• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.34-41
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• 2024

This work presents experimental data and modelling of the release of Mo from high-burnup spent nuclear fuel (63 MWd/kgU) at two different pH values, 8.4 and 13.2 in air. The release of Mo from SF to the solution is around two orders of magnitude higher at pH = 13.2 than at pH = 8.4. The high Mo release at high pH would indicate that Mo would not be congruently released with uranium and would have an important contribution to the Instant Release Fraction, with a value of 5.3%. Parallel experiments with pure non irradiated Mo(s) and XPS determinations indicated that the faster dissolution at pH = 13.2 could be the consequence of the higher releases from metallic Mo in the fuel through a surface complexation mechanism promoted by the OH^- and the oxidation of the metal to Mo(VI) via the formation of intermediate Mo(IV) and Mo(V) species.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.65-70
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• 1997

Computational benchmark calculations have been performed for CANDU DUPIC fuel lattice and core using a Monte Carlo code MCNP-4B with ENDF/B-V library. The eigenvalues of the DUPIC fuel lattice have been predicted by an integral transport code WIMS-AECL using ENDF/B-V library for different burnup steps and lattice conditions. The comparison has shown that the eigenvalues match those of MCNP-4B within 0.20% $\Delta$k difference between WIMS-AECL and MCNP-4B results. The calculation of a 2-dimensional CANDU core loaded with DUPIC fuel has shown that the eigenvalue predicted by a diffusion code RFSP using lattice parameters generated by WIMS-AECL matches that of MCNP-4B within 0.12%Δk and the largest bundle power prediction error is around 7.2%.

• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.111-120
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• 1980

Nuclear physics parameters of the Wolsung CANDU-PHW reactor are computed by use of the PHWCELL computer code that is an improved version of LATREP. The PHWCELL code mainly computes cell parameters of heavy water moderated reactors, and modeling scheme of heavy water reactor cell calculations has been developed with the PHWCELL computer code. The reactor operating conditions considered in the study are cold zero power (CZP) and hot full power (HFP) with equilibrium poison. The cell parameters are also computed as a function of fuel burnup and the numerical results are compared with the results in PSR of the Wolsung unit and in the previous study.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.172-181
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• 1999

Series expansion methods to compute the exponential of a matrix have been compared by applying them to fuel depletion calculations. Specifically, Taylor, Pade, Chebyshev, and rational Chebyshev approximations have been investigated by approximating the exponentials of bum matrices by truncated series of each method with the scaling and squaring algorithm. The accuracy and efficiency of these methods have been tested by performing various numerical tests using one thermal reactor and two fast reactor depletion problems. The results indicate that all the four series methods are accurate enough to be used for fuel depletion calculations although the rational Chebyshev approximation is relatively less accurate. They also show that the rational approximations are more efficient than the polynomial approximations. Considering the computational accuracy and efficiency, the Pade approximation appears to be better than the other methods. Its accuracy is better than the rational Chebyshev approximation, while being comparable to the polynomial approximations. On the other hand, its efficiency is better than the polynomial approximations and is similar to the rational Chebyshev approximation. In particular, for fast reactor depletion calculations, it is faster than the polynomial approximations by a factor of ∼ 1.7.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1998.04a
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• pp.454-460
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• 1998

The vibration behavior of fuel rods has been analyzed by FEM in consideration of axial force and support spring constants. The axial compression force on the fuel rod in reactor decreases with the fuel rod burnup, and its decrease makes the natural frequencies of fuel rod considerably increase. The change of support spring constant can contribute to the remarkable change of the mode shapes, but not greatly to the natural frequencies. The reaction forces of support springs are obtained from normalizing the lst mode with the max. 0.2 mm displacement. The calculated reaction forces are larger than the previous results obtained by disregarding the deflections of the support springs.

• Journal of Powder Materials
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• v.11 no.4
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• pp.314-321
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• 2004

The elastic moduli of simulated dry process fuels with varying composition and density were measured in order to analyze the mechanical properties of a dry process fuel pellet. Resonant ultrasound spectroscopy(RUS) which can determine all elastic moduli with one set of measurements for a rectangular parallelepiped sample was used to measure the elastic moduli of UO$_{2}$ and simulated dry process fuel. The simulated dry process fuel showed a higher value of Young's modulus than UO$_2$ due to the presence of metallic precipitates and solid solution elements in the UO$_{2}$ matrix. The correlation between Young's modulus and porosity(P) of simulated dry process fuel was found to be 231.4-651.8 P (GPa) at room temperature. Dry process fuel with a higher burnup showed higher Young's modulus because total content of fission product element was increased.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.21-23
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• 2005

NARX(Nonlinear AutoRegressive with eXogenous input) neural network was used for prediction of nuclear reactor behavior which was influenced by control rods in short-term period and also by xenon and boron in long-term period in load following operations. The developed model was designed to predict reactor power, xenon worth and axial offset with different burnup rates when control rod and boron were adjusted in load following operations. Data of UCN 3 were collected by ONED94 code. The test results presented exhibit the capability of the NARX neural network model to capture the long term and short term dynamics of the reactor core and seems to be utilized as a handy tool for the use of a plant simulation.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.689-699
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• 2020

In this work, new multi-recycling options of TRU nuclides using PWR fuel assemblies comprised of MOX and FCM (Fully Ceramic Micro Encapsulated) fuels are suggested and neutronically analyzed. These options do not use a fully recycling of TRU but a partial recycling where TRUs from MOX fuels are recycled while the ones from FCM fuels are not recycled due to their high consumption rate resulted from high burnup. In particular, additional external TRU feed in MOX fuels for each cycle was considered to significantly increase the TRU consumption rate and the finally selected option is to use external TRU and enriched uranium feed as a makeup for the heavy metal consumption in MOX fuels. This hybrid external feeding of TRU and enriched uranium in MOX fuel was shown to be very effective in significantly increasing TRU consumption rate, maintaining long cycle length, and achieving negative void reactivity worth during recycling.