• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.150-161
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• 1996

A deformation model was developed to calculate the deformation of the uranium-silicide dispersion fuel (U$_3$Si-Al) elements for research reactors. The model was based on the elasto-plasticity theory and power-law creep theory. Also, isotopic swelling was assumed for the fuel meat and isotropic thermal expansion for the fuel meat and dadding. The new model calculated successfully the deformation of the fuels of HANARO and NRU (in Canada). As the most important result, it was shown that the primary deformation mechanism in the fuel meat was swelling and that in the cladding was creep. For all cases simulated, the maximum hoop stress at cladding outer surface was lass than 5MPa, probably well below the yield stress of the dadding, and finally, the volume change was predicted to be less than 10% in the whole burnup range.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1929-1934
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• 2022

Besides recent progresses in the physics-based modelling of fission gas and helium behaviour, the scarcity of experimental data concerning their combined behaviour (i.e., cocktail) hinders further model developments. For this reason, in this work, we propose a modelling methodology aimed at providing recommendations for accelerated experimental investigations. By exploring a wide range of annealing temperatures and cocktail compositions with a physics-based modelling approach we identify the most interesting conditions to be targeted by future experiments. To corroborate the recommendations arising from the proposed methodology, we include a sensitivity analysis quantifying the impact of the model parameters on fission gas and helium release, in conditions representative of high and low burnup.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.119-125
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• 1994

Transient power level of the fuel rod is one of the key parameters for the transient fuel behavior. Through the analysis of the fuel performance data bases and sensitivity analyses of such parameters as rod power history, fast neutron flux, fuel enrichment and cycle length, which can affect the transient fuel behavior, a methodology generally applicable to find the allowable transient power level during the ANS Conditions I and II below which the mechanical integrity of the fuel rod is maintained was derived, and allowable transient power levels for the 17$\times$17 KOFA fuel rod have been determined as a function of the burnup. With the introduction of this methodology, design analysis of the transient fuel behavior currently being calculated every cycle can be replaced by the simple check of the peak transient power level achievable during the cycle, and an operational flexibility of the reactor can be obtained by allowing higher transient power level up to 689.5 w/cm at low burnup range than current maximum allowable transient power level, 591 w/cm for the 17$\times$17 KOFA fuel.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1031-1043
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• 2017

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Proceedings of the Korean Institute of Resources Recycling Conference
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• 2004.05a
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• pp.138-142
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• 2004

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.44-60
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• 2021

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.