scholarly journals Reactor with metallic fuel and lead-208 coolant

2020 ◽  
Vol 6 (1) ◽  
pp. 23-27
Author(s):  
Georgiy L. Khorasanov ◽  
Anatoliy I. Blokhin
Keyword(s):  
Large Fraction ◽  
Metallic Fuel ◽  
The Core ◽  
Medical Radioisotopes ◽  
Hard Spectrum ◽  
Average Neutron ◽  
High Neutron ◽  
High Neutron Flux ◽  
Average Neutron Energy ◽  
Zr Alloy

The paper considers the concept of a fast lead cooled 25MW reactor for a variety of applications, including incineration of minor actinides, production of medical radioisotopes, testing of radiation-damaged nuclear technology materials, etc. A specific feature of the proposed reactor is rather a high neutron flux of 2.6·1015 n/(cm2·s) at the core center, high average neutron energy of 0.95 MeV at the core center, and a large fraction (40%) of hard neutrons (En > 0.8 MeV). The extremely high estimated reactor parameters are achieved thanks to the small core dimensions (DxH ≈ 0.50×0.42 m2), innovative metallic fuel of the Pu-Am-Np-Zr alloy, and the 208Pb enriched lead coolant. A relatively high probability of 241Am fission (about 50%) is achieved in the reactor core’s hard spectrum, this making it possible to incinerate up to 4 kg of 241Am during one reactor campaign of 1000 effective days.

Anticipated Transient Without Scram Analysis for an ABWR Using RETRAN

2008 ◽  
Author(s):  
Chiung Wen Tsai ◽  
Shu Ming Yang ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Shao Shih Ma ◽  
...  
Keyword(s):  
High Pressure ◽  
Control System ◽  
Neutron Flux ◽  
Transient Analysis ◽  
Valve Closure ◽  
The Third ◽  
The Core ◽  
Main Steam ◽  
High Neutron ◽  
High Neutron Flux

A RETRAN02/MOD5 model was developed for Lungmen ABWR and applied for ATWS transient analysis. Three ATWS events including Main Steam Isolation Valve Closure (MSIVC), Loss of Offsite Power (LOOP), and Inadvertent Opening of all Turbine Bypass Valves (IOTBV) are analyzed in this study. During the first two transients, the vessel pressure is increased as a result of steam flow reduction due to the closure of main steam isolation valves (MSIVs) and Turbine Control Valves (TCVs) respectively. In the third transient, the vessel pressure is reduced because of the open of Turbine Bypass Valves (TBVs) and turns to be increased because of the closure of MSIVs. All of the above transients suffer high neutron flux as a result of void reduction. There are several equipments and procedures to mitigate ATWS transient such as feedwater trip, Reactor Internal Pumps (RIPs) runback and trip, and the depressurization of relief valves. After the ATWS high pressure signal is initiated and permissive for 180 seconds, Standby Liquid Control system is initiated to inject boron liquid into upper plenum to shutdown the reactor. The results conclude that equipments and procedures mitigate the event effectively and the core is brought to shutdown state.

Damage Calculation for First Wall Submitted to High Neutron Flux in a Tokamak

2015 ◽  
Vol 1769 ◽  
Author(s):  
C.E. Velasquez ◽  
M. A. F. Veloso ◽  
A. L. Costa ◽  
C. Pereira
Keyword(s):  
Neutron Flux ◽  
Energy Deposition ◽  
Reaction Rates ◽  
Main Reaction ◽  
First Wall ◽  
Tokamak Reactor ◽  
Mcnp5 Code ◽  
Displacement Per Atom ◽  
High Neutron ◽  
High Neutron Flux

ABSTRACTThe displacement per atom (dpa) has been a specific issue to evaluate the damage in the first wall of the Tokamak. Two different first wall materials were evaluated. In this study, MCNP5 code was used to obtain the neutron flux, the energy deposition and the main reaction rates, on the inboard and outboard first wall. The damage calculations were performed by the SPECTER code using the neutronic parameters obtained by MCNP5. The Tokamak reactor modeled has similar dimensions to the ITER. Tungsten and beryllium alloys were simulated on the outboard first wall. The results indicate which material has a higher resistance to be damage and dpa values for the analyzed material.

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 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.

Correlation femtoscopy in neutron-carbon interactions at average neutron energy of 51 GeV

2005 ◽  
Vol 68 (3) ◽  
pp. 481-487 ◽  
Author(s):  
A. N. Aleev ◽  
N. S. Amaglobeli ◽  
V. P. Balandin ◽  
O. V. Bulekov ◽  
I. M. Geshkov ◽  
...  
Keyword(s):  
Neutron Energy ◽  
Average Neutron ◽  
Average Neutron Energy
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