David Guzonas

Assessment of Candidate Fuel Cladding Alloys for the Canadian Supercritical Water-Cooled Reactor Concept

Selecting and qualifying a fuel cladding material for the Canadian supercritical water-cooled reactor (SCWR) concept remains the most significant materials challenge to be overcome. The peak cladding temperature in the Canadian SCWR concept is predicted to be as high as 800°C. While advanced materials show promise for future deployment, currently, the best options available are austenitic stainless steels and nickel-based alloys. Many of these alloys were extensively studied for use as fuel cladding materials in the 1960s, as part of programs to develop nuclear superheated steam reactors. After extensive out-of-pile testing and consideration of the existing data, five alloys (347 SS, 310 SS, Alloy 800H, Alloy 625, and Alloy 214) were selected for more detailed assessment using a combination of literature surveys and targeted testing to fill in major knowledge gaps. Wherever possible, performance criteria were developed for key materials properties. This paper summarizes the methodology used for the assessment and presents the key results, which show that 310 SS, Alloy 800H, and Alloy 625 would all be expected to give acceptable performance in the Canadian SCWR concept.

PROPERTIES OF AQUEOUS SYSTEMS RELEVANT TO THE SCWR VIA MOLECULAR DYNAMICS SIMULATIONS

Supercritical water (SCW) is the intended heat transfer fluid in the proposed GEN-IV supercritical water cooled reactor (SCWR). The oxidative environment poses challenges in choosing appropriate de...

Combined Effect of Irradiation, Temperature, and Water Coolant Flow on Corrosion of Zr-, Ni–Cr-, and Fe–Cr-Based Alloys

Investigation of the role of irradiation on the corrosion resistance of structural alloys is of vital importance for selection of supercritical water-cooled reactor (SCWR) materials. Gamma heating under SCWR conditions, which induces enhancement of radiolysis and corrosion kinetics at interfaces, can be efficiently simulated by electron beam irradiation over a wide range of deposited dose and temperature. The NSC KIPT-sited Canada–Ukraine Electron Irradiation Test Facility (CU-EITF) still remains the only operating facility capable of in situ irradiation of specimens in a supercritical water (SCW) natural circulation loop. This paper reports the results of postirradiation studies of Zr–1%Nb and Ni–Cr Inconel 690/52MSS alloys after a ~500-h-long exposure in the CU-EITF in the near-critical (23.5 MPa/360–385°C) regime. Results of scanning electron microscopy (SEM) studies of the sample microstructure are presented along with those of the electron-irradiated loop piping, SS X18H10T. The results of corrosion tests under electron-irradiation are correlated to the calculated three-dimensional (3D) fields of absorbed dose and temperature and to the reference data obtained in-pile for topical materials. The paper also discusses the prospects for the use of the CU-EITF facility within a cooperative SCWR program and presents an outlook of the facility development.

Environmentally assisted cracking

Environmentally assisted cracking (EAC) is a complex phenomenon driven by the synergistic interaction of mechanical, chemical and metallurgical factors. The complex interplay between causative factors makes experimental measurements difficult, and the state of knowledge on EAC under supercritical water-cooled reactor (SCWR) conditions is not as well advanced as that of general corrosion. This chapter discusses the effects of the three key causative factors (environment, material, and mechanical) on the occurrence of EAC in supercritical water, focussing on candidate SCWR alloys and expected SCWR in-core conditions. Possible differences in mechanisms in the near-critical and higher temperature regimes are highlighted.

Corrosion phenomena induced by supercritical water in Generation IV nuclear reactors

The various supercritical water-cooled reactor concepts being developed under the Generation IV International Forum are the natural evolution of the water-cooled reactor technology that has successfully supplied the majority of nuclear-based electricity since the dawn of commercial nuclear power generation. The materials challenges that must be addressed in the development of a supercritical water-cooled reactor are in most respects the same as those experienced by the current generations of water-cooled reactors. This chapter summarizes current knowledge of corrosion and environmentally assisted cracking phenomena under the conditions expected in the core of a supercritical water-cooled reactor, with an emphasis on recent advances in experiment and modeling.

Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

Most supercritical water-cooled reactor (SCWR) concepts being considered as part of the Generation IV initiative are direct cycle. In the event of a fuel defect, the coolant will contact the fuel pellet, potentially releasing fission products and actinides into the coolant and transporting them to the turbines. At the high pressure (25 MPa) in an SCWR, the coolant does not undergo a phase change as it passes through the critical temperature in the core, and nongaseous species may be transported out of the core and deposited on out-of-core components, leading to increased worker dose. It is therefore important to identify species with a high risk of release and develop models of their transport and deposition behavior. This paper presents the results of preliminary leaching tests in SCW of U-Th simulated fuel pellets prepared from natural U and Th containing representative concentrations of the (inactive) oxides of fission products corresponding to a fuel burnup of 60 GWd/ton. The results show that Sr and Ba are released at relatively high concentrations at 400°C and 500°C.

Radiolysis of Supercritical Water at 400°C: A Sensitivity Study of the Density Dependence of the Yield of Hydrated Electrons on the (eaq−+eaq−) Reaction Rate Constant

Background: The temperature dependence of the rate constant (k) of the bimolecular reaction of two hydrated electrons (e-aq) measured in alkaline water exhibits an abrupt drop between 150 and 200 oC; above 250 oC, it is too small to be measured reliably. Although this result is well established, the applicability of this sudden drop in k(e-aq + e-aq) above 150 oC, as recommended by Bartels and co-workers, to neutral or slightly acidic solution still remains uncertain. Recent work by Hatomoto et al. combined ultrashort pulse radiolysis experiments with spur diffusion kinetic model simulations; this work suggested that in near-neutral water the abrupt change in k above 150 oC does not occur and that k should increase, rather than decrease, at temperatures greater than 150 oC with roughly the same Arrhenius dependence of the data below 150 oC.Method of approach: In view of this uncertainty of k, Monte Carlo simulations were used to examine the sensitivity of the density dependence of the yield of e-aq in the low-LET radiolysis of SCW (H2O) at 400 oC on variations in the temperature dependence of k. Two different values of the e-aq self-reaction rate constant at 400 oC were used: one based on the temperature dependence of k above 150 oC as measured in alkaline water (4.2 x 108 M-1 s-1) and the other based on an Arrhenius extrapolation of the values below 150 oC as initially proposed by Elliot (2.5 x 1011 M-1 s-1).Results: In both cases, the density dependences of our calculated e-aq yields at 60 ps and 1 ns were found to compare fairly well with the available picosecond pulse radiolysis experimental data (for D2O) for the entire water density range studied (0.15-0.6 g/cm3). Only a small effect of k on the variation of G(e-aq) as a function of density at 60 ps and 1 ns could be observed.Conclusions: Our calculations did not allow us to unambiguously confirm (or deny) the applicability of the predicted sudden drop of k(e-aq + e-aq) at 150 oC in near-neutral water.Keywords: Radiolysis; Linear energy transfer (LET); Supercritical water at 400 oC; Water density; Hydrated electron; Radiolytic yield; Temperature dependence of the e-aq self-reaction rate constant; Monte Carlo simulations.

Effect of Thermal Pretreatment on the Corrosion of Stainless Steel in Flowing Supercritical Water

The effect of high-temperature microstructure degradation (thermal ageing) on the corrosion resistance of austenitic stainless steels in supercritical water (SCW) was evaluated in this study. Mill-annealed (MA) and thermally treated (TT) samples of Type 316L and Type 310S stainless steel were exposed in 25 MPa SCW at 550°C with 8 ppm dissolved oxygen in a flowing autoclave testing loop. The thermal treatments applied to Type 316L (815°C for 1000 hr + water quench) and Type 310S (800°C for 1000 hr + air cool) were successful in precipitating the expected intermetallic phases in each alloy, both within the grains and on the grain boundaries. It was found that a prolonged time at relatively high temperature was sufficient to suppress significant compositional variation across the various intermetallic phase boundaries. This paper presents the results of the gravimetric analysis and oxide scale characterization using scanning electron microscopy (SEM) coupled with X-ray energy-dispersive spectroscopy (EDS). The role played by the fine precipitate structure on formation of the oxide scale, and thus corrosion resistance, is discussed. The combined role of dissolved oxygen and flow (revealed by examining the differences between Type 316L samples exposed in a static autoclave and in the flowing autoclave loop) is also addressed. It was concluded that formation of intermetallic phase precipitates during high-temperature exposure is not likely to have a major effect on the apparent corrosion resistance because of the discontinuous nature of the precipitation.

Radiation effects and mechanical properties

All in-core components in an SCWR will experience irradiation by α and β particles, neutrons and high-energy photons (γ-rays) resulting in damage at the atomic level in the form of ionization and microstructural degradation due to the development of vacancies, interstitials and voids. These microscopic defects induce changes in physical and mechanical properties such as hardening, ductility, swelling, radiation-induced segregation, and creep, and can increase the risk of cracking. In combination with thermal creep, these changes are a major factor in determining long-term component reliability. This chapter discusses the various forms of radiation damage relevant to SCWR concepts, as well as discussing thermal creep of candidate SCWR materials.