Conference Agenda

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Session Overview
Session
10.05: Accident tolerant fuel
Time:
Monday, 16/Mar/2020:
3:15pm - 5:00pm

Session Chair: Changheui Jang, Korea Advanced Institute of Science & Technology-KAIST, Korea, Republic of (South Korea)
Location: R-1013

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Presentations

Interactions between coolant and promising ATF cladding materials – update of the results obtained in the framework of the IL TROVATORE project

Mirco Grosse1, Koba Van Loo2, Chongchong Tang1, Konstantza Lambrinou3

1Karlsruhe Institute of Technology; 2Catholic University Leuven; 3Belgian Nuclear Research Centre SCK∙CEN and University of Huddersfield

In the framework of the Il Trovatore project sponsored by the EU research and innovation program HORIZON 2020 the interaction between promising ATF cladding materials and coolant is investigated. The materials produced by the project partners can be divided into four groups:

- MAX phases

- Oxides for coatings

- Advanced iron bases alloys (FeCrAl and HEA -High Entropy Alloys)

- SiC/SiC composites

The paper gives an overview about the latest results obtained.The corrosion and oxidation performance under operation and accident conditions of the materials investigated will be discussed.

The results show that not only the material type but also the effective surface controls the reaction kinetics. The reaction rates increase strongly if the coolant faces a porous interface as it happens in the porous SANDVIC Cr2AlC material or if the steam reaches the fibers of SiC/SiCf ceramic matrix composites.

At many of the systems both type of processes occur simultaneously:

- Processes reducing the mass (forming of volatile substances like carbon oxides or silicon mono oxide, dissolution in coolant environment like Al2O3 or spallation of coating or oxide parts) and

- Processes resulting in mass gain like the formation of oxide scales or the internal oxidation.

Therefore, the description of the kinetics on basis of mass changes has to be handled carefully. It is very helpful to get additional information about the processes like for instance the hydrogen release or the consumption of coatings. For instance, whereas most of SiC CMC samples behave excellent, specific SiC/SiCf samples with not completely formed protective monolithic surface layer show a low mass change but a very high hydrogen release during oxidation at 1200°C in steam. Obviously, mass gain and mass loss compensate each other in this case.



Microstructure Analysis and Hardness Test of Irradiated Multi-Metallic Layer Composite Material for Accident Tolerant Fuel Cladding

Taeyong Kim1, Jeonghyeon Lee1, Inyoung Song1, Junhyuk Ham1, Michael P Short2, Junho Moon3, Chi Bum Bahn3, Ji Hyun Kim1

1UNIST; 2Massachusetts Institute of Technology; 3Pusan National University

In severe accident conditions of light water reactors, the loss of coolant may cause safety problems with the integrity of Zr fuel cladding. Under these conditions, the Zr cladding is reacted with high temperature steam by producing of hydrogen. In order to avoid these problems, we aim to develop a multi-metallic layered composite (MMLC) fuel cladding. Fe-12Cr-2Si alloy, which is an oxidation-resistant material, can be layered on the outer surface of commercial Zr alloy tube. V- and Ti-metal layers are interposed between Fe-12Cr-2Si layer and Zr layer acting as diffusion barriers to avoid formation of eutectic phases or compounds between Zr and Fe alloys. Therefore, the tube has four layers in total in the order of Zr alloy, Ti, V and Fe-12Cr-2Si alloy from the inside to the outside. However, new intermetallic phases can be created due to interlayer interactions, which can degrade the mechanical properties. In this study, Si ion beam (7 MeV and 8.29×1012 ion #/cm2/s, 15 dpa) was irradiated to the MMLC specimen to simulate reactor environment at 442 ℃. TEM analysis was performed to investigate precipitates and microstructure. The precipitates in Fe-12Cr-2Si layer at Si irradiated MMLC specimen have smaller size and lower number density than that of as-received specimen. This is probably due to back-diffusion caused by the irradiation. And Zr alloy shows segregation of Ti element into grain boundaries. At nano-indentation, the hardness of irradiated Fe-12Cr-2Si is reduced from 3.2 to 2.5 GPa. Nevertheless, the hardness reduction of Zr alloy is from 4.5 to 2.8 GPa. The more hardness reduction of Zr alloy seems to be due to the segregation of Ti causing solid solution hardening. From this results, other layers are less sensitivity to irradiation than Zr alloy, and MMLC materials are expected to be superior to Zr alloys in the irradiation environment.



Thermomechanical Characteristics of Molybdenium for Accident Tolerance Fuel Cladding Under Internal Pressure

Young-Jin Kim1, Bo Cheng2

1FNC Technology; 2Electric Power Research Institute

EPRI developed molybdenum (Mo) and its alloys with an external coating as a candidate fuel cladding to provide enhanced tolerance to severe loss of coolant accidents. Mo alloy has been known to possess good tensile and creep strength at temperatures exceeding 1200oC, and some Mo alloys also have adequate ductility at room temperature suitable for fabrication of thin-wall tubes. However, Mo oxidation kinetics is very rapid even at relatively low oxygen, and thus the ability to increase the performance of Mo under enhanced incident conditions has to be achieved. Thus, in order for protecting Mo under operating and sever coolant loss conditions, a protective coating of Zr alloy or advanced steel alloys or other oxide on Mo surface is proposed.

The pressurized tube test (PTT) system is designed in which biaxial plane stress tests can be carried out on tubular specimens and is used to characterize the mechanical properties of Mo cladding tubes with and without a protective coating for engineering strength and ductility under various temperatures. During the testing, the internal pressure, diameter micrometer, temperature and time is recorded electronically, and hoop stress, plastic and elastic deformation can be calculated.

The elongation and rupture pressure of Mo thin wall tube specimens (0.37”-0.40” diameter with 8-10 mil wall thickness) was measured in-situ at high temperature in an inert environment at 1atm (around the outer surface of test specimens), with up to a 5,000 psi of argon gas within the internal surface of test tubes, operating upwards of 1000oC. This paper will present the preliminary results on the stress-strain behavior of Mo thin wall tubes at various temperatures with and without a protective thin coating. In addition, SEM examination of fracture morphologies of test specimens after PTT will be included.



Optimization of thermo-mechanical processes for an ADSS alloy, a potential ATF cladding material

Chaewon Kim, Chae Won Jeong, Hyeon Bae Lee, Changheui Jang

Korea Advanced Institute of Science & Technology

As accident tolerant fuel (ATF) cladding materials for light water reactors (LWR), alumina-forming duplex stainless steel (ADSS) alloys were developed with nominal composition of Fe--(18−21)Ni (15−21)Cr -(5−7)Al. The microstructure of ADSS alloys exhibited duplex structure (austenite and ferrite phases) with B2-NiAl precipitates. Because of the duplex structure and B2 precipitates, ADSS alloys have high strength (~1 GPa) and enough elongation to be drawn to thin cladding tubes. Here, the fraction and distribution of such three major phases are changed by thermo-mechanical process (TMP), which seems to result in changes of mechanical property. In addition, microstructure would affect the SCC and corrosion resistance at normal condition and accident condition. For example, the phase boundaries would be susceptible to cracking and corrosion behavior would be different in phases with varying Cr and Al contents. Therefore, in this study, one of ADSS alloys, #B51 was treated by various TMP, and microstructure, tensile property and SCC resistance in PWR environment were evaluated to find the optimum conditions.



Accident Tolerant Fuels: Misconceptions and Opportunities

Koroush Shirvan

Massachusetts Institute of Technology

The Accident Tolerant Fuel (ATF) program took nuclear R&D sector by storm in 2011 as all major players in nuclear energy ramped up programs to investigate enhancements and even replacements for current UO2/Zircaloy fuel form that is used in all commercial water cooled reactors. There were about a dozen major concepts being pursued by research laboratories and nuclear vendors in the early days of the program. A common material feature among the ATF cladding concepts was the superior oxidation resistance by proposing concepts such as Cr, FeCrAl and SiC. By 2014, the US nuclear utilities became enamored by ATFs based on simplified projections and highly cited research articles that an oxidation resistant clad alone can significantly improve station blackout (SBO) safety response. This led to placement of several lead test rods in commercial reactors. However, by 2016 the more realistic ATF safety analysis utilizing best estimate codes showed ATF cladding can marginally improve the SBO performance. Given the realities of the long timelines needed for fuel qualification, only modifications to existing fuel concepts in form of coatings and dopants remained the main focus for the US program in the past two years. This work addresses several misconceptions with ATFs such as potential to reduce hydrogen production and avoiding severe accidents. The opportunities with ATFs are outline for both existing reactors as well as future water cooled small modular reactors. Lastly, lessons learned from the past 8 years of development to inform future innovative nuclear technology R&D strategy are listed.



 
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