• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.270-271
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• 2005

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.54-60
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• 2011

A spacer grid, which supports nuclear fuel rods laterally and vertically with a friction grip, is one of the most important structural components in a PWR fuel. The form of grid strap and supporting parts such as grid spring and dimple is known to be closely related with the mechanical/structural performance of spacer grid and nuclear fuel assembly. In this study, reviewing various research results for enhancing the performance of the spacer grid, some structural design considerations and research directions on the spacer grid assembly are suggested for further study.

• Proceedings of the KWS Conference
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• 2005.06a
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• pp.109-111
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• 2005

Nuclear fuel assemblies for pressurized water reactors(PWR) are loaded in the reactor core throughout the residence time of three to five years. A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, laser welding qualities of the spacer grid assembly welded by several welding companies, such as weld strength, weld penetration depth, and weld bead size, are examined and compared.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3563-3579
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• 2021

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.353-362
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• 2014

The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The "Fast Total Monte Carlo" method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on $k_{\infty}$, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.

• Proceedings of the KWS Conference
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• 2003.11a
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• pp.55-57
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• 2003

The fuel assemblies as the nuclear fuel for the pressurized water reactor(PWR) are loaded in the reactor core throughout the residence time of three to five years. The fuel assembly is manufactured using special welding processes and under strict quality assurance and control systems. In this paper welding parts, welding methods, and welding tests for the integrity of the PWR fuel assemblies are introduced.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.212-221
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• 1998

In the present study, a method for the dynamic analysis of a reactor core is developed. Peak responses for the motions induced from earthquake are obtained for a core model. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are investigated. Prediction of fuel assembly stress during an earthquake requires development of a fuel assembly stress analysis model capable of interfacing with the models and results discussed in the dynamic analysis of a reactor core. This analysis uses beam characteristics which describe the overall fuel assembly response. The stress analysis method and its application for the case of an increased seismic level are also presented.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.91-100
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• 2005

Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional $UO_2$ fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M$9^{th}$ equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with $UO_2$. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the $UO_2$ core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to $UO_2$ in terms of economics.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.156-164
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• 2005

The challenges China is facing in energy security are briefly discussed. Then, the development of nuclear power in China in the first half of 21 st century is envisioned, and it is expected that Generation-3 PWR nuclear power plants (NPPs) would be the leading units of nuclear power in the coming $30\~40$ years. As part of the nuclear power program, the R&D work on nuclear fuel cycle is generally proposed.