Weight Gain and Hydrogen Absorption in Supercritical Water At 500°C of Chromium-Coated Zirconium-Based Alloys: Transverse vs Longitudinal Direction

2021 ◽  
Author(s):  
Kittima Khumsa-Ang ◽  
Stephane Rousseau ◽  
Oksana Shiman
Keyword(s):  
Weight Gain ◽  
Supercritical Water ◽  
Hydrogen Absorption ◽  
Corrosion Behaviour ◽  
Reactor Core ◽  
Neutron Cross Section ◽  
Zirconium Alloys ◽  
Longitudinal Direction ◽  
General Corrosion ◽  
Enriched Uranium

Abstract Canadian Nuclear Laboratories (CNL) has an on-going Research & Development programme to support the development of a scaled-down 300 MWe version of the Canadian Super-Critical Water Reactor (SCWR) concept. The 300 MWe and 170-channel reactor core concept uses low enriched uranium fuel and features a maximum cladding temperature of 500°C. Our goal is to test surfacemodified zirconium alloys for use as fuel cladding. Zirconium alloys are attractive as they offer low neutron cross section thereby allowing the use of low enriched fuel. In this paper, we report on the results of general corrosion experiments used to evaluate chromiumcoated zirconium-based alloys in the two chemistries (630 ug/kg O2 in both deaerated and lithiated supercritical water). These experiments were conducted in a refreshed autoclave at 500°C and 23.5 MPa. After exposure, the weight gain and the hydrogen absorption were examined. At adequate coating thickness, longitudinal and transverse coupons show similar corrosion behaviour with improved corrosion resistance compared to uncoated coupons. The measured concentrations of hydrogen absorption are higher for the transverse coupons. Alkaline treatment resulted in higher weight gains than was found in pure oxygenated supercritical water.

General Corrosion of Chromium-Coated Zirconium- and Titanium-Based Alloys in Supercritical Water at 500 °C

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
K. Khumsa-Ang ◽  
M. Edwards ◽  
S. Rousseau
Keyword(s):  
Supercritical Water ◽  
Fuel Cycle ◽  
Neutron Cross Section ◽  
Zirconium Alloys ◽  
Oxide Thickness ◽  
General Corrosion ◽  
Coating Materials ◽  
Grade 5 ◽  
Oxide Phases ◽  
Enriched Uranium

Abstract The 300 MWel small Canadian supercritical water-cooled reactor (SCWR), which is a scaled-down version of the original 1200 MWel concept, has a smaller core, uses low enriched uranium fuel instead of a plutonium–thorium fuel, and features a lower (maximum) cladding temperature of 500 °C. The lower cladding temperature may permit the use of different alloys, including zirconium alloys, which had been ruled out as candidates for the Canadian SCWR, whose cladding temperature may reach 850 °C. The potential to use zirconium alloys is exciting because they have a low neutron cross section, which in turn means that fewer neutrons are lost, and the fuel can be used more efficiently. One advantage, for example,, is that the fuel cycle can be lengthened. In this paper, we report on the results of corrosion experiments used to screen zirconium- and titanium-based alloys as well as corrosion-resistant coating materials such as Cr and Al as potential candidates for fuel cladding in the small Canadian SCWR. These experiments were conducted in a refreshed autoclave in deaerated supercritical water at 500 °C and 23.5 MPa. After exposure, the weight gain was measured, and the oxide thickness and the oxide phases were examined. Of all materials, the coated and uncoated Ti-grade 2 and Ti-grade 5 alloys met our screening qualification criteria, however, Al/Cr-coated zirconium coupons showed notable improvement and will be explored further in future testing.

General Corrosion Properties of Modified PNC1520 Austenitic Stainless Steel in Supercritical Water as a Fuel Cladding Candidate Material for Supercritical Water Reactor

2010 ◽  
Vol 73 ◽  
pp. 72-77
Author(s):  
Yoshihisa Nakazono ◽  
Takeo Iwai ◽  
Hiroaki Abe
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Supercritical Water ◽  
Reactor Core ◽  
General Corrosion ◽  
Fuel Cladding ◽  
Corrosion Properties ◽  
Energy Agency ◽  
Water Reactor ◽  
Supercritical Water Reactor

The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S), the Ti-additional type of PNC1520 (1520Ti) and the Zr-additional type of PNC1520 (1520Zr) by using a supercritical water autoclave. In view of general corrosion, 1520Zr may have larger possibility than 1520S and 1520Ti to adopt a supercritical water reactor core fuel cladding.

Control of hydrogen absorption by nickel films obtained upon magnetic spraying of zirconium alloy using the thermoEMF method

2020 ◽  
Vol 86 (8) ◽  
pp. 32-37
Author(s):  
V. V. Larionov ◽  
Xu Shupeng ◽  
V. N. Kudiyarov
Keyword(s):  
Magnetron Sputtering ◽  
Hydrogen Concentration ◽  
Hydrogen Absorption ◽  
Zirconium Alloy ◽  
Hydrogen Adsorption ◽  
Zirconium Alloys ◽  
Nickel Film ◽  
Protective Film ◽  
Nickel Films ◽  
Hydrogen Penetration

Nickel films formed on the surface of zirconium alloys are often used to protect materials against hydrogen penetration. Hydrogen adsorption on nickel is faster since the latter actively interacts with hydrogen, oxidizes and forms a protective film. The goal of the study is to develop a method providing control of hydrogen absorption by nickel films during vacuum-magnetron sputtering and hydrogenation via measuring thermoEMF. Zirconium alloy E110 was saturated from the gas phase with hydrogen at a temperature of 350°C and a pressure of 2 atm. A specialized Rainbow Spectrum unit was used for coating. It is shown that a nickel film present on the surface significantly affects the hydrogen penetration into the alloy. A coating with a thickness of more than 2 μm deposited by magnetron sputtering on the surface of a zirconium alloy with 1% Nb, almost completely protects the alloy against hydrogen penetration. The magnitude of thermoemf depends on the hydrogen concentration in the zirconium alloy and film thickness. An analysis of the hysteresis width of the thermoEMF temperature loop and a method for determining the effective activation energy of the conductivity of a hydrogenated material coated with a nickel film are presented. The results of the study can be used in assessing the hydrogen concentration and, hence, corrosion protection of the material.

Corrosion Behavior of Model Zirconium Alloys in Deaerated Supercritical Water at 500°C

CORROSION ◽  
2007 ◽  
Vol 63 (6) ◽  
pp. 577-590 ◽  
Author(s):  
Q. Peng ◽  
E. Gartner ◽  
J. T. Busby ◽  
A. T. Motta ◽  
G. S. Was
Keyword(s):  
Corrosion Behavior ◽  
Supercritical Water ◽  
Zirconium Alloys

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

Characterization of Alloy 214 After Exposure to Superheated Steam at 800 °C

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Nick Tepylo ◽  
Rainier Garcia Sanchez ◽  
Xiao Huang
Keyword(s):  
Weight Gain ◽  
Supercritical Water ◽  
Weight Change ◽  
Superheated Steam ◽  
Substrate Surface ◽  
Surface Oxidation ◽  
Metal Substrate ◽  
Cr Content

In this study, an Al-containing alloy 214 was evaluated in superheated steam at 800 °C for a duration of 600 h. The purpose of using superheated steam was to simulate the supercritical water (SCW) condition at higher temperatures where no commercial SCW rig is currently capable of reaching (800 °C and beyond). After exposure to superheated steam, the weight change and surface oxidation were analyzed. Alloy 214 experienced greater weight gain than IN 625 and Ni20Cr5Al, due to its low Cr content. Formation of both Cr2O3 and Al2O3 was observed on the surface after 300 and 600 h of exposure. However, as exposure progressed, more Ni and Mn-containing spinel started to form, signaling Cr and Al depletion on the metal substrate surface.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

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