• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.939-944
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• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1600-1609
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• 2017

The accelerator-driven subcritical system (ADS) is driven by a neutron source from spallation reactions introduced by the injected proton beam. Part of the neutron source has energy as high as a few hundred MeV to a few GeV. The effects of high-energy source neutrons ($E_n$ > 20 MeV) are usually approximated by energy cut-off treatment in practical core calculations, which can overestimate the predicted proton beam current in the ADS design. This article intends to quantize this effect and propose a way to solve this problem. To evaluate the effects of high-energy neutrons in the subcritical core, two models are established aiming to cover the features of current experimental facilities and industrial-scale ADS in the future. The results show that high-energy neutrons with $E_n$ > 20 MeV are of small fraction (2.6%) in the neutron source, but their contribution to the source efficiency is about 23% for the large scale ADS. Based on this, a neutron source efficiency correction factor is proposed. Tests show that the new correction method works well in the ADS calculation. This method can effectively improve the accuracy of the prediction of the proton beam current.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1723-1729
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• 2021

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H^*(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H^*(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.

• The Journal of the Korea Contents Association
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• v.15 no.1
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• pp.331-337
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• 2015

By comparing and evaluating the neutron dose when running cyclotron in fixed type measuring machine and portable type measuring machine, we performed this study to see if there is any difference between the doses in both measuring devices. We performed correction 3 times for the portable type during the experiment. We measured the values 12 times for each correction, so total 36 times by using the fixed type and portable type once every 2 weeks from June 2012 to February 2014. The statistics for the portable type and fixed type for each experiment showed 0.186, 0.511, and 0.057, which are not statistically significant. Therefore, we have come to the conclusion that unless there is deviation based on the correction period of the portable type, correction of fixed type wouldn't be necessary. Also, reduction of social costs may be achieved if we install an alarm meter, which is relatively low priced compared to the high cost neutron measuring machine, when detecting higher level than the interested standard level.

• Journal of the Mineralogical Society of Korea
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• v.16 no.1
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• pp.65-73
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• 2003

In this study we have derived the two correction methods of hydrogen bonding distance. In case of the intermediate or long hydrogen bond(＞2.5 $\AA$), hydrogen bonding distances can be corrected by using the function d(O-H)=exp((2.173-d(O…O))/0.138)+0.958 obtained by least- squares fit to the data from the neutron diffraction at low temperatures. The valence-least-squares method is effective for the distance correction of very short hydrogen bond(<2.5 $\AA$). The distance correction is necessary for the long intermolecular hydrogen bond obtained from X-ray diffraction analysis.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.477-484
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• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.29-33
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• 1997

Neutron room scattering corrections that should be made when neutron detectors are calibrated with a $D_2O$ moderated $^{252}Cf$ neutron source in the center of a calibration room are considered. Such room scattering corrections are dependent on specific neutron source type, detector type, calibration distance, and calibration room configuration. Room scattering corrections for the responses of a thermoluminescence dosimeter and two different types of spherical detectors to neutron source in the Radiation Calibration Laboratory(RCL) neutron calibration facility at the Korea Atomic Energy Research Institute(KAERI) were experimentally determined and are presented. The measured room scattering results are then compared with theoretical results calculated by predicting room scattering effects in terms of parameters related to the specific configuration. Agreement between measured and calculated scattering correction is generally about 10% for three kinds of detectors in the calibration facility.

• Journal of the Korean Society of Radiology
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• v.2 no.3
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• pp.33-35
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• 2008

In the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the present study, we obtained energy dependent neutron absorption ratios of natural rhodium in energy region from 0.003 to 100 eV by MCNP-4B Code. The coefficients for neutron absorption was calculated for several types of thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.299-304
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• 2008

A detection system was set up to measure the neutron generation rate of a recently developed D-D neutron generator. The system is composed of a Si detector, He-3 detector, and electronics for pulse height analysis. The neutron generation rate was measured by counting protons using the Si detector, and the data was crosschecked by counting neutrons with the He-3 detector. The efficiencies of the Si and He-3 detectors were calibrated independently by using a standard alpha particle source $^{241}Am$ and a bare isotopic neutron source $^{252}Cf$, respectively. The effect of the cross-sectional difference between the D(d,p)T and $D(d,n)^3He$ reactions was evaluated for the case of a thick target. The neutron generation rate was theoretically corrected for the anisotropic emission of protons and neutrons in the D-D reactions. The attenuations of neutron on the path to the He-3 detector by the target assembly and vacuum flange of the neutron generator were considered by the Monte Carlo method using the MCNP 4C2 code. As a result, the neutron generation rate based on the Si detector measurement was determined with a relative uncertainty of ${\pm}5%$, and the two rates measured by both detectors corroborated within 20%.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.197-201
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• 1996

In the detail reaction-rate measurements in a critical assembly using the foil activation method, the measured activations of detector foils have inevitably errors caused by detector foil self-shielding effect. If neutron flux could be approximated to Westcott flux: i.e. well thermalized Maxwellian distribution, these activations of detector foil could be corrected to represent the unperturbated flux at any detected position in the cell with using Westcott option and reaction-rate option of the lattice code, WIMS-AECL. These calculated detector material self-shielding correction factors of the tested fuel, CANFLEX provided much information about neutron spectrum of test lattice cell as well as the correction factors themselves. The results could be verified by another lattice calculations.