Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages...Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on. However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge. Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR) in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP). Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M) steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials.展开更多
The fluoride volatility method (FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in R&D of the FVM is corrosion due to the presence of molten salt and corros...The fluoride volatility method (FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in R&D of the FVM is corrosion due to the presence of molten salt and corrosive gases at high temperature. In this work, a frozen-wall technique was proposed to produce a physical barrier between construction materials and corrosive reactants. The protective performance of the frozen wall against molten salt was assessed using FLiNaK molten salt with introduced fluorine gas, which was regarded as a simulation of the FVM process. SS304, SS316L, Inconel 600 and graphite were chosen as the test samples. The extent of corrosion was characterized by an analysis of weight loss and scanning electron microscope studies. All four test samples suffered severe corrosion in the molten salt phase with the corrosion resistance as: Inconel 600>SS316L>graphite>SS304. The presence of the frozen wall could protect materials against corrosion by molten salt and corrosive gases, and compared with materials exposed to molten salt, the corrosion rates of materials protected by the frozen wall were decreased by at least one order of magnitude.展开更多
文摘Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on. However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge. Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR) in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP). Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M) steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials.
基金supported by the Strategic Priority Research Program of the Chinese Academy of Science(No.XDA02030000)
文摘The fluoride volatility method (FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in R&D of the FVM is corrosion due to the presence of molten salt and corrosive gases at high temperature. In this work, a frozen-wall technique was proposed to produce a physical barrier between construction materials and corrosive reactants. The protective performance of the frozen wall against molten salt was assessed using FLiNaK molten salt with introduced fluorine gas, which was regarded as a simulation of the FVM process. SS304, SS316L, Inconel 600 and graphite were chosen as the test samples. The extent of corrosion was characterized by an analysis of weight loss and scanning electron microscope studies. All four test samples suffered severe corrosion in the molten salt phase with the corrosion resistance as: Inconel 600>SS316L>graphite>SS304. The presence of the frozen wall could protect materials against corrosion by molten salt and corrosive gases, and compared with materials exposed to molten salt, the corrosion rates of materials protected by the frozen wall were decreased by at least one order of magnitude.
