• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1109-1114
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• 2009

The evolution of a high burnup structure (HBS) in a light water reactor (LWR) $UO_2$ fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the $UO_2$ matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.924-933
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• 2015

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional $^{235}U$ burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using $^{233}U$, $^{242}Pu$, $^{150}Nd$, and $^{133}Cs$ as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.265-270
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• 1996

The temperature distribution in the pellet was obtained from beginning the general heat conduction equation. The thermal conductivity of pellet used the SIMFUEL data that made clear the effect of burnup on the thermal conductivity degradation. Since the pellet rim acts as the thermal barrier to heat flow. the pellet was subdivided into several rings in which the outer ring was adjusted to play almost the same role as the rim. The local burup in each ring except the outer ring was calculated from the power depression factor based on FASER results. whereas the rim burnup at the outer ring was achieved by the pellet averaged burnup based on the empirical relation. The rim changed to the equivalent Xe film so the predicted temperature shooed the thermal jump across the rim. The observed temperature profiles depended on linear heat generation rate. fuel burnup. and power depression factor. The thermal conductivity degradation modelling can be applied to the fuel performance code to high burnup fuel,

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.311-318
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• 2008

In modem power reactors, nuclear fuels have recently reached 55,000 MWd/MtU from the initial average burnup of 35,000 MWd/MtU to reduce the fuel cycle cost and waste volume. At such high burnups, a fuel pellet produces fission products proportional to the burnup and creates a typical high burnup structure around the periphery region of the pellet, producing the so called 'rim effect'. This rim region of a highly burnt fuel is known to be ca. $200\;{\mu}m$ in width and is known to affect the fuel integrity. To characterize the local burnup in the rim region, solid sampling in the micro meter region by laser ablation is needed so that the distribution of isotopes can be determined by ICP-MS. For this procedure, special radiation shielding is required for personnel safety. In this study, we installed a radiation shielded laser ablation ICP-MS system, and a performance test of the developed system was conducted to evaluate the safe operation of instruments.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.201-210
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• 1997

A thermal conductivity correlation has been proposed which can be applied to high turnup fuel by considering both of thermal conductivity with turnup across fuel pellet and additional degradation at pellet rim due to very high porosity. In addition, a correlation has been developed that can estimate the porosity of rim region as a function of rim burnup under the assumptions that all the produced fission gases are retained in the in porosity and threshold pellet average burnup required for the formation of rim region is 40 MWD/㎏U. Rim width is correlated to rim burnup using measured data. For the RISO experimental data obtained at pellet average turnup of 43.5 MWD/㎏U for three linear heat generation rates of 30, 35 and 40 ㎾/m, radial temperature distributions ore calculated using the present correlation and compared with the measured ones. This comparison shows that the present correlation gives the best agreement with the measured data when it is combined with the HALDEN's correlation for thermal conductivity considering its degradation with burnup. Another comparison with the HALDEN's measured fuel centerline temperature as a function of burnup at 25 ㎾/m up to about 44 MWD/㎾U also suggest that the present correlation yields the best agreement when it is combined with the HALDEN's thermal conductivity.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.351-358
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• 1990

A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the pseudo-element and the pseudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Journal of the Korean institute of surface engineering
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• v.56 no.1
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• pp.84-93
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• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.279-284
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• 1998

Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75${\mu}{\textrm}{m}$ and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.231-241
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• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.