• Title, Summary, Keyword: thermal neutron flux distribution

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Measurements of In-phantom Neutron Flux Distribution at the HANARO BNCT Facility

  • Kim Myong Seop;Park Sang Jun;Jun Byung Jin
    • Nuclear Engineering and Technology
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    • v.36 no.3
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    • pp.203-209
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    • 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.

Conceptual design of a high neutron flux research reactor core with low enriched uranium fuel and low plutonium production

  • Rahimi, Ghasem;Nematollahi, MohammadReza;Hadad, Kamal;Rabiee, Ataollah
    • Nuclear Engineering and Technology
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    • v.52 no.3
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    • pp.499-507
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    • 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.

Monte Carlo Calculation of Thermal Neutron Flux Distribution for (n, v) Reaction in Calandria (몬테칼로 코드를 이용한 중수로 Calandria에서의 $(n,\;{\gamma})$ 반응유발 열중성자속분포 계산)

  • Kim, Soon-Young;Kim, Jong-Kyung;Kim, Kyo-Youn
    • Journal of Radiation Protection and Research
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    • v.19 no.1
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    • pp.13-22
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    • 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.

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Measurement of the fast Neutron Flux Density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II Reactor Using Solid State Track Detector

  • Ro, Seung-Gy;Jun, Jae-Shik;Cho, Sae-Hyung
    • Nuclear Engineering and Technology
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    • v.5 no.4
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    • pp.334-338
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    • 1973
  • The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

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Improving the Neutronic Characteristics of a Boiling Water Reactor by Using Uranium Zirconium Hydride Fuel Instead of Uranium Dioxide Fuel

  • Galahom, Ahmed Abdelghafar
    • Nuclear Engineering and Technology
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    • v.48 no.3
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    • pp.751-757
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    • 2016
  • The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

Evaluation of Neutron Flux Distributions of SMART-P IST Region for the Design of Ex-Core Detector (SMART 연구로 노외계측기 설계를 위한 IST 영역의 중성자속 분포 평가)

  • Koo, Bon-Seung;Kim, Kyo-Youn;Lee, Chung-Chan;Zee, Sung-Quun
    • Journal of Radiation Protection and Research
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    • v.30 no.2
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    • pp.55-60
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    • 2005
  • The evaluation of neutron flux distribution was performed for the ex-core detector design of SMART-P. DORT and MCNP code were used for the calculation of energy-dependent neutron flux distribution at 100% full power condition. Two code results show that maximum thermal flux appears at the $1^{st}$ water region in IST region and agree within 10% difference. In addition, another evaluation was performed code with assumptions that cote was composed of fission source and control rod without fuel assemblies. These assumptions make neutron count rate to be minimized. As a results, maximum thermal flux showed $6.99{\times}10^{-2}(n/cm^2-sec)$, when the strength of initial fission source was assumed as $1.0{\times}10^8(n/sec)$. The main reason of these results is due to the thermalization of fast neutrons in the water region and thermal flux is proportional to 80% of total neutron flux. Therefore, optimization of filler material of detector guide tube, position of installation and axial length of detector segments is necessary for the design of ex-core detector to enhance the neutron count rate and above results could be used in ex-core detector design as a fluence requirement.

SENSITIVITY ANALYSES OF THE USE OF DIFFERENT NEUTRON ABSORBERS ON THE MAIN SAFETY CORE PARAMETERS IN MTR TYPE RESEARCH REACTOR

  • Kamyab, Raheleh
    • Nuclear Engineering and Technology
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    • v.46 no.4
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    • pp.513-520
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    • 2014
  • In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

Study on Determination of Boron using the PGAA Facility at HANARO Research Reactor (하나로의 즉발감마선 방사화분석 장치를 이용한 붕소의 정량에 대한 연구)

  • Chung, Young-Sam;Cho, Hyun-Jae;Moon, Jong-Hwa;Kim, Sun-Ha;Kim, Young-Jin
    • Analytical Science and Technology
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    • v.16 no.5
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    • pp.391-398
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    • 2003
  • Basic research for the determination of boron content in biological sample has been carried out using the PGAA facility of the 24MW research reactor(HANARO). For investigation of characteristics for the measurement condition, neutron flux and its homogeneity were measured at irradiating geometry. The size of thermal neutron beam collimated from beam guide is $2{\times}2cm^2$ at the sample position. The neutron flux measured was the range of $1.0{\sim}6.5{\times}10^7n{\cdot}cm^{-2}{\cdot}s^{-1}$, and flux distribution from center within the radius of 4.5 mm and 9.0 mm was $5.77{\pm}0.71{\times}10^7n{\cdot}cm^{-2}{\cdot}s^{-1}$ and $4.68{\pm}1.64{\times}10^7n{\cdot}cm^{-2}{\cdot}s^{-1}$, respectively. Accordingly, sample size is adjusted within 10 mm for a homogeneous irradiation of high quality. Measurement system is designed to reduce the background source by Compton scattering and to improve the analytical sensitivity. To investigate the energy calibration and Compton suppression effect of gamma-ray counting system, the background conditions on both of Compton and single-mode were measured using NaCl standard. On the other hand, degree of spectral interference for sodium 472 keV peak as a matrix effect in the sample is established for an accurate boron analysis, and then boron content in three certified reference materials (NIST SRM 1570a, 1547, 1573a) was measured by using two modes and the results were compared with each other.

Calculation of Nuclear Characteristics of the TRIGA Mark-III Reactor (TRIGA Mark-III 원자로의 노심특성계산)

  • Chong Chul Yook;Gee Yang Han;Byung Jin Jun;Ji Bok Lee;Chang Kun Lee
    • Nuclear Engineering and Technology
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    • v.13 no.4
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    • pp.264-276
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    • 1981
  • A simulation procedure which can represent time-dependent nuclear characteristics of TRIGA Mark-III reactor is developed. CITATION, a multi-group diffusion-depletion program, has been utilized as calculational tool. The group structure employed in this study consists of 7 groups: -3-fast and 4-thermal-which is conventionally utilized in TRIGA type reactor analysis. Three-dimensional nuclear characteristics are synthesized by combining results from two-dimensional plane calculation and two-dimensional cylinder calculation, since direct three-dimensional approach is not yet possible. An effort ia made to develope a method which can extract effective zone and group dependent bucklings by neutron diffusion theory rather than conventional zone and/or group independent Ducklings by neutron transport theory, since neutron leakage is quite high for small core such as research reactors. It is turned out that the method developed in this study gives satisfactory results. The calculation is performed under assumptions that all control rods are fully withdrawn, that no samples are inserted in the irradiation holes and that the core is located in the center of the reactor pool. Burnup-dependent variation of core excess reactivity, time dependent change of Xe-135 poisoning and reactivity worth of rotary specimen rack are calculated and compared with operation records. Neutron flux and power distribution as well as neutron spectrum in each irradiation .facility are presented.

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