• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.218-227
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• 1986

Studies were conducted to determine the extent of oxidation and same of the mechanical property changes of Zircaloy-4 fuel cladding after it was exposed to hot steam environment. The purpose of these tests was to provide some informations on the embrittlement behavior of CANDU type fuel cladding, which could be experienced under the loss-of-coolant accident conditions. The Zircaloy fuel cladding tubes were exposed in a steam environment at the temperature of 90$0^{\circ}C$, 1,00$0^{\circ}C$. The growth of the ZrO$_2$ layer combined with an oxygen rich $\alpha$-phase layer into the Zircaloy tube material was found as a function of time t and temperature of steam exposure, E=1.1√Dt+0.002 where D is a temperature dependent diffusion coefficient. The tensile strength of the specimens exposed for a short period increased but decreased continuously with further exposure. The circumferential elongation was drastically changed with the exposure time while the hoop strength did't decrease greatly. The X-ray measurement of preferred orientation of the Zircaloy tube material indicated that grains in the as received tube were oriented such that the poles of the basal (0001) planes were predominantly radial, while the poles of the basal plane in the tube materials heattreated at 1,00$0^{\circ}C$ were oriented tangentially. It appears that this reoriented texture may contribute to lessening the decrease of the hoop strength of the heat treated Zircaloy tube material.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.64-71
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• 2003

In order to evaluate the effect of alloying elements (Nb and Sn) on the wear resistance of advanced Zr fuel claddings, sliding wear tests have been performed in room temperature air and water and these results were compared with those of commercial alloys such as Zircaloy-4, A and B alloys. As a result, the advanced Zr fuel claddings have a similar wear resistance compared with the commercial alloys. The wear resistance of the advanced Zr fuel claddings is closely releted to the content of Nb and Sn even though the effects of transition elements are involved in deforming wear properties. In the tested specimens with similar Sn content, wear volume became down to a minimum at $0.4\;wt\;\%$ Nb, then rapidly increased at 1.0 wt Nb. This behavior results in the variation of grain size with alloying contents. But Sn did not have a significant effect on the wear volume of advanced Zr fuel claddings below $1.1\;wt\%$. The relationship between alloying elements and wear behaviour was evaluated and discussed using material compatibility factor.

• Korean Journal of Metals and Materials
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• v.49 no.12
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• pp.937-944
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• 2011

Studies have been carried out in order to reduce fuel-cladding chemical interaction (FCCI) behavior of metallic fuel in sodium-cooled fast reactors (SFR) using an electroplating technique. A $20{\mu}m$ thick Cr layer has been plated by the electrochemical method in the Sargent bath over the HT9 (12Cr-1Mo) clad material and diffusion couple tests of the U-10Zr metallic fuel as well as the rare earth alloy (70Ce-29La) have been conducted. The results show that the Cr plating can prevent FCCI behavior along the fuel-clad interface. However, cracks developed through the thickness during plating, which resulted in the migration of some fuel constituents. Variation of bath temperature, application of pulse current, and post heat treatment have been conducted to control such cracks. We found out that some conditions like the pulse current and the post heat treatment enhanced the layer property by reducing the internal cracks and improving the diffusion couple test.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.184-189
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• 1997

A prompt fission gas release model incorporating the resistance to gas flow in the gap was developed and the effects of gap resistance and failure location on prompt fission gas release from the cladding breach were assessed. The process of prompt fission gas release from the plenum and gap into the coolant was modeled in accordance with three major phenomena: (1) transient gas flow in the gap, (2) the growth of the fission gas bubble while it is still attached to the breach, and (3) the detachment of the fission gas bubble from the breach and mixing with the coolant. The cumulative mass release fraction by the present model was calculated for the case of Young-Gwang 3 & 4 nuclear fuel rod as a typical example. The results showed that the release behavior of prompt fission gas with time was different from the frictionless model which has frequently been used in a simplified approach, and that the location of cladding failure was another key factor for the prompt fission gas release process due to the resistance in the gap.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2578-2590
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• 2023

In this work, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4-5 w/o UO₂ fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er₂O₃) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd₂O₃) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO₂ in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.22-33
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• 1987

The confirmed integrity of nuclear fuel cladding materials is an important object during steady state and transient operations at nuclear power plant. In this context, the clad material yielding behavior is especially important because of pellet-clad gap expansion. During the steep power excursion, the in-pile irradiation behavior differences between uranium-dioxide fuel pellet and zircaloy clad induce the contact pressure between them. If this pressure reaches the zircaloy clad yield pressure, the zircaloy clad will be plastically deformed. After the reactor power resumed to normal state, this plastic permanent expansion of clad tube give rise to the pellet-clad gap expansion. In this paper, the simple mandrel expansion test method which utilizes thermal expansion difference between copper mandrel and zircaloy tube was adopted to simulate this phenomenon. That is, copper mandrel which has approximately three times of thermal expansion coefficient of zircaloy-4 (PWR fuel cladding material) were used in this experiment at the temperature range from 400C to 700C. The measured plastic expansion of zircaloy outer radius and derived mathematical relations give the yield pressure, yield stress of zircaloy-4 clad at the various clad wall temperatures, the activation energy of zircaloy tube yielding, and pellet-clad gap expansion. The obtained results are in good agreement with previous experimental results. The mathematical analysis and simple test method prove to be a reliable and simple technique to assess the yielding behavior and gap expansion measurement between zircaloy-4 tube and uranium-dioxide fuel pellet under biaxial stress conditions.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.650-659
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• 2016

As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.