Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate

2021 ◽  
Vol 6 (1) ◽  
pp. 14-31
Author(s):  
Kien Cuong Nguyenn ◽  
Hai Dang Vo Doan
Keyword(s):  
Conceptual Design ◽  
Nuclear Power ◽  
Nuclear Data ◽  
Critical Assembly ◽  
Design Calculation ◽  
Reactor Physics ◽  
Uranium Fuel ◽  
Calculation Results ◽  
Enriched Uranium ◽  
Accident Conditions

Critical assembly is a very important facility to serve for fundamental reactor physics research, application of neutron source, training and education. In nuclear engineering, critical assembly is a facility for carrying out measurement of reactor physics parameters, creating benchmark problem, validation of neutron physics calculation tool in computer codes and nuclear data. Basing on concept using commercial Nuclear Power Plant (NPP) fuels such as PWR (AP-1000) and VVER-1000 fuel rods with limited 2 meter in length and fully controlled by water level, the conceptual design of the critical assembly has been carried out in neutronic, thermal hydraulics and safety analysis. Ten benchmark critical core configurations of critical assembly are established and investigated to show safety during normal opeartion and accident conditions. Design calculation results show that NPP fuels are fully adequate for critical assembly operating under nominal power 100W and having average neutron flux about 3×109 neutron/cm2.s.

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

A feasibility study on fast reactor with low-enriched uranium fuel at Kyoto University Critical Assembly

2017 ◽  
Vol 100 ◽  
pp. 60-70 ◽  
Author(s):  
Song Hyun Kim ◽  
Cheol Ho Pyeon ◽  
Akito Ohizumi ◽  
Masahiro Fukushima ◽  
Kazufumi Tsujomoto ◽  
...  
Keyword(s):  
Feasibility Study ◽  
Fast Reactor ◽  
Critical Assembly ◽  
Uranium Fuel ◽  
Enriched Uranium

scholarly journals Benchmarking the new ENDF/B-VIII.0 nuclear data library for the first core of Indonesian multipurpose research reactor (RSG GAS)

2020 ◽  
Vol 239 ◽  
pp. 22007 ◽  
Author(s):  
Donny Hartanto ◽  
Victor Gillette ◽  
Tagor Malem Sembiring ◽  
Peng Hong Liem
Keyword(s):  
Research Reactor ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Kinetics Parameters ◽  
Uranium Fuel ◽  
Nuclear Data Library ◽  
Enriched Uranium ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Data Library

The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) is a 30 MWth (max.) pool-type reactor loaded with plate-type low-enriched uranium fuel, using light water as coolant and moderator, and beryllium as reflector. The benchmark of the 1st criticality core of RSG GAS using different nuclear data libraries such as JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 have been performed in the previous work and compared with the experiment result. In this work, the newly released ENDF/B-VIII.0 neutron reaction and thermal neutron scattering libraries will be used and the important neu-tronics parameters such as multiplication factor, kinetics parameters, and fission reaction rate will be calculated using Monte Carlo code MCNP6.2 and compared against the previous work and the experiment result.

Analysis of Critical Experiments Using Medium-Enriched-Uranium Fuel in Kyoto University Critical Assembly (KUCA)

1985 ◽  
Vol 22 (7) ◽  
pp. 507-520 ◽  
Author(s):  
Seiji SHIROYA ◽  
Masatoshi HAYASHI ◽  
Keiji KANDA ◽  
Toshikazu SHIBATA
Keyword(s):  
Critical Assembly ◽  
Uranium Fuel ◽  
Enriched Uranium

scholarly journals THE TCA BENCHMARK FOR VALIDATION OF TEMPERATURE FEEDBACK CALCULATIONS

2021 ◽  
Vol 247 ◽  
pp. 10010
Author(s):  
Steven C. van der Marck ◽  
Nicola L. Asquith
Keyword(s):  
Nuclear Data ◽  
Effect Of Temperature ◽  
Delayed Neutron ◽  
Neutron Data ◽  
Uranium Fuel ◽  
Temperature Feedback ◽  
Critical Configurations ◽  
Enriched Uranium ◽  
Data Libraries ◽  
Nuclear Data Libraries

The TCA benchmark was investigated as a possible candidate for validation of temperature feedback calculations. This benchmark has low-enriched uranium fuel, light water moderation and reflection, and a temperature range of 20–60 °C. The use of three different nuclear data libraries was considered, viz. ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0. Since the results were not as good as hoped for, additional studies were performed to identify the cause(s) of discrepancies. The benchmark values depend on a choice of delayed neutron data, so it was investigated whether this could be the cause of discrepancies. Also, an assessment was made based on critical configurations only, i.e. avoiding the use of delayed neutron data, whether the calculations exhibit a bias in relation to the benchmark in the results for the effect of temperature. Indications were found that such a bias exists. It is concluded that the choice of delayed neutron data has a significant effect on the benchmark values themselves. The use of three major nuclear data libraries leads to a range of benchmark values for each configuration, rather than one value. Also, one has to take into account the possibility of a bias in the calculation of temperature effects. It is not clear at this point what the cause of the bias is.

scholarly journals Some Main Results of Commissioning of The Dalat Research Reactor with Low Enriched Fuel

2014 ◽  
Vol 4 (1) ◽  
pp. 36-45
Author(s):  
Nhi Dien Nguyen ◽  
Ba Vien Luong ◽  
Van Lam Pham ◽  
Vinh Vinh Le ◽  
Ton Nghiem Huynh ◽  
...  
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Experimental Results ◽  
Design Calculation ◽  
Uranium Fuel ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Design Calculations ◽  
Dalat Research Reactor ◽  
Good Agreement

After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements.

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