• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1998-2004
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• 2019

The Multi-Physics (MP) instrument is one of 20 neutron spectrometers planned in the China Spallation Neutron Source (CSNS). This paper presents a shielding calculation for the MP instrument using Monte Carlo codes MCNPX and FLUKA. First, the neutrons that escape from the CSNS decoupled water moderator and are delivered to the beam line of the MP instrument are calculated to use as the source term of the shielding calculation. Then, to validate the calculation method based on multiple variance reduction techniques, a cross check between MCNPX and FLUKA codes is performed by comparing the calculation results of the dose rate distribution on a simplified beam line model. Finally, a complete geometry model of the MP instrument is set up, and the primary parameters for the shielding design are obtained according to the calculated dose rate map considering different worst-case scenarios.

• 한국소음진동공학회논문집
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• 제26권5호
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• pp.501-510
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• 2016

Capabilities of several beamforming techniques are compared for estimating the position of a moving source. Beamforming has enabled to widen our perspective of aeroacoustics in wind tunnel experiments and has provided useful approach in array measurements. Meanwhile beamforming techniques have been developed in a way to improve estimation accuracy and to save ing effort at the same time. In order to achieve reasonable outcome from aeroacoustic measurement, it is important to identify the spectral characteristics of source and to select an appropriate beamformer. Though aeroacoustic sources normally generates broadband noises, many array signal processing have been focused on narrowband processing which makes calculation numerically efficient. However, calculation in frequency-domain requires selection of single frequency of interest which affects spatial resolution and sidelobe level as a consequence. To be able to localize broadband noise source, it is proposed to use broadband beamforming. The formulas implements the deletion of diagonal term from cross spectral matrix. In this study, trajectory of flying source emitting broadband noise was simulated and several beamformers are applied.

• Journal of Radiation Protection and Research
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• 제34권3호
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.179-184
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• 1996

Power density distribution and criticality of a pressurized water reactor are calculated with a Monte Carlo calculation using the MCNP code. The MCNP model is based on one-eighth core symmetry. Individual fuel assemblies are modeled with fullscope three dimensional description except grid spacer. The fuel rod is divided into eight axial segments. Core internals above and below the active fuel region is represented as coolant. After 400 cycle calculations, the system converges to a k value of 1.09151$\pm$0.00066. Fission reaction rate in each rod is also calculated to use as the source term in pressure vessel fluence calculation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.852-853
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• 2005

• 대한조선학회논문집
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• 제29권3호
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• pp.71-79
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• 1992

선체 표면상에 Havelock 쏘오스를 분포시켜 선체 표면상의 경계조건을 만족시키는 N-K 이론을 사용하여 선체에 작용하는 조파저항을 구하였다. 수치계산시 Havelock 쏘오스, 혹은 Green 함수는 Noblesse(1977)가 제시한 형태를 사용하였고, 국부교란항은 Newman(1987), 파도교란항은 Baar & Price(1988)를 따라 각각 수행하였다. 선체표면에 대한 수치적분은 Gauss 구적법을 사용하여 수행하였고, 쏘오스의 세기는 겹선형함수로 선체표면에 걸쳐 연속이라고 가정하였다. 또 조파저항계산은 원장에서의 자유표면을 나타내는 식을 사용하여 de Sendagorta & Grases(1988)의 방법에 따라 구하였다. Wigley선형에 대한 계산을 수행하여 선적분항에 미치는 영향을 고찰하였고, 계산치를 기존의 실험치와 비교한 결과 잘 일치하고 있음을 확인하였다.

• Journal of Radiation Protection and Research
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• 제39권4호
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• pp.176-181
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• 2014

The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.

• Journal of Mechanical Science and Technology
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• 제15권3호
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• pp.392-402
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• 2001

The paper reports a multiple source modeling of low-Reynolds-number dissipation rate equation with aids of DNS data. The key features of the model are to satisfy the wall limiting conditions of the individual source terms in the exact dissipation rate equation using the wall damping functions. The wall damping functions are formulated in term of dimensionless dissipation length scale ι(sup)+(sub)D(≡ι(sub)D($\upsilon$$\xi$)(sup)1/4/$\upsilon$) and the invariants of small and large scale turbulence anisotropy tensors. $\alpha$(sub)ij(=$\mu$(sub)i$\mu$(sub)j/$\kappa$-2$\delta$(sub)ij/3) and e(sub)ij(=$\xi$(sub)ij/$\xi$-2$\delta$(sub)ij/3). The model constants are optimized with aids of DNS data in a plane channel flow. Adopting the dissipation length scale as a parameter of damping function, the applicabilities of $\kappa$-$\xi$ model are extended to the turbulent flow calculation of complex flow passages.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.378-383
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• 2016

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.