Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

D. Guzonas; L. Qiu; S. Livingstone; S. Rousseau

doi:10.1115/1.4031381

Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4031381 ◽

2016 ◽

Vol 2 (2) ◽

Cited By ~ 2

Author(s):

D. Guzonas ◽

L. Qiu ◽

S. Livingstone ◽

S. Rousseau

Keyword(s):

Supercritical Water ◽

Fission Products ◽

Fuel Pellet ◽

Product Release ◽

The Core ◽

Fuel Pellets ◽

Simulated Fuel ◽

Core Components ◽

High Concentrations ◽

Supercritical Water Cooled Reactor

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.

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Role of Lean and Agile Supply Chain Models for Pellet Fuel Technologies

International Journal of Engineering and Management Sciences ◽

10.21791/ijems.2017.3.12. ◽

2017 ◽

Vol 2 (3) ◽

pp. 119-130

Author(s):

János Szendrei ◽

Balázs Kocsi ◽

István Budai ◽

Gábor Grasselli ◽

Edit Szűcs

Keyword(s):

Supply Chain ◽

Fuel Pellet ◽

Process Variables ◽

Wood Pellet ◽

The Core ◽

Fuel Pellets ◽

Agile Supply Chain ◽

Technology Process ◽

Output Market

Business organisations constantly strive to improve their processes, both internal and external. Within the supply chain of a product, different strategies can be applied. This paper aims to answer the basic questions like what is the core of lean and agile SCM strategies, what are the differences of the two models, and how can they be combined. This paper examines further, which of these SCM models and what elements of them can be applied for wood pellet supply chains, also examining the possibility of their combination. The result of these examinations is, that even though fuel pellets can be considered as simple commodity and not a very innovative product, diversity in input characteristics, optimization of fuel pellet technology process variables and changes in output market qualities and quantities need sometimes agile, flexible answer from pellet fuel SCM. The basically lean character of pellet fuel technologies and SCM can further enhanced by implementing agile SCM model elements, as is the increased integration of suppliers into the supply chain.

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Corrosion Behaviour of Bare and NiCrAlY Coated Alloy 214 in Supercritical Water at 700 °C

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4037324 ◽

2017 ◽

Vol 4 (1) ◽

Cited By ~ 1

Author(s):

Suzan Bsat ◽

Xiao Huang ◽

Sami Penttila

Keyword(s):

Supercritical Water ◽

Nuclear Energy ◽

Fossil Fuels ◽

Corrosion Behaviour ◽

Thermodynamic Efficiency ◽

Coated Alloy ◽

Core Components ◽

Corrosion Behaviors ◽

Supercritical Water Cooled Reactor

Concerns with greenhouse gas emissions and the uncertainty of long-term supply of fossil fuels have resulted in renewed interest in nuclear energy as an essential part of the energy mix for the future. Many countries worldwide including Canada, China, and EU are currently undertaking the design of generation IV supercritical water-cooled reactor (SCWR) with higher thermodynamic efficiency and considerable plant simplification. The identification of appropriate materials for in-core and out-of-core components to contain the supercritical water (SCW) coolant is one of the major challenges for the design of SCWR. This study is carried out to evaluate the oxidation/corrosion behaviors of bare alloy 214 and NiCrAlY coated 214 under SCW at a temperature of 700 °C/25 MPa for 1000 h. The results show that chromium and nickel based oxide forms on the bare surface after exposure in SCW for 1000 h. A dense and adhered oxide layer, consisting of Cr2O3 with spinel (Ni(Cr, Al)2O4), was observed on NiCrAlY surface after 1000 h in SCW.

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Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4031247 ◽

2015 ◽

Vol 2 (1) ◽

Cited By ~ 1

Author(s):

Metin Yetisir ◽

Rui Xu ◽

Michel Gaudet ◽

Mohammad Movassat ◽

Holly Hamilton ◽

...

Keyword(s):

Supercritical Water ◽

Thermal Efficiency ◽

Nuclear Reactor ◽

Reactor Core ◽

Operating Conditions ◽

Stress Distributions ◽

Light Water Reactors ◽

Core Components ◽

Water Reactors ◽

Supercritical Water Cooled Reactor

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

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Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4037119 ◽

2017 ◽

Vol 4 (1) ◽

Author(s):

Zhao Chuanqi ◽

Wang Kunpeng ◽

Cao Liangzhi ◽

Zheng Youqi

Keyword(s):

Rare Earth ◽

Fuel Assembly ◽

Supercritical Water ◽

Power Distribution ◽

Rare Earth Oxide ◽

Burnable Poison ◽

The Core ◽

Power Peak ◽

Supercritical Water Cooled Reactor ◽

Control Rods

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

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Dissolution Behavior and Fission Product Release from Irradiated Thoria-Urania Fuel in Groundwater at 90°C

MRS Proceedings ◽

10.1557/proc-757-ii9.3 ◽

2002 ◽

Vol 757 ◽

Author(s):

James L. Jerden ◽

J. C. Cunnane

Keyword(s):

Fission Product ◽

Fission Products ◽

Dissolution Behavior ◽

Well Water ◽

Experimental Conditions ◽

Product Release ◽

J 13 ◽

High Concentrations ◽

Reaction Vessels ◽

Micrometer Scale

ABSTRACTThe dissolution behavior and fission-product release from irradiated thoria-urania fuel was studied by immersing fuel samples in J-13 well water at 90°C. The samples are from the Shippingport Light Water Breeder Reactor and consist of binary solid solutions of (U,Th)O2 with UO2 contents varying from 2.0 to 5.2 Wt.%. The post-irradiation U isotopic composition of the samples used in our experiments is: 87.3% 233U, 10.4% 234U, 1.8% 235U, and <0.5% 238U, 236U, 232U. Burn up values for the samples range from 22.3 to 40.9 megawatt-days per kg-metal. Our tests were performed on polished disks and on crushed and sieved samples in stainless-steel reaction vessels with air-filled head-space. After 196 days of reaction, samples showed no evidence for corrosion at the micrometer scale. Concentration ranges (μgL-1) of key radionuclides in filtered (∼5 nm pore size) leachates were: 0.1 – 15 90Sr, 0.9 – 7.0 99Tc, 0.1 – 35.2 137Cs,<0.2 – 0.8 233U, <0.1 – 0.7 232Th. Concentrations of 237Np, 239Pu, 240Pu and 241Am were all <0.2 μgL-1. The relatively high concentrations of the fission products 90Sr and 137Cs occur early during leaching and decrease for later samplings. Matrix dissolution rates for the irradiated thoria-urania samples range from ∼3x10-3 to <3×10-5 mg m-2day-1 and are at least two orders of magnitude lower than those measured for UO2 spent fuels under similar experimental conditions.

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Iodine release from high-burnup fuel structures: Separate-effect tests and simulated fuel pellets for better understanding of iodine behaviour in nuclear fuels

MRS Advances ◽

10.1557/s43580-021-00175-1 ◽

2021 ◽

Author(s):

Janne Heikinheimo ◽

Teemu Kärkelä ◽

Václav Tyrpekl ◽

Matĕj̆ Niz̆n̆anský ◽

Mélany Gouëllo ◽

...

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Chemical Species ◽

Outer Edge ◽

Fuel Pellet ◽

Fuel Pellets ◽

Simulated Fuel ◽

Fission Gas ◽

High Burnup ◽

Iodine Release

Abstract Iodine release modelling of nuclear fuel pellets has major uncertainties that restrict applications in current fuel performance codes. The uncertainties origin from both the chemical behaviour of iodine in the fuel pellet and the release of different chemical species. The structure of nuclear fuel pellet evolves due to neutron and fission product irradiation, thermo-mechanical loads and fission product chemical interactions. This causes extra challenges for the fuel behaviour modelling. After sufficient amount of irradiation, a new type of structure starts forming at the cylindrical pellet outer edge. The porous structure is called high-burnup structure or rim structure. The effects of high-burnup structure on fuel behaviour become more pronounced with increasing burnup. As the phenomena in the nuclear fuel pellet are diverse, experiments with simulated fuel pellets can help in understanding and limiting the problem at hand. As fission gas or iodine release behaviour from high-burnup structure is not fully understood, the current preliminary study focuses on (i) sintering of porous fuel samples with Cs and I, (ii) measurements of released species during the annealing experiments and (iii) interpretation of the iodine release results with the scope of current fission gas release models. Graphical abstract

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Methodology to Design Simulated Irradiated Fuel by Maximizing Integral Indices (ck, E, G)

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4031074 ◽

2016 ◽

Vol 2 (2) ◽

Cited By ~ 1

Author(s):

Jason R. Sharpe ◽

Adriaan Buijs ◽

Jeremy Pencer

Keyword(s):

Supercritical Water ◽

Fission Products ◽

Optimization Method ◽

Fuel Composition ◽

Reactor Physics ◽

Similarity Indices ◽

Critical Experiment ◽

Supercritical Water Cooled Reactor ◽

Irradiated Fuel

Critical experiments are used for validation of reactor physics codes, in particular, to determine the biases and uncertainties in code predictions. To reflect all conditions present in operating reactors, plans for such experiments often require tests involving irradiated fuel. However, it is impractical to use actual irradiated fuel in critical experiments due to hazards associated with handling and transporting the fuel. To overcome this limitation, a simulated irradiated fuel, whose composition mimics the neutronic behavior of the truly irradiated fuel (TRUFUEL), can be used in a critical experiment. Here, we present an optimization method in which the composition of simulated irradiated fuel for the Canadian supercritical water-cooled reactor (SCWR) concept at midburnup (21.3 MWd kg−1 (IHM)) is varied until the integral indices ck, E, and G are maximized between the true and simulated irradiated fuel. In the optimization, the simulated irradiated fuel composition is simplified so that only the major actinides (U233, Pu238-242, and Th232) remain, while the absorbing fission products are replaced by dysprosia and zirconia. In this method, the integral indices ck, E, and G are maximized while the buckling, k∞ and the relative ring-averaged pin fission powers are constrained, within a certain tolerance, to their reference lattice values. Using this method, maximized integral similarity indices of ck=0.967, E=0.992, and G=0.891 have been obtained.

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