Development of discrete ordinates SN code in three-dimensional (X,Y,Z) geometry for shielding design.

Nishimura, Tatsuo; Tada, Keiko; Yokobori, Hitoshi; Sugawara, Akira

doi:10.3327/jnst.17.539

Cited by 5 publications

(4 citation statements)

References 13 publications

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“…The ENSEMBLE-XYZ calculation< 18 ), which was carried out for demonstration purpose, agrees well with the experiment at the entrance, but underestimates it extremely at the exit. This obvious discrepancy of attenuation factor between the calculation and the experiment is also due to the use of the rough angular quadrature set S 4 .…”

Section: -----------------supporting

confidence: 57%

Development of radiation transport code in three-dimensional (X, Y, Z) geometry for shielding analyses by direct integration method.

Ida¹,

Kondo²,

Togo³

et al. 1987

J. Nucl. Sci. Technol.

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A three-dimensional radiation transport code TRISTAN has been developed by applying the method of direct integration to the group-flux calculation in order to enhance the accuracy of shielding analysis. It uses group-angle transfer matrices derived from the double-differential cross section data instead of the Legendre polynomial expansion.In order to improve the numerical accuracy, a technique to separate radiation into two components (the scattered and the unscattered) is adopted and the balance equation of radiation flux is applied to the solution method. Two new techniques, a stratification of the angular mesh and a kind of the forward-adjoint coupling, are introduced to enhance its applicability to the practical shielding analysis.The validity of the TRISTAN is verified by making comparison with the results of the Monte Carlo code MCNP, and with the experimental values in the JRR-4 streaming experiments.

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Section: -----------------supporting

confidence: 57%

Development of radiation transport code in three-dimensional (X, Y, Z) geometry for shielding analyses by direct integration method.

Ida¹,

Kondo²,

Togo³

et al. 1987

J. Nucl. Sci. Technol.

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“…The thickness, t , along the x ‐direction of each slab can be chosen to be far smaller than the mean free path of the neutrons inside the related material. Thus, neutrons penetrating the slab consist of two main components: the unscattered neutrons and neutrons undergoing only one collision 10,11 . The flux of the penetrating neutrons can be expressed as Equation (): 7–9

\begin{matrix} true \\ right ϕ false(t, E, μ false) & = & left ϕ_{0} (E, μ) e^{- normalΣ (E) \frac{t}{μ}} \\ left + \frac{1}{μ} \int d E^{'} \int d Ω^{'} σ (Ω, {normalΩ}^{'}; E, E^{'}) ϕ_{0} (E^{'}, μ^{'}) \\ left \times e^{- normalΣ (E) \frac{t}{μ}} \int_{0}^{t} d x^{'} e^{- (\frac{Σ (E^{'})}{μ^{'}} - \frac{Σ (E)}{μ}) x^{'}} \end{matrix}

where

ϕ_{0}

( E ,

μ

) is the flux of incident neutrons with energy of E and direction of

μ

, which stands for the cosine value of the angle between the moving direction of the transmitted neutrons and the x ‐direction;

ϕ

( t , E ,

μ

) is for the neutrons penetrating a slab with a thickness of t ; x is the depth of the collision point along the x ‐direction inside the slab;

μ

is the cosine of the angle between the scattered neutron direction, and the x ‐axis after each collision;

normalΣ

( E ) and

normalΣ

( E ) are the total macroscopic cross sections for neutrons with energies of E and E , respectively; and

σ

(

…”

Section: Methodsmentioning

confidence: 99%

“…Thus, neutrons penetrating the slab consist of two main components: the unscattered neutrons and neutrons undergoing only one collision. 10,11 The flux of the penetrating neutrons can be expressed as Equation (1): [7][8][9] 𝜙(t, E, 𝜇) = 𝜙 0 (E, 𝜇)e −Σ(E)…”

Section: Substitution Matrices For Boltzmann Transport Equationmentioning

confidence: 99%

An iterative prediction method for designing the moderator used for the boron neutron capture therapy

Zhang

et al. 2021

Medical Physics

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To conduct research related to slow neutrons, fast neutrons must be mode-rated and shifted to the desired energy region. Methods: In this research, an iterated prediction method, in which the neutron transportation properties of all materials were characterized by a reflection matrix, R, and a transmission matrix, T, was proposed to bypass a time-consuming Monte Carlo simulation and predict the performance of the moderator, including the epithermal neutron flux and the dose of fast neutrons and gamma rays, used for boron neutron capture therapy (BNCT). To find the optimal solution in the huge parameter space, a genetic algorithm combined with transmission and reflection matrices was utilized. Results:The results showed that a 70-loop iteration was able to find a design for the moderator of BNCT with almost 80% higher epithermal neutron flux per kilowatt than that of the empirically optimized moderator that was previously reported in the literature. Compared with the Monte Carlo method, this method had the advantage of reducing the calculation time and statistical errors. Conclusion:The genetic algorithm with matrices (GAM) method can be used to find an optimal solution in a huge parameter space without brute-force calculations. It could be a promising method for designing the moderator for thermal or epithermal neutron usages.

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“…As for 3D Sn codes, the LANL THREETRAN [56], Japan's EMSAMBLE [57], and the ORNL TORT codes were developed in the 1980s. After undergoing various improvements, the TORT code became available for practical use in radiation transport calculation.…”

Section: Progress In Sn Transport Codesmentioning

confidence: 99%

Progress and prospects of calculation methods for radiation shielding

Hirayama

Nakashima

Morishima

et al. 2015

Journal of Nuclear Science and Technology

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Progress in calculation methods for radiation shielding are reviewed based on the activities of research committees related to radiation shielding fields established in the Atomic Energy Society of Japan. A technological roadmap for the field of radiation shielding; progress and prospects for specific shielding calculation methods such as the Monte Carlo, discrete ordinate Sn transport, and simplified methods; and shielding experiments used to validate calculation methods are presented in this paper.

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Development of discrete ordinates SN code in three-dimensional (X,Y,Z) geometry for shielding design.

Cited by 5 publications

References 13 publications

Development of radiation transport code in three-dimensional (X, Y, Z) geometry for shielding analyses by direct integration method.

Development of radiation transport code in three-dimensional (X, Y, Z) geometry for shielding analyses by direct integration method.

An iterative prediction method for designing the moderator used for the boron neutron capture therapy

Progress and prospects of calculation methods for radiation shielding

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