General Corrosion of Chromium-Coated Zirconium- and Titanium-Based Alloys in Supercritical Water at 500 °C

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
K. Khumsa-Ang ◽  
M. Edwards ◽  
S. Rousseau
Keyword(s):  
Supercritical Water ◽  
Fuel Cycle ◽  
Neutron Cross Section ◽  
Zirconium Alloys ◽  
Oxide Thickness ◽  
General Corrosion ◽  
Coating Materials ◽  
Grade 5 ◽  
Oxide Phases ◽  
Enriched Uranium

Abstract The 300 MWel small Canadian supercritical water-cooled reactor (SCWR), which is a scaled-down version of the original 1200 MWel concept, has a smaller core, uses low enriched uranium fuel instead of a plutonium–thorium fuel, and features a lower (maximum) cladding temperature of 500 °C. The lower cladding temperature may permit the use of different alloys, including zirconium alloys, which had been ruled out as candidates for the Canadian SCWR, whose cladding temperature may reach 850 °C. The potential to use zirconium alloys is exciting because they have a low neutron cross section, which in turn means that fewer neutrons are lost, and the fuel can be used more efficiently. One advantage, for example,, is that the fuel cycle can be lengthened. In this paper, we report on the results of corrosion experiments used to screen zirconium- and titanium-based alloys as well as corrosion-resistant coating materials such as Cr and Al as potential candidates for fuel cladding in the small Canadian SCWR. These experiments were conducted in a refreshed autoclave in deaerated supercritical water at 500 °C and 23.5 MPa. After exposure, the weight gain was measured, and the oxide thickness and the oxide phases were examined. Of all materials, the coated and uncoated Ti-grade 2 and Ti-grade 5 alloys met our screening qualification criteria, however, Al/Cr-coated zirconium coupons showed notable improvement and will be explored further in future testing.

Weight Gain and Hydrogen Absorption in Supercritical Water At 500°C of Chromium-Coated Zirconium-Based Alloys: Transverse vs Longitudinal Direction

2021 ◽  
Author(s):  
Kittima Khumsa-Ang ◽  
Stephane Rousseau ◽  
Oksana Shiman
Keyword(s):  
Weight Gain ◽  
Supercritical Water ◽  
Hydrogen Absorption ◽  
Corrosion Behaviour ◽  
Reactor Core ◽  
Neutron Cross Section ◽  
Zirconium Alloys ◽  
Longitudinal Direction ◽  
General Corrosion ◽  
Enriched Uranium

Abstract Canadian Nuclear Laboratories (CNL) has an on-going Research & Development programme to support the development of a scaled-down 300 MWe version of the Canadian Super-Critical Water Reactor (SCWR) concept. The 300 MWe and 170-channel reactor core concept uses low enriched uranium fuel and features a maximum cladding temperature of 500°C. Our goal is to test surfacemodified zirconium alloys for use as fuel cladding. Zirconium alloys are attractive as they offer low neutron cross section thereby allowing the use of low enriched fuel. In this paper, we report on the results of general corrosion experiments used to evaluate chromiumcoated zirconium-based alloys in the two chemistries (630 ug/kg O2 in both deaerated and lithiated supercritical water). These experiments were conducted in a refreshed autoclave at 500°C and 23.5 MPa. After exposure, the weight gain and the hydrogen absorption were examined. At adequate coating thickness, longitudinal and transverse coupons show similar corrosion behaviour with improved corrosion resistance compared to uncoated coupons. The measured concentrations of hydrogen absorption are higher for the transverse coupons. Alkaline treatment resulted in higher weight gains than was found in pure oxygenated supercritical water.

Corrosion Behavior of Model Zirconium Alloys in Deaerated Supercritical Water at 500°C

CORROSION ◽  
2007 ◽  
Vol 63 (6) ◽  
pp. 577-590 ◽  
Author(s):  
Q. Peng ◽  
E. Gartner ◽  
J. T. Busby ◽  
A. T. Motta ◽  
G. S. Was
Keyword(s):  
Corrosion Behavior ◽  
Supercritical Water ◽  
Zirconium Alloys

CAREM X INPRO Case Study

2004 ◽  
Author(s):  
Pablo C. Florido ◽  
Dari´o Delmastro ◽  
Daniel Brasnarof ◽  
Osvaldo E. Azpitarte
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
High Level Waste ◽  
Fuel Cycles ◽  
Surface Intermediate ◽  
Geological Repository ◽  
Enriched Uranium ◽  
Intermediate Storage ◽  
High Level

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.

The Nuclear Fuel Services, Inc. Program to Support Disposition of Enriched Uranium-Bearing Materials

2007 ◽  
Author(s):  
Stephen M. Schutt ◽  
Norman P. Jacob
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Nuclear Material ◽  
Nuclear Materials ◽  
Bearing Materials ◽  
Enriched Uranium

The disposition of surplus nuclear materials has become one of the most pressing issues of our time [1, 2]. Numerous agencies have invoked programs with the purpose of removing such materials from various international venues and dispositioning these materials in a manner that achieves non-proliferability. This paper describes the Nuclear Fuel Services, Inc (NFS) Nuclear Material Disposition Program, which to date has focused on a variety of Special Nuclear Material (SNM), in particular uranium of various enrichments. The major components of this program are discussed, with emphasis on recycle and return of material to the nuclear fuel cycle.

Decommissioning of the Hematite Former Fuel Cycle Facility Using a Decision Flow Logic Based Work Control Process

2013 ◽  
Author(s):  
Keith D. Anderson
Keyword(s):  
Fuel Cycle ◽  
Control Process ◽  
Nuclear Regulatory Commission ◽  
Security Protocol ◽  
Successful Implementation ◽  
Chemical Contamination ◽  
Relative Level ◽  
Decision Logic ◽  
Enriched Uranium ◽  
Regulatory Commission

The remediation and decommissioning of the Hematite Former Fuel Cycle Facility (FFCF), the Hematite Facility, is currently being carried out by Westinghouse Electric Company LLC under the Hematite Decommissioning Project (HDP). The Hematite Facility is located near the town of Hematite, Missouri, USA. The Hematite Facility consists of 228 acres of land with primary operations historically being conducted within the central portion of the property that is roughly 10 acres including Burial Pits and the Site Pond area. Decommissioning and remediation activities are being performed with the eventual objective of the release of the property. Primary contaminants include the legacy disposal and contamination of natural and enriched uranium from the nuclear fuel cycle, as well as chemicals used during the facility operations. Two major regulatory bodies, the U.S. Nuclear Regulatory Commission (NRC) and the Missouri Department of Natural Resources (MDNR), provide critical roles in the approval and oversight of the current regulatory path to remediation, decommissioning and eventual release. Further, remediation and decommissioning activities are performed under the implementing policies, plans, and procedures under the Hematite Decommissioning Plan (DP) and the Record of Decision (ROD). Remediation and decommissioning tasks at the Hematite Former Fuel Cycle Facility, referred to as the Hematite Facility, are performed against a disciplined decision logic flow that applies accumulated technical and monitoring data to determine each step of the excavation, exhumation, and removal of wastes from the Burial Pits and the remaining Areas of Concern (AOC). Decision flow logic is based upon the nuclear criticality safety controls and threshold conditions, relative level of radioactive and chemical contamination, security protocol, and final waste stream disposition. The end result is to remediate the residual radioactive and chemical contamination to approved dose-based and risk-based cleanup criteria as negotiated with U.S. Federal and State Regulators. The purpose of the paper is to provide a summary of the successful implementation of the decision flow logic to the remediation and decommissioning tasks performed to date.

Possible scenarios for the transition to thorium fuel cycle in molten salt reactor by using enriched uranium

2018 ◽  
Vol 104 ◽  
pp. 75-84 ◽  
Author(s):  
D.Y. Cui ◽  
X.X. Li ◽  
S.P. Xia ◽  
X.C. Zhao ◽  
C.G. Yu ◽  
...  
Keyword(s):  
Molten Salt ◽  
Fuel Cycle ◽  
Molten Salt Reactor ◽  
Enriched Uranium

General Corrosion Properties of Modified PNC1520 Austenitic Stainless Steel in Supercritical Water as a Fuel Cladding Candidate Material for Supercritical Water Reactor

2010 ◽  
Vol 73 ◽  
pp. 72-77
Author(s):  
Yoshihisa Nakazono ◽  
Takeo Iwai ◽  
Hiroaki Abe
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Supercritical Water ◽  
Reactor Core ◽  
General Corrosion ◽  
Fuel Cladding ◽  
Corrosion Properties ◽  
Energy Agency ◽  
Water Reactor ◽  
Supercritical Water Reactor

The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S), the Ti-additional type of PNC1520 (1520Ti) and the Zr-additional type of PNC1520 (1520Zr) by using a supercritical water autoclave. In view of general corrosion, 1520Zr may have larger possibility than 1520S and 1520Ti to adopt a supercritical water reactor core fuel cladding.

Achieving Resource Sustainability in China Through the Thorium Fuel Cycle in the Candu Reactor

2010 ◽  
Author(s):  
Peter G. Boczar ◽  
Bronwyn Hyland ◽  
Keith Bradley ◽  
Sermet Kuran
Keyword(s):  
Fuel Cycle ◽  
Close Approach ◽  
Natural Uranium ◽  
Fissile Material ◽  
Energy Potential ◽  
Candu Reactors ◽  
Fuel Cycles ◽  
Resource Sustainability ◽  
Enriched Uranium

The CANDU® reactor is the most resource-efficient reactor commercially available. The features that enable the CANDU reactor to utilize natural uranium facilitate the use of a wide variety of thorium fuel cycles. In the short term, the initial fissile material would be provided in a “mixed bundle”, in which low-enriched uranium (LEU) would comprise the outer two rings of a CANFLEX® bundle, with ThO2 in the central 8 elements. This cycle is economical, both in terms of fuel utilization and fuel cycle costs. The medium term strategy would be defined by the availability of plutonium and recovered uranium from reprocessed used LWR fuel. The plutonium could be used in Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. Recovered uranium could also be effectively utilized in CANDU reactors. In the long term, the full energy potential from thorium could be realized through the recycle of the U-233 (and thorium) in the used CANDU fuel. Plutonium would only be required to top up the fissile content to achieve the desired burnup. Further improvements to the CANDU neutron economy could make possible a very close approach to the Self-Sufficient Equilibrium Thorium (SSET) cycle with a conversion ratio of unity, which would be completely self-sufficient in fissile material (recycled U-233).

scholarly journals INVESTIGATION ON INHERENT SAFETY OF ONE FLUID-MOLTEN SALT REACTOR (OF-MSR) WITH VARIOUS STARTING FUEL

2020 ◽  
Vol 22 (2) ◽  
pp. 54
Author(s):  
R. Andika Putra Dwijayanto ◽  
Dedy Prasetyo Hermawan
Keyword(s):  
Molten Salt ◽  
Temperature Coefficient ◽  
Fuel Cycle ◽  
Minor Actinides ◽  
Molten Salt Reactor ◽  
Inherent Safety ◽  
Safety Aspects ◽  
Safe Option ◽  
Enriched Uranium ◽  
Coefficient Of Reactivity

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

Long Term Storage of Nuclear Spent Fuel as Key Role of Japan’s Nuclear Fuel Cycle Until 2100: Cost and Benefit

2010 ◽  
Author(s):  
Tadahiro Katsuta
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Storage Site ◽  
Spent Fuel ◽  
Final Disposal ◽  
Enriched Uranium ◽  
Interim Storage ◽  
Reprocessing Plant

Political and technical advantages to introduce spent nuclear fuel interim storage into Japan’s nuclear fuel cycle are examined. Once Rokkasho reprocessing plant starts operation, 80,000 tHM of spent Low Enriched Uranium (LEU) fuel must be stored in an Away From Reactor (AFR) interim storage site until 2100. If a succeeding reprocessing plant starts operating, the spent LEU will reach its peak of 30,000 tHM before 2050, and then will decrease until the end of the second reprocessing plant operation. Throughput of the second reprocessing plant is assumed as twice of that of Rokassho reprocessing plant, indeed 1,600tHM/year. On the other hand, tripled number of final disposal sites for High Level Nuclear Waste (HLW) will be necessary with this condition. Besides, large amount of plutonium surplus will occur, even if First Breeder Reactors (FBR)s consume the plutonium. At maximum, plutonium surplus will reach almost 500 tons. These results indicate that current nuclear policy does not solve the spent fuel problems but rather complicates them. Thus, reprocessing policy could put off the problems in spent fuel interim storage capacity and other issues could appear such as difficulties in large amount of HLW final disposal management or separated plutonium management. If there is no reprocessing or MOX use, the amount of spent fuel will reach over 115,000 tones at the year of 2100. However, the spent fuel management could be simplified and also the cost and the security would be improved by using an interim storage primarily.

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