COUNTERING NEUTRON CLUSTERING IN MONTE CARLO WITH A NEUTRON SOURCE INJECTION

Neutron clustering is a recently identified problem with Monte Carlo eigenvalue calculations which can produce significantly erroneous results. Previous work by Sutton & Mittal (2017) considered neutron clustering as a problem of maintaining ‘genetic diversity’ within the neutron population. This paper proposes reducing the extent of neutron clustering by replacing fission neutrons in the source bank with uncorrelated neutrons, sampled from a uniform source distribution – effectively adding new neutron genealogies to the population. The efficacy of the method is demonstrated on a number of simple problems, showing improved behaviour of the Shannon entropy and neutron centre-of-mass. Although currently limited in scope, this paper intends to provide a route to reducing clustering effects in more general problems.

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Experimental and Computational Dose Rate Evaluation Using SN and Monte Carlo Method for a Packaged 241AmBe Neutron Source

Nuclear Science and Engineering ◽

10.1080/00295639.2021.1906587 ◽

2021 ◽

pp. 1-22

Author(s):

Meng-Jen (Vince) Wang ◽

Glenn E. Sjoden

Keyword(s):

Monte Carlo ◽

Monte Carlo Method ◽

Dose Rate ◽

Neutron Source ◽

Rate Evaluation

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Conceptual Shield Design for Boron Neutron Capture Therapy Facility Using Monte Carlo N-Particle Extended Simulator with Kartini Research Reactor as Neutron Source

ASEAN Journal on Science and Technology for Development ◽

10.29037/ajstd.532 ◽

2020 ◽

Vol 35 (3) ◽

pp. 177-181

Author(s):

Afifah Hana Tsurayya ◽

Azzam Zukhrofani Iman ◽

R. Yosi Aprian Sari ◽

Arief Fauzi ◽

Gede Sutresna Wijaya

Keyword(s):

Monte Carlo ◽

Radiation Dose ◽

Dose Rate ◽

Neutron Source ◽

Gamma Ray ◽

Outer Part ◽

Radiation Shielding ◽

Radiation Dose Rate ◽

Boron Neutron Capture ◽

Radiation Workers

The research aims to measure the radiation dose rate over the radiation shielding which is made of paraffin and aluminium and to determine the best shield material for the safety of radiation workers. The examination used MCNP (Monte Carlo N-Particle) simulator to model the BNCT neutron source and the shield. The shield should reduce radiation to less than the dose limit of 10.42 µSv/h, which is assumed to be the most conservative limit when the duration of workers is 1920 h. The first design resulted in a radiation dose rate which was still greater than the limit. Therefore, optimization was done by adding the lead on the outer part of the shield. After optimization by adding the lead with certain layers, the radiation dose rate decreased, with the largest dose being 57.60 µSv/h. Some locations over the limit could be overcome by other radiation protection aspects such as distance and time. The paraffin blocks were covered by aluminium to keep the shield structure. The lead was used to absorb the gamma ray which resulted from the interaction between the neutrons and aluminium.

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Monte Carlo optimization simulation of neutron shielding performance of iron/polyethylene combined structure

MATEC Web of Conferences ◽

10.1051/matecconf/201818902001 ◽

2018 ◽

Vol 189 ◽

pp. 02001

Author(s):

Yinghong Zuo ◽

Jinhui Zhu ◽

Shengli Niu ◽

Honggang Xie ◽

Peng Shang

Keyword(s):

Monte Carlo ◽

Neutron Source ◽

Thickness Ratio ◽

Optimal Thickness ◽

Neutron Shielding ◽

Monte Carlo Optimization ◽

Optimization Simulation ◽

Combined Structure ◽

Shielding Design ◽

Different Thickness

This study aims to get the optimization neutron shielding design of iron/polyethylene combined shield structure. The neutron transmission coefficient with various energies for different thickness of iron and polyethylene combined shield structure were calculated by using Monte Carlo method. The simulation results show that the optimization effect of iron/polyethylene combined shield is not obvious when the neutron energy is low or the shield is thin, there is an optimal thickness ratio of iron to polyethylene adopted to get the best neutron shielding performance when the energy of neutron source is above 2 MeV and the total thickness of combined shielding structure is more than 20 cm. The optimal thickness ratio of iron to polyethylene increases with the increasing energy of neutron source; with the increasing of neutron source energy ranging from 4 MeV to 14 MeV, the optimal thickness ratio of iron to polyethylene trends from 0.11 to nearly 1.6.

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Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

10.1063/1.1945318 ◽

2005 ◽

Author(s):

Olena Gritzay

Keyword(s):

Monte Carlo ◽

Neutron Source ◽

Research Reactor ◽

Monte Carlo Calculation

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Feasibility Study Using GEANT4 Monte Carlo Simulations of a PET Cyclotron-Based Neutron Source for Boron Neutron Capture Therapy

New Physics Sae Mulli ◽

10.3938/npsm.67.1319 ◽

2017 ◽

Vol 67 (11) ◽

pp. 1319-1326

Author(s):

Jae Won CHOI ◽

Bo Sun KANG*

Keyword(s):

Monte Carlo ◽

Monte Carlo Simulations ◽

Neutron Source ◽

Feasibility Study ◽

Boron Neutron Capture Therapy ◽

Neutron Capture ◽

Neutron Capture Therapy ◽

Boron Neutron Capture

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Acceleration of fission source convergence in the Serpent 2 Monte Carlo code using a response matrix based solution for the initial source distribution

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.12.044 ◽

2019 ◽

Vol 128 ◽

pp. 63-68 ◽

Cited By ~ 4

Author(s):

Jaakko Leppänen

Keyword(s):

Monte Carlo ◽

Source Distribution ◽

Monte Carlo Code ◽

Response Matrix ◽

Source Convergence ◽

Initial Source

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Determination of an Effective Detector Position for Pulsed-Neutron-Source Alpha Measurement by Time-Dependent Monte Carlo Neutron Transport Simulations

Science and Technology of Nuclear Installations ◽

10.1155/2018/2350458 ◽

2018 ◽

Vol 2018 ◽

pp. 1-7 ◽

Cited By ~ 1

Author(s):

Sang Hoon Jang ◽

Hyung Jin Shim

Keyword(s):

Monte Carlo ◽

Neutron Source ◽

Neutron Transport ◽

Signal Amplitude ◽

Time Dependent ◽

Convergence Time ◽

Simple Method ◽

Detector Position ◽

Pulsed Neutron ◽

Pulsed Neutron Source

A simple method using the time-dependent Monte Carlo (TDMC) neutron transport calculation is presented to determine an effective detector position for the prompt neutron decay constant (α) measurement through the pulsed-neutron-source (PNS) experiment. In the proposed method, the optimum detector position is searched by comparing amplitudes of detector signals at different positions when their α estimates by the slope fitting are converged. The developed method is applied to the Pb-Bi-zoned ADS experimental benchmark at Kyoto University Critical Assembly. The α convergence time estimated by the TDMC PNS simulation agrees well with the experimental results. The α convergence time map and the corresponding signal amplitude map predicted by the developed method show that polyethylene moderator regions adjacent to fuel region are better positions than other candidates for the PNS α measurement.

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Detection of Combined n/γ Fission Signatures Induced by an Epithermal Neutron Source

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4036698 ◽

2017 ◽

Vol 3 (3) ◽

Author(s):

A. Ocherashvili ◽

T. Bogucarska ◽

A. Beck ◽

G. Heger ◽

M. Mosconi ◽

...

Keyword(s):

Neutron Source ◽

Proper Time ◽

Pulse Shape Discrimination ◽

Fast Neutrons ◽

Nuclear Materials ◽

Test Assembly ◽

Fission Neutrons ◽

Pulsed Neutron ◽

Prompt Fission ◽

Multiple Detectors

In this paper, a method is presented for the detection of special nuclear materials (SNMs) in shielded containers, which is both sensitive and applicable under field conditions. The method uses an external pulsed neutron source to induce fission in SNM and subsequent detection of the fast prompt fission neutrons. The detectors surrounding the container under investigation are liquid scintillation detectors able to distinguish gamma rays from fast neutrons by means of pulse shape discrimination method (PSD). One advantage of these detectors, besides the ability for PSD analysis, is that the analog signal from a detection event is of very short duration (typically few tens of nanoseconds). This allows the use of very short coincidence gates for the detection of the prompt fission neutrons in multiple detectors, while benefiting from a low background coincidence rate, yielding a low detection limit. Another principle advantage of this method derives from the fact that the external neutron source is pulsed. By proper time gating, the interrogation can be conducted by epithermal source neutrons only. These neutrons do not appear in the fast neutron signal following the PSD analysis, thus providing a fundamental method for separating the interrogating source neutrons from the sample response in the form of fast fission neutrons. This paper describes laboratory tests with a configuration of eight detectors in the Pulsed Neutron Interrogation Test Assembly (PUNITA). Both the photon and neutron signature for induced fission is observed, and the methods used to isolate these signatures are described and demonstrated.

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