Test Reactor Irradiation Fuel Assembly Development in China

2017 ◽  
Author(s):  
Jie Ding ◽  
Yixiong Zheng ◽  
Yang Ding ◽  
Song Liu ◽  
Libing Zhu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Zirconium Alloys ◽  
Reactor Irradiation ◽  
Fuel Rods ◽  
Post Irradiation ◽  
Irradiation Growth ◽  
Corrosion Creep ◽  
Test Reactor ◽  
Irradiation Performance ◽  
Test Fuel

During the development of zirconium alloys, the irradiation in the test reactor is a critical step to comparison the irradiation properties of candidate alloys, such as corrosion, creep and irradiation growth. In this paper, a small scaled fuel assembly for test reactor irradiation is designed, which meets the needs of new zirconium alloys development. The irradiation fuel assembly (IFA) can be easily disassembled, and the test fuel rods or irradiation specimen can be easily replaced, which makes it possible to do the further post-irradiation examination in the hot cell to obtain the irradiation performance data. Now the IFA has finish fabrication and the test reactor irradiation program is planned to launch in 2017.

Post-irradiation examinations of dispersion fuel rods with modified zirconium alloys claddings

2019 ◽  
pp. 206-212
Author(s):  
G. V. Kulakov ◽  
Y. V. Konovalov ◽  
A. A. Kosaurov ◽  
M. M. Peregud ◽  
V. Y. Shishin ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Corrosion Resistance ◽  
Master Alloy ◽  
Zirconium Alloys ◽  
High Strength ◽  
Fuel Rods ◽  
Dispersion Fuel ◽  
Post Irradiation ◽  
Strength And Ductility

Modified zirconium alloys E635Mand E635opt based on E635 alloy (E635 was selected as master alloy) have been developed at Bochvar Institute. Fuel rods with such claddings were manufactured at Bochvar Institute and were irradiated at MIR reactor (SC RIAR, Dimitrovgrad). The results from the PIE performed at RIAR are presented. Such features of claddings as microstructures, corrosion resistance (width and structure of oxide), hydrogen contents, distribution of hydrides, mechanical properties were examined and discussed. Modifications of the alloy E635opt and E635M showed higher resistance to corrosion and hydrogen pick-up compared to the E635 alloy, while maintaining high strength and ductility. They have confirmed their prospects for use as cladding for fuel rods with enhanced characteristics.

scholarly journals POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

2013 ◽  
Vol 45 (7) ◽  
pp. 847-858 ◽  
Author(s):  
H.J. RYU ◽  
J.M. PARK ◽  
Y.J. JEONG ◽  
K.H. LEE ◽  
Y.S. LEE ◽  
...  
Keyword(s):  
Fuel Rods ◽  
Dispersion Fuel ◽  
Post Irradiation

2D Natural Convection and Radiation Heat Transfer Simulations of a PWR Fuel Assembly Within a Constant Temperature Support Structure

2006 ◽  
Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Fuel Assembly ◽  
Heat Generation ◽  
Temperature Difference ◽  
Wall Temperature ◽  
Radiation Heat Transfer ◽  
Radiation Heat ◽  
Fuel Rods ◽  
Water Reactor

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14×14 pressurized water reactor (PWR) spent nuclear fuel assembly within a square basket tube of a typical transport package were conducted using a commercial computational fluid dynamics package. The assembly is composed of 176 heat generating fuel rods and 5 larger guide tubes. The maximum cladding temperature was determined for a range of assembly heat generation rates and uniform basket wall temperatures, with both helium and nitrogen backfill gases. The results are compared with those from earlier simulations of a 7×7 boiling water reactor (BWR). Natural convection/radiation simulations exhibited measurably lower cladding temperatures only when nitrogen is the backfill gas and the wall temperature is below 100°C. The reduction in temperature is larger for the PWR assembly than it was for the BWR. For nitrogen backfill, a ten percent increase in the cladding emissivity (whose value is not well characterized) causes a 4.7% reduction in the maximum cladding to wall temperature difference in the PWR, compared to 4.3% in the BWR at a basket wall temperature of 400°C. Helium backfill exhibits reductions of 2.8% and 3.1% for PWR and BWR respectively. Simulations were performed in which each guide tube was replaced with four heat generating fuel rods, to give a homogeneous array. They show that the maximum cladding to wall temperature difference versus total heat generation within the assembly is not sensitive to this geometric variation.

scholarly journals Test reactor irradiation coordination

1995 ◽  
Author(s):  
D W Heartherly ◽  
I I Siman Tov ◽  
D W Sparks
Keyword(s):  
Reactor Irradiation ◽  
Test Reactor

Modeling of MOX Fuel Pellet-Clad Interaction Using ABAQUS

2002 ◽  
Author(s):  
Richard G. Ambrosek ◽  
Robert C. Pedersen ◽  
Amanda Maple
Keyword(s):  
Axial Length ◽  
Fuel Pellet ◽  
Solution Technique ◽  
Outer Diameter ◽  
Symmetric Model ◽  
Post Irradiation ◽  
Fuel Pellets ◽  
Mox Fuel ◽  
Test Reactor

Post-irradiation examination (PIE) has indicated an increase in the outer diameter of fuel pins being irradiated in the Advanced Test Reactor (ATR) for the MOX irradiation program. The diameter increase is the largest in the region between fuel pellets. The fuel pellet was modeled using PATRAN and the model was evaluated using ABAQUS, version 6.2. The results from the analysis indicate the non-uniform clad diameter is caused by interaction between the fuel pellet and the clad. The results also demonstrate that the interaction is not uniform over the pellet axial length, with the largest interaction occurring in the region of the pellet-pellet interface. Results were obtained for an axi-symmetric model and for a 1/8 pie shaped segment, using the coupled temperature-displacement solution technique.

Thermal Evaluations for Next Generation Nuclear Plant Fuel Testing

Volume 4 ◽  
2004 ◽  
Author(s):  
Richard G. Ambrosek ◽  
Debbie J. Utterbeck ◽  
Brandon Miller
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Nuclear Plant ◽  
Light Water Reactors ◽  
Next Generation ◽  
Preliminary Information ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Test Reactor ◽  
Test Fuel

The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility.

Projection of the Neutronic and Thermal Fuel Rod Behavior in a BWR

2004 ◽  
Author(s):  
Hector Hernandez Lopez ◽  
Javier Ortiz Villafuerte
Keyword(s):  
Fuel Assembly ◽  
Mechanical Performance ◽  
Nuclear Research ◽  
Operating Conditions ◽  
Fast Neutrons ◽  
Two Phase ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Thermal Limits ◽  
Operation Conditions

Currently, at the Instituto Nacional de Investigaciones Nucleares (National Institute for Nuclear Research) in Mexico, it is being developed a computational code for evaluating the neutronic, thermal and mechanical performance of a fuel element at several different operation conditions. The code is referred as to MCTP (Multigrupos con Temperaturas y Potencia), and is benchmarked against data from the Laguna Verde Nuclear Power Plant (LVNPP). In the code, the neutron flux is approximated by six groups of energy: one group in the thermal region (E < 0.625 eV), four in the resonances region (0.625 eV < E < 0.861 MeV), and one group in the fast region (E > 0.861 MeV). Thus, the code is able to determine the damage to the cladding due to fast neutrons. The temperature distribution is approximated in both axial and radial directions taking into account the changes in the coolant density, for both the single and two-phase regions in a BWR channel. It also considerate the changes in the thermal conductivity of all materials involved for the temperature calculations, as well as the temperature and density effects in the neutron cross sections. In the code, fuel rod burnup is evaluated. Also, plutonium production and poison production from fission. In this work, the neutronic and thermal performance of fuel rods in a 10×10 fuel assembly is evaluated. The fuel elements have a content of 235U. The fuel assembly was introduced to the unit 1 of LVNPP reactor core in the cycle 9 of operation, and will stay in during three cycles. In the analysis of fuel rod performance, the operating conditions are those for the cycle 9 and 10, whereas for the current cycle (cycle 11) the reactor is projected to operate during 460 days. The analysis for cycle 11 uses the actual location of the fuel assembly that will have in the core. The results show that the fuel rods analyzed did not reach the thermal limits during the cycles 9 and 10, as expected, and for cycle 11 the same thermal limits are not predicted to be reached.

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