Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCS

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.

Download Full-text

Processing and application of nuclear data for low temperature criticality assessment

EPJ Web of Conferences ◽

10.1051/epjconf/202023914006 ◽

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.

Download Full-text

Analysis of the CREOLE experiment on the reactivity temperature coefficient of the UO2 light water moderated lattices using Monte Carlo transport calculations and ENDF/B-VII.1 nuclear data library

Nuclear Engineering and Technology ◽

10.1016/j.net.2019.11.018 ◽

2020 ◽

Vol 52 (6) ◽

pp. 1120-1130

Author(s):

S. El Ouahdani ◽

L. Erradi ◽

H. Boukhal ◽

E. Chakir ◽

T. El Bardouni ◽

...

Keyword(s):

Monte Carlo ◽

Temperature Coefficient ◽

Nuclear Data ◽

Light Water ◽

Nuclear Data Library ◽

Data Library ◽

Monte Carlo Transport

Download Full-text

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

EPJ Web of Conferences ◽

10.1051/epjconf/202023922006 ◽

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).

Download Full-text

Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries

Nuclear Energy and Technology ◽

10.3897/nucet.7.68951 ◽

2021 ◽

Vol 7 (2) ◽

pp. 103-109

Author(s):

Olga N. Andrianova ◽

Yury Ye. Golovko ◽

Gleb B. Lomakov ◽

Yevgeniya S. Teplukhina ◽

Gennady M. Zherdev

Keyword(s):

Nuclear Data ◽

Monte Carlo Code ◽

Fast Reactors ◽

Neutron Data ◽

Critical Configurations ◽

Nuclear Data Library ◽

Data Libraries ◽

Nuclear Data Libraries ◽

Data Library ◽

Evaluated Nuclear Data

The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear data library and other evaluated nuclear data libraries (ENDF, JEFF, JENDL) from different years. A set of integral experiments on BFS critical assemblies carried out in different years at the Institute of Physics and Power Engineering (60 different critical configurations) was analyzed. The considered integral experiments are included in the database of evaluated experimental neutronic data used to justify the neutronic performance of sodium and lead cooled fast reactors, to verify codes and nuclear data as well as to estimate uncertainties in neutronic parameters due to the nuclear data uncertainties. It has been shown that the ROSFOND evaluated nuclear data library is a library that minimizes the calculation and experimental discrepancies for the considered set of integral experiments. The paper also presents the results of criticality calculations for models of sodium and lead cooled fast reactors based on different evaluated neutron data libraries and provides estimates for the uncertainty in criticality associated with nuclear data.

Download Full-text

VVER-1000 BENCHMARK INTERPRETATION WITH MONTE CARLO CODE MCS

EPJ Web of Conferences ◽

10.1051/epjconf/202124710006 ◽

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.

Download Full-text

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

Nuclear Science and Techniques ◽

10.1007/s41365-020-00831-5 ◽

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

Download Full-text

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

Applied Sciences ◽

10.3390/app11115234 ◽

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.

Download Full-text