• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.319-328
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• 1992

The Galerkin formulation of the finite element method is applied to the integral law of the first-order form of the one-group neutron transport equation in one-dimensional spherical geometry. Piecewise linear or quadratic Lagrange polynomials are utilized in the integral law for the angular flux to establish a set of linear algebraic equations. Numerical analyses are performed for the scalar flux distribution in a heterogeneous sphere as well as for the criticality problem in a uniform sphere. For the criticality problems in the uniform sphere, the results of the finite element method, with the use of continuous finite elements in space and angle, are compared with the exact solutions. In the heterogeneous problem, the scalar flux distribution obtained by using discontinuous angular and spatical finite elements is in good agreement with that from the ANISN code calculation.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.289-298
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• 1997

Time-consuming experiments have been required in the development of neutron moisture gauge to induce a relation between the water content in soil and the neutron counts. Applying a monte carlo computer code to simulate the experiments of neutron moisture gauging may contribute to reduce time and efforts for experiments and produce a calibration equation which is more applicable to soil in general. In this study MCNP4A, a monte carlo computer code, was employed to simulate soil experiments and the simulated results were compared with experimental ones. The comparative study showed that MCNP4A is applicable to simulate the experiments and calibration equation can be obtained through simulations. Effects of dry density changes were also studied.

• Journal of the Korean Society of Radiology
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• v.16 no.5
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• pp.527-536
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• 2022

The radiation shielding characteristic of neutron shielding material has been studied as the preliminary study in order to design cosmic-ray shielding material. Specially, Soft Magnetic Material, known to be effective in EMP and radiation shielding, has been investigated to check if the material would be applicable to cosmic-ray shielding. In this work, thermal neutron shielding experiment was conducted and the Monte Carlo N-Particle(MCNP) was applied to employ skymap.dat, which is cosmic-ray data embedded in MCNP. As a result, polyethylene, borated polyethylene, and carbon nano tube, containing carbon or hydrogen, have been found to be effective in reduction of neutron flux below 20 MeV (including thermal, epithermal, evaporation). In contrast, the materials composed of iron such as SS316 and Soft Magnetic Material show a good shielding performance in the cascade energy range (above 20 MeV). Since Soft Magnetic Material is consisting of 13% of boron, it can also decrease thermal neutron flux, so it is expected that it would show a significant reduction on the entire range of neutron energy if the Soft Magnetic Material is used with hydrogen and carbon, so called low Z material.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.148-154
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• 1985

Radiation shield design for a shipping cask, KSC-1, was evaluated to verify that the cask can be used in the transportation of a spent fuel assembly discharged from KNU 5 & 6. Radiation source term of the spent fuel assembly was calculated with the computer program ORIGEN-79, QAD-CG, ANISN-KA and DOT 3.5 codes Were used in the shielding calculations and the nuclear cross section data needed was extracted from the DLC-23/CASK library. It is concluded that KSC-1 shipping cask satisfies the requirements specified in the relevant regulations under normal conditions of transport and under accident conditions in transport.

• Journal of Energy Engineering
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• v.16 no.1 s.49
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• pp.46-52
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• 2007

The piecewise constant angular approximation is developed to replace the conventional angular quadrature sets in the solution of the second-order, multi-dimensional $S_{N}$ neutron transport equations. The newly generated quadrature sets by this method substantially mitigate ray effects and can be used in the same manner as the conventional quadrature sets are used. The discrete-ordinates and the piecewise-constant approximations are applied to both the first-order Boltzmann and the second-order form of neutron transport equations in treating angular variables. The result is that the mitigation of ray effects is only achieved by the piecewise-constant method, in which new angular quadratures are generated by integrating angle variables over the specified region. In other sense, the newly generated angular quadratures turn out to decrease the contribution of mixed-derivative terms in the even-parity equation that is one of the second-order neutron transport equation. This result can be interpreted as the entire elimination or substantial mitigation of ray effect are possible in the simplified even-parity equation which has no mixed-derivative terms.

• Nuclear Engineering and Technology
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• v.3 no.3
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• pp.121-128
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• 1971

A unified modal-nodal expansion of tile angular distribution of neutron flux in one spatial dimension is considered, following the proposal of Harms. Several standard nodal and/or modal methods of analysis are shown to be specializations of this technique. The modal-nodal moment from of the mono-energetic transport equation with isotropic sources and scattering is derived and the infinite medium eigenvalue problem solved. The technique is shown to yield results which approximate the exact value of the inverse diffusion length in non-multiplying media more accurately than standard methods of equal or somewhat greater computational complexity.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

• Journal of Radiation Protection and Research
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• v.4 no.1
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• pp.1-4
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• 1979

Effective multiplication factor has been calculated for one PWR fresh fuel assembly immersed in a spent fuel storage vault on the basis of the neutron transport theory. A numerical calculation has been carried out by means of Sn approximation. The method employed in this study is that the energy domain is broken into 16 groups, the angular variable is divided into four discrete direction, i.e., $S_4$, and the spatial variable which is divided into fine meshes at the interface between different materials is discretized into 27 mesh points. The calculated $K_{eff}$ value of 0.6145 seems to be far small in comparison with the value obtained by other author for an infinite array of fuel assemblies.

• Journal of Radiation Protection and Research
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• v.38 no.1
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• pp.37-43
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• 2013

In case of neutron dose measurement using TLDs (thermo-luminescence dosimeters), because the neutron energy dependence of the TLD is very high, the calibration of the energy response according to the characteristics of the neutron spectrum of workplace is required. In the present study, the ambient dose equivalent rates inside and around the Long-Counter (neutron detector) with narrow and complex inside in the neutron field of $^{252}Cf$ were evaluated. The calibration factors to account for the neutron energy dependence of TLDs were established for both the bare and $D_2O$ modulated $^{252}Cf$ neutron beams, respectively. The values of the TLD's measurement were compared with the computational results of the MCNPX (Monte Carlo N-Particles transport code). When using the two calibration factors of the TLD than a single calibration factor, the measured and the calculated values at the point of verification outside and inside the Long-Counter were in more good agreement. This results show that TLD should be calibrated in the reference neutron field similar to workplace situation.

• Journal of Radiation Protection and Research
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• v.8 no.1
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• pp.33-37
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• 1983

Neutron reflecting effects in terms of effective multiplication factor have been calculated with varying water or concrete thickness, and gap distance between concrete reflector and a fissile solution system. A numerical calculation of effective multiplication factors has been carried out by using the discrete ordinates method with the help of the computer code, ANISN. It is revealed that the reflecting .effect by thin concrete is lower than that of the identical thickness of water while the effect by thick water is low compared to that of the identical thickness of concrete. It seems that the effective multiplication factors are first decreasing rapidly with gap distance, which is filled with water, between concrete reflector and the fissile solution system, and then decrease slowly over the distance of about 15cm.