• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1525-1531
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• 2019

The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.189-190
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• 2017

For the concept design of the device, a tool was made to test the simulated fuel rods and cutting force and the cutting force was measured. When 2-CUT and 3-CUT modules were used, the maximum force in 2-CUT at 12.5 mm/s speed change was $197.5kg_f$ and the maximum force at 3-CUT was $363.2kg_f$. The change of force in 2-CUT rapidly increases from about 1 second, and you can see that there are increase and decrease of the force change from about 5 seconds to 18 seconds, and it was rapidly decreased and the cut was made. The force change in 3-CUT has higher force at about 5 seconds later than 2-CUT at the speed of 12.5 mm/s, and you can see that it has the same tendency afterwards. If you search for the force at adequate speed from this cutting force test, 2-CUT module requires less slitting force than 3-CUT module, and the cutting time for 250 mm at 12.5 mm/s was 21 seconds, which can cut 4 m fuel rod in 5 minutes. But, there are cases of not completely slitting with 2-CUT module, so it is necessary to supplement this in the future through experiments.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.3
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• pp.281-287
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• 2011

We consider the uncertainty in the elastic buckling formula for a thin tube. We take into account the measurement uncertainty of Young's modulus and Poisson's ratio and the tolerance of the tube thickness and diameter. Elastic buckling must be prohibited for a thin tube such as a nuclear fuel rod that must satisfy a self-stand criterion. Since the predicted critical buckling pressure overestimated that found in the experiment, the determination of the minimum safety factor is crucial. The uncertainty in each parameter (i.e., Young's modulus, Poisson's ratio, thickness, and diameter) is mutually independent, so the safety factor is evaluated as the sum of the inverse of each uncertainty. We found that the thickness variation greatly affects the uncertainty. The minimum safety factor of a thin tube of Zirconium alloy is evaluated as 1.547 for a thickness of 0.87 mm and 3.487 for a thickness of 0.254 mm.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.87-93
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• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.178-182
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• 1992

Spent PWR fuel rods have been scanned axially and sectionally to measure the relative gamma-ray intensity of Cs-137 and then bumups of the scanned rods determined by measuring Nd-148 which has been chemically separated. From these experimental results, a linear relation(LR) between the gamma-ray intensity of Cs-137 and the bumup in the range of 10∼35 GWD/MTU was obtained. In order to validate the LR, the Cs-137 gamma-ray intensity of unknown sample was nondestructively measured and the bumup obtained by the LR was compared with that of the Nd-148 method. It is revealed that the results from both methods are in good agreement, and thus it seems to be possible to estimate the bumup of spent PWR fuel rod by measuring nondestructively gamma-ray of fission product Cs-137.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.313-323
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• 2003

The sliding wear behaviors of Zircaloy-4 nuclear fuel rod were investigated using two support springs with convex and concave shapes in room temperature air and water. The main focus is to compare the wear behavior of various test variables such as slip amplitude, environment, contact contours with different spring shape and a number of cycles. The results indicated that wear volume and maximum wear depth increased with slip amplitude in both air and water, but their trends tended to change according to the spring shapes and test environments. In air condition, the wear volume was controlled by wear debris behavior generated on worn surface. As a result, final wear volume and maximum wear depth decreased if a ratio of protruded wear volume to worn area $(D_p)$ would be saturated to specific value. This is because wear particle layer could accommodate large strain by accumulating and transforming wear particle layer. However, in water condition, metal-to metal contact was more dominant and wear volume was greatly affected by changed mechanical behavior between contact surfaces since wear debris should be generated after repeated plastic deformation and fracture. After wear test, worn surfaces were examined using optical microscope and SEM and details of wear mechanism were discussed using a ratio of wear volume to worn area $(D_e)$ at each test condition.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.1-13
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• 2020

At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.