• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1896-1906
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• 2020

The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark calculations were performed by DeCART stand-alone and DeCART/MATRA multi-physics coupled code system to verify their accuracy. The solutions of DeCART stand-alone calculations for the control rod bank worth, detector signal, isothermal temperature coefficient, and critical boron concentration agreed very well with the measurements. The root-mean-square errors of the boron letdown curves for two-cycles were less than about 20 ppm, while the individual and total control rod bank worth agreed well within 7.3% and 2.4%, respectively. For the BEAVRS benchmark calculations at the beginning of burnup, the difference between DeCART simplified thermal-hydraulic stand-alone and DeCART/MATRA coupled calculations were not significantly large. Therefore, it is concluded that both the DeCART stand-alone code and the DeCART/MATRA multi-physics coupled code system have the capabilities to generate high fidelity transport solutions at core follow calculations.

• Korean Journal of Materials Research
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• v.14 no.7
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• pp.451-456
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• 2004

Thermal creep properties of the zirconium tube which was developed for high burnup application were evaluated. The creep test of cladding tubes after various final heat treatment was carried out by the internal pressurization method in the temperature range from $350^{\circ}C to 400^{\circ}C$ and from 100 to 150 MPa in the hoop stress. Creep tests were lasted up to 900days, which showed the steady-state secondary creep rate. The creep resistance of zirconium claddings was higher than that of Zircaloy-4. Factors that affect creep resistance, such as final annealing temperature, applied stress and alloying element were discussed. Tin as an alloying element was more effective than niobium due to solute hardening effect of tin. In case of advanced claddings, the optimization of final heat treatment temperature as well as alloying element causes a great influence on the improvement of creep resistance.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.482-493
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• 2002

The MOX fuel for LWR is fabricated either by direct mechanical blending of UO$_2$ and PuO$_2$ or by two stage mixing. Hence Pu-rich particles, whose Pu concentrations are higher than pellet average one and whose size distribution depends on a specific fabrication method, are inevitably dispersed in MOX pellet. Due to the inhomogeneous microstructure of MOX fuel, the thermal conductivity of LWR MOX fuel scatters from 80 to 100 % of UO$_2$ fuel. This paper describes a mechanistic thermal conductivity model for MOX fuel by considering this inhomogeneous microstructure and presents an explanation for the wide scattering of measured MOX fuel＇s thermal conductivity. The developed model has been incorporated into a KAERI＇s fuel performance code, COSMOS, and then evaluated using the measured in-pile data for MOX fuel. The database used for verification consists of homogeneous MOX fuel at beginning-of-life and inhomogeneous MOX fuel at high turnup. The COSMOS code predicts the thermal behavior of MOX fuel well except for the irradiation test accompanying substantial fission gas release. The over-prediction with substantial fission gas release seems to suggest the need for the introduction of a recovery factor to a term that considers the burnup effect on thermal conductivity.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.17 no.9
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• pp.819-825
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• 2007

Vibration characteristics of an duel-cooling cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.242-242
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• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.559-566
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• 2015

The thermalehydraulic characteristics for the CANadian Deuterium Uranium Flexible (CANFLEX)-burnable poison (BP) fuel channel, which is loaded with a BP at the center ring based on the CANFLEX-RU (recycled uranium) fuel channel, are evaluated and compared with that of standard 37-element and CANFLEX-NU (natural uranium) fuel channels. The distributions of fuel temperature and critical channel power for the CANFLEX-BP fuel channel are calculated using the NUclear Heat Transport CIRcuit Thermohydraulics Analysis Code (NUCIRC) code for various creep rate and burnup. CANFLEX-BP fuel channel has been revealed to have a lower fuel temperature compared with that of a standard 37-element fuel channel, especially for high power channels. The critical channel power of CANFLEX-BP fuel channel has increased by about 10%, relative to that of a standard 37-element fuel channel for 380 channels in a core, and has higher value relative to that of the CANFLEX-NU fuel channel except the channels in the outer core. This study has shown that the use of a BP is feasible to enhance the thermal performance by the axial heat flux distribution, as well as the improvement of the reactor physical safety characteristics, and thus the reactor safety can be improved by the use of BP in a CANDU reactor.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.663-667
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• 2017

Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below $300^{\circ}C$. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor ($K_{IH}$) to initiate DHC as a function of temperature in Zry-4 for temperatures between $227^{\circ}C$ and $315^{\circ}C$. The experimental technique used in this study was the pin-loading testing technique. To determine the $K_{IH}$, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around $300^{\circ}C$, there was a sharp increase in $K_{IH}$ indicating the upper temperature limit for DHC. The value for $K_{IH}$ at $227^{\circ}C$ was determined to be $2.6{\pm}0.3MPa$ ${\surd}$m.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.280-291
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• 2018

In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010-2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1079-1092
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• 2020

A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.