scholarly journals VVER-1000 BENCHMARK INTERPRETATION WITH MONTE CARLO CODE MCS

2021 ◽  
Vol 247 ◽  
pp. 10006
Author(s):  
Fathurrahman Setiawan ◽  
Matthieu Lemaire ◽  
Hyunsuk Lee ◽  
Peng Zhang ◽  
Deokjung Lee
Keyword(s):  
Monte Carlo ◽  
Nuclear Research ◽  
Concrete Block ◽  
Nuclear Data ◽  
Thermal Reactor ◽  
Deep Penetration ◽  
Monte Carlo Code ◽  
Nuclear Research Institute ◽  
Spacer Grids ◽  
Nuclear Data Library

An interpretation of the NEA-1517/82 benchmark from the SINBAD shielding database has been conducted with the MCS Monte Carlo code developed at the Ulsan National Institute of Science and Technology (UNIST) and the ENDF/B-VII.1 nuclear data library. The NEA-1517/82 benchmark corresponds to experiments on a VVER-1000 critical mock-up (thermal reactor with hexagonal fuel lattice) inside the LR-0 research reactor operated by the Nuclear Research Institute (NRI) in the Czech Republic. A new 3D model of the VVER-1000 mock-up core is developed for MCS based on the SINBAD documentation. The model includes the top and bottom parts of fuel pins, the spacer grids and core components: baffle, barrel, downcomer, tank, reactor pressure vessel (RPV) and concrete block used as biological shielding. The quality of the model is verified first by code/code comparison of MCS against MCNP6 for criticality and power distributions (pin-by-pin and axial power). The validation of MCS results is then performed against six critical cases, 260 measured pin powers and benchmark calculations of the axial power profile. Finally, a comparison of calculated and measured neutron spectra inside the mock-up core is presented as a preliminary study for upcoming works on the deep-penetration shielding capability of MCS.

scholarly journals Processing and application of nuclear data for low temperature criticality assessment

2020 ◽  
Vol 239 ◽  
pp. 14006
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Tara Hanlon ◽  
Richard Hiles ◽  
Malcolm Lingard ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Low Temperature ◽  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Monte Carlo Code ◽  
Temperature Range ◽  
Nuclear Data Library ◽  
Criticality Assessment ◽  
Good Agreement

Until recently, criticality safety assessment codes had a minimum temperature at which calculations can be performed. Where criticality assessment has been required for lower temperatures, indirect methods, including reasoned argument or extrapolation, have been required to assess reactivity changes associated with these temperatures. The ANSWERS Software Service MONK® version 10B Monte Carlo criticality code, is capable of performing criticality calculations at any temperature, within the temperature limits of the underlying nuclear data in the BINGO continuous energy library. The temperature range of the nuclear data has been extended below the traditional lower limit of 293.6 K to 193 K in a prototype BINGO library, primarily based on JEFF-3.1.2 data. The temperature range of the thermal bound scattering data of the key moderator materials was extended by reprocessing the NJOY LEAPR inputs used to produce bound data for JEFF-3.1.2 and ENDF/B-VIII.0. To give confidence in the low temperature nuclear data, a series of MONK and MCBEND calculations have been performed and results compared against external data sources. MCBEND is a Monte Carlo code for shielding and dosimetry and shares commonalities to its sister code MONK including the BINGO nuclear data library. Good agreement has been achieved between calculated and experimental cross sections for ice, k-effective results for low temperature criticality benchmarks and calculated and experimentally determined eigenvalues for thermal neutron diffusion in ice. To quantify the differences between ice and water bound scattering data a number of MONK criticality calculations were performed for nuclear fuel transport flask configurations. The results obtained demonstrate good agreement with extrapolation methods. There is a discernible difference in the use of ice and water data.

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 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.

Organ Dose Calculations Using ICRP Adult Voxel Phantoms and TRIPOLI-4 Monte Carlo Code

2020 ◽  
Vol 6 (4) ◽  
Author(s):  
Yi-Kang Lee
Keyword(s):  
Monte Carlo ◽  
Computational Models ◽  
Radiation Transport ◽  
Organ Dose ◽  
Radiation Shielding ◽  
Radiological Protection ◽  
Monte Carlo Code ◽  
Coupled Transport ◽  
Nuclear Data Library ◽  
Voxel Phantoms

Abstract The ICRP 110 adult male and female voxel phantoms are the official computational models representing the ICRP (International Commission on Radiological Protection) Reference Male and Reference Female. In 2018, the Working Group 6 (WG6) of European Radiation Dosimetry Group (EURADOS) organized a study on the usage of the ICRP voxel reference phantoms. Organ dose calculation tasks with radiation transport codes were proposed in occupational, environmental, and medical dosimetry. The TRIPOLI-4 Monte Carlo radiation transport code has been widely used in radiation shielding, criticality safety, and reactor physics fields for supporting French nuclear energy research and industrial applications. To enhance the application fields of TRIPOLI-4, the 2018 EURADOS-WG6 tasks are being taken into account by using different features of the TRIPOLI-4 code. In this work, the ICRP reference voxel phantoms were first adapted into TRIPOLI-4. More than 14 × 106 voxels were represented in a mixed lattice geometry including 140 organs-tissues and 52 tissue media. Diverse exposure scenarios were then investigated by using 60Co and 241Am gamma-ray sources, 16N beta source, and 10 keV neutron source. The TRIPOLI-4 standard nuclear data library was utilized on these neutron, photon, electron, and positron-coupled transport calculations. Energy deposition estimators for electron, positron, neutron, and photon coupled with mesh tally options were used to calculate the organ absorbed dose DT and the effective dose E. TRIPOLI-4 calculation methods and primary results for the EURADOS-WG6 voxel phantom exercise on organ dose study tasks are reported here.

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