Advances in Materials Science for Environmental and Energy Technologies III

Tatsuki Ohji, Josef Matyas, Navin Jose Manjooran, Gary Pickrell, Andrei Jitianu, Tatsuki Ohji, Josef Matyáš, Navin Jose Manjooran, Gary Pickrell, Andrei Jitianu, Tatsuki Ohji, Josef Matyáš, Navin Jose Manjooran, Gary Pickrell, Andrei Jitianu

• Publisher: Wiley-American Ceramic Society
• Year: 2014
• Print ISBN: 9781118996683
• eBook ISBN: 9781118996706
• Edition: 1
• Topic: Technology & Engineering
• Subtopic: Materials Science
• Index: Technology & Engineering

This proceedings contains a collection of 26 papers from the following six 2013 Materials Science and Technology (MS&T'13) symposia:

• Green Technologies for Materials Manufacturing and Processing V
• Materials Development and Degradation Management in Nuclear Applications
• Materials Issues in Nuclear Waste Management in the 21st Century
• Energy Storage III: Materials, Systems and Applications
• Nanotechnology for Energy, Healthcare and Industry
• Hybrid Organic – Inorganic Materials for Alternative Energy

Materials Issues in Nuclear Waste Management

ADVANCED STEELS FOR ACCIDENT TOLERANT FUEL CLADDING IN COMMERCIAL NUCLEAR REACTORS

Raul B. Rebak

GE Global Research One Research Circle, CEB2551, Schenectady, NY 12309, USA

ABSTRACT

After the March 2011 events at the Fukushima site, the U.S. congress directed the Department of Energy (DOE) to focus efforts on the development of fuels with enhanced accident tolerance. In comparison with the standard UO2–Zircaloy system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. A GE Global Research led program is investigating the behavior of advanced steels both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam. Results show that, under accident conditions, the advanced steels: (1) have low reactivity with steam, (2) would generate less hydrogen than the current zirconium based alloys, (3) offer higher mechanical properties, and (4) offer excellent retention of fission products. These advanced steels are also highly resistant to environmental cracking under normal operation conditions and they are not susceptible to suffer irradiation damage such as void swelling.

Keywords: light water reactors, fuel cladding, advanced steels, steam, crack propagation, mechanical properties

INTRODUCTION

The commercial nuclear energy in the United States had its origin in the nuclear navy. The navy originally adopted zirconium based alloys over stainless steels for the fuel cladding mainly because of the higher transparency to neutrons of the former making the reactors more compact for submarine applications [Terrani et al. 2013]. In spite of the thermal neutron cross section of stainless steels being approximately 12–16 times higher than for zirconium alloys it is now understood that the fuel enrichment penalty incurred by the use of stainless steel cladding can be partially overcome by using thinner wall advanced stainless cladding because they are stronger than the zirconium alloys [Terrani et al 2013].

At the beginning of the nuclear navy program, the susceptibility to cracking of sensitized stainless steels in high temperature water initially relegated the stainless steels in research and development as compared to the zirconium alloys. Six decades later, it is now understood that high strength ferritic stainless steels are resistant to environmental cracking and irradiation damage, which would not limit their application as fuel cladding in light water reactors. Moreover, current significant progress in steelmaking practices shows that the chemical purity in modern steels can be highly controlled. Similarly, there is an increased ability in the fabrication of these advanced steels into thin walled tubes, including readily joining (welding) by several techniques.

Terrani et al. [2013] cite several uses in the industry of non-zirconium alloys as fuel cladding, including type 304, 316 and 347 austenitic stainless steels and austenitic nickel based alloys such as Inconel 600 and Incoloy 800. Type 304 SS fuel cladding was used for some time in US commercial light water reactors, for example at the Connecticut Yankee and San Onofre 1 power stations [Rivera and Meyer 1980]. The early cracking of type of 304 SS was linked to sensitization due to welding of high carbon alloys. The cracking phenomenon of sensitized stainless steels is now well understood and controlled and it is not a current concern in light water reactors.

In the post Fukushima Daiichi scenario there is an increased effort by the international community to find a material or materials that would an alternative to the zirconium alloy for fuel cladding. This material should have

• Improved reaction kinetics with steam;
• Slower hydrogen production rate;
• Improved cladding and fuel properties;
• Enhanced retention of fission products.

One of these alternatives could be advanced steels since earlier concerns about stainless steels such as cracking, radiation damage and neutron economy can be now retired by using advanced ferritic steels (e.g. Fe-Cr-Al alloys). The advanced steels are also far superior to zirconium based alloys regarding higher strength, higher retention of fission products, and higher resistance to reaction with steam.

Behavior of Ferritic and Austenitic Steels in Light Water Reactors

Austenitic stainless steels such as types 304 and 316 are highly susceptible to stress corrosion cracking in chloride containing environments, especially at temperatures higher than 60°C [Sedriks 1996]. The most common tests used in the industry to determine susceptibility to chloride cracking are immersion of U-bend specimens (ASTM G 30) into hot solutions of chloride salts including magnesium chloride or sodium chloride (ASTM G 36 and G 123). Ferritic stainless steels such as types 405 and 430 are highly resistant to SCC in hot chloride solutions [Sedriks 1996, Bond et al 1967].

Austenitic stainless steels such as type 304, 308, 316, 321 and 347 are used worldwide as construction materials for light water power reactors [Andresen 2012]. In the USA the most common austenitic alloy may be type 304 SS (UNS S30403) and in Japan the preferred stainless steel is type 316 (S31603). European countries such as Germany may prefer to use titanium (Ti) or niobium (Nb) stabilized types of stainless steel such as type 321 (S32100) and 347 (S34700). Austenitic stainless steels are susceptible to stress corrosion cracking (SCC) in boiler water reactor (BWR) service and in a lesser extent in pressurized water reactor (PWR) service [Andresen 2012].

Austenitic stainless steel (SS) core internals components are susceptible to irradiation assisted stress corrosion cracking (IASCC) during service in nuclear power plants light water reactors [Chung and Shack 2006, Cookson et al 1993, Jacobs et al 1993]. One of the effects of irradiation is the hardening of the SS due to modifications in the dislocation distribution in the alloy [Was 2003, Bruemmer 2002]. Irradiation also alters the local chemistry of these austenitic alloys, for example in the vicinity of grain boundaries by a mechanism of radiation induced segregation (RIS). The segregation or depletion phenomena at or near grain boundaries may enhance the susceptibility of these irradiated alloys to stress corrosion cracking (SCC) [Was et al 2006, Yonezawa et al 2000]. The effect of the IASCC on austenitic stainless steels may impact the life extension of currently operating light water reactors due to the progressive dose accumulation [Hojná 2012].

In nuclear power plant applications, ferritic steels have superior void swelling resistance because they experience delayed void nucleation and they sustain less than 2% swelling even at irradiation levels close to 200 dpa [Raj and Vijayalakshmi 2010]. On the other hand, austenitic stainless steels such as type 304 undergo the onset of significant void swelling and possible embrittlement at dose rates in the order of 20 dpa [Was et al 2006]. Besides the higher resistance of ferritic steels to radiation damage, other benefits that could make these steels more attractive than the austenitic stainless steels in nuclear applications include: (1) Ferritic materials have lower cost since they do not contain nickel (Ni), and generally contain lower chromium (Cr), (2) They do not contain Ni or cobalt (Co) that could be become activated in commercial reactors, (3) They offer a lower coefficient of thermal expansion (CTE) that matches the CTE of pressure vessel ferritic alloys such as type A508, A516, or A533 [Ren et al 2008], and (4) Ferritic steels have higher thermal conductivity for heat transfer capabilities (Table 1).

Table 1 – Physical Properties of Ferritic and Austenitic Steels

Steel	CTE (0-538°C) μm/m/°C	Thermal Conductivity at 100°C (W/m.K)
Zircaloy-2	8.32 & 15.7 (orientation dependent)	13.8
Ferritic type 430 (16% Cr)	11.4	23.9
Austenitic type 304L (18% Cr)	18.4	16.2

Reaction of Cladding Materials with Steam

In the case of a loss of coolant accident (LOCA), such as in the Fukushima Daiichi situation, the cladding of the fuel will be exposed to steam. The zirconium alloy plus steam reaction has been widely studied under loss of coolant accident scenarios (Whitmarsh 1962, Baker and Just 1962, Leistikow and Schanz 1987, IAEA 1992, Grandjean and Hache 2008, Terrani et al 2013). Zircaloy oxidizes in presence of steam to form zirconia and hydrogen following an exothermic reaction:

(1)

According to Baker and Just the chemical heat generated by the reaction of zirconium and steam in (Equation 1) could exceed the nuclear heat generation during a destructive nuclear transient. Moreover, the hydrogen generated by the reaction could give rise to a pressure surge and might subsequently react explosively with air [Baker and Just 1962].

All the ferrous materials listed in Table 2, including ferritic steel T91, have lower reaction k...

Table of contents

• Cover
• Half Title page
• Title page
• Copyright page
• Preface
• Green Technologies for Materials Manufacturing and Processing
• Materials Issues in Nuclear Waste Management
• Materials and Systems for Energy Applications
• Nanotechnology for Energy, Healthcare and Industry
• Author Index