• Title/Summary/Keyword: radiation transport

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Hot Gas Analysis of Circuit Breakers By Combining Partial Characteristic Method with Net Emission Coefficient

  • Park, Sang-Hun;Bae, Chae-Yoon;Jung, Hyun-Kyo
    • KIEE International Transaction on Electrical Machinery and Energy Conversion Systems
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    • v.3B no.3
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    • pp.115-121
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    • 2003
  • This paper proposes a radiation model, which considers radiation transport as an important component in hot gas analysis. This radiation model is derived from combining the method of partial characteristics (MPC) with net emission coefficient (NEC), and it covers the drawbacks of existing models. Subsequently, using this proposed model, the arc-flow interaction in an arcing chamber can be efficiently computed. The arc is represented as an energy source term composed of ohmic heating and the radiation transport in the energy conservation equation. Ohmic heating term was computed by the electric field analysis within the conducting plasma region. Radiation transport was calculated by the proposed radiation model. Also, in this paper, radiation models were introduced and applied to the gas circuit breaker (GCB) model. Through simulation results, the efficiency of the proposed model was confirmed.

Measurements and Assessments on Shielding Performance of FCTC10 60Co Transport Container

  • Zhuang, Dajie;Zhang, Guoqing;Li, Guoqiang;Wang, Renze
    • Journal of Radiation Protection and Research
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    • v.41 no.3
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    • pp.310-314
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    • 2016
  • Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

A PRACTICAL LOOK AT MONTE CARLO VARIANCE REDUCTION METHODS IN RADIATION SHIELDING

  • Olsher Richard H.
    • Nuclear Engineering and Technology
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    • v.38 no.3
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    • pp.225-230
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    • 2006
  • With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of variance reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered in the areas of source definition, skyshine, streaming, and transmission.

RADIATION SAFETY ASSESSMENT FOR KN-12 SPENT NUCLEAR FUEL TRANSPORT CASK USING MONTE CARLO SIMULATION

  • Kim, J.K.;Kim, G.H.;Shin, C.H.;Choi, H.S.
    • Journal of Radiation Protection and Research
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    • v.26 no.3
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    • pp.207-214
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    • 2001
  • The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

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Multilevel acceleration of scattering-source iterations with application to electron transport

  • Drumm, Clif;Fan, Wesley
    • Nuclear Engineering and Technology
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    • v.49 no.6
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    • pp.1114-1124
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    • 2017
  • Acceleration/preconditioning strategies available in the SCEPTRE radiation transport code are described. A flexible transport synthetic acceleration (TSA) algorithm that uses a low-order discrete-ordinates ($S_N$) or spherical-harmonics ($P_N$) solve to accelerate convergence of a high-order $S_N$ source-iteration (SI) solve is described. Convergence of the low-order solves can be further accelerated by applying off-the-shelf incomplete-factorization or algebraic-multigrid methods. Also available is an algorithm that uses a generalized minimum residual (GMRES) iterative method rather than SI for convergence, using a parallel sweep-based solver to build up a Krylov subspace. TSA has been applied as a preconditioner to accelerate the convergence of the GMRES iterations. The methods are applied to several problems involving electron transport and problems with artificial cross sections with large scattering ratios. These methods were compared and evaluated by considering material discontinuities and scattering anisotropy. Observed accelerations obtained are highly problem dependent, but speedup factors around 10 have been observed in typical applications.

Radiation Detection System for Prevention of Illicit Trafficking of Nuclear and Radioactive Materials

  • Kwak, Sung-Woo;Chang, Sung-Soon;Yoo, Ho-Sik
    • Journal of Radiation Protection and Research
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    • v.35 no.4
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    • pp.167-171
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    • 2010
  • Fixed radiation portal monitors (RPMs) deployed at border, seaport, airport and key traffic checkpoints have played an important role in preventing the illicit trafficking and transport of nuclear and radioactive materials. However, the RPM is usually large and heavy and can't easily be moved to different locations. These reasons motivate us to develop a mobile radiation detection system. The objective of this paper is to report our experience on developing the mobile radiation detection system for search and detection of nuclear and radioactive materials during road transport. Field tests to characterize the developed detection system were performed at various speeds and distances between the radioactive isotope (RI) transporting car and the measurement car. Results of measurements and detection limits of our system are described in this paper. The mobile radiation detection system developed should contribute to defending public's health and safety and the environment against nuclear and radiological terrorism by detecting nuclear or radioactive material hidden illegally in a vehicle.

Optimization of radiation shields made of Fe and Pb for the spent nuclear fuel transport casks

  • V.G. Rudychev;N.A. Azarenkov;I.O. Girka;Y.V. Rudychev
    • Nuclear Engineering and Technology
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    • v.55 no.2
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    • pp.690-695
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    • 2023
  • Recommendations are given to improve the efficiency of radiation protection of transport casks for SNF transportation. The attenuation of ${\gamma}$-quanta of long-lived isotopes 134Cs, 137mBa(137Cs), 154Eu and 60Co by optimizing the thicknesses and arrangement of layers of Fe and Pb radiation shields of transport casks is studied. The fixed radiation shielding mass (fixed mass thickness) is chosen as the main optimization criterion. The effect of the placement order of Fe and Pb layers in a combined two-layer radiation shield with an equivalent thickness of 30 cm is studied in detail. It is shown that with the same mass thicknesses of the Fe and Pb layers, the placement of Fe in the first layer, and Pb - in the second one provides more than twofold attenuation of ${\gamma}$-quanta compared to the reverse placement: Pb - in the first layer, Fe - in the second. The increase in the efficiency of attenuation of ${\gamma}$-quanta for TC with combined shielding of Fe and Pb is shown to be achieved by designing the first layer of radiation shielding around the canister with SNF from Fe of the maximum possible thickness.