Analysis of OECD/NEA medium 1000 MWth sodium-cooled fast reactor using the Monte Carlo serpent code and ENDF/B-VIII.0 nuclear data library

2020 ◽  
Vol 31 (12) ◽  
Author(s):  
Fatima I. Al-Hamadi ◽  
Bassam A. Khuwaileh ◽  
Peng Hong Liem ◽  
Donny Hartanto
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Nuclear Data ◽  
Nuclear Data Library ◽  
Data Library

scholarly journals Benchmarking the new ENDF/B-VIII.0 nuclear data library for OECD/NEA medium 1000 MWth sodium-cooled fast reactor

2020 ◽  
Vol 239 ◽  
pp. 22006
Author(s):  
Donny Hartanto ◽  
Bassam Khuwaileh ◽  
Peng Hong Liem
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Nuclear Data ◽  
Control Rod ◽  
Continuous Energy ◽  
Nuclear Data Library ◽  
And Control ◽  
Benchmark Evaluation ◽  
Data Library

This paper presents the benchmark evaluation of the new ENDF/B-VIII.0 nuclear library for the OECD/NEA Medium 1000 MWth Sodium-cooled Fast Reactor (SFR). There are 2 SFR cores: metallic fueled (MET-1000) and oxide fueled (MOX-1000). The continuous-energy Monte Carlo Serpent2 code was used as the calculation tool. Various nuclear libraries such as ENDF/B-VII.1 and JENDL-4.0 were included to be compared with the newest ENDF/B-VIII.0. The evaluated parameters are k,βeff, sodium void reactivity (∆ρNa), Doppler constant (∆ρDoppler), and control rod worth (∆ρCR).

scholarly journals Evolution of the importance of neutron-induced reactions along the cycle of an LFR

2020 ◽  
Vol 239 ◽  
pp. 22010
Author(s):  
Pablo Romojaro ◽  
Francisco Álvarez-Velarde
Keyword(s):  
Fast Reactor ◽  
Safety Assessment ◽  
Nuclear Energy ◽  
Nuclear Data ◽  
Point Of View ◽  
Sensitivity Analyses ◽  
Safety Level ◽  
Energy Needs ◽  
Nuclear Data Library ◽  
Data Library

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. Several LFR concepts are now in design phase, such as MYRRHA and ALFRED, and accurate nuclear data are required for the neutronic and safety assessment of the fast reactor designs. In this work, an assessment of the evolution of the importance of neutron-induced reactions along the cycle of a reference LFR design (i.e., ALFRED) with the state-of-the-art JEFF-3.3 nuclear data library is performed. Sensitivity analyses have been carried out with MCNP6 code in order to identify the most relevant isotopes and reactions from the neutronic point of view at BoL, BoC and EoC. Furthermore, an uncertainty quantification has been performed with the SUMMON system to study the evolution of uncertainties in the keff along the reactor cycle. The results from this work provide an exhaustive picture on the influence of nuclear data on core criticality performance, identifying key quantities and nuclear data needs relevant to achieve an improved safety level for LFR.

scholarly journals Neutronics performances of gas-cooled fast reactor for 300-600 MWt Output Power with Modified CANDLE burn-up scheme in radial direction

2021 ◽  
Vol 2072 (1) ◽  
pp. 012013
Author(s):  
F H Irka ◽  
Z Suud ◽  
D Irwanto ◽  
S N Khotimah ◽  
H Sekimoto
Keyword(s):  
Output Power ◽  
Fast Reactor ◽  
Energy Production ◽  
Radial Direction ◽  
Fuel Cycle ◽  
Nuclear Data ◽  
Closed Fuel Cycle ◽  
Nuclear Data Library ◽  
Reactor Type ◽  
Data Library

Abstract Gas-Cooled Fast Reactor-GFR is a Generation IV reactor that is helium-cooled and has a closed fuel cycle. Due to the target operation on 2022-2030, this reactor type still needs further research and development technologies. We investigated the neutronics performances of a GFR balance type core with some modification of CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) burn-up scheme in the radial direction. The output power varied from 300 to 600 MWt. The neutronics calculation was performed using SRAC 2002 with JENDL 4.0 nuclear data library. The analysis indicate the reactor could operate critically for ten years without refueling with burn-up level 20% HM.

scholarly journals Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties

2021 ◽  
Vol 11 (11) ◽  
pp. 5234
Author(s):  
Jin Hun Park ◽  
Pavel Pereslavtsev ◽  
Alexandre Konobeev ◽  
Christian Wegmann
Keyword(s):  
Monte Carlo ◽  
Fusion Reactor ◽  
Nuclear Data ◽  
Data File ◽  
Nuclear Model ◽  
Model Parameters ◽  
Tritium Breeding ◽  
Tritium Breeding Ratio ◽  
Nuclear Data Library ◽  
Breeding Ratio

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively.

scholarly journals CENDL project, the chinese evaluated nuclear data library

2017 ◽  
Vol 146 ◽  
pp. 02002 ◽  
Author(s):  
Zhigang Ge ◽  
Haicheng Wu ◽  
Guochang Chen ◽  
Ruirui Xu
Keyword(s):  
Nuclear Data ◽  
Nuclear Data Library ◽  
Data Library ◽  
Evaluated Nuclear Data
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