scholarly journals A Simplified Supercritical Fast Reactor with Thorium Fuel

2014 ◽  
Vol 2014 ◽  
pp. 1-9
Author(s):  
Peng Zhang ◽  
Kan Wang ◽  
Ganglin Yu
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Mechanical Design ◽  
Inherent Safety ◽  
Fuel Conversion ◽  
Reactor Physics ◽  
Mox Fuel ◽  
Feasible Option ◽  
Gen Iv ◽  
Simulation Scheme

Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

Development and Application of the Neutronics/Thermal-Hydraulics Coupling Code for Safety Analysis of EBR-II Loss of Heat Sink Tests Without Scram

2016 ◽  
Author(s):  
Daogang Lu ◽  
Chao Guo ◽  
Danting Sui
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Fuel Cycle ◽  
Transient Behavior ◽  
Safety Analysis ◽  
Cold Pool ◽  
Thermal Hydraulics ◽  
Closed Fuel Cycle ◽  
Zone Method ◽  
Gen Iv

In the GEN IV technology evaluations, the LMFBR (Liquid Metal Fast Breeder Reactor) system which includes SFR (Sodium-cooled Fast Reactor) and LFR (Lead-cooled Fast Reactor) was top-ranked in sustainability due to its closed fuel cycle and it is top-ranked in proliferation resistance and physical protection because it employs a long-life core. It is necessary to conduct the coupled neutronics and thermal-hydraulics simulation when the feedback effects are significant in the safety analysis of Anticipated Transients Without Scram (ATWS) in LMFBR. Thus, a neutronics-thermalhydraulics coupling code for safety analysis of LMFBR was developed and used to analyze whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under Loss of Heat Sink Without Scram (LOHSWS) tests in this paper. The two mixing zone method for cold pool coupled with SAC-CFR was used and the predicted results agree well with measurements which are taken from EBR-II LOHSWS test data.

Hard Spectrum STFR (Spray Type Fast Reactor) Cooled by Water at 7.5 MPa

2004 ◽  
Author(s):  
Asashi Kitamoto ◽  
Yasunori Ohoka
Keyword(s):  
Fast Reactor ◽  
Void Ratio ◽  
Cross Flow ◽  
Hot Water ◽  
Future Performance ◽  
Mox Fuel ◽  
Fuel Burn ◽  
Hard Spectrum ◽  
Two Phases ◽  
Gen Iv

New concept of fast reactor (FR), i.e. the STFR (Spray Type FR), is proposed here to perform high burn-up of UO2 fuel or MOX fuel by the use of BWR technology, and to improve the backend process of discharged fuel. STFR can be realized by some important changes of BWR system, at 7.50MPa. In case of STFR, heat produced in the core is removed by the evaporation of sprayed hot water jetted to fuel with cross flow at 7.50MPa, and two phases (liquid and vapor) of coolant at high void ratio goes down to the bottom of PV (pressure vessel). This is an improved concept of BWR, which can be regarded as a breakthrough of FBR. This concept has not been listed in GEN IV. Future performance of STFR are as follows, (1) STFR can increase the fraction of direct fission of 238U with neutron reaction of higher energy than 1MeV, (2) STFR can burn the nuclear fuel to the higher burn-up (80 to 200GWd/Mg-HM) compared with BWR fuel burn-up. (3) Higher burn-up of fuel will reduce the frequency of reprocessing, so STFR can reduce the reprocessing cost per power production. (4) STFR can reduce the remains of Pu and MA (Minor Actinides: Np, Am, Cm etc.) in discharged fuel.

Sub-Channel Analysis of Pb-Bi-Cooled Reactor With Modified COBRA-EN

2014 ◽  
Author(s):  
Yonghong Tian ◽  
Wenxi Tian ◽  
Zhaoming Meng ◽  
Yingwei Wu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Experimental Results ◽  
Inherent Safety ◽  
Nuclear Wastes ◽  
Light Water ◽  
Driven System ◽  
Channel Analysis ◽  
Accelerator Driven System ◽  
Gen Iv ◽  
Good Agreement

Lead-bismuth eutectic (LBE) cooled fast reactor, one of the six types of reactors in Gen-IV, has very good inherent safety and significant advantages in reducing and burning nuclear wastes, enhancing economy. Also LBE cooled accelerator driven system (ADS) has been a very innovative and potential waste burner. COBRA-EN is a mature, stable and widely-used sub-channel analysis code for light water cooled reactor but it couldn’t be applied in Pb-Bi-cooled reactor directly. Some modifications were made for COBRA-EN in the present work, then the code was named COBRA-PB and was suitable for the sub-channel analysis of Pb-Bi-cooled reactor. The modified code was verified and validated with CFX and experimental results. There was a good agreement between the two results. Then sub-channel analysis of Pb-Bi-cooled reactor was done with the modified code.

scholarly journals A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

2013 ◽  
Vol 2013 ◽  
pp. 1-10 ◽  
Author(s):  
Hangbok Choi ◽  
Robert W. Schleicher ◽  
Puja Gupta
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
Matrix Material ◽  
Reactor Core ◽  
Operation Time ◽  
Fuel Load ◽  
Fissile Material ◽  
Reactor Physics

In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2), which is a compact gas-cooled fast reactor (GFR). The EM2augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

scholarly journals Gas Cooled Fast Reactor Research and Development in the European Union

2009 ◽  
Vol 2009 ◽  
pp. 1-7 ◽  
Author(s):  
Richard Stainsby ◽  
Karen Peers ◽  
Colin Mitchell ◽  
Christian Poette ◽  
Konstantin Mikityuk ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Energy System ◽  
Outlet Temperature ◽  
Framework Programme ◽  
The European Union ◽  
Closed Fuel Cycle ◽  
Gen Iv ◽  
Actinide Transmutation

Gas-cooled fast reactor (GFR) research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV), that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5) GCFR project in 2000, through FP6 (2005 to 2009) and looking ahead to the proposed activities within the 7th Framework Programme (FP7).

Comparative Study of Basic Reactor Physics of the DFR Concept Using U-Pu and TRU Fuel Salts

2017 ◽  
Author(s):  
Xiang Wang ◽  
Rafael Macian-Juan
Keyword(s):  
Steady State ◽  
Liquid Metal ◽  
Molten Salt ◽  
Fast Reactor ◽  
Flux Distribution ◽  
Molten Salt Reactor ◽  
Reactor Physics ◽  
Fuel Salt ◽  
Gen Iv ◽  
First Time

The Dual Fluid Reactor (DFR) is a molten salt fast reactor developed by the IFK1 based on the Gen-IV Molten-Salt Reactor (MSR) and the Liquid-Metal Cooled Reactor (SFR, LFR) concepts. The analysis reported focuses on the comparison between previous neutronic calculations with the default fuel salt of U-Pu mixture and new ones with a transuranium (TRU) salt fuel option under steady state conditions. They include criticality, neutron spectrum, spatial flux distribution and temperature coefficient values. Fuel based on molten TRU salts has already been considered for the MSFR and other molten salt reactor designs. Therefore, the DFR for the first time has a comparable baseline with other molten salt reactors, so that its performance with TRU salt fuel can be assessed.

Current Status of the Development of the Fast Reactor Fuel Cycle in India

2009 ◽  
pp. 120-126
Author(s):  
K.V. Govindan Kutty ◽  
P.R. Vasudeva Rao ◽  
Baldev Raj
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Fuel ◽  
Current Status ◽  
Fast Reactor Fuel

scholarly journals The Effect Assessment for Fast Reactor Fuel Cycle Deployment-improvement of the Assessment Method-

2013 ◽  
Vol 39 ◽  
pp. 43-51
Author(s):  
Kyoko Mukaida ◽  
Hiroki Shiotani ◽  
Kiyoshi Ono ◽  
Takashi Namba
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Assessment Method ◽  
Reactor Fuel ◽  
Effect Assessment ◽  
Fast Reactor Fuel

Component design and accident analysis of fuel assembly for prototype GEN-IV sodium-cooled fast reactor

2018 ◽  
Vol 340 ◽  
pp. 133-145 ◽  
Author(s):  
K.H. Yoon ◽  
H.-K. Kim ◽  
H.S. Lee ◽  
J.S. Cheon
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Accident Analysis ◽  
Component Design ◽  
Gen Iv

Progressive Evolution of the Core of the Fast Breeder Test Reactor

2010 ◽  
Author(s):  
S. Varatharajan ◽  
K. V. Sureshkumar ◽  
K. V. Kasiviswanathan ◽  
G. Srinivasan
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Power Reactor ◽  
Test Bed ◽  
Fast Breeder Test Reactor ◽  
The Core ◽  
Linear Heat ◽  
Mox Fuel ◽  
Test Reactor ◽  
Carbide Fuel

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

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