• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4084-4094
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• 2022

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Nondestructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.941-950
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• 2013

The Korean Nuclear-Hydrogen Technology Development (NHTD) Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about $1030^{\circ}C$ in BOC (Beginning of Cycle) during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about $1.7{\times}10^{21}\;n/cm^2$, PIE (Post-Irradiation Examination) of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility). This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1602-1611
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• 2021

A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.3
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• pp.151-164
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• 2004

The changes of microstructural and mechanical properties were evaluated for the high burn-up fuel claddings after the neutron irradiation of $1.8{\sim}3.1{\times}10^{20}n/cm^2$ (E>1.0 MEV) in HANARO research reactor. After the irradiation, the spot-type dislocations (a-type dislocations) were easily observed in most claddings, and the density of the dislocations was different depending on the grains and was higher at grain boundaries than within grains. As the final annealing temperature increased, the density of spot-type dislocations increased and the line-type dislocations (c-type dislocations) which was perpendicular to the <0002> direction, appeared sporadically in some claddings. However, the types of precipitates in the fuel claddings after the irradiation were not changed from that in unirradiated claddings. The mechanical properties including the hardness, strength and elongation after the irradiation were changed due to the formation of spot-type dislocations. That is, the increase in hardness and strength as well as the decrease in elongation after the irradiation was occurred simultaneously with increasing the final annealing temperature. Owing to the Nb contribution to the formation of spot-type dislocation during the irradiation, the increase in hardness and strength in higher Nb-contained Zr alloys after the irradiation was higher than that in lower Nb-contained Zr alloys.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• Journal of Welding and Joining
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• v.21 no.6
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• pp.20-25
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• 2003

Various welding methods such as Gas Tungsten Arc Welding(GTAW), magnetic force electrical resistance welding and Laser Beam Welding(LBW) are now available for end cap closure of nuclear fuel rods. Even though the resistance and GTA welding processes are widely used in manufacturing commercial fuel rods, they can not be recommended for the remote seal welding of fuel rods in the hot cell Facility due to the complexity of the electrode alignment, the difficulty in replacing parts in a remote manner and the large heat input for the thin sheath. Therefore, the Nd:YAG laser system using optical fiber transmission was selected for the end cap welding of irradiation fuel rods in the hot cell. The remote laser welding apparatus in the hot cell Facility was developed using a pulsed Nd:YAG laser of 500 watt average power with an optical fiber transmission. The weldment quality such as microstructure and mechanical strength was satisfactory. The optimum conditions of laser welding for encapsulating irradiation fuel rods in the hot cell were obtained.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.