In-situ and post irradiation behavior of cryogenic temperature sensors under fast neutron fluence

2002 ◽  
Author(s):  
Y. P. Filippov
Keyword(s):  
Fast Neutron ◽  
Cryogenic Temperature ◽  
Neutron Fluence ◽  
Temperature Sensors ◽  
Post Irradiation ◽  
Irradiation Behavior

scholarly journals Proton irradiation and resistivity recovery in stages I & II of pure and carbon-doped Fe

2020 ◽  
Vol 27 ◽  
pp. 148
Author(s):  
Andreas Theodorou ◽  
M. Syskaki ◽  
Z. Kotsina ◽  
M. Axiotis ◽  
G. Apostolopoulos
Keyword(s):  
Electrical Resistivity ◽  
Cryogenic Temperature ◽  
Proton Irradiation ◽  
Radiation Defects ◽  
Tandem Accelerator ◽  
Carbon Doped ◽  
Post Irradiation ◽  
Solute Atoms ◽  
Mev Protons

Pure and C-doped Fe specimens were irradiated with 5 MeV protons at cryogenic temperature at the NCSR-"Demokritos" TANDEM accelerator in order to investigate the interactions between carbon atoms and radiation defects. During the subsequent post-irradiation isochronal annealing up to 180 K the defects start to migrate and interact either mutually or with the C impurities. The defect evolution is observed by in-situ electrical resistivity recovery measurements. Comparison of results from pure and C-doped Fe specimens reveals the effect of C solute atoms on the defect kinetics.

scholarly journals Thermal Predictions of the AGR-3/4 Experiment With Post Irradiation Examination Measured Time-Varying Gas Gaps

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki ◽  
Binh T. Pham
Keyword(s):  
Thermal Analysis ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Element Model ◽  
Post Irradiation ◽  
Measured Time ◽  
Dimensional Measurements ◽  
Test Reactor ◽  
Experiment Individual

A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) varying gas gaps. The experiment was irradiated at the advanced test reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program, which supports the development of the very high-temperature gas-cooled reactor under the advanced reactor technologies project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) tristructural-isotropic-fueled compacts were inserted into 12 separate capsules for the experiment. The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using PIE-measured time (fast neutron fluence) varying gas gaps and compare with experimentally measured thermocouple (TC) data. PIE-measured experimental data were used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all the fuel compacts, graphite holders, and all of the graphite rings used. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each nonfuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for each capsule.

scholarly journals Operando and in situ in a TEM imaging in a cryogenic temperature range

2021 ◽  
Vol 27 (S1) ◽  
pp. 386-387
Author(s):  
Martial Duchamp ◽  
Joseph Vas ◽  
Reinis Ignatans ◽  
Aaron David Mueller ◽  
Rohit Medwal ◽  
...  
Keyword(s):  
Cryogenic Temperature ◽  
Temperature Range

scholarly journals A Fast Neutron Fluence Reduction Strategy for Pressure Vessel Lifetime Extension

2000 ◽  
Vol 37 (sup1) ◽  
pp. 120-124 ◽  
Author(s):  
Jong Kyung Kim ◽  
Chang Ho Shin ◽  
Bo Kyun Seo ◽  
Myung Hyun Kim ◽  
Goung Jin Lee
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reduction Strategy ◽  
Lifetime Extension

scholarly journals A Method to Estimate the Fast-Neutron Fluence for the Hiroshima Atomic Bomb

1994 ◽  
Vol 63 (10) ◽  
pp. 3546-3547 ◽  
Author(s):  
Tokushi Shibata ◽  
Mineo Imamura ◽  
Seiichi Shibata ◽  
Yoshitomo Uwamino ◽  
Tohru Ohkubo ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Atomic Bomb ◽  
Neutron Fluence

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

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