A Comparison of Fast Thorium Breeder Reactor Designs With Oxide and Metallic Fuels

2008 ◽  
Author(s):  
V. Jagannathan ◽  
Usha Pal ◽  
R. Karthikeyan ◽  
Devesh Raj
Keyword(s):  
Unit Volume ◽  
Core Region ◽  
Metallic Fuel ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Depletion Rate ◽  
Metallic Fuels ◽  
Core Characteristics

Loading of seedless thoria rods in internal blanket regions and using them later as part of seeded fuel assemblies is the central theme of the thorium breeder reactor (ATBR) concept [1]. The fast reactors presently consider seedless blanket region surrounding the seeded core region. This results in slower fissile production rate in comparison to fissile depletion rate per unit volume. The overall breeding is achieved mainly by employing blanket core with more than double the volume of seeded core. The blanket fuel is discharged with fissile content of ∼30g/kg, which is much less than the asymptotic maximum possible fissile content of 100g/kg. This is due to smaller coolant flow provided for in the blanket regions. In a newly proposed fast thorium breeder reactor (FTBR) [2], the blanket region is brought in and distributed through out the core. By this the fissile depletion and production rates per unit volume become comparable. The core considered simultaneous breeding from both fertile thoria and depleted uranium and hence the concept can be called as fast twin breeder reactor as well. Sodium is used as coolant. The blanket fuel rods achieve nearly 80% of the seed fuel rod burnup and also contain nearly the maximum possible fissile content at the time of discharge. In this paper a comparison of FTBR core characteristics with oxide and metallic fuel are compared.

scholarly journals Engineering Margin in Calculations of Energy Release in VVER-1000 core

2018 ◽  
pp. 20-26
Author(s):  
A.M. Abdullayev ◽  
A.I. Zhukov ◽  
S.V. Maryokhin ◽  
S.D. Riabchykov
Keyword(s):  
Energy Release ◽  
Three Dimensional ◽  
Measurement Data ◽  
Reactor Core ◽  
Small Scale ◽  
Two Dimensional ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies

A method for calculating the engineering margin factor (EMF) in calculations of the energy release in the core of VVER-1000 reactors is proposed in the paper. The analysis of various approaches in the calculation of EMF is carried out and various factors influencing EMF and the ways of their consideration —deterministic and statistical — are determined. The main attention is paid to the influence of gaps between the fuel assemblies on the energy release of fuel rods and the contribution of this factor to the EMF. The limitations and conservatism of two-dimensional small-scale calculations of the energy release of fuel rods in case of deviation of the gap size between the fuel assemblies from the design one are shown. A three-dimensional approach to calculating the contribution of gaps to the EMF is proposed. The approach is based on detailed measurements of the shape of fuel assemblies removed from the core performed at Zaporizhzhya NPP [13]; simulation of the distribution of gaps in the reactor core [16] using measurement data; two-dimensional calculations of the energy release of fuel rods in separate fuel assemblies, surrounded by gaps of different widths, with mirroring boundary conditions; three-dimensional calculations of energy release of fuel rods in fuel assemblies in the reactor core. Two-dimensional and three-dimensional calculations are performed by the wellknown ALPHA-H/PHOENIX-H/ANC-H codes. The proposed approach allows considering not only the change in the fuel rod power, particularly of the peripheral rods, which is inherent in the currently used methods of calculating EMF, but also takes into account the change in the power of the fuel assemblies in the core, which makes the proposed method more realistic and removes the excessive conservatism of EMF calculations and, thereby, allows improving fuel efficiency. For fuel assemblies produced by Westinghouse, it is proposed to use full EMF: for fuel rod power (FΔH) 1.111 and for fuel rod linear power (FQ) 1.173. The use of the BEACONTM monitoring system makes it possible to further reduce the EMF: for fuel rod power (FΔH) - up to 1.084 and for fuel rod linear power (FQ) - up to 1.121.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

scholarly journals Development of End Plug Welding Technique for SFR Fuel Rod Fabrication

2016 ◽  
Vol 2016 ◽  
pp. 1-9
Author(s):  
Jung Won Lee ◽  
Jong Hwan Kim ◽  
Ki Hwan Kim ◽  
Jeong Yong Park ◽  
Sung Ho Kim
Keyword(s):  
Arc Welding ◽  
Reactor Performance ◽  
Metallic Fuel ◽  
Fuel Rods ◽  
Welding Equipment ◽  
Fuel Rod ◽  
Gas Tungsten Arc ◽  
Welding Technique ◽  
Zr Alloy

In Korea, R&D on a sodium-cooled fast reactor (SFR) was begun in 1997, as one of the national long-term nuclear R&D programs. As one fuel option for a prototype SFR, a metallic fuel, U-Zr alloy fuel, was selected and is currently being developed. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of the end plug to the cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions, and parameters were developed for the end plug welding of SFR metallic fuel rods. A gas tungsten arc welding (GTAW) technique was adopted and the welding joint design was developed. In addition, the optimal welding conditions and parameters were established. Based on the establishment of the welding conditions, the GTAW technique was qualified for the end plug welding of SFR metallic fuel rods.

A Study on SCWR Core Design and Fuel Management

2013 ◽  
Author(s):  
Yong Zhang ◽  
Jue Yang ◽  
Dongsheng Li ◽  
Liangzhi Cao ◽  
Fei Wang ◽  
...  
Keyword(s):  
Thermal Power ◽  
Coolant Temperature ◽  
Equivalent Diameter ◽  
Fuel Management ◽  
Double Row ◽  
Fuel Rods ◽  
Average Temperature ◽  
The Core ◽  
Fuel Assemblies ◽  
Operation Pressure

A type of SCWR core design with Double-row fuel rods assembly will be introduced in this paper. The Double-row fuel rods SCWR (DL-SCWR) core is designed for reaching a thermal power of 2280 MW with an equivalent diameter of 3.16 m and length of 4.20 m. There are 249 fuel assemblies in the core, and 180 fuel rods in each assembly. The inlet coolant temperature is 280 °C and the operation pressure is 25 MPa. There are 80 fresh assemblies in the equilibrium cycle. The numerical results show that the average discharged burnup of the equilibrium is 37 GWD/tU, and the average temperature of outlet coolant is 500°C. In the equilibrium cycle, the maximum cladding surface temperature is less than 650 °C.

Identification of Non-Linear Parameters of a Nuclear Fuel Rod

2017 ◽  
Author(s):  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Prabakaran Balasubramanian ◽  
Marco Amabili ◽  
Brian Painter ◽  
...  
Keyword(s):  
Circular Cylindrical Shell ◽  
Harmonic Balance Method ◽  
Excitation Amplitude ◽  
Resonance Phenomenon ◽  
Response Curves ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Safety Margins ◽  
Fuel Assemblies

In Pressurized Water Reactors (PWR), fuel assemblies are made up of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the nonlinear response of the structure, and, particularly, its dissipation, is of paramount importance for the choice of safety margins, in the design of fuel assemblies, to ensure their functionality and safety in the worst external condition scenarios. To model the nonlinear dynamic response of fuel rods, the identification of the nonlinear stiffness and damping parameters is required. A tool based on harmonic balance method was developed to identify these parameters from the experimentally obtained force-response curves, considering one-to-one internal resonance phenomenon present in axisymmetric structures such as cylindrical tubes and shells. To validate the tool, it was applied to the reference case of circular cylindrical shell filled with water, which revealed an increase of damping with the excitation amplitude. In the following paper, the more realistic case of a single fuel rod with clamped-clamped boundary condition was investigated by applying harmonic excitation at various force levels. The nonlinear parameters including damping were extracted from experimental results by means of the adapted tool. An increase in damping with excitation amplitude has been shown according to earlier studies.

Development of Approaches to Simulate Fuel Rod Destruction With Different Fuel Type

2020 ◽  
Author(s):  
Eduard Usov ◽  
Nikolay Pribaturin ◽  
Vladimir Chukhno ◽  
Ilya Klimonov ◽  
Anton Butov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Liquid Metals ◽  
Numerical Algorithms ◽  
Severe Accident ◽  
Theoretical Research ◽  
Detailed Calculation ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core

Abstract Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.

scholarly journals Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

2013 ◽  
Vol 2013 ◽  
pp. 1-7 ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Power Distribution ◽  
Outer Core ◽  
Monte Carlo Code ◽  
The Core ◽  
Fuel Burn ◽  
Fuel Assemblies ◽  
Core Characteristics ◽  
Power Flattening ◽  
Reactivity Insertion ◽  
Operational Time

An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

Research and Engineering Application of Repaired Fuel Assembly Reusing Technique

2017 ◽  
Author(s):  
Lihua Wang ◽  
Qingxiang Yang ◽  
Ping Yang ◽  
Jiazheng Liu ◽  
Libing Zhu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Engineering Application ◽  
Evaluation Technique ◽  
Fuel Rods ◽  
The Core ◽  
Fuel Assemblies

Due to debris in the coolant against clad, fuel clad wear, fuel handling fault and so on, fuel rods maybe be damaged during the operation of nuclear power plants, in order that the fuel assemblies with damaged fuel rods are discharged before scheduled. If the damaged fuel assemblies are not reloaded into the core of the nuclear power plant, the fuel utilization decreases and the economy of the nuclear power plant is partly lost. For retrieving the loss of the economy, the damaged fuel assemblies can be repaired by replacing damaged fuel rods with dummy rods which don’t include fissile nuclides. Then, the repaired fuel assemblies can be reloaded into the core. As the repaired fuel assemblies are different with the normal fuel assemblies, especially the number of the damaged fuel rods is considerable, a whole quantitative analysis is very necessary to evaluate the effects from the reuse of the repaired fuel assemblies. In this paper, a full scope evaluation of reload design are performed including nuclear design, fuel design, thermal hydraulic design and safety evaluation, and some necessary improvements are done for the software system, design methods and progress which have been used in the normal reload design. As results, an integrated evaluation technique is developed to evaluate the feasibility and safety of reusing the repaired fuel assemblies, and the key effects due to the reuse of the repaired fuel assemblies are extracted, and the different effects are studied for the different materials of the dummy rods which can be used to conduct how to choose the proper material of dummy rods. In addition, this technique has been successfully applied in the engineering and the loss of economy due to the damage of fuel assemblies was retrieved partly. Therefore, the integrated evaluation technique has also important directive to other nuclear power plants if the repaired fuel assemblies are planned to reuse.

A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

2002 ◽  
Author(s):  
Hiroyuki Yoshida ◽  
Akira Ohnuki ◽  
Hajime Akimoto
Keyword(s):  
Pressure Vessel ◽  
Volume Ratio ◽  
Normal Operation ◽  
Primary System ◽  
Fuel Rods ◽  
The Core ◽  
Fuel Assemblies ◽  
Core Cooling ◽  
Lattice Core ◽  
Tight Lattice

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation PWR can be maintained.

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