)as forge sources. The experiment was arranged in a split plot design with four replications from Feb. 1995. to Jan. 1996 at Hohenheim university. Main plots consisted of three forage species, Festulolirtrn braunii(Paulita), Lolium multiJomm(Lemtla), and Secale cereale(Halo) and subplots consisted of three maturity, boot, heading, and anthesis. The growth characteristics, DM yields, nutritive value, N and energy balance were observed. The heading date of Festulolium braunii was about two days earlier than that of Loliunz mrtltiflorum, farthermore Fesrulolium braunii retained higher leaf weight ratio(LWR) than other species. Festulolium brarrnii failed to show any significant differences in grass length, number of tillers, LAI, CGR, and DM accumulation when harvested at the same stage of maturity as compare with Loliurn mulriiflorutn or Setale cerede, bur the content of CP, OMD, and mineral elements and NEL value of Festulolium braunii compare to othes were slightly increased(P<0.05), while the content of NDF, ADF, and lignin decreased(P<0.05). The yield of CP was greater for Festulolium braunii than for Lolium multfimm and yields of DOM and NEL per unit area(ha) for Festulolium braunii were greater than for other species(P<0.05). The equivalent or higher DM production and nutritive value are obtained when Festulolium braunii are cut during the period from boot to anthesis stages, therefor DM intake, preference, and N and energy balance were also higher for Festulolirtm braunii compare to for other species. It could be suggested that Lolium multrfirum would be a more suitable substitute to Festuloliron braunii without decreasing forage production and quality, but the substitutive effect of Secale cereale would be difficult because of the maor differences in growth type and period compare to Fesrulolium braunii. From above the results of this experiment it appears that Festulolium braunii have a potential to provide forage sources. but funher research is needed to ascertain their potential under the climate of Korea.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.307-314
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• 2001

The changes of weight gain, structure, morphology and uranium oxidation states in l0wt% G $d_2$ $O_3$-doped U $O_2$ during the oxidation below 475$^{\circ}C$ and heat treatment at 130$0^{\circ}C$ in air were investigated using TGA, XRD, SEM, EPMA and XPS. The room temperature ( $U_{0.86}$G $d_{0.14}$) $O_2$Cubic Phase Converted to highly distorted ( $U_{0.86}$G $d_{0.14}$)$_3$ $O_{8}$ -type sing1e Phase by oxidation at 475 $^{\circ}C$ in air. This oxidized phase was reduced by annealing at 130$0^{\circ}C$ in air. The room temperature XRD pattern of the 130$0^{\circ}C$ annealed powder revealed that ( $U_{0.86}$G $d_{0.14}$)$_3$ $O_{8}$ -type single phase was separated into Gd-depleted $U_3$ $O_{8}$ and Gd-enriched ( $U_{0.7}$G $d_{0.3}$) $O_2$$_{+x}$ type cubic phase. The reduction and phase separation by the high temperature annealing of kinetically metastable and highly deformed ( $U_{0.86}$G $d_{0.14}$)$_3$ $O_{8}$ -type phase are interpreted in terms of cation size difference between G $d^3$$^{+}$ and U according to the oxidation state of U.U.U.U.U.te of U.U.U.U.U.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1626-1637
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• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.375-382
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• 2006

Characterization tests (load vs. displacement curve) are conducted for the springs of Zirconium alloy spacer grids for an advanced LWR fuel assembly. Twofold testing is employed: strap-based and assembly-based tests. The assembly-based test satisfies the in situ boundary conditions of the spring within the grid assembly. The aim of the characterization test via the aforementioned two methods is to establish an appropriate assembly-based test method that fulfills the actual boundary conditions. A characterization test under the spacer grid assembly boundary condition is also conducted to investigate the actual behavior of the spring in the core. The stiffness of the characteristic curve is smaller than that of the strap-wised boundary condition. This phenomenon may cause the strap slit condition. A spacer grid consists of horizontal and vertical straps. The strap slit positions are differentiated from each other. They affords examination of the variation of the external load distribution in the grid spring. Localized legions of high stress and their values are analyzed, as they may be affected by the spring shape. Through a comparison of the results of the test and FE analysis, it is concluded that the present assembly-based analysis model and procedure are reasonably well conducted and can be used for spring characterization in the core. Guidelines for improving the mechanical integrity of the spring are also discussed.

• Journal of the Korean Ceramic Society
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• v.50 no.6
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• pp.359-363
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• 2013

Silicon carbide and its composites are being considered as a nuclear fuel cladding material for LWR nuclear reactors because they have a low neutron absorption cross section, low hydrogen production under accident conditions, and high strength at high temperatures. The SiC composite cladding tube considered in this study consists of three layers, monolith CVD SiC - $SiC_f$/SiC composite -monolith CVD SiC. The volume fraction of SiC fiber and surface roughness of the composite layer affect mechanical and corrosion properties of the cladding tube. In this study, various types of SiC fiber preforms with tubular shapes were fabricated by a filament winding method using two types of Tyranno SA3 grade SiC fibers with 800 filaments/yarn and 1600 filaments/yarn. After chemical vapor infiltration of the SiC matrix, the surface roughness and fiber volume fraction were measured. As filament counts were changed from 800 to 1600, the surface roughness increased but the fiber volume fraction decreased. The $SiC_f$/SiC composite with a bamboo-like winding pattern has a smaller surface roughness and a higher fiber volume fraction than that with a zigzag winding pattern.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.499-508
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• 2011

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.42-59
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• 2002

In order to transmute long-lived radioactive nuclides such as transuranics(TRU), Tc-99, and I- l29 in LWR spent fuel, a preliminary conceptual design study has been performed for the accelerator driven subcritical reactor system, called HYPER(Hybrid Power Extraction Reactor) The core has a hybrid neutron energy spectrum: fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU is loaded into the HYPER core as a TRU-Zr metal form because a metal type fuel has very good compatibility with the pyre- chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb-Bi is chosen as a primary coolant because Pb-Bi can be a good spallation target and produce a very hard neutron energy spectrum. As a result, the HYPER system does not have any independent spallation target system. 9Cr-2WVTa is used as a window material because an advanced ferritic/martensitic steel is known to have a good performance under a highly corrosive and radiation environment. The support ratios of the HYPER system are about 4∼5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4 ∼5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is faced with.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.