Synthesis of Zr-Nb-Mo-Y-Sn alloys for fuel cladding material for PWR small-medium nuclear reactor

2021 ◽  
Author(s):  
Hakimul Wafda ◽  
Djoko Hadi Prajitno ◽  
Very Trisnawan
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cladding ◽  
Cladding Material

scholarly journals Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

2008 ◽  
Vol 2008 ◽  
pp. 1-9 ◽  
Author(s):  
Enrico Zio ◽  
Francesco Di Maio
Keyword(s):  
Confidence Interval ◽  
Order Statistics ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Fuel Cladding ◽  
Complete Group ◽  
Bootstrap Technique ◽  
Safety Margins ◽  
Group Distribution

In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

Aging Effect of Chromium Coated Zircaloy-4 Accident Tolerant Fuel Cladding Material

2019 ◽  
Author(s):  
C. Castano ◽  
J. Rojas ◽  
R. Uhorchuk ◽  
R. Umretiya ◽  
S. Vargas
Keyword(s):  
Aging Effect ◽  
Fuel Cladding ◽  
Cladding Material

Results of a New Production of Nuclear-Grade 1.4970 ‘15-15Ti’ Stainless Steel Fuel Cladding Tubes for GEN IV Reactors

2014 ◽  
Author(s):  
Rémi Delville ◽  
Erich Stergar ◽  
Marc Verwerft
Keyword(s):  
Stainless Steel ◽  
Microstructural Analysis ◽  
Nuclear Grade ◽  
Fuel Cladding ◽  
Fast Reactors ◽  
New Production ◽  
Steel Grades ◽  
Gen Iv ◽  
Past Experiences ◽  
Cladding Material

Titanium stabilized 1.4970 ‘15-15Ti’ stainless steel cladding is the primary choice for fuel cladding of several current fast spectrum research reactor projects. The choice of cladding material is based on past experiences and the availability of material databases from similar steel grades that have proven their reliability in past sodium-cooled fast reactors programs. However the last production in Europe of nuclear-grade 15-15Ti was more than 20 years ago and it remained to be seen if the know-how to produce such steel with the strict specifications for nuclear fuel cladding was still available. Results of a new production of nuclear-grade 15-15Ti cladding tubes at Sandvik for SCK•CEN is presented in this paper. It is shown that materials properties are within the strict specifications similar to the ones used during past sodium-cooled fast reactors programs. Special attention is given to microstructural analysis of the newly produced steel which contains a large number of stabilizing Ti(C-N) precipitates known for their beneficial effect on in-pile material properties and thermal creep. Results from metallography, SEM and TEM investigations are presented.

Development of Diffusion Barrier Coatings for Mitigation of Fuel-Cladding Chemical Interactions

2012 ◽  
Vol 507 ◽  
pp. 3-7 ◽  
Author(s):  
Vahid Firouzdor ◽  
Lucas Wilson ◽  
Kumar Sridharan ◽  
Brandon Semerau ◽  
Benjamin Hauch ◽  
...  
Keyword(s):  
Diffusion Couple ◽  
Diffusion Barrier ◽  
Nuclear Reactor ◽  
Fuel Cladding ◽  
Oxide Coatings ◽  
Chemical Interactions ◽  
Barrier Coatings ◽  
Deposition Parameters ◽  
Diffusion Characteristics ◽  
Melting Point Eutectic

Fuel Cladding Chemical Interactions (FCCI) in a nuclear reactor occur due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and the formation of low melting point eutectic compounds. Deposition of diffusion barrier coatings of a thin oxide on the inner surface of the cladding can potentially reduce or delay the onset of FCCI. This study examines the feasibility of using nanofluid-based electrophoretic deposition (EPD) process to deposit coatings of titanium oxide, yttria-stabilized zirconia (YSZ) and vanadium oxide. The deposition parameters, including the nanofluid composition, current, and voltage were optimized for each coating material using test flat substrates of T91 ferritic-martensitic steel. Diffusion characteristics of the coatings were investigated by diffusion couple experiments using the fuel surrogate cerium. These diffusion couple studies performed in the temperature range of 560°C and 585°C showed that the oxide coatings significantly reduce the solid state inter-diffusion between cerium to steel.

scholarly journals Effect Of -Quenching On Oxidation Resistance Of Zirconium Alloyzrnbmoge For Fuel Cladding Material

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
B. Bandriyana
Keyword(s):  
Oxidation Resistance ◽  
Growth Characteristic ◽  
Optical Microscope ◽  
Magnetic Suspension ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Corrosion Properties ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Cladding Material

<p>Effect of b-quenching of Zr-2.5Nb-0.5Mo-0.1Ge alloy used for advanced fuel cladding material of Pressurized Water Reactor (PWR) was investigated. The aim of this research isto improve the mechanical and corrosion properties through modificationof the alloy with regard to high reactor burn up. The quenching process was conducted by heating the sample at temperature of 950 <sup>o</sup>Cand soaking 2.5 hours,followed by quenching in water at room temperature and then continued with annealing process at 500 and 600<sup>o</sup>C. The change of hardness and oxidation resistance were characterized using optical microscope and scanning electron microscope (SEM). The effect on the oxidation resistance was investigated by the high temperature oxidation test using the MSB (Magnetic Suspension Balance) at 700 <sup>o</sup>C for 5 hours. The hardness increased from 217 VHN to 265 VHN after quenching due to grain refinement and precipitation hardening. The oxidation rate followed the typical parabolic growth characteristic. The formation of thin layer was considered to be a stable oxide ZrO<sub>2</sub>that influenced the oxidation characteristic and increasing the hardness of the alloy.</p>

Postirradiation Ductility Demonstration Tests of Weapons-Derived MOX Fuel Cladding

2004 ◽  
Author(s):  
W. J. McAfee ◽  
W. R. Hendrich ◽  
T. E. McGreevy ◽  
C. A. Baldwin ◽  
N. H. Packan
Keyword(s):  
Synergistic Effects ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Method ◽  
Test Results ◽  
Fuel Cladding ◽  
Oak Ridge ◽  
Mox Fuel ◽  
Uranium Plutonium ◽  
Cladding Material

The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (FMDP) is pursuing reactor irradiation of mixed uranium-plutonium oxide (MOX) fuel for disposal of surplus weapons-usable plutonium. Since most of the MOX fuel utilization experience has been with reactor-grade plutonium, it is desired to demonstrate that the unique properties of the surplus weapons-derived or weapons-grade (WG) plutonium do not compromise the applicability of this MOX experience base. A related question to be addressed for weapons-derived MOX fuel is that of ductility loss of the cladding. While irradiation induced loss of ductility has long been known and quantified for many cladding materials, the potential synergistic effects of irradiation and the unique constituents (i.e., gallium) of weapons-derived MOX fuel are not known. As part of an extensive fuel qualification research program conducted by Oak Ridge National Laboratory (ORNL), a new test method was developed and validated to measure the room temperature ductility and hoop tensile properties of MOX fuel cladding. The cladding material is a zirconium alloy designated as Zr-4 manufactured by Sandvick Corporation. This paper is a summary of the test method developed and of demonstration test results obtained for MOX cladding irradiated to 21 GWd/MT [7 × 1020 n/cm2 (E &gt; 1 MeV)].

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