Benchmark Tests of Radiation Transport Computer Codes for Reactor Core and Shield Calculations

Asaoka, T.; ASANO, Norio; Nakamura, Hisashi; Mizuta, Hiroshi; Chichiwa, Hiroshi; Miyasaka, Shun-ichi; Zukeran, Atsushi; Tsutsui, Tsuneo; Fujimura, Toichiro; Katsuragi, S.

doi:10.1080/18811248.1978.9733200

Cited by 4 publications

(2 citation statements)

References 4 publications

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“…However, the differences in the properties and the types of geometries typically make these benchmark solutions less applicable for comparing the performance of various angular discretization techniques. Many of these benchmarks [83,84] are relevant to shielding applications and tend to feature regions with strong scattering as well as void regions which can look very different than the opacity distributions typical of heat transfer applications. Other examples [85] include the transient term which is often neglected in heat transfer applications but which may be important for microscale systems [86] or optical tomography [87].…”

Section: Comparisons Among Different Angular Discretization Methodsmentioning

confidence: 99%

Coupling radiative heat transfer in participating media with other heat transfer modes

Tencer

Howell

2015

J Braz. Soc. Mech. Sci. Eng.

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transfer modes are minimized or eliminated as in Dewar flasks and cryogenic storage systems. In this review, we discuss the common and not-so-common methods for treating radiation heat transfer in participating (absorbing/emitting and scattering) media, and how these methods are coupled with the overall energy equation for treating thermal transfer problems.Because radiation is almost inevitably coupled with other heat transfer modes, an overall thermal analysis requires solution of the energy equation, which is actually a balance of energy rates. Here, we follow the development in Howell et al. [1].The general form for the energy equation for a compressible fluid is where D/Dt is the substantial derivative, the left-hand side accounts for changes in stored energy within the medium, and the terms on the right-hand side account for changes in pressure work, contributions by conduction and radiation, internal sources (chemical, electrical, nuclear, etc.,), and viscous dissipation. In most practical thermal analysis problems, the pressure work and viscous dissipation terms can be neglected.Of greatest interest here is the evaluation of the local divergence of radiative flux, −∇ · q r . This is the difference in the local absorbed radiative energy minus the local emitted radiation, 4π ∞ =0 κ I b d , where κ λ is the local spectral absorptivity of the medium and I λb is the blackbody intensity given by the Planck distribution. In an absorbing-emitting-scattering medium, the absorbed radiative energy is found by first finding the local radiative intensity I λ (Ω), which is defined as the spectral energy propagating in a given direction Ω per unit solid angle, per unit area AbstractThe common methods for finding the local radiative flux divergence in participating media through solution of the radiative transfer equation are outlined. The pros and cons of each method are discussed in terms of their speed, ability to handle spectral properties and scattering phenomena, as well as their accuracy in different ranges of media transport properties. The suitability of each method for inclusion in the energy equation to efficiently solve multi-mode thermal transfer problems is discussed. Finally, remaining topics needing research are outlined.

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Section: Comparisons Among Different Angular Discretization Methodsmentioning

confidence: 99%

Coupling radiative heat transfer in participating media with other heat transfer modes

Tencer

Howell

2015

J Braz. Soc. Mech. Sci. Eng.

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“…In this section, we evaluate the behaviour of our solver in controlled settings using several classic testing cases which serve as benchmarks for other similar codes that use deterministic methods to obtain the solution. We have selected the tests of Reed [16,22], Alcouffe [16], a deep penetration analysis with two energy groups [23] and one case of anisotropic scattering with three energy groups in a series of consecutive materials [24,25].…”

Section: Classic Literature Testsmentioning

confidence: 99%

Validating NEUTRO, a deterministic finite element neutron transport solver for fusion applications, with literature tests, experimental benchmarks and other neutronic codes

Goldberg¹,

Duxans²,

Gelabert³

et al. 2022

Plasma Phys. Control. Fusion

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Neutron reactions on fusion reactor materials are key phenomena to be understood to enable fusion as a feasible energy source for future reactors. We present significant improvements and validations of NEUTRO, a deterministic neutron transport code dedicated to solving the Boltzmann Transport Equation. The code is integrated as a module in the Alya system developed by the Barcelona Supercomputing Center and uses the Discrete Ordinates Method over an angular portion, multigroup for energy discretization and the Finite Element Method over unstructured meshes to treat special complex domains. Material anisotropy of the scattering medium is introduced into the scattering kernel using real base expressions for spherical harmonics. In order to build the total cross-section and the respective group matrix for the elastic cross-section, we use the NJOY code. We test the solver using different geometries and materials with varying levels of scattering properties. We compare our results on classic tests, benchmarks obtained from a Nuclear Energy Agency database and test cases using other neutron transport codes.

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Considerations of core storage arising from the finite element analysis of radiation diffusion‐transport benchmark problems

Ziver

Goddard

Ackroyd

1982

Numerical Meth Engineering

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SUMMARYProblems involving the diffusion and transport (angular dependence considered) of neutron radiation are frequently encountered in radiation physics and nuclear engineering. Neutron diffusion problems require substantial computer storage arising from the energy and spatial dependence, while in the case of transport problems the angular dependence of the neutron flux and of the neutron scattering process gives rise to a further considerable increase in storage. In the present work, a method has been developed which greatly eases the bandwidth problem in the solution of large systems of linear equations and algorithms have been developed to assemble and solve such systems of equations in one and two dimensions. The formalism used is the variational method and, for simplicity, linear interpolating functions have been used to derive a symmetric banded system of equations. Finite element algorithms have been implemented in the codes FEED1 and FEEDZ, which treat one and two space dimensions, respectively. Three benchmark problems are analysed in this paper and results are compared with finite difference discrete ordinate method solutions. It is shown that the finite element method provides fast and accurate solutions to neutron diffusion and transport problems.

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Benchmark Tests of Radiation Transport Computer Codes for Reactor Core and Shield Calculations

Cited by 4 publications

References 4 publications

Coupling radiative heat transfer in participating media with other heat transfer modes

Coupling radiative heat transfer in participating media with other heat transfer modes

Validating NEUTRO, a deterministic finite element neutron transport solver for fusion applications, with literature tests, experimental benchmarks and other neutronic codes

Considerations of core storage arising from the finite element analysis of radiation diffusion‐transport benchmark problems

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