scholarly journals Experimental determinations of the pre- and postirradiation thermal transport and thermal expansion properties of simulated fuel rods for an HTGR

1978 ◽  
Author(s):  
J.P. Moore ◽  
T.G. Godfrey ◽  
R.S. Graves ◽  
W.P. Eatherly ◽  
E.L. Jr. Long ◽  
...  
Keyword(s):  
Thermal Expansion ◽  
Thermal Transport ◽  
Fuel Rods ◽  
Simulated Fuel

scholarly journals Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

2020 ◽  
Vol 2020 ◽  
pp. 1-12
Author(s):  
Young-Hwan Kim ◽  
Yung-Zun Cho ◽  
Jin-Mok Hur
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Energy Research ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Compressive Stresses ◽  
Simulated Fuel ◽  
Ceramic Fuel ◽  
Experimental Approaches

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

Uncertainty evaluation of the thermal expansion of simulated fuel

2006 ◽  
Vol 447 (2) ◽  
pp. 197-201 ◽  
Author(s):  
Chang Je Park ◽  
Kweon Ho Kang ◽  
Kee Chan Song
Keyword(s):  
Thermal Expansion ◽  
Uncertainty Evaluation ◽  
Simulated Fuel

Uncertainty assessment for the thermal expansion of simulated fuel via the Monte Carlo method

2007 ◽  
Vol 455 (1-2) ◽  
pp. 114-118
Author(s):  
Chang Je Park ◽  
Kweon Ho Kang ◽  
Kee Chan Song ◽  
Myung Seung Yang
Keyword(s):  
Monte Carlo ◽  
Thermal Expansion ◽  
Monte Carlo Method ◽  
Uncertainty Assessment ◽  
Simulated Fuel ◽  
The Monte Carlo Method

Thermal Expansion of a Simulated Fuel with Fission Products Forming Solid Solutions

2006 ◽  
Vol 27 (1) ◽  
pp. 161-170 ◽  
Author(s):  
K. H. Kang ◽  
K. C. Song ◽  
M. S. Yang ◽  
S. H. Lee ◽  
J. B. Ko ◽  
...  
Keyword(s):  
Thermal Expansion ◽  
Solid Solutions ◽  
Fission Products ◽  
Simulated Fuel

scholarly journals KRITZ-1-Mk CRITICAL MEASUREMENTS AT TEMPERATURES FROM 20 °C TO 250 °C

2021 ◽  
Vol 247 ◽  
pp. 09028
Author(s):  
Dennis Mennerdahl
Keyword(s):  
Thermal Expansion ◽  
Temperature Dependence ◽  
Nuclear Data ◽  
Large Temperature ◽  
Positive Temperature ◽  
Fuel Rods ◽  
Water Reactor ◽  
Fuel Rod ◽  
Fuel Assemblies ◽  
Water Height

Benchmarks are needed to validate methods to account for temperature-dependence of nuclear data. An evaluation of 37 KRITZ-1-Mk critical water height measurements, together with associated iso-reactivity temperature effects and coefficients, is released with the 2019 Handbook of the International Reactor Physics Experiment Evaluation Project (IRPhEP). The KRITZ zero-power research reactor, operated between 1969 and 1975 in Studsvik (Sweden), was contained in a pressure vessel, allowing full size fuel assemblies or fuel rods in light water at temperatures up to 250 °C without boiling. Preliminary results were published in 1971 and 1972 for four series of altogether 37 measurements with Marviken (Boiling Heavy Water Reactor) UO2 fuel rods, each containing a 235U isotopic mass fraction of 1.35 %. Temperature was the predictor variable, while critical water height was the response variable. Each series was characterized by the fuel rod lattice design and by the soluble boron concentration in water. The KRITZ measurements were focused on temperature-dependence (differences). High measurement correlations reduced the ?k uncertainties, typically from 195 pcm to 40 pcm for a large temperature change. Thermal expansion of fuel and reactor components was not measured. Detailed and simple benchmarks include estimated thermal expansion as a simplification. Benchmark calculation results using JEFF-3.3 nuclear data reduce the large biases observed for older libraries but a remarkable positive temperature trend is observed for series 4. In 2019, Studsvik Nuclear released information on KRITZ-1-Mk and on other KRITZ-1 and KRITZ-2 critical measurements with Boiling Water Reactor fuel assemblies and fuel rod clusters.

scholarly journals Thermal expansion and its impacts on thermal transport in the FPU-α-β model

AIP Advances ◽  
2015 ◽  
Vol 5 (5) ◽  
pp. 053203 ◽  
Author(s):  
Xiaodong Cao ◽  
Dahai He ◽  
Hong Zhao ◽  
Bambi Hu
Keyword(s):  
Thermal Expansion ◽  
Thermal Transport
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