• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1075-1080
• /
• 2019

This paper presents a study on thermal neutron reflection properties of neutron guide for cylinder, spindle, elliptic and parabolic geometries using $^{241}Am-Be$ neutron source (5.2 Ci) and $BF_3$ detector, whereas neutron guide is important instrument for transportation of neutrons. To this goal, the required inner and outer radii of neutron guide have been calculated to achieve the highest guided thermal neutron flux based on MCNPX Monte Carlo code. The maximum flux of cylinder geometry with a length 50 cm has been obtained at an inner radius 9 cm and an outer radius 21 cm. Also, the maximum value of thermal neutron albedo is $0.46{\pm}0.001$ at 12 cm thickness of parabolic guide.

• Nuclear Engineering and Technology
• /
• v.36 no.3
• /
• pp.203-209
• /
• 2004

In-phantom neutron flux distribution is measured at the HANARO BNCT irradiation facility. The measurements are performed with Au foil and wires. The thermal neutron flux and Cd ratio obtained at the HANARO BNCT facility are $1.19{\times}10^9\;n/cm^{2}s$ and 152, respectively, at 24 MW reactor power. The measured in-phantom neutron flux has a maximum value at a depth of 3 mm in the phantom and then decreases rapidly. The maximum flux is about $25\%$ larger than that of the phantom surface, and the measured value at a depth of 22 mm in the phantom is about a half of the maximum value. In addition, the neutron beam is limited well within the aperture of the neutron collimator. The two-dimensional in-phantom neutron flux distribution is determined. Significant neutron irradiation is observed within 20 mm from the phantom surface. The measured neutron flux distribution can be utilized in irradiation planning for a patient.

• Bulletin of the Korean Chemical Society
• /
• v.27 no.12
• /
• pp.2061-2063
• /
• 2006

• Nuclear Engineering and Technology
• /
• v.49 no.4
• /
• pp.787-791
• /
• 2017

Americium-beryllium (Am-Be; n, ${\gamma}$) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4226-4230
• /
• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1425-1430
• /
• 2024

This study develops superheated emulsion detectors that are both sensitive to fast neutrons, and thermal neutrons owing to the exergonic ⁶₃Li(n, α)³₁H capture reaction caused by the ⁶Li-containing compound dispersed throughout the gel-like medium. The experimental research was conducted on two SEDs. One detector was an ordinary Freon-12 detector and the other was a Freon-12 detector containing 3.4 % (by weight) LiCl. In order to investigate the sensitivity of lithium-containing SEDs to thermal neutrons, two types of SEDs were simultaneously exposed to various flux levels of thermal neutrons from ²⁴¹Am-Be neutron source inside a cylindrical tank filled with water. A Boron-lined proportional counter was used to estimate the thermal neutron flux and the relevant MCNP code was developed for flux and dose calculations in the prepared set-up around ²⁴¹Am-Be source. The results demonstrate that there is a proportional relationship between the variations of SED response and the change in thermal neutron flux and dose. Also, the sensitivity of SED was estimated.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.575-600
• /
• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.499-507
• /
• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• Journal of Radiation Protection and Research
• /
• v.19 no.1
• /
• pp.13-22
• /
• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• Journal of the Korean Institute of Telematics and Electronics
• /
• v.9 no.5
• /
• pp.1-5
• /
• 1972

In order to improve the response time of nelltron theromopile for reactor control a neutron thermopile made use of a vacuunl evaporated thin film thor mocouple was fablicated and tested. The test results were compared with a wire-type neutron thermopile. Good linearities between the response of the neutron thermopile and the thermal flux has been shown in the ranges from n/$\textrm{cm}^2$/sec. Thermal neutron flux distributions in the core of TRIGA Mark-II reactor were measured using the fabricated neutron thermopile, and the results were conpared with data obtained by the acrivatin foil measurement.