Thermal neutron albedo and flux for different geometries neutron guide

S. Azimkhani,D. Rezaei Ochbelagh,F. Zolfagharpour 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.4

This paper presents a study on thermal neutron reflection properties of neutron guide for cylinder,spindle, elliptic and parabolic geometries using 241Am-Be neutron source (5.2 Ci) and BF3 detector,whereas neutron guide is important instrument for transportation of neutrons. To this goal, the requiredinner and outer radii of neutron guide have been calculated to achieve the highest guided thermalneutron flux based on MCNPX Monte Carlo code. The maximum flux of cylinder geometry with a length50 cm has been obtained at an inner radius 9 cm and an outer radius 21 cm. Also, the maximum value ofthermal neutron albedo is 0.46 ± 0.001 at 12 cm thickness of parabolic guide.

Monte Carlo Calculation of Thermal Neutron Flux Distribution for (n, γ) Reaction in Calandria

Kim,Jong-Kyung,Kim,Kyo-Youn,Kim,Soon-Young 대한방사선 방어학회 1994 방사선방어학회지 Vol.19 No.1

CANDU 6 중수형 원자로 운전중에 Calandria Shell내에서 발생하는 (n, γ) 반응유발 열중성자속분포와 CANDU 6 발전소의 측면 및 하단 차폐구조에서의 방사선 선량률을 계산하기 위하여 몬테칼로 방법을 이용한 MCNP 4.2 코드를 사용하였다. 계산결과와 비교해 볼 때 약간 큰 값들의 분포를 보여주고 있다. 이 계산결과의 응용으로서 작업자 접근가능지역(Worker Accessible Areas)에서의 감마 선량률을 계산해본 결과 설계목표치인 6 μ㏜/h 보다 낮은 값을 주는 것으로 나타났다. (n, γ) 반응유발 열중성자속분포에 대한 MCNP 4.2 코드의 계산결과는 CANDU 6형 원자로의 방사선 차폐해석에 중요한 자료로 널리 이용될 수 있을 것이다. The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for (n, γ)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∼10 neutrons/㎠-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 μ㏜/h. The methodology used in this study to evaluate the thermal neutron flux distribution for(n, γ)reaction can be applied to radiation shielding analysis of CANDU 6 type plants.

Thermal-hydraulic safety analysis of radioisotope production in HANARO using MCNP6 and COMSOL multiphysics: A feasibility study

김태윤,한보영,양성우,이재기,선광민,박병건,예성준 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.11

The High-flux Advanced Neutron Application Reactor (HANARO) produces radioisotopes (RIs) (131I, 192Ir, etc.) through neutron irradiation on various RI production targets. Among them, 177Lu and 166Ho are particularly promising owing to their theranostic characteristics that facilitate simultaneous diagnosis and treatment. Prior to neutron irradiation, evaluating the nuclear heating of the RI production target is essential for ensuring the thermal-hydraulic safety of HANARO. In this study, the feasibility of producing 177Lu and 166Ho using irradiation holes of HANARO was investigated in terms of thermal-hydraulic safety. The nuclear heating rates of the RI production target by prompt and delayed radiation were calculated using MCNP6. The calculated nuclear heating rates were used as an input parameter in COMSOL Multiphysics to obtain the temperature distribution in an irradiation hole. The degree of temperature increase of the 177Lu and 166Ho production targets satisfied the safety criteria of HANARO. The nuclear heating rates and temperature distribution obtained through the in silico study are expected to provide valuable insight into the production of 177Lu and 166Ho using HANARO.

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

Ahmed Abdelghafar Galahom 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.3

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO2) and uranium zirconium hydride (UZrH1.6) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO2 contains 8 × 8 fuel rods while that fueled with UZrH1.6 contains 9 × 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. UZrH1.6 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.

The Distribution of Thermal Neutron Fluxes into the Self-Shielded Wall Equipped with a Medical Self-Shielded PET (Positron Emission Tomography) Cyclotron

M. Sakama,T. Saze,K. Maeda,K. Akamatsu,E. Honda,H. Nishitani 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23

The role of systematic evaluations of thermal neutron fluxes originated in various medical small radiation accelerator facilities has become more and more important to establish uniformly a clearance system about radioactivities of radioactive waste materials produced, for example, when those facilities would be decommissioned and improved. The purpose of this paper was to investigate the distribution of thermal neutron fluxes into the self-shielded wall at the PET cyclotron in Tokushima University Hospital.The distribution of thermal neutron fluxes was determined using the activation Au foil method of ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction and the visualization was observed using the imaging plate (IP) of a long ribbon shaped activation Au thin foil. It was found that the thermal neutron fluxes were distributed from 6.65 × 10^6 cm^(-2) s^(-1) (at 1.5 cm distance from an inside wall into a polyethylene layer) to 1.20 × 10^2 cm^(-2) s^(-1) (at 71.0 cm distance into a heavy concrete layer), and also that the distribution trend of thermal neutron fluxes will be approximately consistent with that of the calculated data led by MCNP code. We have confirmed that it will be possible for the IP visualization of thermal neutron fluxes into the self-shielded wall at the PET cyclotron to reproduce those distribution quantitatively and over a wide area.

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 Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.3

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.

Effect of a Metal Electrode on the Radiation Tolerance of a SiC Neutron Detector

박준식,신희성,김호동,김한수,박세환,이철호,김용균 한국물리학회 2012 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.61 No.3

The Korea Atomic Energy Research Institute (KAERI) has developed a silicon carbide (SiC) diode as a neutron detector that can be used in harsh environments such as nuclear reactor cores and spent fuel. The radiation tolerance of the SiC detector was studied in the present work. Especially, the effect of a metal electrode on the radiation tolerance of the SiC detector was studied. Four different types of SiC detectors were fabricated, and the operation properties of the detectors were measured and compared before and after neutron irradiations of 2.16 ?10<SUP>15</SUP> n/cm<SUP>2</SUP> and 5.40 ?10<SUP>17</SUP> n/cm<SUP>2</SUP>. From the comparison, the detector with Ti/Au electrode structure shows the highest radiation tolerance among detectors. A detector assembly was fabricated using two types of SiC p-i-n diode detectors, which showed highest radiation tolerance. Signals from the detectors were measured with current mode to minimize the noise of the detector. Signal currents from detectors were measured for neutron fluxes ranging from 5.54 ?10<SUP>6</SUP> n/cm<SUP>2</SUP> s to 2.86 ?10<SUP>8</SUP> n/cm<SUP>2</SUP> s and gamma doses up to 100 Gy/h.

Optimized Design of a Facility for Measuring Sodium-24 in Blood by Using Monte Carlo Simulations

Jian-Bo Yang,Rui Li,Zhi Liu,Hong Huang 한국물리학회 2016 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.68 No.8

The Monte Carlo N-Particle Transport Code is adopted for numerical calculations and simulation analyses of the parameters for the designed irradiation facility for human blood (including the neutron moderator and its thickness, the neutron source’s location, and the collimator’s radius) and of the parameters for the unit for measuring human blood (such as the crystal thickness and radius of NaI) to measure the activity of 24Na in the human body exposed to neutron irradiation in nuclear accidents. Calculation results show that the most suitable parameters for the irradiation facility for human blood include 6-cm polyethylene as the neutron moderator, collimator radius of 7 mm, and a neutron source placed at the bottom of the collimator. However, the parameters for the unit to measure human blood are as follows: both the thickness and the radius of the crystal at the bottom of the NaI detector are 5 cm. The effectiveness of the design parameters was verified by using actual experiments.