• 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.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3235-3241
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• 2022

The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR) for stable supply of ⁹⁹Mo. The Fission ⁹⁹Mo Production Process (FMPP) of KJRR produces solid waste such as spent uranium cake and alumina cake, and liquid waste in the form of intermediate level liquid waste (ILLW) and low level liquid waste (LLLW). This study thus established the operating range and optimum operating conditions for the cementation of ILLW from FMPP. It also evaluated whether cement waste form samples produced under optimum operational conditions satisfy the waste acceptance criteria (WAC) of a disposal facility in Korea (Korea radioactive waste agency, KORAD). Considering economic feasibility and safety, optimum operational conditions were achieved at a w/c ratio of 0.55, and the corresponding salt content was 5.71 wt%. The cement waste form samples prepared under optimum operational conditions were found to satisfy KORAD's WAC when tested for structural stability and leachability. The results indicate that the proposed cementation conditions for the disposal of ILLW from FMMP can be effectively applied to KJRR's disposal facility.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.387-394
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• 2014

A Slowing Down Time Spectrometer (SDTS) system is a highly efficient technique for isotopic nuclear material content analysis. SDTS technology has been used to analyze spent nuclear fuel and the pyro-processing of spent fuel. SDTS requires an external neutron source to induce the isotopic fissile fission. A high intensity neutron source is required to ensure a high for a good fissile fission. The electron linear accelerator system was selected to generate proper source neutrons efficiently. As a first step, the electron generator of an 80-keV electron gun was manufactured. In order to produce the high beam power from electron linear accelerator, a proper beam current is required form the electron generator. In this study, the beam current was measured by evaluating the performance of the electron generator. The beam current was determined by five parameters: high voltage at the electron gun, cathode voltage, pulse width, pulse amplitude, and bias voltage at the grid. From the experimental results under optimal conditions, the high voltage was determined to be 80 kV, the pulse width was 500 ns, and the cathode voltage was from 4.2 V to 4.6 V. The beam current was measured as 1.9 A at maximum. These results satisfy the beam current required for the operation of an electron linear accelerator.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.965-974
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• 2016

The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel $(U_{0.8}Pu_{0.2})O_2$ fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel $(U_{0.8}Pu_{0.2})O_2$.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.99-111
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• 1998

Chemical composition of the insoluble residue in a simulated spent PWR fuel(SIMRJEL) were studied. SIMFUELS were prepared by adding calculated amount of FP(fission product) elements with a burnup of 3.6% FIMA(fission per initial metal atom) to uranium in nitrate solution, evaporating the mixed solution to dryness, calcining at 90$0^{\circ}C$ in a stream of 4% H$_2$ + 96% He, and heating the pellet at 140$0^{\circ}C$ under high and low oxygen potentials. Insoluble residue was obtained from the dissolution of the SIMFUEL with HNO$_3$(1 : 1). The chemical composition of the SIMFUELs and the insoluble residues was determined by EPMA(electron probe microanalysis), XPS(X-ray photoelectron spectroscopy) and by XRD (X-ray diffraction) measurements. All of the insoluble residues suspended and precipitated were composed mainly of Mo, Ru with a small amount of Zr, Rh, Pd and Cd. The amount of insoluble residue(<1 wt.%) and a Mo/Ru ratio decreased with increasing oxygen potential. Formation of the zirconium molybdate precipitate, ZrMo$_2$O$_{7}$(OH)$_2$($H_2O$)$_2$, was observed in the residues. The possible role of Mo on the phase formation was discussed in regard to oxygen potential.l.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4049-4061
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• 2022

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.281-290
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• 2014

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.144-147
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• 1985

The spontaneous fission rate of $^{238}U$ has been determined using a solid state track recorder that was a pre-etched mica. Counting the tracks revealed in mica made it possible to calculate the spontaneous fission rate. The mica remained in close contact with a $^{238}UO_2$ foil for about five years. The resulting fission rate was $5.21{\pm}0.33$ fissions/g-sec.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.381-390
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• 2009

This article provides selects of outstanding problems in computational neutron transport, with some suggested approaches thereto, as follows: i) ray effect in discrete ordinates method, ii) diffusion synthetic acceleration in strongly heterogeneous problems, iii) method of characteristics extension to three-dimensional geometry, iv) fission source and $k_{eff}$ convergence in Monte Carlo, v) depletion in Monte Carlo, vi) nuclear data evaluation, and vii) uncertainty estimation, including covariance data.