Corrosion and stress corrosion cracking in supercritical water

Was, Gary S.; Ampornrat, Pantip; Gupta, Gaurav; Teysseyre, S.; West, E. A.; Allen, Todd R.; Sridharan, Kumar; Tan, Lizhen; Chen, Y.; Ren, Xuan; Pister, C.

doi:10.1016/j.jnucmat.2007.05.017

Cited by 355 publications

(189 citation statements)

References 13 publications

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“…Even though a lot of supercritical technology for the secondary side (such as turbines) already exists because of the fossil fuel based power plants, there still are large uncertainties with regards to corrosion and thermo-hydraulics of supercritical water in these tight lattice bundles; see e.g. [19]- [20].…”

Section: Introductionmentioning

confidence: 99%

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

T’Joen

Rohde

2012

Nuclear Engineering and Design

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The HPLWR (High Performance Light Water Reactor) is the European concept design for a SCWR (SuperCritical Water Reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction -mass flow rate or gravity -mass flow rate feedback of the system), or they can be coupled thermo-hydraulic -neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic -neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 seconds. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating conditions. A clear instability zone was found within the stability plane, which seems to be similar to that of a BWR. Experimental data on the stability of a supercritical loop is rare in open literature, and these data could serve as an important benchmark tool for existing codes and models.

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Section: Introductionmentioning

confidence: 99%

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

T’Joen

Rohde

2012

Nuclear Engineering and Design

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“…These changes are caused by radiation-induced evolution of compositions and microstructures. The main effects of radiation on reactor materials are: (1) dimensional change associated with gas bubble swelling, void swelling, grain growth, and creep; [1][2][3][4][5] (2) loss of ductility and increase in ductile-brittle transition temperature (DBTT) due to the formation of secondphase precipitates, self-interstitial atomic (SIA) loops, and dislocation networks; 6,7 (3) oxidation and corrosion accelerated by high temperature, fission products, and radiation damage; [8][9][10] and (4) local and bulk changes in chemical composition, including irradiation-enhanced segregation of alloy components and phase separation.…”

Section: Introductionmentioning

confidence: 99%

“…9,[18][19][20][21] The radiation-induced heterogeneity of the microstructures depends on the initial phase and defect structure of the fresh (non-irradiated) materials and on the type and severity of the radiation environment. Fundamental understanding of heterogeneous three-dimensional microstructure evolution is crucial to the development of advanced radiation tolerant materials that can significantly improve the life extension of current reactor fleet and impact advanced reactor designs.…”

Section: Introductionmentioning

confidence: 99%

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

Sun

et al. 2017

npj Comput Mater

114

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Complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the phase field method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiated nuclear materials are reviewed. The review shows that (1) Phase field models can correctly describe important phenomena such as spatial-dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; (2) The phase field method can qualitatively and quantitatively simulate two-dimensional and three-dimensional microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and (3) The Phase field method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the phase field method, as applied to irradiation effects in nuclear materials. npj Computational Materials (2017) 3:16 ; doi:10.1038/s41524-017-0018-y INTRODUCTIONHigh energy particle (such as neutron, ion, and electron) radiation can create major changes in the shape and thermo-mechanical properties of nuclear fuels and structural components of nuclear reactors. These changes are caused by radiation-induced evolution of compositions and microstructures. The main effects of radiation on reactor materials are: (1) dimensional change associated with gas bubble swelling, void swelling, grain growth, and creep; 1-5 (2) loss of ductility and increase in ductile-brittle transition temperature (DBTT) due to the formation of secondphase precipitates, self-interstitial atomic (SIA) loops, and dislocation networks; 6, 7 (3) oxidation and corrosion accelerated by high temperature, fission products, and radiation damage; 8-10 and (4) local and bulk changes in chemical composition, including irradiation-enhanced segregation of alloy components and phase separation.11-17 Figure 1 shows typical microstructures observed in irradiated materials.9, 18-21 The radiation-induced heterogeneity of the microstructures depends on the initial phase and defect structure of the fresh (non-irradiated) materials and on the type and severity of the radiation environment. Fundamental understanding of heterogeneous three-dimensional microstructure evolution is crucial to the development of advanced radiation tolerant materials that can significa...

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“…The temperature limit in oxidizing environments for these materials to maintain their oxidation resistance and mechanical properties can be as high as 1000°C [8]. In austenitic stainless steels, particularly, the protective surface oxide generally assumes a 2 or 3 layered structure [9] with magnetite on the outer surface and an inner layer of oxide with elevated chromium content forming either a corundum-type (Cr,M) 2 For personal use only.…”

Section: Introductionmentioning

confidence: 99%

“…Usually there is weight gain in the samples after exposure to SCW because of oxidation. The weight gains over time, commonly used to quantify oxidation rate, are higher with the increase in temperature, pressure, and oxygen content in water for almost all the alloys studied [9]. Otoguro et al [11] studied the behaviour of austenitic stainless in SCW and subcritical conditions.…”

Section: Introductionmentioning

confidence: 99%

Effect of Water or Steam Pressure on the Oxidation Behaviour of Alloy 625 and A286 at 625 °c

Huang¹,

Sanchez²

2016

CNL Nuclear Review

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Corrosion and stress corrosion cracking in supercritical water

Cited by 355 publications

References 13 publications

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

Effect of Water or Steam Pressure on the Oxidation Behaviour of Alloy 625 and A286 at 625 °c

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