Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water—Effect of Chromium Content in Alloys and Dissolved Hydrogen—

Terachi, Takumi; Yamada, Takuyo; Miyamoto, Tomoki; Arioka, Koji; Fukuya, Koji

doi:10.1080/18811248.2008.9711883

Cited by 128 publications

(34 citation statements)

References 23 publications

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“…The results obtained show a very low oxide layer thickness comparing to experimental data obtained under PWR conditions [9] but are in agreement with the results observed in the first PWR which operated using SS 304 as cladding material [4]. This study must be extended to evaluate the SS behavior under loss-of-coolant accident and reactivity-initiated accident.…”

Section: Resultssupporting

confidence: 74%

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Evaluation of corrosion on the fuel performance of stainless steel cladding

Gomes

Abe

Silva

et al. 2016

EPJ Nuclear Sci. Technol.

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Abstract. In nuclear reactors, the use of stainless steel (SS) as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4) under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

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Section: Resultssupporting

confidence: 74%

“…The chromium content plays an important role to define the composition of the oxide layer formed on the SS cladding [9].…”

Section: Corrosion Of Stainless Steel (Ss)mentioning

confidence: 99%

Evaluation of corrosion on the fuel performance of stainless steel cladding

Gomes

Abe

Silva

et al. 2016

EPJ Nuclear Sci. Technol.

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“…The inner oxide layer is chromium rich, protective and described as a mixed iron/chromium spinel ((Ni,Fe)Cr 2 O 4 ) while the outer oxide is found to be crystallites of geometric shapes of magnetite (Fe 3 O 4 ). Many authors observed a nickel enrichment below the metal/ oxide interface [5][6][7]. The inner oxide layer is formed by anionic diffusion whereas the outer crystallites grow from a cationic diffusion mechanism coupled with a precipitation mechanism of dissolved cations in the medium, originating from a diffusion process within the inner layer [1,7,8].…”

Section: Introductionmentioning

confidence: 99%

Role of irradiation and irradiation defects on the oxidation first stages of a 316L austenitic stainless steel

et al. 2019

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The role of irradiation and irradiation defects on the oxidation first stages of 316 L alloy was investigated. A sample with both a proton pre-irradiated and an unirradiated area was exposed to a simulated PWR environment during 24 hours. Irradiation defects and Radiation Induced Segregation at grain boundary and on irradiation defects were characterized and quantified and their effect on the oxidation was evaluated. Irradiation affects the morphology, thickness and chemistry of the oxide layers formed. It enhances the oxidation kinetic and induces the formation of an inner oxide richer in chromium. Defects induced by irradiation act as preferential nucleation sites.

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“…pH control is achieved by controlling the boric acid (H 3 BO 3 ) concentration (∼500 ppm) and the lithium (LiOH) concentration (∼2.2 ppm). Corrosion experiments with simulated PWR water chemistry generally follow these guidelines [25]. In addition, minimization of dissolved oxygen to <5 ppb is desired.…”

Section: Chemistry Controlmentioning

confidence: 99%

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

Sabharwall¹,

O’Brien²,

Yoon³

et al. 2015

EPJ Nuclear Sci. Technol.

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Abstract. A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/ steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750°C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF 4 ) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼450°C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be used for code and model verification and validation, one of the key purposes for the loop. The experimental facility will provide a much-needed database for successful development of advanced reactors and provide insight into the needs and challenges in instrumentation for advanced high-temperature reactors.

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Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water—Effect of Chromium Content in Alloys and Dissolved Hydrogen—

Cited by 128 publications

References 23 publications

Evaluation of corrosion on the fuel performance of stainless steel cladding

Evaluation of corrosion on the fuel performance of stainless steel cladding

Role of irradiation and irradiation defects on the oxidation first stages of a 316L austenitic stainless steel

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

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