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.

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.

Influence of Changes in Pressure and Temperature of Supercritical Water on the Susceptibility to Stress Corrosion Cracking of 316L Austenitic Stainless Steel

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Alberto Sáez-Maderuelo ◽  
Dolores Gómez-Briceño ◽  
César Maffiotte
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Stress Corrosion ◽  
Supercritical Water ◽  
Stress Corrosion Cracking ◽  
High Efficiency ◽  
Austenitic Stainless Steels ◽  
Corrosion Cracking ◽  
General Corrosion ◽  
316L Austenitic Stainless Steel

The supercritical water reactor (SCWR) is one of the Generation IV designs. The SCWR is characterized by its high efficiency, low waste production, and simple design. Despite the suitable properties of supercritical water as a coolant, its physicochemical properties change sharply with pressure and temperature in the supercritical region. For this reason, there are many doubts about how changes in these variables affect the behavior of the materials to general corrosion or to specific types of corrosion such as stress corrosion cracking (SCC). Austenitic stainless steels are candidate materials to build the SCWR due to their optimum behavior in the light water reactors (LWRs). Nevertheless, their behavior under the SCWR conditions is not well known. First, the objective of this work was to study the SCC behavior of austenitic stainless steel 316 type L in deaerated supercritical water at 400°C/25  MPa and 30 MPa and 500°C/25  MPa to determine how variations in pressure and temperature influence its behavior with regard to SCC and to make progress in the understanding of mechanisms involved in SCC processes in this environment. Second, the oxide layer formed at 400°C/30  MPa/<10  ppb O2 was analyzed to gain some insight into these processes.

scholarly journals Desain Inti Reaktor SCWR (Supercritical Water Reactor) Model Teras Silinder (r, z) dengan Bahan Bakar Thorium Hasil Daur Ulang

2019 ◽  
Vol 7 (2) ◽  
pp. 215-222
Author(s):  
Setiyaningsih Setiyaningsih ◽  
◽  
Yanti Yulianti ◽  
Simon Sembiring ◽  
◽  
...  
Keyword(s):  
Power Density ◽  
Supercritical Water ◽  
Power Factor ◽  
Critical Condition ◽  
Peak Power ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Reactor Model ◽  
Water Reactor ◽  
Supercritical Water Reactor

The Research of the supercritical water reactor (SCWR) core design of the cylindrical core model (r, z) using the SRAC program has been done. The SRAC basic code was PIJ and CITATION. PIJ was used to calculate the fuel level and CITATION was used to calculate the reactor core level. The calculation of the reactor core has been done on the 1/4 cylinder core (r, z) and the geometry of the fuel cell was the cylindrical cell. Reactor fuel material was thorium burned 40 GWd/t and 30 GWd/t. The neutron parameters in this research were fuel enrichment, burn up, reactor core size, reactor core configurations, multiplication factor, and power density distribution. Multiplication factor (k-effective) in this research was 1.000004, which is reactor was in a critical condition. The reactor core in critical condition had the size of radius (r) was 130 cm, height (z) was 270 cm and fuel enrichment 2.8262%. The maximum power density was 130.0808 Watts /cm3 which was located at a radius of 25 cm and 135 cm high. The peak power factor in the radial direction was 1.6063 and the peak power factor in the axial direction was 1.3189.

GIBSON TUBE DUPLEX 2205

Alloy Digest ◽  
2001 ◽  
Vol 50 (1) ◽  
Keyword(s):  
Stainless Steel ◽  
Corrosion Resistance ◽  
Austenitic Stainless Steel ◽  
Heat Exchangers ◽  
Heat Treating ◽  
General Corrosion ◽  
Bend Strength ◽  
Duplex Alloy ◽  
Resistance To Stress ◽  
Desalination Plants

Abstract Duplex alloy 2205 is a nitrogen-enhanced, ferritic-austenitic stainless steel. It has high resistance to stress-corrosion cracking (SCC); pitting; crevice and general corrosion. It has roughly twice the yield strength of austenitic stainless grades and it has good weldability. Some uses include heat exchangers, downhole instrument tubing, and applications in desalination plants. All Gibson Tube Duplex 2205 products meet the tighter composition limits of UNS S32205 to enhance corrosion resistance. This datasheet provides information on composition, physical properties, elasticity, tensile properties, and bend strength. It also includes information on corrosion resistance as well as heat treating and machining. Filing Code: SS-813. Producer or source: Gibson Tube.

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