Response Matrix Method–Based Importance Solver and Variance Reduction Scheme in the Serpent 2 Monte Carlo Code

The transport of neutrons in long beamlines at spallation neutron sources presents a unique challenge for Monte-Carlo transport calculations. This is due to the need to accurately model the deep-penetration of high-energy neutrons through meters of thick dense shields close to the source and at the same time to model the transport of low- energy neutrons across distances up to around 150 m in length. Typically, such types of calculations may be carried out with MCNP-based codes or alternatively PHITS. However, in recent years there has been an increased interest in the suitability of Geant4 for such types of calculations. Therefore, we have implemented supermirror physics, a neutron chopper module and the duct-source variance reduction technique for low- energy neutron transport from the PHITS Monte-Carlo code into Geant4. In the current work, we present a series of benchmarks of these extensions with the PHITS software, which demonstrates the suitability of Geant4 for simulating long neutron beamlines at a spallation neutron source, such as the European Spallation Source, currently under construction in Lund, Sweden.

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Calculation of response matrix of CaSO4:Dy based neutron dosimeter using Monte Carlo code FLUKA and measurement of 241Am–Be spectra

Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms ◽

10.1016/j.nimb.2010.06.031 ◽

2010 ◽

Vol 268 (17-18) ◽

pp. 2825-2830 ◽

Cited By ~ 3

Author(s):

S. Chatterjee ◽

A.K. Bakshi ◽

S.P. Tripathy

Keyword(s):

Monte Carlo ◽

Monte Carlo Code ◽

Response Matrix ◽

Neutron Dosimeter

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On a New Variance Reduction Technique: Neural Network Biasing - A Study of Two Test Cases with the Monte Carlo Code Tripoli4

Nuclear Technology ◽

10.13182/nt09-a9308 ◽

2009 ◽

Vol 168 (3) ◽

pp. 793-798 ◽

Cited By ~ 2

Author(s):

Eric Dumonteil

Keyword(s):

Neural Network ◽

Monte Carlo ◽

Variance Reduction ◽

Reduction Technique ◽

Test Cases ◽

Monte Carlo Code ◽

Variance Reduction Technique

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ENHANCING THE PERFORMANCE OF FISSION SOURCE CONVERGENCE WITH FUNCTIONAL EXPANSION TALLIES IN SERPENT 2 MONTE CARLO CODE

EPJ Web of Conferences ◽

10.1051/epjconf/202124704001 ◽

2021 ◽

Vol 247 ◽

pp. 04001

Author(s):

Ana Jambrina ◽

Jaakko Leppänen ◽

Heikki Suikkanen

Keyword(s):

Monte Carlo ◽

Form Factors ◽

Convergence Acceleration ◽

Initial Guess ◽

Monte Carlo Code ◽

Response Matrix ◽

Forward Solution ◽

Functional Expansion ◽

Source Convergence ◽

New Feature

This paper presents an upgrade to the built-in response matrix based solver implemented in Serpent 2 Monte Carlo code aiming to improve the fission source convergence when obtaining the forward solution to the k-eigenvalue criticality source problems. The functional expansion tallies are introduced in an attempt to improve the accuracy of the cell-wise form factors that feed the response matrix solver, replacing the current mesh-based approach. The functional expansion tallies reconstruct the binning surface and collision tallies, by using high-order series expansion to represent the original and continuous spatial distributions. This new feature is implemented to Serpent 2 and tested by single-assembly and full-core PWR calculations (BEAVRS benchmark). The results show enhanced performance of the convergence acceleration methodology based on an improved initial guess of the fission source.

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Verification of Shielding Calculation Capability of RMC With H.B.Robinson-2 Pressure Vessel Benchmark

Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation ◽

10.1115/icone26-81694 ◽

2018 ◽

Author(s):

Junjie Rao ◽

Xiaotong Shang ◽

Kan Wang

Keyword(s):

Monte Carlo ◽

Neutron Flux ◽

Pressure Vessel ◽

Variance Reduction ◽

Fission Neutron ◽

Deep Penetration ◽

Monte Carlo Code ◽

Calculation Methods ◽

Neutron Distribution ◽

Window Method

RMC is a 3-D continuous energy Monte Carlo code developed by REAL team in Tsinghua University, China. Besides the capability of fuel cycle burnup calculation, hybrid MPI/OpenMP parallelism strategy, sensitivity and uncertainty analysis, N-TH coupling calculation, shielding calculation methods including general source description, regional importance method, weight window method and source biasing method have been also developed for deep penetration problems. H.B.Robinson-2 Pressure Vessel Benchmark (HBR-2 benchmark) is used for the qualification of pressure vessel neutron flux calculation methods and shielding calculations based on this model have been performed by Monte Carlo codes such as SCALE, MCNPX and deterministic transport code DORT. In this work, the verification calculation of shielding calculation capability of RMC is conducted based on HBR-2 benchmark. The total calculation consists of two stages. Criticality calculation is performed first to obtain the fission neutron distribution in the reactor core assemblies. Then the fission neutron distribution is regarded as the initial neutron source in the following fixed source calculation. Variance reduction techniques such as source biasing and regional importance methods are combined together to be able to reduce the variance of the neutron flux in regions within and outside the pressure vessel including the downcomer and cavity regions. The preliminary calculation results show good agreement with MCNP and the shielding calculation of RMC is justified and applicable for deep penetration problems.

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Analysis of the Signal over Noise Ratio of the hodoscope determined by Monte Carlo calculation

EPJ Web of Conferences ◽

10.1051/epjconf/202125306001 ◽

2021 ◽

Vol 253 ◽

pp. 06001

Author(s):

J. Di Salvo ◽

S. Mirotta ◽

V. Chevalier

Keyword(s):

Monte Carlo ◽

Variance Reduction ◽

Nuclear Research ◽

Scattering Coefficient ◽

Fast Neutrons ◽

Monte Carlo Code ◽

Power Pulse ◽

Reduction Techniques ◽

Test Loop ◽

Noise Ratio

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA) for the purpose of the CABRI International Program (CIP), managed by IRSN in the framework of an OECD/NEA agreement. The hodoscope equipment installed in the CABRI reactor is an almost unique online fuel motion monitoring system, thanks to the measurement of the fast neutrons emitted during a power pulse by a tested rod positioned inside a dedicated test loop reproducing PWR conditions. This system is dedicated to the analysis of fuel displacement. Hence, one of the most important parameter measured by the hodoscope detectors is the Signal over Noise Ratio (SNR), characterizing the fraction of neutrons directly coming from the test rod (“signal”) over neutrons coming from the core (“noise”). It is interesting to calculate the SNR in order to define some quantitative criterions to improve hodoscope measurements and to understand if any modification linked to the test loop may significantly change this essential parameter. Another parameter of interest is the so-called “scattering coefficient”, which corresponds to the fraction of neutrons coming from the test rod and being scattered between their birth and their detection. This parameter is used to enhance the analysis of the fuel displacement which may happen during the power transient. In this article, the method used to calculate the SNR using MCNP6.2 Monte Carlo code will be detailed. Because the hodoscope detectors are located far away from the test rod (up to 4 meters), a 2D model of CABRI core and instrumentation has been implemented. No variance reduction techniques have been used to solve this problem in order to record the place of birth of neutron which contributes to the different scores with the goal to perform a detailed analysis of the SNR. This strategy allows to access numerically to the “scattering coefficient”. Finally, the comparison between calculated and measured SNR for a case study will be carried out. A quite good agreement between the 2D simulations and experiments recently performed in the CABRI reactor has been obtained.

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A hybrid approach for neutronics calculations in the neutron shielding of sodium fast reactors

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2021016 ◽

2021 ◽

Vol 7 ◽

pp. 17

Author(s):

Amine Hajji ◽

Christine Coquelet-Pascal ◽

Patrick Blaise

Keyword(s):

Monte Carlo ◽

Variance Reduction ◽

Hybrid Approach ◽

Complex Problem ◽

Gas Production ◽

Monte Carlo Code ◽

Hybrid Scheme ◽

Neutron Shielding ◽

Reduction Techniques ◽

Computational Performance

Neutron calculations in the neutron shielding of fast neutron reactors are a complex problem as deterministic schemes are usually not suited for such calculations while Monte-Carlo codes have poor computational performance due to the very low flux levels in neutron shields. In this article, both methods are studied, as well as a hybrid scheme on the neutron shielding of the ASTRID fast reactor benchmark. This hybrid scheme uses a fission source calculated by a deterministic code in order to precisely calculate neutron fluxes in the shielding with a Monte-Carlo code using variance reduction techniques. This provides reference results in order to validate deterministic calculations. Comparisons between deterministic codes and this hybrid reference show that large biases are obtained, up to 50%. Further studies are made to reduce the biases, showing that many physical phenomena should be treated, including anisotropy of the scattering law at high energies and spatial self-shielding inside the boron carbide shielding. These improvements reduce the biases to less than 10%. Finally, some applications to designing criteria for the neutron shielding are presented, such as gas production in the neutron shielding and activation of secondary sodium at the intermediate heat exchanger (IHX).

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